L-MT-20-036, Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times

Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times

ML20356A131
Person / Time
Site:	Monticello
Issue date:	12/21/2020
From:	Conboy T Northern States Power Company, Minnesota
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-20-036
Download: ML20356A131 (146)

Text

2807 West County Road 75 Monticello, MN 55089 December 21, 2020 L-MT-20-036 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" (EPID L-2020-LLA-0062)

References:

1) Letter (L-MT-20-003) from NSPM to the NRC, "License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," dated March 30, 2020 (ADAMS Accession No. ML20090F820)

2) Letter from the Technical Specification Task Force (TSTF) to the NRC, "TSTF Comments on Draft Safety Evaluation for Traveler TSTF-505, Provide Risk-Informed Extended Completion Times and Submittal of TSTF-505, Revision 2," Revision 2, dated July 2, 2018 (ADAMS Accession No. ML18183A493)

3) Email from the NRC to NSPM, "Monticello Request for Additional Information RE: TSTF-505 license amendment request," dated October 26, 2020 (ADAMS Accession No.ML20302A197)

In Reference 1, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), submitted a license amendment request to the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP). The proposed amendment would modify TS requirements to permit the use of Risk-Informed Completion Times in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" (Reference 2). The NRC identified the need for additional information and provided the Request for Additional Information (RAI) in Reference 3. The enclosure to this letter provides NSPMs response to the NRC RAI.

The information provided in this letter does not alter the evaluations performed in accordance with 10 CFR 50.92 in Reference 1.

fl Xcel Energy8

Document Control Desk Page 2 NSPM is notifying the State of Minnesota of this request by transmitting a copy of this letter and enclosures to the designated State Official.

Please contact Mr. Ron Jacobson at 612-330-6542 or ronald.g.jacobson@xcelenergy.com if there are any questions or if additional information is needed.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

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

Executed on December 21, 2020.

Thomas A.

Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC State of Minnesota

L-MT-20-036 NSPM Enclosure Page 1 of 75 Response to Request for Additional Information License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b"

1.0 BACKGROUND

In Reference 1, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), submitted a license amendment request for the Technical Specifications (TS) of the Monticello Nuclear Generating Plant (MNGP). The proposed amendment would modify TS requirements to permit the use of Risk-Informed Completion Times (RICTs) in accordance with Technical Specification Task Force (TSTF)-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" (Reference 2).

The NRC identified the need for additional information and provided the Request for Additional Information (RAI) in Reference 3. The following section provides NSPMs response to the NRC RAI.

2.0 RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION RAI 1 - Probabilistic Risk Analysis (PRA) Upgrades Associated with PRA Model Updates Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," provides guidance for addressing PRA acceptability (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090410014). It describes a peer review process using the ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (the PRA Standard), as one acceptable approach for determining the technical acceptability of the PRA.

The PRA Standard defines PRA upgrade as "the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences." Section 1-5 of the PRA Standard states that upgrades of a PRA "shall receive a peer review in accordance with the requirements specified in the peer review section of each respective part of this standard."

Appendix 1-A of the PRA Standard identifies PRA upgrades as satisfying one of three criteria:

(1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, or (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

Section 4 of Enclosure 2 to the LAR states that the internal events and internal flooding PRA was subject to a full-scope peer review in April 2013 against RG 1.200, Revision 2. The LAR further states a closure review of facts and observations (F&Os) was completed in

L-MT-20-036 NSPM Enclosure Page 2 of 75 October 2017 in accordance with the process documented in Appendix X to Nuclear Energy Institute (NEI) 05-04/07-12/12-06, "Close Out of Facts and Observations (F&Os)" (ADAMS Accession No. ML17086A431). Section 2 of Enclosure 2 of the LAR states that one model change constituted a PRA upgrade, which became the subject of a separate focused-scope peer review. Section 4 of Enclosure 2 of the LAR states that the internal events PRA model has since been updated. It is unclear to the NRC staff if any changes incorporated into the model, unrelated to F&O closures, were evaluated for potential PRA upgrades since the April 2013 peer review. In light of these observations:

a)

Summarize the model changes performed for the internal events, including internal flooding, PRA since April 2013 that are not associated with the resolutions of closed F&Os. This description should be of sufficient detail to determine whether the changes are considered PRA maintenance or PRA upgrades as defined in the PRA Standard, Section 1-5.4, as qualified by RG 1.200. For each change, indicate whether the change was PRA maintenance or a PRA upgrade, along with justification for this determination.

b)

For any of these changes that are determined to be a PRA upgrade, confirm that focused-scope peer review(s) have been conducted. Describe the peer review(s) and status of the resulting F&Os. Provide any of these F&Os that are not yet closed, along with their disposition with respect to this application.

NSPM Response to RAI 1:

a)

The following notable changes to the MNGP Internal Events PRA model have occurred since the April 2013 Peer Review. Each change was evaluated using the NSPM PRA model maintenance process to determine if the change constituted an update or an upgrade as defined by Appendix 1-A of the PRA Standard (Reference 12). Except as noted, the changes described in this response were determined to be PRA updates.

Accident Sequence o Transient event tree updated to credit the High Pressure Coolant Injection (HPCI) long term for late injection due to the revision to the Torus Temperature branch of Emergency Operating Procedure (EOP), "Primary Containment Control," advising operators not to depressurize the reactor pressure vessel (RPV) if it would result in a loss of injection needed for core cooling. Model updates to reflect the as-operated plant that do not lead to a significant impact on the accident sequences are considered updates.

o Added potential for feedwater overfill event to the internal events model to align with the Fire PRA model. Change was determined to be model maintenance to allow for an All Hazards model to support RICT.

Data o PRA component boundaries were updated to match NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S.

Commercial Nuclear Power Plants" (Reference 4). Data updates are considered part of model maintenance.

L-MT-20-036 NSPM Enclosure Page 3 of 75 o Common Cause Frequencies were updated to ensure all appropriate component groups were included. Data updates are considered part of model maintenance.

Human reliability analysis (HRA) - Created new operator actions, refined existing actions, and combined similar actions. HRA methodologies for quantification and dependency analysis did not change. Model updates to reflect the as-operated plant that do not lead to a significant impact on the accident sequences are considered updates.

Initiating Events - No notable changes to initiating events were incorporated.

Internal Flooding Accident Sequence o Incorporated maintenance-induced flooding events into the PRA model.

Determined to be an extension of existing methodology that did not lead to a significant impact on the accident sequences and is therefore an update.

o Transitioned from using AddEvent to FRANX software for inserting flood logic into the model. PRA modeling tool change is considered part of model maintenance.

Internal Flooding Initiating Events - Updated selected significant flood initiator frequencies by using drawings as the source of pipe lengths instead of using estimates from plant walkdowns. Refinement of frequencies to reflect the as-built plant is classified as an update.

Level 2 - Level 2 event tree simplifications were made consolidating and combining similar event trees that made minimal contributions to the Large Early Release Frequency (LERF). Event tree simplifications that do not lead to a significant impact on the accident sequences are considered updates.

Quantification - Incorporation of a convolution analysis to estimate the probability of AC power recovery prior to run failure of the equipment supporting onsite power. This convolution analysis was determined to be a PRA upgrade and is described in the response to RAI 1b.

Success Criteria - No notable changes to the success criteria were incorporated.

Systems Analysis o Incorporation of plant modifications such as the Hard Pipe Vent modifications for Diverse and Flexible Coping Strategies (FLEX), pinning condensate demineralizer valves closed to prevent spurious movement, and a new train added for the fuel oil system. Model updates to reflect the as-built plant that do not lead to a significant impact on the accident sequences are considered updates.

o Modeled FLEX DC powered fuel oil transfer pumps in the PRA model following the NEI 16-06 guidance (Reference 5). These pumps replaced the previously modeled, portable, gas-powered fuel oil transfer pump. Additional description of the change is included in the response to RAI 11. Model updates to reflect the

L-MT-20-036 NSPM Enclosure Page 4 of 75 as-built/as-operated plant that do not lead to a significant impact on the accident sequences are considered updates.

Uncertainty - No notable changes to the uncertainty analysis were incorporated.

b)

One change to the internal events PRA model was self-determined to be a change in methodology and constituted a model upgrade. This upgrade was the incorporation of a convolution analysis to estimate the probability of AC power recovery prior to run failure of the equipment supporting onsite power. This PRA upgrade was reviewed in a focused-scope peer review which included supporting requirements AS-B7, SC-A5, SY-A10, and QU-A1 of the ASME/ANS RA-Sa-2009 Standard. All supporting requirements were met to Capability Category II/III. No finding level F&Os were identified during the focused-scope peer review.

RAI 2 - Generic Fire PRA Questions Relatively extensive and detailed reviews of fire PRAs were undertaken in support of each LAR to transition to National Fire Protection Association Standard 805 (NFPA-805)

"Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants, 2001 Edition" which has not been completed for Monticello.

The NRC staff evaluates the acceptability of the PRA for each new risk-informed application.

RG 1.200 states that "NRC reviewers... [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application." As discussed in RG 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," (ADAMS Accession No. ML17317A256) the staff recognizes that the acceptable technical adequacy of risk analyses necessary to support regulatory decision making may vary with the relative weight given to the risk assessment element of the decision-making process. Using more defensible and less simplified assumptions could substantively affect the fire risk and fire risk profile of the plant. The NRC staff notes that the calculated results of the PRA are used to estimate a risk-informed completion time (RICT), which subsequently determines how long systems, structures, and components (SSCs) controlled by TSs (both individual SSCs and multiple, unrelated SSCs) can remain inoperable. Therefore, the PRA results are given a very high weight in an application for TSTF-505 "Provide Risk-Informed Extended Completion Times" (ADAMS Accession No. ML18183A493). The NRC staff requests additional information on the following issues that have been previously identified as potentially key fire PRA assumptions:

a)

Use of unreviewed methods:

The LAR provides the history of the fire PRA peer review. Section 5 of LAR Enclosure 2 states that the March 2015 peer review determined that no unreviewed analysis methods were used. It is unclear to the NRC staff what criteria were used by the peer review team to determine what constitutes an unreviewed method. Methods may have been used in

L-MT-20-036 NSPM Enclosure Page 5 of 75 the fire PRA that deviate from NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," (ADAMS Accession Nos. ML052580075 and ML052580118) or other acceptable guidance that may have been used, e.g., frequently asked questions, NUREGs, or interim guidance documents.

i) Identify methods used in the fire PRA that deviate from guidance in NUREG/CR-6850 or other acceptable guidance.

ii) If such deviations exist, then justify their use in the fire PRA and impact on the RICT.

iii) As an alternative to item ii above, add an implementation item to replace those methods with a method acceptable to NRC prior to the implementation of the RICT program. If an implementation item is proposed, include a description of the replacement method along with justification that it is consistent with NRC-accepted guidance.

b)

Reduced transient heat release rates:

The key factors used to justify using transient fire reduced heat release rates (HRRs) below those prescribed in NUREG/CR-6850 are discussed in a letter from the NRC to NEI (ADAMS Package Accession No. ML12172A406). If any reduced transient HRRs below the bounding 98% HRR of 317 kW from NUREG/CR-6850 were used, discuss the key factors used to justify the reduced HRRs. Include in this discussion:

i) Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied.

ii) A description for each location where a reduced HRR is credited, and a description of the administrative controls that justify the reduced HRR including how location-specific attributes and considerations are addressed. Include a discussion of the required controls for ignition sources in these locations and the types and quantities of combustible materials needed to perform maintenance. Also, include discussion of the personnel traffic that would be expected through each location.

iii) The results of a review of records related to compliance with the transient combustible and hot work controls.

c)

Obstructed plume model:

NUREG-2178, "Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE -FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume," (ADAMS Accession No. ML16110A140) contains refined peak HRRs, compared to those presented in NUREG/CR-6850, and guidance on modeling the effect of plume obstruction. Additionally, NUREG-2178 provides guidance that indicates that the obstructed plume model is not applicable to cabinets in which the fire is assumed to be located at elevations of less than one-half of the cabinet.

i) If obstructed plume modeling was used, then indicate whether the base of the fire was assumed to be located at an elevation of less than one-half of the cabinet.

L-MT-20-036 NSPM Enclosure Page 6 of 75 ii) Justify any modeling in which the base of an obstructed plume is located at less than one-half of the cabinet's height.

iii) As an alternative to item ii above, add an implementation item to remove credit for the obstructed plume model in the fire PRA prior to the implementation of the RICT program.

d)

Systems not credited in the fire PRA:

The NRC staff notes that some conservative PRA modeling could have a nonconservative impact on the RICT calculations. If an SSC is part of a system not credited in the fire PRA or it is supported by a system that is assumed to always fail, then the risk increases due to taking that SSC out of service is masked. Therefore, address the following:

i) Identify the systems or components that are assumed to be always failed in the PRA or not included in the PRA (due to lack of cable tracing or other reasons). Justify that this assumption has an inconsequential impact on the RICT calculations.

ii) As an alternative to item (i), above, propose a mechanism to ensure that a sensitivity study is performed for applicable SSCs which accounts for the impact on the RICT of the 1) nonconservative PRA assumption of failed SSCs or 2) SSCs not included in the PRA model. The proposed mechanism should also ensure that any additional risk from correcting the false assumption that the SSC is always failed is either accounted for in the RICT calculation or is compensated for by applying additional risk management actions (RMAs) during the RICT.

e)

Well-sealed Motor Control Center (MCC) cabinets:

Guidance in Frequently Asked Question 08-0042, "Fire Propagation from Electrical Cabinets" (from Supplement 1 of NUREG/CR-6850) applies to electrical cabinets below 440 V. With respect to Bin 15 as discussed in Chapter 6, it clarifies the meaning of "robustly or well-sealed." Thus, for cabinets of 440 V or less, fires from well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440 V and higher, the original guidance in Chapter 6 remains and recommends that Bin 15 panels which house circuit voltages of 440 V or greater are counted because an arcing fault could compromise panel integrity (an arcing fault could burn through the panel sides, but this should not be confused with the high energy arcing fault type fires)."

Fire PRA FAQ 14-0009, "Treatment of Well-Sealed MCC Electrical Panels Greater than 440V" (ADAMS Accession No. ML15119A176) provides the technique for evaluating fire damage from MCC cabinets having a voltage greater than 440 V. Therefore, propagation of fire outside the ignition source panel must be evaluated for all MCC cabinets that house circuits of 440 V or greater.

i) Describe how fire propagation outside of well-sealed MCC cabinets greater than 440 V is evaluated.

L-MT-20-036 NSPM Enclosure Page 7 of 75 ii) If well-sealed cabinets less than 440 V are included in the Bin 15 count of ignition sources, provide justification for using this approach.

f)

Fire PRA methods for outdated fire PRA and peer review:

LAR Enclosure 9, Section 3 states that the Monticello fire PRA was developed using consensus methods outlined in NUREG/CR-6850 and interpretations of technical approaches as required by the NRC. Part (e) of proposed TS 5.5.16 states that the approaches and methods used in the RICT program shall be acceptable to the NRC.

Methods to assess risk must be those used to support the LAR or other methods approved by NRC for generic use.

There have been some changes to the fire PRA methodology since the development of the Monticello fire PRA that was peer reviewed. The integration of NRC-accepted fire PRA methods and studies described in NUREG-2180 "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE)," (ADAMS Accession No. ML16343A058) that are relevant to this submittal could potentially impact the TSTF-505 results, core damage frequency (CDF), or large early release fraction (LERF).

Section 2.5.5 of RG 1.174 provides guidance that indicates additional analysis is necessary to ensure that contributions from the above influences would not change the conclusions of the LAR.

i) Provide a detailed justification for why the integration of the above NRC-accepted fire PRA methods and studies would not significantly impact the RICT calculation. As part of this justification, identify potential fire PRA methodologies used in the fire PRA that are no longer accepted by the NRC staff. Provide technical justification for methods in your PRA that have not yet been accepted by the NRC staff and evaluate the significance of their use on the RICT estimates.

ii) Alternatively, if the above guidance has been implemented in your PRA, provide the following:

1) Indicate whether the changes to the fire PRA are PRA maintenance or a PRA upgrade as defined in the PRA Standard, Section 1-5.4, as qualified by RG 1.200, along with justification for the determination.

2) Discuss any focused-or full-scope peer review performed to evaluate the changes that were determined in item 1 above to constitute a PRA upgrade. Including when the peer review was performed and when the peer review report that evaluated the upgrade was approved.

NSPM Response to RAI 2:

a)

The MNGP method for apportionment of Main Control Board ignition frequency differs from the method described in NUREG/CR-6850 Appendix L (Reference 6). The MNGP method was peer reviewed and is appropriate because the ignition frequency is apportioned based on the density of cables and thus the concentration of sources (e.g.,

L-MT-20-036 NSPM Enclosure Page 8 of 75 connections, switches) that are more likely to cause a fire. A comparison of the method used in the Monticello Fire PRA model against the NUREG/CR-6850 Appendix L method showed that the method used in the Monticello Fire PRA calculates a higher likelihood for the different fire scenario sequences modeled.

b)

Reduced transient fire HRR below the bounding 98th percentile is not credited in the Monticello Fire PRA model.

c)

The update for obstructed plume guidance from NUREG-2178 (Reference 7) has not been implemented in the MNGP Fire PRA model.

d)

The following systems and components were not credited in the Monticello Fire PRA model due to lack of cable selection:

Standby Liquid Control (SBLC),

Anticipated Transient Without Scram (ATWS),

the K-1E air compressor, the Screen Wash pump (P-104),

the Main Condenser including support components, Reactor and Turbine Building non-safety related HVAC units, Multiple Spurious Operation AC loads, Control Rod Drive Hydraulic (CRDH) pump suction support equipment, and the L-41 AC panel.

Since the SBLC system has associated Limiting Condition for Operations (LCOs) in scope for the RICT program, adding it to the Monticello Fire PRA model will be included as an implementation item.

Also, it was found that the majority of the risk contribution for the not credited equipment above was attributed to the L-41 AC panel which supports the pneumatic supply to the SRVs. Therefore, modeling the L-41 AC panel in the Monticello Fire PRA model also will be added as an implementation item.

A sensitivity study was performed that removed all fire induced failures for the remaining systems (Condenser, HVAC, ATWS, etc.). This sensitivity study showed that the delta risk for CDF and LERF is 0.7% and 1.3% respectively when compared to the base RICT model values. This small delta CDF/LERF shows that the impact of assuming that these remaining systems are failed is negligible and will not significantly affect calculated RICTs.

e)

For MNGP, NSPM follows the NUREG/CR-6850 guidance on fire propagation for electrical cabinets above 440V (i.e., no credit for well-sealed cabinets is taken).

f)

MNGP does not have an incipient detection system at the plant therefore the NUREG-2180 (Reference 8) guidance does not apply.

L-MT-20-036 NSPM Enclosure Page 9 of 75 RAI 3 - Joint Human Error Probability Floor NUREG-1921 "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines-Final Report,"

(ADAMS Accession No. ML12216A104) discusses the need to consider a minimum value for the joint probability of multiple human failure events (HFEs) in human reliability analyses (HRAs). NUREG-1921 refers to Table 2-1 of NUREG-1792, "Good Practices for Implementing Human Reliability Analysis (HRA)," April 2005 (ADAMS Accession No. ML051160213), which recommends that joint human error probability (HEP) values should not be below 1E-5. Table 4-4 of EPRI 1021081, "Establishing Minimum Acceptable Values for Probabilities of Human Failure Events," provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. Therefore, the guidance in NUREG-1921 allows for assigning joint HEPs that are less than 1E-5, but only through assigning proper levels of dependency.

TSTF-505 evaluations use the fire PRA and the internal events PRA. The LAR does not provide information about whether and, if so what, minimum joint HEP value is currently assumed. Also, even if the assumed minimum joint HEP values are shown to have no impact on the PRA risk estimates, it is not clear to the NRC staff how it will be ensured that the impact remains minimal for future PRA model revisions. In light of these observations:

a)

For the internal events PRA:

i) Explain what minimum joint HEP value was assumed.

ii) If a minimum joint HEP value less than 1E-06 was used, describe the sensitivity study that was performed and the quantitative results that justify that the minimum joint HEP value has no impact on the RICT application.

iii) If the minimum joint HEP value evaluated in ii cannot be shown to have no impact on the application, confirm that it is justified by demonstrating that the EPRI 1021081 lower value guideline does not apply (e.g., using such criteria as the dependency factors identified in NUREG-1921 to assess level of dependence).

b)

For the fire PRA:

i) Explain what minimum joint HEP value was assumed.

ii) If a minimum joint HEP value less than 1E-05 was used, describe the sensitivity study that was performed and the quantitative results that justify that the minimum joint HEP value has no impact on the RICT application.

iii) If the minimum joint HEP value evaluated in ii cannot be shown to have no impact on the application, confirm that it is justified by demonstrating that the EPRI 1021081 lower value guideline does not apply (e.g., using such criteria as the dependency factors identified in NUREG-1921 to assess level of dependence).

c)

Estimate how many joint HEP values addressed in a and b fall below the guideline values. Discuss the range of values used. For each PRA, provide examples where this justification is applied (at least two if they exist).

L-MT-20-036 NSPM Enclosure Page 10 of 75 NSPM Response to RAI 3:

a)

For the Internal Events PRA:

i)

The Internal Events PRA model of record utilizes a minimum joint HEP value of 1E-06.

ii)

Not applicable - see response to RAI 3.a.i.

iii)

Not applicable - see response to RAI 3.a.i.

b)

For the Fire PRA:

i)

The Fire PRA model of record default joint HEP floor is 1E-05, except for the combinations containing either a long term LERF or decay heat removal HFE. In those long-term cases, the joint HEP floor was lowered to 1E-06. This change in joint HEP floor was done as part of resolving F&O 1-19 and documented in the Fire HRA notebook. This resolution to the finding F&O was subsequently reviewed and closed as part of the finding closure review described in Enclosure 2 of the LAR.

ii)

No sensitivity study has been performed - see response to RAI 3.b.i. for a discussion of the justification for lowering the HEP floor for long term actions.

iii)

EPRI 1021081 guidance was not utilized to determine when use of a joint HEP floor below 1E-05 is appropriate. The long term criteria discussed previously was applied during resolution of finding F&O 1-19. Using a lower than 1E-05 joint HEPs for those combinations that contain long term LERF or decay heat removal HFE has been reviewed and accepted as part of peer review finding closure process. Although the combinations with long term LERF or decay heat removal HFE may meet the criteria to be considered "independent" or "very low" dependence from the EPRI report, this guidance has not been directly applied.

c)

The PRA model used in the CRM model is a single fault tree that contains a single top events for all hazard CDF and LERF. Due to this all hazard construction, the all hazard joint HEP floor is initially set to 1E-05 for all combinations regardless of initiator (Internal Events, Internal Flooding, or Fire). The joint HEP floor is then lowered to 1E-06 for combinations, containing either a long term LERF or decay heat removal HEP as described in 3.b.i. The table below shows the breakdown of the range of joint HEP values.

L-MT-20-036 NSPM Enclosure Page 11 of 75 Frequency Number of Combinations Percent 1.0E-06 4816 60.61%

1.0E <1.0E-05 615 7.74%

>= 1.0E-05 2515 31.65%

Note: During the RICT calculation process an error was discovered that resulted in some joint HEP values inappropriately being lowered to 1E-06 which is below their dependent value as calculated by the HRA Calculator. A sensitivity study was performed with corrected joint HEP values to determine the impact of this issue on the base RICT model as well as sample RICTs. The sensitivity showed less than 1% change in sample RICT calculations and no impact on the total CDF/LERF base values. This issue was added to the PRA change tracking process and will be resolved before RICT program implementation. The values in this table reflect the distribution of the corrected combinations.

The combinations with the joint HEP floor lowered to 1E-06 will be reviewed and the justifications will be documented before RICT program implementation (as mentioned earlier in this response). This justification will include a review of human failure events (HFE) combinations to determine if any below 1E-05 are both >1% and in the top 95% of results of CDF and LERF. Each combination that meets the criteria will be reviewed to determine if it is based on the correct dependence. Combinations with dependence that is not correct due to software limitations will be justified for their acceptability and documented.

RAI 4 - PRA Model Update Process Section 2.3.4 of NEI 06-09, "Risk-Managed Technical Specifications (RMTS) Guidelines" (ADAMS Accession No. ML12286A322), specifies that "[c]riteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations."

LAR Enclosure 7 states that if a plant change or a discovered condition is identified with potential significant impact on the RICT calculations then an unscheduled update of the PRA models will be implemented. More specifically, the LAR states that if the plant changes meet specific criteria defined in the plant PRA and update procedures, including criteria associated with consideration of the cumulative risk impact, then the change will be incorporated into applicable PRA models without waiting for the next periodic PRA update.

The LAR does not explain under what conditions an unscheduled update of the PRA model will be performed or the criteria defined in the plant procedures that will be used to initiate the update.

L-MT-20-036 NSPM Enclosure Page 12 of 75 a)

Describe the conditions under which an unscheduled PRA update (i.e., more than once every two refueling cycles) would be performed and the criteria that would be used to require an unscheduled PRA update.

b)

Describe how the impact on the RICT program is considered when reviewing plant changes or conditions for implementation in the PRA. In the response define what is meant by "significant impact to the RICT Program calculations."

NSPM Response to RAI 4:

a)

NSPM uses a living model concept for updates to the PRA. If PRA changes are implemented that are determined to meet certain cumulative thresholds (e.g., Mitigating System Performance Index (MSPI) Birnbaum, delta CDF/LERF) or if a significant impact to a PRA application is predicted, the current living model is documented and quantified as an application specific model. The newly documented model is then used to update impacted applications as needed.

b)

The impact on PRA applications, including the RICT program, is reviewed on a quarterly basis by assembling and quantifying the latest living model files. The results of this quantification are compared against cumulative thresholds to determine if an update to one or more applications is required.

Significant impact is defined as a greater than 25% change in all hazards total CDF or LERF or an item that calls into question the technical adequacy or capability of the PRA.

The quarterly quantification may result in an update to the RICT program. An item that has a significant impact will result in an update to the RICT program.

RAI 5 - Implementation Items The NRC safety evaluation (ADAMS Accession No. ML071200238) approving NEI 06-09 states that "RG 1.174, Revision 1, and RG 1.200, Revision 1, define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change." NEI 06-09 states that the "PRA shall be reviewed to the guidance of [RG 1.200]

Rev. 0 for a PRA which meets Capability Category (CC)-II for the supporting requirements of the ASME internal events at power PRA Standard. Deviations from these capability categories relative to the RMTS program shall be justified and documented." NEI 06-09 further clarifies that the "PRA shall be maintained and updated in accordance with approved station procedures to ensure it accurately reflects the as-built, as-operated plant."

LAR Attachment 5 lists three implementations items that must be complete prior to implementation of the program for RICT to satisfy the guidance that the PRA reflect the as-built, as-operated plant and that the PRA technical adequacy is acceptable prior to implementation of the RICT program. The three implementation items are:

L-MT-20-036 NSPM Enclosure Page 13 of 75 NSPM shall ensure that reactor protection system (RPS) instrumentation is modeled in the Monticello PRA with sufficient detail to accurately calculate a RICT.

NPSM shall ensure that mechanical vacuum pump system (MVP) and isolation instrumentation are modeled in the MNGP PRA with sufficient detail to accurately calculate a RICT.

NPSM shall ensure that the automatic depressurization system (ADS) and instrumentation is modeled in the Monticello PRA with sufficient detail to accurately calculate a RICT.

also states that if implementation of any of these changes constitute a PRA upgrade as defined in the PRA Standard as endorsed by RG 1.200, then a focused-scope peer review will be completed and that any findings will be resolved and incorporated in the PRA prior to the implementation of the RICT program. However, it is unclear to the NRC staff how the addition of these system models will meet CC II of the PRA Standard for internal events and fire PRA.

Provide the details how each of the above systems will be adequately modeled and in accordance with the PRA Standard Capability Category II. Justify how the proposed modeling is sufficient for the RICT program. Include in this discussion how mechanical components, instrument channels, logic components, and other relevant system components will be modeled. For the RPS system, also address the items in RAI 6.a below.

NSPM Response to RAI 5:

The RPS instrumentation, MVP isolation instrumentation, and ADS instrumentation will be added to the internal events and fire PRA models in accordance with CC-II for the supporting requirements of the ASME internal events at power PRA Standard as endorsed by RG 1.200, Rev. 2 (Reference 13). This means that for each of these systems their impacts on success criteria, accident sequence, human reliability, initiating events, preventive and corrective maintenance, common cause, and data analysis will be assessed and documented in accordance with the PRA Model Update Process described in the LAR Enclosure 7. System modeling in accordance with RG 1.200, Rev. 2, meets the PRA quality requirements for the RICT program as described in NEI 06-09.

RPS logic, as will be modeled in PRA, encompasses various sensors and relays that monitor signals from several functions as are listed in TS Table 3.3.1.1-1. All channels of each of these functions will be modeled in the PRA. NUREG/CR-6928 (Reference 4) will be used for the RPS component boundaries and data. TS functions for which a RICT could be applied are described in Table E1-1.

For each MVP instrumentation function where a RICT could be applied as described in Table E1-1, all channels will be modeled in the PRA. The component boundaries and data from NUREG/CR-6928 will be used.

L-MT-20-036 NSPM Enclosure Page 14 of 75 Similarly, for each ADS instrumentation function where a RICT could be applied as described in Table E1-1, all channels will be modeled in the PRA. The component boundaries and data from NUREG/CR-6928 will be used.

RAI 6 - Instrumentation and Controls Concerning the quality of the PRA model, NEI 06-09 states, "RG 1.174, Revision 1, and RG 1.200, Revision 1 define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change."

Based on documentation in the LAR, it is not clear to the NRC whether instrumentation and controls (I&C) are always modeled in sufficient detail to support implementation of TSTF-505.

The following additional information is requested:

Explain how I&C is modeled in the PRA. Include the following:

a) the scope of the I&C equipment that is explicitly included (e.g., bistables, relays, sensors, integrated circuit cards) b)

description of the level of detail that is modeled (e.g., are all channels of an actuation circuit modeled?)

c) discussion of what data and whether plant-specific data is used d) discussion of the associated TS functions for which a RICT can be applied.

NSPM Response to RAI 6:

The I&C components such as relays and sensors are explicitly modeled in the PRA. For example, Emergency Core Cooling System (ECCS) logic, as modeled in the PRA, is comprised of various sensors that monitor reactor pressure, reactor water level, and drywell pressure. Input from these instruments is processed through relay logic, which in turn provides output signals that control operation of the Residual Heat Removal (RHR),

Core Spray (CS), HPCI, and Reactor Core Isolation Cooling (RCIC) systems. Reactor low level logic is used to provide start signals to RHR, CS, HPCI, and RCIC. Reactor high level logic provides a trip signal to HPCI and RCIC. Reactor low pressure logic provides pump start and valve open permissive for RHR and CS. Drywell high pressure logic provides automatic start signals to RHR, CS, and HPCI, as well as signals to isolate primary containment.

The PRA model includes relays and sensors for all trains and channels for the systems that are modeled in the PRA. For example, both trains of RHR and CS are modeled such that ECCS Logic System Division I actuation circuits actuate Division I components and ECCS Logic System Division II actuation circuits actuate Division II components. HPCI operates as a single train system and is actuated from ECCS logic system Division I and/or ECCS logic system Division II sensors. The low-low reactor level logic actuates one set of relays and the high drywell pressure logic actuates another set of relays.

L-MT-20-036 NSPM Enclosure Page 15 of 75 Either set of relays then actuate motor operated valves (MOVs), HPCI pump speed and suction controls, and the pump start relay. Similarly, RCIC operates as a single train system and is actuated from ECCS Logic System Division I and/or ECCS Logic System Division II sensors. The low-low reactor level logic actuates RCIC relays. Auxiliary contacts of these relays actuate the RCIC MOVs, pump speed controls, and the pump start circuit.

The PRA utilizes generic failure data directly for non-risk significant basic events, and also utilizes generic data combined with plant specific data in the Bayesian update process used to determine the risk significant failure rates / probabilities. The data used to develop the common cause failure (CCF) parameter values is periodically updated by the NRC through its research organization and the data was available through the NRC website. The primary source of generic component failure data is the report NUREG/CR-6928. Although NUREG/CR-6928 has a great deal of equipment failure data, the MNGP model has some equipment /failure modes that are not covered in it. In these cases, the following secondary data sources were examined for suitability:

Failure to scram generic data were taken from NUREG-5500, Volume 3 (Reference 9).

Digital feedwater and electrical pressure regulator (EPR) control system generic failure data (type codes CK E and DF E) were taken from the PRA calculation "Type Codes," which refers to NUREG-75 (WASH-1400) (Reference 25).

Spurious operation of a manual switch (failure to remain closed, type code SM L) generic failure rate is taken from the "Emergency Diesel Generator System Quantification" calculation, which refers to NUREG-75 (WASH-1400).

TS functions for which a RICT could be applied are described in the revised Table E1-2 in of this Enclosure.

RAI 7 - Digital Instrumentation and Controls Section 2.3.4 of NEI 06-09 states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT program. The NRC safety evaluation for NEI 06-09 states that this consideration is consistent with Section 2.3.5 of RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications" (ADAMS Accession No. ML100910008). NEI 06-09 further states that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties which could potentially impact the results of a RICT calculation and that sensitivity studies should be used to develop appropriate compensatory RMAs.

The NRC staff understands that Monticello has digital control systems. Regarding digital I&C, NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed applications. In addition, known modeling

L-MT-20-036 NSPM Enclosure Page 16 of 75 challenges exist such as the lack of industry data for digital I&C components, the difference between digital and analog system failure modes, and the complexities associated with modeling software failures including common-cause software failures. Though reliability data from vendor tests may be available, this source of data is not a substitute for in-the-field operational data. Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT program.

a)

Provide the results of the sensitivity study performed for each digital system in the PRA model justifying that the uncertainty associated with modeling the system does not impact the RICT calculations. Also, present the baseline failure probability for the digital system (that is increased by a factor of 50 in the sensitivity case).

b)

For any other digital systems that are credited in the PRA models, provide the results of a sensitivity study on the SSCs in the RICT program demonstrating that the uncertainty associated with modeling digital I&C systems has inconsequential impact on the RICT calculations.

c)

As an alternative to item (b) above, identify which Limiting Conditions for Operation (LCOs) are determined to be impacted by digital I&C system modeling for which RMAs will be applied during a RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation require additional RMAs.

NSPM Response to RAI 7:

a)

The MNGP PRA model includes the digital feedwater control system, digital pressure controller for RHR service water, electric pressure regulator for turbine steam admission and bypass valves, and the HPCI and RCIC flow controllers. Also, the power range neutron monitoring system includes a digital control monitoring system, consisting of six subsystems described below.

The digital feedwater control system is designed to regulate the reactor feedwater system in supplying water to the reactor primary vessel such that proper reactor vessel water level is maintained. The level of the water in the reactor is controlled by a feedwater controller which receives inputs from reactor vessel water level, steam flow, and feedwater flow transmitters. The reactor feedwater flow regulating valves fail "as is" and the valves may be switched to manual control in the event of failure. The instrumentation for the reactor feedwater system is separate from reactor protection system instrumentation, thereby preventing control system failure from affecting the operation of the protection system.

Digital pressure controller (DPC-4103) is used to manually adjust the differential pressure across RHR heat exchanger. DPC-4103 provides redundant pressure control of the service water side of the RHR Service Water (SW) heat exchanger to ensure that service water pressure remains higher than RHR pressure to prevent release of radioactive material from the primary system to the environment.

L-MT-20-036 NSPM Enclosure Page 17 of 75 The digital electric pressure controller is a proportional type programmable controller (micro-processor based which compares the main steam line pressure at the inlet of the turbine with a setpoint to position the turbine steam admission valves and bypass valves.

The HPCI and RCIC flow controllers maintains the required system flow. The HPCI flow controller provides a flow signal to the turbine control logic (turbine governor) and HPCI flow indication in the control room over the range of 0-3500 gpm. The HPCI flow controller compares the actual system flow to an operator specified setpoint and output a signal to the turbine control logic (turbine governor) to control system flow. Similarly, the RCIC flow controller provides a RCIC pump discharge flow signal to the turbine control logic (turbine governor) and provides indication and manual adjustment capability to turbine control. The governor valve and the electric governor magnetic (EGM) are included in the HPCI turbine component boundary and the RCIC speed control system and the turbine governor valve are included within the component boundary of the RCIC pump.

The power range neutron monitoring systems, is a digital control monitoring system, consists of six subsystems: the local power range monitor (LPRM), the average power range monitor (APRM), the rod block monitor (RBM), the 2-of-4 voter logic system, the oscillation power range monitor (OPRM), and the recirculation flow instrumentation. The power range neutron monitoring system measures the neutron flux in the reactor from 0% power to 125% of full power operation.

1)

The LPRM subsystem is designed to continuously monitor the neutron flux level in the reactor.

2)

The APRM is designed to provide a continuous, accurate indication of the core average power and to SCRAM the reactor when power limits are exceeded. There are four APRMs and each monitors flux inputs from 24 LPRM detectors and the recirculation flow from each recirculation loop.

3)

The RBM is an operational aid designed to prevent violation of the fuel integrity safety criteria during withdrawal of a single control rod. The RBM also provides a local relative power communication signal for operator evaluation during control rod movement.

4)

The 2-of-4 voter logic system functions as an interface between the four APRM/OPRM channels and the RPS. Each 2-of-4 Voter is designed to provide a trip output signal to an RPS channel when it has received trip signals from at least two of the four APRMs or two of the four OPRMs.

5)

The OPRM System is designed to monitor groups of LPRM detector signals to detect core thermal-hydraulic instabilities and SCRAM the reactor when limits are reached. Each of the four OPRM channels will generate an RPS trip signal to terminate the event.

L-MT-20-036 NSPM Enclosure Page 18 of 75

6)

Each APRM calculates a value for total recirculation flow rate based on flow input from both recirculation loops. The value is used to determine alarms and trips for Simulated Thermal Power (STP). The OPRM System uses the total flow value and the power level to define where the OPRM System oscillation trips and alarms are enabled. The RBMs compare the total flow values from each of the APRM instruments to alert operators to gross flow differences between APRM instruments.

A sensitivity study was performed where the digital feedwater control system signal failure basic event, digital pressure controller (DPC-4103) fails to function basic event, and digital electric pressure controller basic event were increased by a factor of fifty to account for the potential for uncertainty related to digital equipment failure rates.

Although, HPCI and RCIC flow controllers are not explicitly modeled in the PRA, a sensitivity study was performed by adding Process Logic (level) fail to operate failure probability from NUREG-6928 to the steam driven turbine fail to start probability. This change increased the original HPCI and RCIC pump failure to start probability by 14%.

The current PRA model has a basic event with a point estimate of 3.80E-6 that represents RPS electrical system failure. The failure probability for this basic event was estimated in accordance with NUREG-5500, Volume 3. It should be noted that NUREG-5500, Volume 3 provides this point estimate by only considering reactor high pressure trip, reactor low water level trip, and actuation of manual scram switch. Since the current point estimate for the electrical system failure does not include all the potential reactor trip signals, it is reasonable to consider that in most initiators the failure probability would be lower than currently modeled if all the reactor protection trip signals were included.

Furthermore, power range neutron monitoring system is a completely independent neutron flux trip system. Therefore, detailed modeling of the power range neutron monitoring system per implementation item 1 is expected to lower the total failure probability of the reactor protection system. While no data specifically associated with the digital power range neutron monitoring system or equivalent analog system is currently available in NUREG-6928, a range of values associated with process logic are available.

A failure rate at the approximate midpoint of the failure rates for process logic was assumed for the baseline failure rate of the neutron monitoring system to estimate the impact of detailing modeling. Using this assumed value, a combined system failure probability was calculated by multiplying the assumed failure rate (i.e., Process Logic (level) fail to operate failure probability) from NUREG-6928 and the existing point estimate. This had the net effect of lowering the base failure probability of the RPS electrical failure. Then the assumed neutron monitoring system failure rate was multiplied by a factor of fifty to account for the potential for uncertainty related to digital equipment failure rates to recalculate a "sensitivity" failure probability. This "sensitivity" failure probability remains below the original point estimate from NUREG-5500.

The following table present the baseline failure probability for the digital system and the failure probability used in the sensitivity study:

L-MT-20-036 NSPM Enclosure Page 19 of 75 Digital System Baseline Failure Probability Probability used for the Sensitivity Study Feed Water System 6.43E-05 3.22E-03 RHR SW Pressure Controller 1.97E-05 9.85E-04 Digital EPR Turbine Controller 6.43E-05 3.22E-03 HPCI Flow Controller 4.48E-03 5.11E-03 RCIC Flow Controller 4.48E-03 5.11E-03 RPS Electrical 2.38E-09 1.19E-07 (3.80E-06 used for conservatism)

The sensitivity study shows the baseline CDF value was increased by 0.33% and the baseline LERF value was increased by 0.20%. Furthermore, this sensitivity study re-calculated the RICT values for all out-of-service (OOS) equipment cases while applying the increased event probabilities (since the sensitivity failure probability for the RPS electrical is lower than the current point estimate, the existing failure probability was conservatively used in the sensitivity). The results showed that the impact on RICTs for the OOS equipment cases changed was small. RICTs for most cases were the same duration in days. RICTs for four cases changed by 1 day. This sensitivity shows that the individual failure rates used for the digital feedwater control system signal failure, digital pressure controller (DPC-4103) failure, digital electric pressure controller failure, HPCI and RCIC flow controller failure are not key sources of uncertainty with respect to RICT calculations.

b)

No other digital systems are credited in the PRA model.

c)

Not applicable.

RAI 8 - Surrogate Events in the PRA Models The NRC safety evaluation for NEI 06-09 specifies that the LAR should provide a comparison of the TS functions to the functions modeled in the PRA. Justification should be provided to show that the scope of the PRA model is consistent with the licensing basis assumptions.

Table E1-1 in Enclosure 1 to the LAR identifies each LCO proposed for inclusion in the RICT program. It describes how the systems and components covered in the TS LCO are implicitly or explicitly modeled in the PRA. For some TS LCO Conditions, the table explains that the associated SSCs are not modeled in the PRAs but will be conservatively represented using a surrogate event.

Note 6 to the table states "[low pressure coolant injection] LPCI loop select logic failure was used as a conservative surrogate. This basic event represents the probability that LPCI loop select fails in such a way that it causes LPCI injection to occur on the loop where the line break occurred." This note applies to several functions listed in Table E1-1:

L-MT-20-036 NSPM Enclosure Page 20 of 75 LCO 3.3.5.1.B Function 2.h four Reactor Steam Dome Pressure - Low (Break Detection) channels Function 2.k two Reactor Steam Dome Pressure - Time Delay Relay (Break Detection) channels LCO 3.3.5.1.C Function 2.i, eight Recirculation Pump Differential Pressure - High (Break Detection) channels Function 2.j, four Recirculation Riser Differential Pressure - High (Break Detection) channels Function 2.l, two Recirculation Pump Differential Pressure - Time Delay Relay (Break Detection) channels Function 2.m, two Recirculation Riser Differential Pressure - Time Delay Relay (Break Detection) channels Explain why the proposed surrogate is conservative.

NSPM Response to RAI 8:

The LPCI mode of operation is automatically initiated and is dependent on the initiation logic. Due to the potential for injected water to be lost through certain recirculation loop loss of coolant accidents (LOCAs), logic is provided to determine which recirculation pump loop is injected into and therefore which RHR injection valves are signaled to open.

LPCI operation is therefore dependent on the proper operation of the LPCI loop selection logic (LLSL). Successful initiation requires selection of an unbroken recirculation loop for LPCI, opening of the appropriate injection valves, closing the recirculation pump discharge valve in the selected loop, and starting the RHR pumps.

The LPCI loop select logic is modeled in the PRA as a single basic event for simplification purposes. The instrumentation associated with the functions described in the RAI are inputs to the LPCI loop select logic. Since the PRA models the entire LPCI loop select logic as a single basic event, the failure of the LPCI loop select logic basic event bounds the individual failures of each function. The LPCI loop select basic event will be used as a surrogate event for the functions 2.h, 2.i, 2.j, 2.k, 2.l, and 2.m in TS Table 3.3.5.1-1.

This basic event represents the probability that LPCI loop select fails to select the correct loop. Thus, this surrogate basic event is a conservative representation for any of the functions described in the RAI.

L-MT-20-036 NSPM Enclosure Page 21 of 75 RAI 9 - PRA Model Uncertainty Analysis Process The NRC staff safety evaluation to NEI 06-09 specifies that the LAR should identify key assumptions and sources of uncertainty and licensees should assess and disposition each as to their impact on the RMTS application.

NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Main Report" (ADAMS Accession No. ML17062A466),

presents guidance on the process of identifying, characterizing, and qualitative screening of model uncertainties.

LAR Enclosure 9 states that the process for identifying key assumptions and sources of uncertainties for the internal events and fire PRAs was performed using the guidance in NUREG-1855. The LAR explains that to identify key assumptions and sources of PRA modeling uncertainty (1) the internal events and fire PRA models and notebooks were reviewed for plant-specific issues and (2) generic sources of uncertainty identified in EPRI 1016737, "Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments," and 1026511, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," were also reviewed for applicable issues. The LAR concludes for both the internal events and fire PRAs that "no specific uncertainty issues have been identified that would impact the RICT application," and no candidate key assumption and sources of uncertainty were presented in the LAR.

Based on the discussion in the LAR, it is not clear to NRC staff what specific process and criteria were used to screen uncertainties from an initial comprehensive list of assumptions and sources of PRA modeling uncertainty (including those associated with plant-specific features, modeling choices, and generic industry concerns), in order to conclude that no uncertainty issues could impact the RICT calculations. It is also not clear whether certain key assumptions and sources of uncertainty were initially identified but found to be unimportant through use of sensitivity studies per guidance described in LAR Enclosure 9, Section 1.0.

Therefore, address the following:

a)

Describe the specific Monticello process used to screen uncertainties from the initial comprehensive lists of PRA uncertainties (including those associated with plant-specific features, modeling choices, and generic industry concerns), in order to eventually conclude that the uncertainty issues could not impact the RICT calculations.

b)

Include description of the criteria that was used to screen from a comprehensive listing of sources of uncertainty to a smaller set of key candidate assumptions and sources of uncertainty; and also describe the criteria used to justify that none of the key candidate assumptions and sources of uncertainty could have an impact on the RICT calculations.

As part of this description, explain whether use of the results of sensitivity studies were included as part of the criteria that was used.

L-MT-20-036 NSPM Enclosure Page 22 of 75 c)

Include description of plant or PRA procedures, practices or processes that are used to support the identification and dispositioning PRA modeling uncertainty concerns (e.g., a PRA change database).

d)

During the review of licensees PRA uncertainty notebooks provided during the audit (see ADAMS Accession No. ML20154K763 for the NRC staffs audit plan), the staff noted three PRA assumptions that may impact the application but did not appear to be examined or dispositioned for the application. For two of the assumptions, results of sensitivity studies reported in the uncertainty notebook showed a significant impact on the base CDF (up to 200%). The first item regarded an assumption that the operators will vent containment to below 50 psig (for [reactor] core isolation cooling (RCIC))

backpressure trip setpoint even though the Emergency Operating Procedures (EOPs) only direct the operators to vent below 56 psig. The second item was regarding the assumption that only rapidly evolving overpressure events lead to a rupture of containment and gradually evolving events, like the loss of containment heat removal (CHR), would create smaller leaks in containment to level off pressure so that a rupture would not occur. The third assumption was that RCIC is credited in the PRA model after battery depletion.

i) To illustrate the process addressed in items a through c above, discuss how the three assumptions described were considered in the process for reviewing key assumptions and sources of uncertainties for the application.

ii) Provide a disposition of the impact of these assumptions on the application.

NSPM Response to RAI 9:

a)

The evaluation of sources of uncertainty for the RICT application built upon the activities performed in support of the MNGP application for implementation of 10 CFR 50.69 (Reference 16). However, it was recognized that the impacts of any uncertainties on the RICT application could differ from those evaluated for the 10 CFR 50.69 application. In particular, RICT calculations involve primarily delta-risk evaluations, whereas 10 CFR 50.69 categorization evaluations rely on risk ranking of structures, systems, and components (SSCs) in the plant.

The process used to evaluate sources of uncertainty for the RICT application follows the guidance illustrated in Figure 4-1 of EPRI TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments" (Reference 17). The sources of uncertainty evaluation for the baseline internal events PRA considers both plant-specific sources of uncertainty and the generic uncertainties identified in EPRI TR-1016737.

During the development of the 10 CFR 50.69 LAR, the Sources of Uncertainty Notebooks for the PRA models of record were reviewed to collect a listing of all sources of uncertainty that were identified as having potential impacts on the base PRA model or risk-informed applications. If the Sources of Uncertainty Notebook already provided a justification that the model uncertainty need not be evaluated further as a potential source of uncertainty for the base model or for applications (e.g., negligible contribution, best-estimate modeling, etc.), then those model uncertainties were not considered

L-MT-20-036 NSPM Enclosure Page 23 of 75 further. This information represents the input from the "base model assessment" as shown in Figure 4-1. It should be noted that the additional lists of potential generic sources of uncertainty from Table A-4 of EPRI TR-1016737 were also considered in the sources of uncertainty notebooks for both the internal events and fire PRAs. The potential uncertainty items noted in Table A-3 of EPRI TR-1016737 overlap some of the issues already shown in Table A-4 or evaluated in the base model evaluation of sources of uncertainty. Other items in Table A-3 are noted as being adequately assessed through the industry peer review process or are properly modeled in the PRA. Lastly, some items are not applicable to MNGP (e.g., PWR-specific issues, use of alternate injection systems, etc.). Application-specific uncertainties, as shown in Figure 4-1, are addressed in Section 4.0 of LAR Enclosure 9.

Twenty significant assumptions and uncertainties were identified from the internal events PRA model for consideration for RICT impacts. Thirteen relevant sources of uncertainty from the fire PRA model were also identified.

A total of 33 candidate sources of uncertainty (20 internal events/internal flooding PRA and 13 fire PRA) were then evaluated for their specific impacts on the calculation of RICTs. The calculation of a RICT is based on Incremental Core Damage Probability (ICDP) and Incremental Large Early Release Probability (ILERP). These are delta-risk measures that evaluate the change in risk over the baseline "zero maintenance" risk for the plant. In reviewing each of the candidate sources of uncertainty for the internal events/internal flooding and fire PRAs, the following considerations were applied to determine if a RICT impact could exist:

Candidate uncertainties that are qualitatively shown to have a very small impact on total risk, and would be expected to have a negligible impact on delta-CDF and delta-LERF (particularly uncertainties that pertain to parts of the model that would not impact components that are in the RICT program, such as changes to non-support system initiating event frequencies, human error probabilities not related to RICT-eligible equipment, etc.). Ten of the 33 candidate sources of uncertainty were dispositioned using this criterion.

Candidate uncertainties that are represented through conservative PRA modeling that would be expected to have a negligible impact on delta-risk RICT calculations (e.g.,

conservatisms in non-equipment data values that would tend to cancel out in a RICT calculation.) Eight of the 33 candidate sources of uncertainty were screened using this criterion.

Candidate uncertainties that were identified, but for which current industry-accepted approaches and data were used, are not considered as key sources of uncertainty.

This is consistent with the ASME/ANS PRA Standard definition of a "source of modeling uncertainty" which states: "a source is related to an issue in which there is no consensus approach or model and where the choice of approach or model is known to have an effect on the PRA model." A number of these candidates were

L-MT-20-036 NSPM Enclosure Page 24 of 75 derived from the EPRI list of generic PRA uncertainties. Fourteen of the 33 candidate sources of uncertainty were screened using this criterion.

Candidate uncertainties that were examined via sensitivity studies to confirm that the impact on baseline CDF and LERF are negligibly small are not considered as key sources of uncertainty for the RICT program. One of the 33 candidate sources of uncertainty was dispositioned using this criterion.

If a candidate source of uncertainty could be shown to satisfy the above considerations using a qualitative evaluation, then this was considered to be adequate. One potential source of uncertainty was initially dispositioned on the basis of similar qualitative arguments and is described in 9.d.

As a result of this comprehensive review of potential uncertainties generally following the guidance provided in NUREG-1855, Revision 1 (Reference 19), none of the 33 items were identified as having a significant impact on the calculation of RICTs.

b)

As described in RAI 9.a, specific criteria were used to assess each source of uncertainty from the baseline PRA models. As discussed in the response to RAI 9.d, a quantitative sensitivity study was performed to evaluate the RICT impacts of assuming a higher likelihood of failure to operate RCIC following battery depletion. For some of the other potential uncertainties that were evaluated for the internal events PRA model, an estimation of risk impact was able to be obtained through inspection of the PRA results.

For example, uncertainties were screened on the basis of:

Negligible probability of occurrence of the failure scenarios of concern (e.g., several orders of magnitude below other scenarios with similar impacts).

The impacted system(s) having very low importance in the PRA results.

Surrogate events that demonstrate the impact of a potential uncertainty had negligible risk significance.

For the fire PRA, most of the identified uncertainties pertain to fire PRA modeling issues that utilize current industry and NRC-accepted methods. However, as many of these methods are acknowledged to have inherent conservatisms, they were identified as potential uncertainties for consideration.

c)

As previously discussed in the MNGP LAR to adopt TSTF-425, "Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specifications Task Force (RITSTF) Initiative 5b" (Reference 24), NSPM employs a multi-faceted, structured approach in establishing and maintaining the technical adequacy and plant fidelity of the PRA models for its nuclear generation sites. This approach includes a proceduralized PRA maintenance and update process, as well as the use of independent peer reviews.

The following information describes this approach as it applies to the MNGP PRA.

L-MT-20-036 NSPM Enclosure Page 25 of 75 The NSPM risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plants. This process includes a governing Corporate Directive and subordinate implementing procedures. The NSPM PRA maintenance and update process is described in the following procedures:

"PRA Change Database Use and Application Guide" addresses the following elements:

o Identifies and tracks ongoing evaluation of plant changes and collecting new information including identified errors in the PRA model, Peer Review Findings and suggestions.

o The qualification of PRA personnel.

o Documentation of disposition of PRA impacts (PRA Change Database).

"PRA Guideline for Model Maintenance and Update" addresses the following elements:

o Maintenance and update of the PRA to be consistent with the as-built, as-operated plant, including closure of peer review findings.

o Consideration of the cumulative impact of pending changes on the PRA.

o Impact of plant changes on the PRA models.

o Control of software used for the PRA models.

o Documentation of the PRA Maintenance and Update process.

The overall model update process defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operating experience), and for controlling the model and associated computer files.

PRA Change Database Forms A PRA Change Form is stored electronically in the PRA Change Database (PCD),

governed by PRA procedure. A PCD is created for all issues that are identified that could impact the PRA model, including resolving concerns associated with modelling assumptions and uncertainties. Each open PCD Form documents details of each identified issue, evaluates the risk impact of that specific issue, and identifies affected PRA systems, analyses, and risk-informed applications.

During resolution of the PCD issue, the preparer will follow model revision instructions delineated in procedure and revise the appropriate portions of the PRA model or documentation to address the issue. A summary of the resolution of the issue is documented in the PCD form under the Actual Solution section. As appropriate, the preparer will include which methodology was chosen and why, if there were any additional details of the problem discovered during resolution, a summary of areas of the model that were revised (e.g., Feedwater fault trees, etc). At a minimum, all points addressed in the Detailed Problem Description, Proposed Solution, or Risk Level

L-MT-20-036 NSPM Enclosure Page 26 of 75 Description should be addressed. In addition, the PCD form section related to Affected Applications, Model Revision number and Record of Analysis will be filled in as necessary.

The reviewer of the PCD form will review the Affected Applications, Model Revision number and Record of Analysis sections for completeness and accuracy to ensure that the solution addressed the problem and to verify that the change was incorporated and documented in accordance with the ASME/ANS PRA Standard.

d)

Appendix C of the MNGP Sources of Uncertainty Notebook included a discussion of various sensitivity studies that had been performed using an earlier version of the PRA models for CDF. The subjects of these sensitivity studies were not necessarily identified sources of uncertainty; however, several of these were noted as potentially having a "moderate" or "significant" impact on the PRA results as noted in the RAI. The two sensitivity study items noted in the RAI are further discussed below concerning RICT calculation impacts:

Reactor Inventory Control - A CDF sensitivity was noted in the MNGP Sources of Uncertainty notebook since RCIC requires that RPV and containment pressures remain within limits to ensure that RCIC and HPCI pumps can operate if needed for inventory control. The sensitivity study examined the risk impacts if venting occurred that would disable RCIC injection. However, the plants Operations Manual, "Primary Containment Control," includes guidance to operators to ensure that depressurization actions do not impact the operability of injection systems if they are needed to support key safety functions. In the event that the backpressure trip was actuated, operators can re-open the trip throttle valve remotely from the control room per RCIC operating procedure to allow RCIC to restart. Therefore, the likelihood of a scenario in which RCIC would be inadvertently disabled by venting operations is negligible. Therefore, this potential sensitivity issue is not of concern for RICT calculations. The sources of uncertainty evaluations presented in Appendices A and B of the Sources of Uncertainty Notebook specifically evaluated potential uncertainties pertaining to inventory control during venting operations and RPV depressurization and determined that these were not significant sources of uncertainty.

Containment Heat Removal - The MNGP PRA assumes that only rapidly evolving overpressure events (e.g., ATWS, vapor suppression failure) lead to a rupture of containment that leads to a significant degradation of the environment in the reactor building and loss of mitigating systems located in the building. Gradually evolving events, such as loss of containment heat removal, are assumed to lead to leakage from the containment sufficient to level off containment pressure at its ultimate capacity but not rupture the containment. This assumption was based on an evaluation performed in support of the Individual Plant Examination (IPE) which considered potential containment failure modes under gradual and rapid overpressure conditions. The sources of uncertainty evaluations presented in Appendices A and B of the Sources of Uncertainty Notebook categorized the treatment of overpressure containment failure mechanisms as not a significant uncertainty for the PRA based on

L-MT-20-036 NSPM Enclosure Page 27 of 75 the high likelihood of drywell head leakage being the dominant failure mode for the gradual overpressure sequences. A sensitivity study in Appendix C of the MNGP Sources of Uncertainty Notebook examined the risk impacts of assuming that all containment overpressure scenarios resulted in a large rupture that disabled mitigating systems. However, that sensitivity study was performed using an older revision of the PRA model. Refinements in the MNGP PRA model incorporated in the current model has reduced the overall CDF. A new sensitivity study was performed using the current PRA model, assuming a 56% probability of a large containment rupture (based on plant-specific containment failure analysis performed for the IPE) in the overpressure scenarios. This updated sensitivity study showed that CDF increases by 12% and LERF increases by 20%, which is much less than the ~200%

increase in CDF reported in the previous sensitivity study. The risk increases in the updated sensitivity study primarily result from scenarios in which control rod drive (CRD) injection is failed as a result of a containment rupture while other injection systems are failed due to other environmental impacts of the scenario (e.g. flooding-induced failures). CRD system components are not included in the RICT program, so this potential uncertainty would have negligible impact on the RICT results.

Operation of RCIC following battery depletion is addressed in the MNGP EOPs and the likelihood of operator failure to use RCIC under these conditions was assessed using industry-standard HRA practices. In response to the NRCs question concerning this operator action, a sensitivity study was performed in which the HEP was increased by a factor of 3 from the PRA-assumed value. This is a proceduralized operator action, so the factor of 3 sensitivity is an appropriate adjustment factor to assume for this uncertainty.

The results show that overall CDF increased by a maximum of about 5% for most of the RICT-eligible LCOs, with no or negligible decreases in the calculated RICTs. Some specific LCOs had CDF increases of up to 18%; however, for most the RICT impacts were minimal (i.e., the worst case being a 7% reduction for a 15-day RICT down to 14 days). One other LCO Condition (3.3.5.1.B) has its RICT reduced from 17 days to 15 days and another LCO Condition (3.8.4.B) shows a reduction in calculated RICT from 8 days to 6 days. Given the limited impact on most RICTs given the conservative assumption of a factor of 3 increase in the HEP (roughly corresponding to the HEP being set to its 95th percentile value), this potential uncertainty would not have a significant impact on the RICT results.

RAI 10 - Potential Loss of Function Condition (ECCS)

TSTF-505 Revision 2 does not allow for TS loss of function conditions (i.e., those conditions that represent a loss of a specified safety function or inoperability of all required trains of a system required to be OPERABLE) in the RICT program.

LAR Enclosure 1, Table E1-1 appears to include an LCO that could represent TS loss of function because it allows a configuration that does not meet the design-basis success criteria:

L-MT-20-036 NSPM Enclosure Page 28 of 75 TS LCO 3.5.1 (Emergency Core Cooling System). There are two LPCI subsystems and two Core Spray subsystems and that the design success criteria are "One LPCI subsystem and two Core Spray subsystems or Two LPCI subsystems and one Core Spray subsystem." Therefore, Condition D (two LPCI subsystems inoperable) and Condition E (One Core Spray Subsystem inoperable AND (One LPCI subsystem inoperable OR One or two LPCI pump(s) inoperable)) appear to be a loss of function.

Explain why this Condition does not represent a TS loss of function or remove the LCO Condition from the RICT program.

NSPM Response to RAI 10:

The design success criteria for LCO 3.5.1 was stated incorrectly in table E1-1 of the LAR.

The table E1-1 in Attachment 2 provides revised LCO 3.5.1 design success criteria.

LCO 3.5.1 Condition D allows "Two LPCI subsystems" to be inoperable for reasons other than Condition C or G. Condition C applies to one LPCI pump in each subsystem.

Condition G applies to open RHR intertie return line isolation valves. In this Condition both HPCI and ADS with CS remain Operable and will ensure that no loss of ECCS function exists. If HPCI or one or more ADS valves were inoperable at the same time that Condition D was entered, Condition L would be required to be entered due to the loss of function. Therefore, Condition D does not represent a loss of function.

LCO 3.5.1 Condition E allows "One Core Spray subsystem" AND "One LPCI subsystem OR one or two LPCI pumps" to be inoperable. In this Condition HPCI and ADS with one LPCI subsystem and ADS with one CS subsystem remain Operable and will ensure that no loss of ECCS function exists. If HPCI or two or more ADS valves were inoperable at the same time that Condition E was entered, Condition L would be required to be entered due to the loss of function. If HPCI and one or more ADS valves were inoperable at the same time that Condition E was entered, Condition M would be required to be entered due to the loss of function. Therefore, Condition E does not represent a loss of function.

RAI 11 - Credit for FLEX Equipment and Actions The NRC staff assessed challenges to incorporating FLEX equipment and strategies into a PRA model (ADAMS Accession No. ML17031A269). With respect to equipment failure probability, the NRC staff drew the following conclusion:

The uncertainty associated with failure rates of portable equipment should be considered in the PRA models consistent with the ASME/ANS PRA Standard as endorsed by RG 1.200. Risk-informed applications should address whether and how these uncertainties are evaluated.

With respect to HRA, NEI 16-06 Section 7.5 recognizes that the current HRA methods do not translate directly to human actions required for implementing mitigating strategies. Sections

L-MT-20-036 NSPM Enclosure Page 29 of 75 7.5.4 and 7.5.5 of NEI 16-06 describe actions to which the current HRA methods cannot be directly applied, such as: debris removal, transportation of portable equipment, installation of equipment at a staging location, routing of cables and hoses; and those complex actions that require many steps over an extended period, multiple personnel and locations, evolving command and control, and extended time delays. In its assessment, the NRC staff drew the following conclusion:

Until gaps in the human reliability analysis methodologies are addressed by improved industry guidance, [Human Error Probabilities] HEPs associated with actions for which the existing approaches are not explicitly applicable, such as actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06, along with assumptions and assessments, should be submitted to NRC for review.

With regard to uncertainty, Section 2.3.4 of NEI 06-09 states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT program. It states that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties which could potentially impact the results of a RICT calculation. It also states that the insights from the sensitivity studies should be used to develop appropriate RMAs, including highlighting risk significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions. Uncertainty exists in PRA modeling of FLEX, related to the equipment failure probabilities for FLEX equipment used in the model, the corresponding operator actions, and preinitiator failure probabilities. Therefore, FLEX modeling assumptions can be key assumptions and sources of uncertainty for RICTs proposed in this application.

Section 3 of Enclosure 2 of the LAR states that a limited amount of FLEX equipment is addressed in the PRA model in accordance with NEI 16-06. Specifically, two FLEX transfer cubes are credited to refill the diesel fire pump tank. Based on a sensitivity study, credit for this limited amount of FLEX equipment reduces CDF by ~1% and has no impact on LERF.

Due to very small CDF and LERF impact, inclusion of the limited amount of FLEX equipment in the PRA model will have a minimal impact on the calculated RICT. However, no RICT-associated LCO sensitivity studies were provided to confirm the LAR statement.

Perform, justify, and provide results of LCO-specific sensitivity studies that assess impact on RICT due to FLEX failure probabilities and FLEX HEP. Part of the response include the following:

a)

Justify values selected for the sensitivity studies, including justification of why the chosen values constitute bounding realistic estimates.

b)

Provide numerical results on specific selected RICTs and discussion of the results.

L-MT-20-036 NSPM Enclosure Page 30 of 75 c)

If applicable, describe how the results of the sensitivity studies will be used to identify RMAs prior to the implementation of the RICT program, consistent with the guidance in Section 2.3.4 of NEI 06-09.

NSPM Response to RAI 11:

a)

The all hazard RICT model credits two FLEX Fuel Oil Transfer Cubes to refill the Diesel Fire Pump tank. The FLEX sensitivity study described in the LAR completely removed credit for the FLEX equipment to refill Diesel Fire Pump tank from the all hazard PRA model when calculating the updated sample RICT values. This sensitivity shows the maximum potential impact on RICT calculation.

b)

This sensitivity showed a 3% or less impact to the sample RICT values in all but a few RICTs. For the RICTs that showed a greater than 3% change, the RICT was already limited by the front stop or remained at the 30-day backstop. A few of the calculated RICT values (for 3.3.2.2, 3.5.1.E and 3.8.1.B/D) were increased by greater than 10%.

However, the available number of days for these sample RICTs were not impacted. For example, RICT samples for TS 3.5.1.E and 3.8.1.D were limited to their front stop for both cases. The number of days for sample RICTs TS 3.3.2.2.A and 3.8.1.B was not impacted by removing credit for the FLEX equipment because they were already at the 30-day backstop. The table below lists the LCOs that were changed by greater than 3%:

LCO Condition Component Results (Base)

Results (No-FLEX Credited)

% Change RICT Base (days)

RICT No-FLEX Credited (days) 3.3.2.2.A 6A_K20B 1.28E-04 1.47E-04 15 30 Days 30 Days 3.5.1.B CS 1.60E-04 1.69E-04 6 28 Days 28 Days 3.5.1.E 1LPCI+1CS 7.33E-04 8.33E-04 14 5 Days 5 Days 3.5.1.E Div2_LPCI 7.34E-04 8.33E-04 13 5 Days 5 Days 3.5.1.E Div2 1.67E-03 1.80E-03 8 72 Hours 72 Hours 3.8.1.B EDG12 9.45E-05 1.27E-04 34 30 Days 30 Days 3.8.1.D 1R+2R+EDG11 5.70E-04 6.48E-04 14 6 Days 6 Days 3.8.1.D 1R+2R+EDG12 5.70E-04 6.48E-04 14 6 Days 6 Days c)

No RMAs will be pre-determined to account for the potential epistemic uncertainties associated with inclusion of the FLEX equipment described in the response to RAI 11.a in the PRA model. The sensitivity study described in the response to RAI 11.b showed that the impact of the FLEX equipment on calculated RICTs was very small. RMAs will be developed on a case-specific basis using the insights available in the configuration risk management (CRM), as described in LAR Enclosure 12.

L-MT-20-036 NSPM Enclosure Page 31 of 75 RAI 12 - Real-Time Risk Model Regulatory Position 2.3.3 of RG 1.174 states that the level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of the change should include establishing a cause-effect relationship to identify portions of the PRA affected by the change being evaluated. Full-scale applications of the PRA should reflect this cause-effect relationship in a quantification of the impact of the proposed licensing basis change on the PRA elements.

Section 4.2 of NEI 06-09 describes attributes of the tool used for configuration risk management (CRM) including:

Model translation from the PRA to a separate CRM tool is appropriate; CRM fault trees are traceable to the PRA. Appropriate benchmarking of the CRM tool against the PRA model shall be performed to demonstrate consistency.

CRM application tools and software are accepted and maintained by an appropriate quality program.

The CRM tool shall be maintained and updated in accordance with approved station procedures to ensure it accurately reflects the as-built, as-operated plant.

of the LAR describes the attributes of the CRM tool for use in RICT calculations.

The LAR explains that the internal events, internal flooding events, and fire events PRA models are maintained as separate models. The LAR also describes several changes made to the PRA models to support calculation of configuration-specific risk and mentions approaches for ensuring the fidelity of the real-time risk to the PRAs including real-time risk maintenance, documentation of changes, and testing.

Describe the benchmarking activities performed to confirm consistency of the real-time risk model results to the results of each PRA model of record (MOR), including periodicity of real-time risk updates compared to the MOR updates. Address each of the MORs (i.e.,

internal events, internal flooding events, and internal fire events) in the response.

NSPM Response to RAI 12:

The CRM is initially based upon the all-hazards average maintenance PRA model (Internal Events, Internal Flood, and Fire). The all-hazard average maintenance model is updated to remove the availability factor (i.e., set to 1.0), add the seismic penalty factor, and optimize for quantification speed.

After the all hazards fault tree is optimized for use in the CRM the model is quantified and the baseline results (i.e., zero-maintenance, zero equipment OOS) are reviewed for consistency with the original average maintenance PRA model results. A number of OOS configurations are quantified and results are reviewed for reasonable results. Any

L-MT-20-036 NSPM Enclosure Page 32 of 75 discrepancies are investigated. This review ensures that no errors have been introduced during model speed optimization.

The PRA model maintenance and update procedures specify that the CRM shall be updated after a new model of record is released. The periodicity of MOR releases is as specified in Enclosure 7 of the LAR (Reference 1), nominally once every two fuel cycles and applies to Internal Events, Internal Flooding, and Fire PRA models. Enclosure 7 of the LAR also describes that PRA changes with potential for significant impact will be incorporated in an unscheduled update in an application-specific PRA model. An application-specific model may include updates to Internal Events (with Internal Flooding),

Fire PRA, or both. Application-specific model documentation includes a review of the impact on risk-informed applications and identifies any applications that require update.

The CRM will be updated for application specific models that significantly impact the results, including significant changes to RICT calculations. Any update to the CRM utilizes the most recent MOR or application specific model for each hazard model.

RAI 13 - Unspecified RICT Estimates NEI 06-09 states the following with regard to high-risk configurations.

RMTS evaluations shall evaluate the instantaneous core damage frequency (CDF),

instantaneous large early release frequency (LERF). If the SSC inoperability will be due to preplanned work, the configuration shall not be entered if the CDF is evaluated to be greater or equal than 10-3 events/year or the LERF is evaluated to be greater or equal to 10-4 events/year. If the SSC inoperability is due to an emergent event, if these limits are exceeded, the plant shall implement appropriate risk management actions to limit the extent and duration of the high risk configuration.

NEI 06-09 prohibits voluntary entry into a high-risk configuration but it allows entry in such configurations due to emergent events with implementation of appropriate RMAs.

Table E1-2 of Enclosure 1 of the LAR provides RICT estimates for TS actions proposed to be in the scope of the RICT program. However, RICT estimates for several LCO actions (3.5.1.D and E, 3.8.4.B, 3.8.7.A and B) are not provided. In addition, Note 1 of Table E1 2 states:

Several quantification results exceed the risk cap level of 1E-03 (CDF) or 1E-04 (LERF).

Those LCOs are listed as "No Entry" given the quantified risk. However, it is possible that the LCO could be entered for a partial failure and would result in lower quantified risk. In a lower risk condition, entry into the RICT program would be allowed.

In light of these observations:

a)

Clarify the intent of your note and whether NEI 06-09 will be followed with regard to involuntary entries into high-risk configurations.

L-MT-20-036 NSPM Enclosure Page 33 of 75 b)

Explain what is meant by "LCO could be entered for a partial failure and would result in lower quantified risk." Provide examples and associated RICT estimates.

NSPM Response to RAI 13:

a)

The requirements of NEI 06-09 Revision 0-A (Reference 10) will be followed regarding involuntary entries into high risk configurations.

Note 1 was intended to explain that the calculated RICT estimate was conservative for the configuration and that less significant trains or sub-components within scope for the LCO could result in "Inoperable" equipment but with a lower calculated risk, depending on what equipment was out of service.

b)

See section a) for explanation of the note. A discussion of each LCO referencing Note 1 is as follows:

LCO 3.8.4.B: The RICT estimate for LCO 3.8.4.B was based on the worst case of the four DC subsystems out of service. The calculated risk varied significantly between the subsystems due to asymmetries in the PRA model caused by plant design. The quantification results showed that the D11 DC subsystem exceeded the risk cap but the D100, D21, and D31 DC subsystems would have allowed a voluntary RICT entry. The worst-case result was provided in the LAR.

The RICT estimate for LCO 3.5.1.D was erroneously calculated. Subsequent update to the calculations resulted in a RICT of 15 days. Total risk was below the risk cap so this RICT could be voluntarily entered.

The RICT estimate for LCO 3.5.1.E was based on the worst case of several combinations of Core Spray and LPCI components out of service. The calculated risk varied significantly depending on the combination of components. For example, the quantification results showed that the combination of Core Spray Loop B and LPCI Division 2 (both pumps) resulted in exceeding the risk cap. However, the combination of Core Spray Loop A with 1 LPCI pump from each division resulted in a 30-day RICT.

The RICT estimates for LCO 3.8.7.A, and 3.8.7.B were calculated based on taking the main 4kV or DC subsystem out of service. The LCOs in question apply to multiple components within the AC and DC distribution systems. A bus, MCC, or panel downstream of the main bus/panel would likely result in lower calculated risk because less equipment would be impacted. In this case, voluntary entry into a RICT may be possible depending on the magnitude of the calculated risk.

RICT estimates for the different cases evaluated for each LCO are as follows:

L-MT-20-036 NSPM Enclosure Page 34 of 75 Tech Spec LCO Condition Example Component /

Subsystem RICT Estimate 3.5.1.D Two LPCI subsystems inoperable for reasons other than Condition C or G 2 LPCI subsystems 15 Days 3.5.1.E One Core Spray subsystem inoperable.

AND One LPCI subsystem inoperable.

OR One or two LPCI pump(s) inoperable.

Core Spray Loop A & LPCI Div 1 (2 pumps) 27 Days Core Spray Loop B & LPCI Div 2 (1 pump) 5 Days Core Spray Loop A & LPCI Div 1 & 2 (1 pump each Div) 30 Days Core Spray Loop B & LPCI Div 1 & 2 (1 pump each Div) 5 Days Core Spray Loop B & LPCI Div 2 (2 pumps) 2 Days (no voluntary entry) 3.8.4.B One Division 1 or Division 2 DC electrical power subsystem inoperable for reasons other than Condition A.

D11 DC Subsystem 2 Days (no voluntary entry)

D21 DC Subsystem 8 Days D31 DC Subsystem 11 Days D100 DC Subsystem 16 Days 3.8.7.A One or more AC electrical power distribution subsystems inoperable.

Bus 15 3 Days (no voluntary entry)

Bus 16 2 Days (no voluntary entry) 3.8.7.B One or more DC electrical power distribution subsystems inoperable.

D11 DC Subsystem 2 Days (no voluntary entry)

D21 DC Subsystem 8 Days D31 DC Subsystem 11 Days D100 DC Subsystem 16 Days RICT Estimates are based on Rev. 2 of the sample RICT calculation. The most limiting RICT estimate for each TS LCO Condition is specified in Table E1-2 (Attachment 5).

RAI 14 - PRA Modeling Regulatory Position 2.3.3 of RG 1.174 states that the level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of the change should include establishing a cause-effect relationship to identify portions of the PRA affected by the change being evaluated. Full-scale applications of the PRA should reflect this cause-effect relationship in a quantification of the impact of the proposed licensing basis change on the PRA elements.

L-MT-20-036 NSPM Enclosure Page 35 of 75 The NRC staffs safety evaluation for NEI 06-09 specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions and that justification be provided to show that the scope of the PRA model is consistent with the licensing basis assumptions.

Regarding unmodeled SSCs, the evaluation states the following:

NEI 06-09, Revision 0, specifically applies the RMTS only to those SSCs which mitigate core damage or large early releases. Where the SSC is not modeled in the PRA, and its impact cannot otherwise be quantified using conservative or bounding approaches, the RMTS are not applicable, and the existing frontstop CT would apply.

Further, Item 11 in Section 2.3 of TSTF-505, Revision 2, states:

The traveler will not modify Required Actions for systems that do not affect core damage frequency (CDF) or large early release frequency (LERF) or for which a RICT cannot be quantitatively determined.

LAR Enclosure 1, Table E1-1, lists LCOs and corresponding PRA functions.

TS LCO 3.6.1.7.A: One required suppression chamber-to drywell vacuum breaker inoperable for opening. The design criterion is six out of eight suppression chamber-to-drywell vacuum breakers OPERABLE for Opening. The PRA success criteria is one suppression chamber-to drywell vacuum breaker OPERABLE for Opening.

TS LCO 3.7.2.A: One Emergency Service Water (ESW) subsystem inoperable. The comments state that "hydraulic analysis has been performed to show that ESW is not required to prevent CDF and LERF."

a)

Describe and justify the modeling of vacuum breakers in the PRA and the analysis performed to support the PRA success criteria.

b)

Justify inclusion of TS LCO 3.7.2.A in the scope of the RICT program if it does not impact CDF and LERF.

NSPM Response to RAI 14:

a)

The chief purpose of the vacuum breakers opening is to prevent excessive water level variation in the downcomers submerged in suppression pool water, and to prevent backflow of water from the suppression chamber to the drywell when the suppression chamber is pressurized due to non-condensable gases. Failure of the drywell vacuum breakers to open could also damage the drywell due to excessive vacuum. Currently PRA success criteria requires one out of eight drywell vacuum breakers to successfully open. However, the basis for PRA success criteria for drywell vacuum breakers to open is unclear. Thus, this issue was added to the PRA change tracking process and will be resolved before RICT program implementation. The resolution will determine the

L-MT-20-036 NSPM Enclosure Page 36 of 75 appropriate PRA success criteria, document the basis for the criteria, and update the PRA model as needed.

b)

The Emergency Service Water system supplies the Control Room Emergency Filtration Train (EFT-ESW), the ECCS room coolers, and ECCS pump motor cooling. The ESW is not modeled in the PRA. Neither the EFT-ESW, the ECCS room coolers nor ECCS pump motor cooling are credited to provide any mitigating function for either CDF or LERF in either the internal events or fire PRA models. The hydraulic analysis was performed to support determination of which HVAC systems must be modeled in the PRA. The results of this analysis demonstrate that the temperature in the room for the 24-hour PRA scenario will remain below the equipment threshold in each room, and the equipment will perform satisfactorily.

The analysis presented in this calculation was based on the room loss of HVAC temperatures for a 24-hour period. Changes in plant configuration during operation do not change the outcome of this analysis. A RICT will be applied to non-modeled ESW system that are screened due to this thermal hydraulic analysis. Mapping will be added to the CRM model to allow operators to enter the LCO and take affected equipment OOS, but there will be no change in risk specific to the impacted equipment themselves. The RICT will be calculated based on the delta-risk associated with plant state during the period of time that LCO 3.7.2.A is active. If no other equipment is OOS at the time, the delta risk will be based on the CDF and LERF seismic penalty factors alone. If other equipment is OOS, delta-risk will increase and the RICT will decrease commensurate with the increase in risk.

RAI 15 - RICT Estimates During the audit the NRC staff noted discrepancies in reported RICT estimates between LAR Table E1-2 and new revisions to the PRA documentation presented during the audit.

a)

Confirm that the RICT estimates provided in the LAR are correct or provide updated RICT estimates. If updated estimates are provided, explain and justify any differences from the estimates provided in LAR Table E1-2.

b)

For TS 3.5.1.D (Two LPCI subsystems inoperable for reasons other than Condition C or G) and TS 3.5.1.E (One Core Spray subsystem inoperable AND One LPCI subsystem inoperable OR One or two LPCI pump(s) inoperable) explain why LAR Table E1-2 specifies "No Entry," implying these conditions are high-risk configurations.

NSPM Response to RAI 15 a)

RICT estimates have been re-calculated using the updated seismic penalty factor specified in RAI 19 and to support comparison to the sensitivity studies needed to support the RAI responses. The updated RICT estimates are shown in the revised Table E1-2 provided in Attachment 5 to this enclosure, which replaces Table E1-2 in the LAR.

L-MT-20-036 NSPM Enclosure Page 37 of 75 b)

"No Entry" term was intended to explain that the calculated RICT for the most conservative configuration would not allow voluntary entry.

The RICT estimate for TS 3.5.1 Condition D is incorrect in Table E1-2 in the LAR. The updated RICT estimate for TS 3.5.1 Condition D is shown in the revised Table E1-2 in to this Enclosure, which replaces Table E1-2 in the LAR.

The RICT estimate for TS 3.5.1 Condition E was calculated based on worst case combination of the Core Spray pumps and LPCI pumps out of service. The calculated risk varied significantly between the trains due to asymmetries in the PRA model caused by plant design. For instance, Core Spray and two LPCI pumps in Division 2 exceeded the risk cap but the Division 1 pumps would have allowed a voluntary RICT entry. The worst-case result was provided in the LAR. Table E1-2 in Attachment 5 to this Enclosure provides the updated RICT estimate for the worst-case result for TS 3.5.1 Condition E.

RAI 16 - Fire Modeling The LAR referred to risk evaluation and fire modeling analysis. The NRC staff was unable to fully evaluate the fire modeling performed as part of the fire probabilistic risk assessment (fire PRA). Regarding the acceptability of the fire PRA approach, methods, and data, describe the fire modeling calculational model or numerical methods (e.g., fire modeling tools and techniques) used in support of the fire PRA model.

NSPM Response to RAI 16:

The Monticello Fire PRA model, which is peer reviewed, uses various industry standard fire modeling tools to determine initiator to target damage. This includes the NUREG-1805 FDTs, CFAST, and FDS. The NUREG-1805 Fire Dynamics Tools (FDTs) imbedded in the Fire Modeling Database (FMDB) were used mainly to determine the severity factor and initial screening of target damage. Consolidated Model of Fire Growth and Smoke Transport (CFAST) was used to determine time to hot gas layer for single and multi-compartment analysis. Fire Dynamic Simulator (FDS) was used to determine if structural steel damage would occur for turbine generator fires and the impact of fires on sensitive electronics in the Main Control Room.

RAI 17 - Damage Thresholds Part 4 of the PRA Standard (ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications") indicates that damage thresholds should be established to support the fire PRA. The standard further indicates that thermal impact(s) must be considered in determining the potential for thermal damage of structures, systems, and

L-MT-20-036 NSPM Enclosure Page 38 of 75 components and appropriate temperature and critical heat flux criteria must be used in the analysis.

Provide the following information:

a)

Describe how the installed cabling in the fire areas was characterized, specifically with regard to the critical damage threshold temperatures and critical heat fluxes for thermoset and thermoplastic cables.

b)

Explain how the damage thresholds for non-cable components (i.e., pumps, valves, electrical cabinets, etc.) were determined. Identify any non-cable components that were assigned damage thresholds different from those for thermoset and thermoplastic cables and provide a technical justification for these damage thresholds.

c)

Describe the damage criteria that were used for exposed temperature-sensitive electronic equipment. Explain how temperature-sensitive equipment inside an enclosure was treated and provide a technical justification for these damage criteria.

NSPM Response to RAI 17:

a)

Due to the abundance to thermoplastic cabling at MNGP, all cable targets (conduits or cable trays) in the plant were assumed to be thermoplastic therefore the damage criteria of 400°F and 6kW/m2 from NUREG/CR-6850 was used.

b)

For non-cable targets the following approach is taken based on guidance provided by NUREG/CR-6850, Section H.2:

For major components such as motors, valves, etc., the fire vulnerability should be assumed to be limited to the vulnerability of the power, control, or instrument cables connected to the components and subjected to fire-generated conditions in a postulated fire scenario.

Equipment constructed of ferrous metal with no cable connections such as pipes and water tanks should be considered invulnerable to fire.

Passive components such as flow check valves should be considered invulnerable to fire.

c)

For fires inside an electrical cabinet, regardless of the fire size all equipment inside the electrical cabinet was assumed failed at time zero. For fires outside of an electrical cabinet such as transients the transient areas were chosen to incorporate the electrical cabinets and associated cable trays therefore the thermal based failures of the cabinets are accounted for.

L-MT-20-036 NSPM Enclosure Page 39 of 75 RAI 18 - Heat Soak Method The LAR states that in Revision 4.0 of the Monticello fire PRA model, enhanced fire modeling methods (heat soak) were added. Describe the heat soak method that was used to convert the damage times in Appendix H of NUREG/CR-6850 to a percent of damage function for targets exposed to a time-varying heat flux.

NSPM Response to RAI 18 The time to damage for a cable is determined from the data presented in NUREG/CR-6850 Tables H-5 through H-8. In practice, the analyst determines the maximum exposure temperature and finds the corresponding failure time in the table.

It should be noted that the NUREG/CR-6850 tables contain only constant exposure data and NUREG/CR-6850 does not provide guidance on their use for time-varying exposures.

This leaves the conservative approach of starting a timer when the exposure exceeds the minimum threshold and imposing cable failure when the timer exceeds the minimum duration from the tables. A brief example is given below in Table 1 for a thermoplastic cable. At 0 minutes the temperature exceeds 205°C and the timer starts. At 2 minutes the temperature has risen to 350°C which has a damage duration of 3 minutes. This is followed by a drop in temperature, but conservatively the 3 minute duration is kept (as identified in bold, italicized font in the table below). The cable then fails at 3 minutes (3 minutes after exceeding the 205°C threshold). With this conservative approach, a brief excursion at high temperature results in early cable failure.

Table 1: Example of Applying Appendix H Constant Exposure Tables Time (min)

Temperature

(°C)

Duration (min)

Timer (min) 0 210 30 0

1 320 5

1 2

350 3

2 3

320 3

3 4

320 3

4 Heat Soak (Damage Integral) Approach A less conservative approach than the one discussed above would account for time-varying exposure while still being applicable to the broad classes of a generic thermoplastic or thermoset cable. Since the actual damage mechanism results from heating the cable jacket until it fails, it would be appropriate to credit periods of exposure at low temperature with having a lesser impact on the time to failure than periods of exposure at higher temperatures.

The integrated approach to determining cable damage described in this section begins with any temperature greater than 25°C and therefore accounts for "pre-heating" of cables. Equation 4 is used to calculate the damage accumulated for temperatures below

L-MT-20-036 NSPM Enclosure Page 40 of 75 the non-heat soak cable failure temperatures for thermoset and thermoplastic (330°C and 205°C). Above these temperatures, heat soak method follows the guidance provided in NUREG/CR-6850, Appendix H which uses a minimum failure time of 1 minute for temperatures exceeding 490°C for thermoset and 370°C for thermoplastic. Therefore, there is no upper or lower bound of applicability for temperature using the heat soak methodology. Furthermore, there is no consideration for cable size using the heat soak methodology since it is not mentioned as a factor effecting this methodology described in NUREG/CR-6850, Appendix H. The only known cable failure tool that does employ cable size is the THIEF model described in NUREG/CR-6931 (Reference 11) which is not currently used in the Monticello Fire PRA.

The cable damage problem is analogous to evaluating the potential for skin burns. In both cases one is evaluating an in-depth temperature (inside of the cable jacket or the epidermis-dermis boundary) and imposing a harm (cable failure or burn) based upon the temperature solution. As with the Appendix H tables in NUREG/CR-6850, there are a number of approaches for skin burns that have developed simple times to burn based on constant exposures. Skin burn modeling has also utilized the concept of a damage integral:

(1) where is the damage integral and is a time dependent reaction rate. Damage occurs when crosses a threshold value. In the case of skin burn modeling, the reaction rate is assumed to be an Arrhenius process. The data in the Appendix H tables contain exposure durations as a function of temperature ranges. This can be modeled in a damage integral by assuming the reaction rate is the inverse of the exposure duration.

That is:

(2) where is the exposure temperature at time and

is the time until damage from NUREG/CR-6850 Tables H-5 through H-8. Cable failure would then occur with a damage integral of 1 or:

1 (3)

The example in Table 1 is shown below in Table 2 with the damage integral applied using trapezoidal integration. The cable now fails at 4 minutes rather than 3 minutes.

Table 2: Example of Applying the Heat Soak Approach Time (min)

Temperature

(°C)

Duration (min)

Damage Integral (min) 0 210 30 0.0 1

320 5

0.1 2

350 3

0.4 3

320 5

0.7 4

320 5

1.1

L-MT-20-036 NSPM Enclosure Page 41 of 75 One consideration in the damage integral approach is the treatment of the time when cables are below the minimum damage threshold. For example, consider a thermoplastic cable that spends a couple of hours at 200°C and then makes a transient to just over 205°C. It would be inappropriate to credit the cable with having 30 minutes of survival time as two hours of preheating should put the cable on the edge of failure. That is, if a cable spends time near but not over the minimum threshold and then rises over the minimum threshold, it would be reasonable to assume that the preheated cable would have a somewhat lower time to damage.

Figure 1 below shows the integrated exposure required to cause damage using the lower and upper temperature bound for each damage interval in NUREG/CR-6850 Tables H-5 and H-6. The integrated exposure, ", in J/m2 is computed as

(4) where is the Stefan-Boltzmann constant (5.67 10-11 W/m²-K4), is the ambient temperature in K, is the exposure temperature in K, and is convective heat transfer coefficient which can conservatively be assumed to be 50 W/(m2K).

Figure 1: Integrated exposure needed to cause cable damage for a given constant exposure temperature As seen in the figure, the net exposure required increases as the temperature decreases.

A similar result is seen for the heat flux tables. One approach would be to fit a curve to the data in the figures and extrapolate to lower exposure temperatures. This approach has the significant issue that extrapolating an empirical dataset could result in non-conservative behavior. The only consideration that one can be confident on is that the total integrated exposure as a function of temperature should continue to increase as temperature decreases. Therefore, at exposures temperatures below the threshold exposure, a conservative assumption would be that the integrated exposure at the 0

5 10 15 20 200 300 400 Exposure(MJ/m2)

ExposureTemperature(°C)

Thermoplastic LowerBound UpperBound 0

5 10 15 20 25 30 300 400 500 Exposure(MJ/m2)

ExposureTemperature(°C)

Thermoset LowerBound UpperBound 0

L-MT-20-036 NSPM Enclosure Page 42 of 75 threshold is required for damage to occur. This will conservatively estimate the damage integral component for temperatures or heat fluxes below their respective thresholds.

As an example, if Eq 4. is used to compute the incident fluxes for a temperature of 100°C and 205°C, the result is a ratio of 2.2:1. With the assumption of a constant integrated exposure, the

at 100°C would be 2.2 x 30 minutes or 66 minutes. This would apply similarly to thermoset cables using the threshold temperature of 330°C.

It is noted that this assumption means that a cable that sits long enough at low exposure, even at ambient conditions, would eventually reach a cable damage integral of 1. Since clearly a cable at ambient would not actually be damaged, in addition to a damage integral of 1, this heat soak method will also require that the cable exposure be over the threshold exposure when the integral of 1 is reached.

The assumption of constant integrated exposure below the Appendix H threshold temperature introduces some uncertainty of unknown magnitude to the calculation.

However, the trend in Figure 1 demonstrates that the assumption is conservatively biased and will not result in an undue lengthening of the predicted time to failure.

The implementation of this heat soak approach was done via two sets of VBA subroutines

- one for temperature exposure and one for heat flux exposure. Both routines are called with a table of values of time vs. exposure and the cable type. The routines output the damage integral along with time to damage.

Verification Three sets of verification exercises were performed for the VBA subroutines:

1.

Reproduce the NUREG/CR-6850 Appendix H tables

2.

Demonstrate the approach does not fail cables at low exposure

3.

Demonstrate the approach yields the expected value for a time-dependent exposure Results of the verification were that the NUREG/CR-6850 Appendix H tables were reproduced. Three scenarios were run for both heat flux and temperature exposures to test low exposure failures and in each case damage was not reached prior to the exposure exceeding the threshold. Each type of cable was exposed to an exponentially increasing temperature or heat flux exposure. The damage integral was computed by hand as well as by the VBA subroutines and after comparing the two they produced identical results.

Validation Validation of the heat soak method was performed using intermediate scale test data from NUREG/CR-6931. This report contains data from a series of cable exposure tests that were used to develop and validate the THIEF model. Tests were performed using a range of cable types (sizes and jacket and insulator materials). The end results of the THIEF validation were a bias of -15% (faster time to failure) and a standard deviation of

L-MT-20-036 NSPM Enclosure Page 43 of 75 33%. The validation results demonstrated that the heat soak method reduces the conservatism of the strict application of Appendix H while still maintaining a slight degree of conservatism when compared to the higher fidelity THIEF method.

RAI 19 - Bounding Seismic LERF Estimate Section 2.3.1, Item 7, of NEI 06-09 states that the "impact of other external events risk shall be addressed in the Risk-Managed Technical Specifications (RMTS) program." It explains that one method to do this is by "performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated Risk-Informed Completion Time (RICT)." The NRC staffs safety evaluation for NEI 06-09 states that "[w]here PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT."

A seismic PRA model is not available for Monticello and the seismic hazard cannot be screened out for the RICT application. Section 3 of Enclosure 4 to the LAR stated that a seismic CDF and seismic LERF "penalty" was determined for this application using the current Monticello seismic hazard curve developed and reported in response to Recommendation 2.1 of the Near-Term Task Force (NTTF 2.1) provided by letter dated May 14, 2014 (ADAMS Accession No. ML14136A288). Section 3.1 of Enclosure 4 to the LAR stated that the total Monticello seismic CDF (SCDF) is estimated to be 3.0E-05 per year using Monticello value for high confidence in low probability of failure (HCLPF), the spectral ratios in the safety assessment for GI-199 "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment" (ADAMS Accession No. ML100270639), and the hazard curves developed in response to NTTF 2.1. It is unclear to the staff why the licensee used the spectral ratios determined from the Individual Plant Examination of External Events submittals in GI-199 instead of developing them from the more recent hazard curves developed in response to NTTF 2.1 to determine the SCDF estimate.

Details of the approach for determining the seismic LERF "penalty" are provided in LAR, Section 3.3 using the conditional large early release probability (CLERP) for internally initiated events with some adjustment (i.e., the contribution of certain containment bypass events that would not be expected from a seismic event were not included in the CLERP). The LAR states that the CLERP determined using this approach was chosen as an "adequately conservative" estimate. In addition, NRC staff notes that LERF-to-CDF ratio for seismic events can be significantly higher than the same ratio for internal events due to the unique nature of seismically induced failures. It is unclear that the selected CLERP of 5%

represents a conservative or bounding estimate for use as the seismic LERF "penalty" in the proposed RICT calculations.

a)

Justify the use of the GI-199 spectral ratios instead of spectral ratios developed from the Monticello seismic hazard curve in the response to NTTF 2.1 or update the Monticello SCDF "penalty" for the proposed RICTs using the spectral ratios from the recent seismic hazard curves.

L-MT-20-036 NSPM Enclosure Page 44 of 75 b)

Justify that the seismic LERF "penalty" provided in the submittal to support RICT calculations for the Monticello is conservative for this application. Include rationale that deriving seismic LERF-to-CDF ratio using the internal events LERF-to-CDF ratio is conservative or bounding for seismically induced events, given that internal events random failures do not capture seismically induced failures that may uniquely contribute to LERF.

c)

If the approach to estimating the seismic LERF penalty cannot be justified as bounding for this application in response to item (b) above, then provide, with justification, the bounding seismic LERF "penalty" for use in RICT calculations.

NSPM Response to RAI 19:

Subsequent to the submittal of the LAR, the approach used to calculate the seismic penalty values for CDF and LERF was updated using a more refined methodology that considered the NTTF seismic spectra and a more detailed method for estimating the seismic LERF penalty. The revised evaluation results in a reduction in the proposed seismic CDF penalty; however, the updated seismic LERF to CDF ratio is now approximately 37%.

The information provided in Attachment 1 replaces the information provided in the original MNGP LAR (Reference 1) Enclosure 4, Sections 3.0 and 6.0.

RAI 20 - Screening of Snow Section 2.3.1, Item 7, of NEI 06-09 states that the "impact of other external events risk shall be addressed in the RMTS program," and explains that one method to do this is by documenting prior to the RMTS program that external events that are not modeled in the PRA are not significant contributors to configuration risk. The safety evaluation for NEI 06-09 states, "Other external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk."

LAR Enclosure 4, Section 4 concludes that external hazards other than seismic events can be screened from consideration in the RICT program including snow. The LAR also states that hazards are evaluated for plant configurations allowed under the RICT program. LAR, Table E4-2, indicates that criterion "C1" (event damage potential is less than events for which plant is designed) and criterion "C4" (event is included in the definition of another event) was used to screen the snow hazard. The LAR further states that the design-basis roof live load is 50 pounds per square foot (psf), the average snowfall per year in Monticello, Minnesota is 46.3 inches, and the maximum recorded snowfall from a single storm in Minnesota occurred near Finland, Minnesota and measured 46.5 inches with an estimated weight of 46.5 psf, which is within the design basis. However, considering the small margin between the design-basis roof live load and the average and maximum recorded snowfalls, it

L-MT-20-036 NSPM Enclosure Page 45 of 75 is unclear to the NRC staff how the assumptions that resulted in the screening based on criterion C1 and C4 will continue to remain valid during the proposed RICTs.

Discuss how existing procedures or RMAs will ensure that the assumptions for screening the risk from snowfall will be maintained during the proposed RICTs.

NSPM Response to RAI 20 Industry practice in PRA is to assume that design limits have been properly calculated and that plant structures have been constructed to meet those limits. Postulating failure at or near a design basis limit in the PRA would be a very conservative assumption, as actual failure would be expected at a value significantly beyond the design limit.

The quoted maximum snowfall accumulation presented in LAR Enclosure 4 is for Finland, MN, which is located in the extreme northern part of the state near the Lake Superior shoreline. That area receives far more snowfall than the Monticello, MN area near where the plant is located. Monticello, MN is located between Minneapolis and St. Cloud, MN.

As noted in Section 2 of the USAR, the maximum snowfall accumulation in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> recorded in Minneapolis is 16.2 inches, and the maximum recorded in St. Cloud is 12.2 inches (per Table 2.3-3). Snowfall averages about 42 inches per year (42.2 inches for Minneapolis and 42.4 inches for St. Cloud). A conservative estimate for the weight of one inch of snow would be 1.75 pounds per square foot (based on 62.4 lb/cu ft of water with a moisture density of 33%). This value for an approximate 16-inch snowfall is well below the 50 psf design limit for the MNGP safety-related structures. Therefore, the screening of snowfall events from consideration in the RICT program is appropriate.

RAI 21 - Operating Electrical Systems The LAR includes proposed changes to TS 3.8.1, "AC Sources - Operating," TS 3.8.4, "DC Sources - Operating," and TS 3.8.7, "Distribution Systems - Operating."

The LAR proposed to add the alternate RICT to the completion times of TS 3.8.7, "Distribution Systems - Operating," Conditions A and B. The changed portion is indicated in italics.

Condition A: One or more AC electrical power distribution subsystems inoperable Required Action A.1: Restore AC electrical power distribution subsystem(s) to OPERABLE status Completion Time: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> OR In accordance with the Risk-Informed Completion Time Program

L-MT-20-036 NSPM Enclosure Page 46 of 75 Condition B: One or more DC electrical power distribution subsystems inoperable Required Action B.1: Restore DC electrical power distribution subsystem(s) to OPERABLE status Completion Time: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> OR In accordance with the Risk-Informed Completion Time Program The TS LCO 3.8.7 requires Division 1 and Division 2 AC and DC electrical power distribution subsystems to be operable.

However, TSTF-505 excludes conditions for loss of function from the RICT program.

Specifically, it recommends adding a note to certain RICTs in the TS to exclude the case for a loss of function if the TS Condition is applicable to "one or more" inoperable channels, subsystems, or trains.

The NRC staff notes that in some cases, the Monticello TS 3.8.7 Condition A or Condition B would be a loss of function condition. For example, if both 125/250-Volt DC distribution panels in the Division 1 and Division 2 electrical power distribution subsystems were inoperable, the LAR Table E1-1 design success criterion (i.e., one AC electrical power distribution subsystem capable of supporting minimum safety functions) for Condition A would be defeated resulting in a loss of function. Also, if both 4.16-Kilovolt essential buses in Division 1 and Division 2 electrical power distribution subsystems were inoperable, the LAR Table E1-1 design success criterion (i.e., one DC electrical power distribution subsystem capable of supporting minimum safety functions) for Condition B would be defeated resulting in a loss of function. However, the proposed alternate RICTs do not include the note, as recommended by TSTF-505, to exclude the cases for loss of function of TS 3.8.7 Condition A or Condition B from the RICT program.

The staff requests the following information to address this discrepancy:

a)

Discuss with supporting documentation how the Monticello design-basis functions are met without loss of safety function of the AC and DC electrical power distribution systems in Monticello TS 3.8.7 Condition A and Condition B, respectively.

b)

If in the response to item a. above, any of the plant configurations of TS 3.8.7 Conditions A and B result in a TS loss of function condition, explain this apparent inconsistency with TSTF-505 or provide an updated TS markup that excludes the TS loss of function conditions from the proposed RICT program, as recommended.

NSPM Response to RAI 21 a)

LCO 3.8.7 Condition A T.S. LCO 3.8.7 Condition A allows "one or more" AC electrical power distribution subsystems to be inoperable for the required completion time. However, this Condition does not apply to a loss of function because Condition D is required to be entered upon a loss of function.

L-MT-20-036 NSPM Enclosure Page 47 of 75 LCO 3.8.7 Condition B T.S. LCO 3.8.7 Condition B allows "one or more" DC electrical power distribution subsystems to be inoperable for the required completion time. However, this Condition does not apply to a loss of function because Condition D is required to be entered upon a loss of function.

LCO 3.8.7 Condition D LCO 3.8.7 Condition D, two or more electrical power distribution subsystems inoperable that result in a loss of function, is entered if a loss of function occurs. This loss of function could be due to both divisions of AC, both divisions of DC, or one division of AC with the opposite division of DC. Therefore, the presence of Condition D precludes the application of a RICT to Condition A or B when a loss of function occurs.

b)

Not applicable RAI 22 - ECCS Success Criteria Item 7 in Section 2.4 of the Monticello LAR indicates that TS LCO 3.5.1, Emergency Core Cooling System (ECCS) - Operating, requires each ECCS injection/spray subsystem and the ADS function of three safety/relief valves be OPERABLE, and that the remaining OPERABLE ECCS subsystems would provide adequate core cooling during a loss-of-coolant accident (LOCA) for the following Conditions:

1. Condition B: One Low Pressure Coolant Injection (LPCI) subsystem inoperable for reasons other than Condition A, or, one Core Spray subsystem inoperable;

2. Condition C: One LPCI pump in both LPCI subsystems inoperable;

3. Condition D: Two LPCI subsystems inoperable for reasons other than Condition C or G;

4. Condition E: One Core Spray subsystem inoperable and one LPCI subsystem inoperable; or one Core Spray subsystem inoperable and one or two LPCI pump(s) inoperable.

LAR Table E1-1, "In-scope TS/LCO Conditions to Corresponding PRA Functions," lists the remaining OPERABLE ECCS subsystems in the column of Design Success Criteria for TS 3.5.1 Conditions B, C, D, and E in the columns of MNGP TS 3.5.1.B, 3.5.1.C, 3.5.1.D, and 3.5.1.E, respectively.

Discuss the analysis of record (AOR) that demonstrates that the remaining OPERABLE ECCS subsystems could provide adequate core cooling during a LOCA for TS 3.5.1 Conditions B, C, D, and E. Reference the NRC documents approving the AOR relevant to this concern or address the acceptability of the AOR if not previously approved by the NRC.

NSPM Response to RAI 22:

LAR Table E1-1, "In-scope TS/LCO Conditions to Corresponding PRA Functions," lists the remaining OPERABLE ECCS subsystems in the column of Design Success Criteria

L-MT-20-036 NSPM Enclosure Page 48 of 75 for TS 3.5.1 Conditions B, C, D, and E, respectively. In preparing the response to this RAI, it was determined that the success criteria required an update.

The remaining systems for LCO Conditions 3.5.1.B, 3.5.1.C, 3.5.1.D, and 3.5.1.E were compared to the bounding events in USAR Table 14.7-11. The remaining ECCS capability for each Condition has equivalent or greater capacity than the limiting analyzed configuration in the AOR, adequate core cooling is demonstrated, and no loss of function occurs.

Technical Specification 3.5.1 Conditions LCO Condition Inoperable Systems Remaining Systems Limiting Event in USAR Table 14.7-11 and AOR Criteria B

1 LPCI subsystem or 2 CS pumps + 2 LPCI pumps + 3 ADS valves +

HPCI Loss of DC Battery 1 CS pump + 2 LPCI pumps + 3 ADS valves 1 CS subsystem 1 CS pump + 4 LPCI pumps + 3 ADS valves +

HPCI C

1 LPCI pump in each subsystem 2 CS pumps + 2 LPCI pumps + HPCI + 3 ADS valves Loss of DC Battery 1 CS pump + 2 LPCI pumps + 3 ADS valves D

2 LPCI subsystems 2 CS pumps + HPCI + 3 ADS valves LPCI Injection Valve Failure 2 CS pumps + HPCI + 3 ADS valves E

1 CS subsystem +

1 LPCI subsystem or 1 CS pump + 2 LPCI pumps + HPCI + 3 ADS valves Loss of DC Battery 1 CS pump + 2 LPCI pumps + 3 ADS valves 1 CS subsystem +

1 or 2 LPCI pumps 1 CS pump + 2 or 3 LPCI pumps + HPCI + 3 ADS valves The AOR that demonstrates adequate core cooling during a LOCA has been reviewed and approved is discussed below.

On July 10, 2009, License Amendment No. 162 (Reference 20) was issued revising the Required Actions and the Completion Times in TS 3.5.1, "Emergency Core Cooling System [ECCS] - Operating," to provide a 72-hour completion time (CT) to restore a low-pressure ECCS subsystem(s) to operable status after discovery of two low-pressure ECCS subsystems inoperable.

Specifically, the existing Condition D was revised (1) to apply to two entire Low-Pressure Core Injection (LPCI) subsystems being inoperable (the previous Condition only applied when two LPCI subsystems were inoperable due to inoperable injection paths); (2) added a new Condition E to provide a 72-hour CT when one Core Spray subsystem and one

L-MT-20-036 NSPM Enclosure Page 49 of 75 LPCI subsystem (or one or two LPCI pump(s) were inoperable); and (3) added a new Condition F providing a 72-hour Completion Time for when both Core Spray subsystems were inoperable. License Amendment 162 was based upon the MNGP LOCA analysis, the corresponding ECCS single failure analysis (discussed in USAR Table 14.7-11), and precedents for several other Boiling Water Reactors. Note, Condition F was later removed in November 2014 by License Amendment No. 184 (Reference 21) after it was determined that operation of at least one Core Spray subsystem was necessary post-accident to maintain adequate long-term core cooling (top-down spray cooling versus flooding required for advanced fuel designs).

The ECCS LOCA analysis results and single failure assumptions have subsequently been re-reviewed by the NRC several times, e.g., as part of the licensing approval process for an extended power uprate (EPU) (Amendment No. 176 - Reference 22), and most recently in June 2015 for the transition to ATRIUM 10XM fuel and utilization of AREVA (now Framatome) safety analysis methodologies (Amendment No. 188 -

Reference 18), and determined acceptable by the NRC. An AREVA licensing report, ANP-3211(P), "Monticello EPU LOCA Break Spectrum Analysis for ATRIUM' 10XM Fuel (Reference 23), submitted in support of the AREVA fuel transition, approved in Amendment No. 188, discusses the ECCS analysis. The ANP-3211(P) report included results of a single-failure of battery (DC) power, single-failure of diesel generator, single-failure of the HPCI system, single-failure of an LPCI injection valve, and single-failure of one automatic depressurization system (ADS) valve.

The analyses in support of the MNGP licensing basis demonstrate that the Design Success Criteria for Technical Specification 3.5.1 Conditions B, C, D and E have been met. The Design Success Criteria use a complement of equipment that is equal to or more limiting than the Technical Specification Conditions require.

RAI 23 - ADS Success Criteria TSTF-505 excludes loss of function conditions from the RICT program. LAR Table E1-1, "In-scope TS/LCO Conditions to Corresponding PRA Functions," lists TS LCO 3.5.1.K, a Condition with one ADS valve inoperable, and indicates that three ADS valves are required to be OPERABLE. The column of "Design Success Criteria" indicates that three ADS valves are available.

Clarify for TS 3.5.1.K Condition with one of three required ADS valves inoperable, that the design success criteria need two or three available ADS valves. Discuss the AOR that demonstrated adequacy of two or three ADS valves for reactor pressure vessel rapid depressurization to mitigate the LOCA consequences and reference the NRC documents approving the AOR of the concern or address the acceptability of the AOR if it was not previously approved by the NRC.

L-MT-20-036 NSPM Enclosure Page 50 of 75 NSPM Response to RAI 23:

The design success criteria for ADS in support of TS LCO Condition 3.5.1.K is discussed in USAR Section 4.4.2.2, Automatic Depressurization System, which states:

The Automatic Depressurization System (ADS) feature of the Reactor Pressure Relief System serves as a backup to the High Pressure Coolant Injection (HPCI)

System under LOCA conditions. If the HPCI System does not operate, the system is depressurized sufficiently to permit the Low Pressure Coolant Injection (LPCI) and Core Spray System to operate to protect the fuel cladding... Three safety/relief valves are included in the ADS. All three ADS valves are required to provide sufficient capacity for the ADS... Emergency depressurization within the Emergency Operating Procedures requires manual opening of three valves.

The EOPs control the use of ADS manually. This requires opening 3 ADS valves. If an ADS valve does not open, the procedure directs opening other safety relief valves (SRVs) until 3 valves are open. Manual backup for loss of an ADS valve is provided.

USAR Table 14.7-11 summarizes the ECCS systems available for failure of an ADS valve. ADS is a backup system to HPCI for provision of core cooling with high reactor pressure as would exist in a small break accident. If HPCI is available in this case, it provides adequate core cooling.

Technical Specification 3.5.1 Conditions Limiting Event in USAR Table 14.7-11 and AOR LCO Condition Inoperable Systems Remaining Systems Design Success Criteria K

1 ADS Valve 2 ADS, 2 CS, HPCI, 4 LPCI Loss of ADS Valve (Framatome) 2 ADS, 2 CS, HPCI, 4 LPCI As discussed in the response to RAI 22, AREVA licensing report ANP-3211(P), submitted in support of the AREVA fuel transition and approved in License Amendment No. 188, discusses the ECCS analysis and the impact of a single failure of the HPCI System or an ADS valve. The report states that "[a]ll three ADS valves are assumed operable during the LOCA except when a single failure is assumed to prevent one ADS valve from opening." ANP-3211(P) provided an analysis for a single failure of an ADS valve.

USAR Section 4.4.2.2 indicates if the HPCI System does not operate, the reactor is depressurized sufficiently using ADS to permit the LPCI and Core Spray System operation to protect the fuel cladding. Condition I provides a 14 day period for HPCI restoration when the system is inoperable. Similarly, Condition K provides the same restoration period for the ADS System when it is inoperable, e.g., when only two of the three ADS OPERABLE. Therefore, either HPCI System operation or ADS together with low pressure system(s) operation (LPCI and/or Core Spray) are sufficient to address a small break LOCA. If HPCI (Condition I) or one or more ADS valves (Condition K) were

L-MT-20-036 NSPM Enclosure Page 51 of 75 Inoperable at the same time, Condition L would be required to be entered which requires shutdown and depressurization to 150 psi. Therefore, Condition K does not represent a loss of function.

The analyses in support of the MNGP licensing basis demonstrate that the Design Success Criteria for Technical Specification 3.5.1 Conditions have been met. The Design Success Criteria use a complement of equipment that is equal to or more limiting than the Technical Specification Conditions require. See also the response for RAI 10 for information on required response for the loss of HPCI in conjunction with the loss of an ADS valve.

RAI 24 - Diversity and Manual Actuation Enclosure 1 Section 3 of the LAR provides descriptions of the I&C system design features including its redundancy. This section, however, does not provide adequate information to demonstrate sufficient diversity is maintained. This diversity position was established in the TSTF-505, Revision 2. In addition, the RG 1.174 Revision 2 states the licensee should "assess whether the proposed LB [licensing basis] change meets the defense-in-depth principle" by not "over-relying on programmatic activities as compensatory measures associated with the change in the LB". The RG 1.174, Revision 3, further elaborates that human actions (e.g., manual system actuation) are considered as one type of compensatory measure.

a)

Please provide additional information to demonstrate that at least one redundant or diverse means (e.g., other automatic features or manual action) to accomplish the safety functions (e.g., reactor trip, safety injection, or containment isolation) remain available during the use of the RICT for all Design Basis Accidents analyzed in Updated Safety Analysis Report Chapter 15. An example of the diversity justification that has been previously accepted by staff can be found in the precedent LAR at ADAMS Accession No. ML19280C844, or RAI at ADAMS accession No. ML19224B705.

b)

If the diverse means the licensee identified are manual actuations, confirm that these "manual actuations" are modeled in the plant PRA, defined in plant operation procedures to which operators are trained, and confirm the completion times associated with these actions are evaluated as adequate.

NSPM Response to RAI 24:

a) The following 8 tables (Table 24-1 through Table 24-8) provide the information to demonstrate that at least one redundant or diverse means to accomplish the safety functions remain available during the use of the RICT.

L-MT-20-036 NSPM Enclosure Page 52 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 1.a - Intermediate Range Monitors, Neutron Flux -

High High Reactor Trip Initiation (SCRAM)

Rod Withdrawal Error - High Low Power (USAR 14.4.3)

1. Automatic Initiation

a. IRM Neutron Flux - High,

b. APRM Neutron Flux - High,

Setdown,

c. 2-of-4 Voter,

d. Rod Block Monitor

2. Manual SCRAM Control Rod Drop Accident (USAR 14.7.1)

1. Automatic Initiation

a. IRM Neutron Flux - High,

b. APRM Neutron Flux - High,

Setdown,

c. 2-of-4 Voter

2. Manual SCRAM 1.b - Intermediate Range Monitors, INOP SCRAM None

1. Manual SCRAM 2.a - Average Power Range Monitors, Neutron Flux - High, (Setdown)

SCRAM Rod Withdrawal Error - Low Power (USAR 14.4.3)

1. Automatic Initiation

a. APRM Neutron Flux - High, Setdown Trip,

b. IRM Neutron Flux - High,

c. 2-of-4 Voter,

d. Rod Worth Minimizer

2. Manual SCRAM Control Rod Drop Accident (USAR 14.7.1)

1. Automatic Initiation

a. APRM Neutron Flux - High,

Setdown,

b. IRM Neutron Flux - High,

c. 2-of-4 Voter

2. Manual SCRAM 2.b - Average Power Range

Monitors, Simulated Thermal Power - High SCRAM Generator Load Rejection (USAR 14.4.1)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. Reactor Vessel Steam Dome Pressure High,

d. Turbine Control Valve Fast

Closure,

e. APRM 2-of-4

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 53 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips SCRAM Loss of Feedwater Heater (USAR 14.4.2)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM - Fixed Neutron Flux -

High,

c. 2-of-4 Voter

2. Manual SCRAM SCRAM Rod Withdrawal Error - Low Power (USAR 14.4.3)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Neutron Flux - High,

c. 2-of-4 Voter

d. Rod Block Monitor

2. Manual SCRAM SCRAM Feedwater Controller Failure -

Maximum Demand (USAR 14.4.4) -

high level trip results in turbine trip

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. TSV Closure,

d. 2-of-4 Voter

2. Manual SCRAM SCRAM Turbine Trip without Bypass (USAR 14.4.5)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. Reactor Vessel Steam Dome Pressure High,

d. TSV Closure Trip,

e. 2-of-4 Voter

2. Manual SCRAM SCRAM Vessel Pressure ASME Code Compliance (USAR 14.5.1)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. Reactor Vessel Steam Dome Pressure High,

d. 2-of-4 Voter

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 54 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 2.c - Average Power Range Monitors, Neutron Flux - High SCRAM Generator Load Rejection (USAR 14.4.1)

1. Automatic Initiation

a. APRM Fixed Neutron Flux - High,

b. APRM Simulated Thermal Power

- High,

c. Reactor Vessel Steam Dome Pressure High,

d. Turbine Control Valve Fast

Closure,

e. 2-of-4 Voter

2. Manual SCRAM SCRAM Loss of Feedwater Heater (USAR 14.4.2)

1. Automatic Initiation

a. APRM - Fixed Neutron Flux -

High,

b. APRM Simulated Thermal Power

- High,

c. 2-of-4 Voter

2. Manual SCRAM SCRAM Rod Withdrawal Error - Low Power (USAR 14.4.3)

1. Automatic Initiation

a. APRM - Fixed Neutron Flux -

High,

b. APRM Simulated Thermal Power

- High,

c. 2-of-4 Voter

d. Rod Block Monitor

2. Manual SCRAM SCRAM Feedwater Controller Failure -

Maximum Demand (USAR 14.4.4)

1. Automatic Initiation

a. APRM Fixed Neutron Flux - High,

b. TSV Closure,

c. APRM Simulated Thermal Power

- High,

d. 2-of-4 Voter

2. Manual SCRAM SCRAM Turbine Trip without Bypass (USAR 14.4.5)

1. Automatic Initiation

a. APRM Fixed Neutron Flux - High,

b. Reactor Vessel Steam Dome Pressure High,

c. TSV Closure Trip,

d. APRM Simulated Thermal Power

- High,

e. 2-of-4 Voter

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 55 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips SCRAM Vessel Pressure ASME Code Compliance (USAR 14.5.1)

1. Automatic Initiation

a. APRM Fixed Neutron Flux - High,

b. APRM Simulated Thermal Power

- High,

c. Reactor Vessel Steam Dome Pressure High,

d. 2-of-4 Voter

2. Manual SCRAM 2.d - Average Power Range Monitors, Inop.

SCRAM None

1. Manual SCRAM 2.e - Average Power Range

Monitors, 2-Out-Of-4 Voter SCRAM Loss of Feedwater Heater (USAR 14.4.2)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM - Fixed Neutron Flux -

High,

c. 2-of-4 Voter

2. Manual SCRAM SCRAM Rod Withdrawal Error - Low Power (USAR 14.4.3)

1. Automatic Initiation

a. IRM Neutron Flux - High,

b. APRM Neutron Flux - High,

Setdown,

c. APRM Neutron Flux - High,

d. APRM Simulated Thermal Power

- High,

e. 2-of-4 Voter

f.

Rod Block Monitor

2. Manual SCRAM SCRAM Feedwater Controller Failure -

Maximum Demand (USAR 14.4.4)

1. Automatic Initiation

a. 2-of-4 Voter,

b. APRM Fixed Neutron Flux - High,

c. APRM Simulated Thermal Power

- High,

d. TSV Closure

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 56 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips SCRAM Turbine Trip without Bypass (USAR 14.4.5)

1. Automatic Initiation

a. 2-of-4 Voter,

b. APRM Fixed Neutron Flux - High,

c. APRM Simulated Thermal Power

- High,

d. Reactor Vessel Steam Dome Pressure High,

e. TSV Closure Trip

2. Manual SCRAM SCRAM Vessel Pressure ASME Code Compliance (USAR 14.5.1)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. Reactor Vessel Steam Dome Pressure High,

d. 2-of-4 Voter

2. Manual SCRAM SCRAM Generator Load Rejection (USAR 14.4.1)

1. Automatic Initiation

a. APRM Fixed Neutron Flux -

High,

b. APRM Simulated Thermal Power

- High,

c. Reactor Vessel Steam Dome Pressure High,

d. Turbine Control Valve Fast

Closure,

e. 2-of-4 Voter

2. Manual SCRAM 2.f - Average Power Range Monitors, OPRM Upscale SCRAM None

1. Automatic Initiation

a. APRM Oscillation Power Range Monitor Upscale

2. Manual SCRAM 2.g - Average Power Range

Monitors, Extended Flow Window Stability -

High (an extension of the Simulated Thermal Power -

High function)

SCRAM Generator Load Rejection (USAR 14.4.1)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. Reactor Vessel Steam Dome Pressure High,

d. Turbine Control Valve Fast

Closure,

e. APRM 2-of-4

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 57 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips SCRAM Loss of Feedwater Heater (USAR 14.4.2)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM - Fixed Neutron Flux -

High,

c. 2-of-4 Voter,

2. Manual SCRAM SCRAM Rod Withdrawal Error - Low Power (USAR 14.4.3)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Neutron Flux - High,

c. 2-of-4 Voter

d. Rod Block Monitor

2. Manual SCRAM SCRAM Feedwater Controller Failure -

Maximum Demand (USAR 14.4.4) -

high level trip results in turbine trip

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. TSV Closure,

d. 2-of-4 Voter

2. Manual SCRAM SCRAM Turbine Trip without Bypass (USAR 14.4.5)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. Reactor Vessel Steam Dome Pressure High,

d. TSV Closure Trip,

e. 2-of-4 Voter

2. Manual SCRAM SCRAM Vessel Pressure ASME Code Compliance (USAR 14.5.1)

1. Automatic Initiation

a. APRM Simulated Thermal Power

- High,

b. APRM Fixed Neutron Flux - High,

c. Reactor Vessel Steam Dome Pressure High,

d. 2-of-4 Voter

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 58 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 3 - Reactor Vessel Steam Dome Pressure - High SCRAM Generator Load Rejection (USAR 14.4.1)

1. Automatic Initiation

a. Reactor Vessel Steam Dome Pressure High,

b. Turbine Control Valve Fast

Closure,

c. APRM Fixed Neutron Flux - High,

d. APRM Simulated Thermal Power

- High,

e. APRM 2-of-4 Voter APRM Upscale/INOP

2. Manual SCRAM SCRAM Turbine Trip without Bypass (USAR 14.4.5)

1. Automatic Initiation

a. Reactor Vessel Steam Dome Pressure High

b. TSV Closure Trip

c. APRM Fixed Neutron Flux - High,

d. APRM Simulated Thermal Power

- High,

e. 2-of-4 Voter

2. Manual SCRAM SCRAM Vessel Pressure ASME Code Compliance (USAR 14.5.1)

1. Automatic Initiation

a. Reactor Vessel Steam Dome Pressure High,

b. APRM Simulated Thermal Power

- High,

c. APRM Fixed Neutron Flux - High,

d. 2-of-4 Voter

2. Manual SCRAM SCRAM Pressure Regulator Failure - Closed (ATWS USAR 14.8.2)

1. Automatic Initiation

a. Reactor Vessel Steam Dome Pressure High

b. APRM Neutron Flux - High

c. APRM Simulated Thermal Power

- High,

d. 2-of-4 Voter

2. Manual SCRAM 4 - Reactor Vessel Water Level - Low SCRAM None

1. Automatic Initiation

a. Low Water Level

2. Manual SCRAM 5 - Main Steam Isolation Valve -

Closure SCRAM Main Steam Line Break Accident (USAR 14.7.3)

1. Automatic Initiation

a. Main Steam Isolation Valve -

Closure

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 59 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips SCRAM Pressure Regulator Failure - Closed (ATWS USAR 14.8.2)

1. Automatic Initiation

a. Main Steam Isolation Valve -

Closure,

b. Reactor Steam Dome Pressure -

High,

c. APRM Fixed Neutron Flux - High

d. APRM Simulated Thermal Power

- High,

e. 2-of-4 Voter

2. Manual SCRAM 6 - Drywell Pressure - High SCRAM None

1. Automatic Initiation

a. Drywell Pressure - High,

b. Reactor Vessel Level - Low

2. Manual SCRAM 7.a - Scram Discharge Volume Water Level -

High, Resistance Temperature Detector SCRAM None

1. Automatic Initiation

a. Four diverse sensors volume

2. Manual SCRAM 7.b - Scram Discharge Volume Water Level -

High, Float Switch SCRAM None

1. Automatic Initiation

a. Four diverse sensors volume

2. Manual SCRAM 8 - Turbine Stop Valve - Closure SCRAM Feedwater Controller Failure -

Maximum Demand (USAR 14.4.4)

1. Automatic Initiation

a. TSV Closure,

b. APRM Fixed Neutron Flux - High,

c. APRM Simulated Thermal Power

- High,

d. 2-of-4 Voter

2. Manual SCRAM SCRAM Turbine Trip without Bypass (USAR 14.4.5)

1. Automatic Initiation

a. TSV Closure Trip

b. Reactor Vessel Steam Dome Pressure High

c. APRM Fixed Neutron Flux - High,

d. APRM Simulated Thermal Power

- High,

e. 2-of-4 Voter

2. Manual SCRAM

L-MT-20-036 NSPM Enclosure Page 60 of 75 Table 24-1: RPS Actuation Instrument Diversity TS 3.3.1.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips SCRAM Pressure Regulator Failure - Closed (ATWS USAR 14.8.2)

1. Automatic Initiation

a. TSV Closure Trip

b. Reactor Water Level - Low

c. Reactor Vessel Steam Dome Pressure - High

2. Manual SCRAM SCRAM Loss of Condenser Vacuum (Discussed in USAR 7.6.1.2.6)

1. Automatic Initiation

a. Reactor Vessel Steam Dome Pressure High

b. APRM Fixed Neutron Flux - High,

c. APRM Simulated Thermal Power

- High,

d. 2-of-4 Voter

2. Manual SCRAM 9 - Turbine Control Valve Fast

Closure, Acceleration Relay Oil Pressure - Low SCRAM Generator Load Rejection (USAR 14.4.1)

1. Automatic Initiation

a. Turbine Control Valve Fast

Closure,

b. Reactor Vessel Steam Dome Pressure High,

c. APRM Fixed Neutron Flux - High,

d. APRM Simulated Thermal Power

- High,

e. 2-of-4 Voter

2. Manual SCRAM 10 - Reactor Mode Switch -

Shutdown Position SCRAM None

1. Manual SCRAM

a. Mode Switch to Shutdown,

b. Manual SCRAM Pushbuttons,

c. RPS Channel Test Switches,

d. Manually initiate Alternate Rod Insertion (ARI) 11 - Manual Scram SCRAM None

1. Manual SCRAM

a. Manual SCRAM Pushbuttons,

b. RPS Channel Test Switches,

c. De-energize Trip System Power,

d. Manually initiate ARI

L-MT-20-036 NSPM Enclosure Page 61 of 75 Table 24-2: Feedwater Pump and Main Turbine High Water Level Trip Actuation Instrument Diversity TS 3.3.2.2 Function Safety Function Plant Condition/Accident Diverse Reactor Trips Reactor Vessel Water Level -

High Trip of Feedwater Pumps and Main Turbine Feedwater Controller Failure -

Maximum Demand (USAR 14.4.4)

1. Automatic Initiation

a. Reactor Vessel Water Level - High

2. Manual Reactor Feed Pump (RFP) and Turbine Trip Table 24-3: Anticipate Transient Without SCRAM Recirculation Pump Trip Instrument Diversity TS 3.3.4.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips a - Reactor Vessel Water Level - Low Low Trip both Recirculation Pumps Anticipated Transient Without SCRAM (USAR 14.8.2 and 14.7.6.2)

1. Automatic Initiation

a. Reactor Vessel Water Level-Low

Low,

b. Reactor Pressure - High

2. Manual Recirculation Pump Trip (RPT) b - Reactor Vessel Steam Dome Pressure - High Trip both Recirculation Pumps Anticipated Transient Without SCRAM (USAR 14.8.2 and 14.7.6.2)

1. Automatic Initiation

a. Reactor Pressure - High,

b. Reactor Vessel Water Level-Low Low

2. Manual RPT Table 24-4: Emergency Core Cooling System (ECCS) Instrumentation Diversity TS 3.3.5.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 1.a - Core Spray (CS) System, Reactor Vessel Water Level -

Low Low Actuate both CS system divisions and the associated emergency diesel generator (EDG)

Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Water Vessel - Low

Low,

b. Drywell Pressure - High

2. Manual Spray Initiation 1.b - CS System, Drywell Pressure

- High Actuate both CS system divisions and the associated EDG Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Drywell Pressure - High,

b. Reactor Water Vessel - Low Low

2. Manual Spray Initiation

L-MT-20-036 NSPM Enclosure Page 62 of 75 Table 24-4: Emergency Core Cooling System (ECCS) Instrumentation Diversity TS 3.3.5.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 1.c - CS System, Reactor Steam Dome Pressure -

Low (Injection Permissive)

Permit CS System Actuation Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Pressure-Low (Injection Permissive)

2. Manual Spray Initiation 1.d - CS System, Reactor Steam Dome Pressure Permissive - Low (Pump Permissive)

Permit CS System Actuation Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Pressure-Low (Pump Permissive)

2. Manual Spray Initiation 1.e - CS System, Reactor Steam Dome Pressure Permissive -

Bypass Timer (Pump Permissive)

Permit CS System Actuation Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Pressure-Low (Pump Permissive - Bypass Timer)

2. Manual Spray Initiation 1.f - CS System, Core Spray Pump Start - Time Delay Relay Permit CS System Actuation Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Pump Start Time Delay Relay

2. Manual Spray Initiation 2.a - Low Pressure Coolant Injection (LPCI)

System, Reactor Vessel Water Level - Low Low Actuate both LPCI system divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Water Vessel - Low

Low,

b. Drywell Pressure - High

2. Manual Injection 2.b - LPCI System, Drywell Pressure - High Actuate both LPCI system divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Drywell Pressure - High,

b. Reactor Water Vessel - Low Low

2. Manual Injection 2.c - LPCI System, Reactor Steam Dome Pressure - Low (Injection Permissive)

Permit actuation of both LPCI divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Pressure-Low (Injection Permissive) with contacts in both LPCI divisions

2. Manual Injection

L-MT-20-036 NSPM Enclosure Page 63 of 75 Table 24-4: Emergency Core Cooling System (ECCS) Instrumentation Diversity TS 3.3.5.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 2.d - LPCI System, Reactor Steam Dome Pressure Permissive - Low (Pump Permissive)

Permit actuation of both LPCI divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Pressure-Low (Pump Permissive) with contacts in both LPCI divisions

2. Manual Injection 2.e - LPCI System, Reactor Steam Dome Pressure Permissive -

Bypass Timer (Pump Permissive)

Permit actuation of both LPCI divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Pressure-Low (Pump Permissive - Bypass Timer) with contacts in both LPCI divisions

2. Manual Injection 2.f - LPCI System, Low Pressure Coolant Injection Pump Start - Time Delay Relay Actuate both LPCI system divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Pump Start Time Delay Relay

2. Manual Injection 2.g - LPCI System, Low Pressure Coolant Injection Pump Discharge Flow -

Low (Bypass)

Provide LPCI Pump Protection Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. LPCI Discharge Flow - Low (Bypass)

2. Manual Injection 2.h - LPCI System, Reactor Steam Dome Pressure - Low (Break Detection)

Actuate both LPCI system divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Steam Dome Pressure

- Low (Break Detection) with contacts in both LPCI divisions

2. Manual Injection 2.i - LPCI

System, Recirculation Pump Differential Pressure - High (Break Detection)

Actuate either LPCI pump in each LPCI division Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Recirc Pump dP - High (Break Detection) with contacts in both LPCI divisions

2. Manual Injection

L-MT-20-036 NSPM Enclosure Page 64 of 75 Table 24-4: Emergency Core Cooling System (ECCS) Instrumentation Diversity TS 3.3.5.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 2.j - LPCI

System, Recirculation Riser Differential Pressure - High (Break Detection)

Actuate both LPCI Divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Recirc Riser dP - High (Break Detection) with contacts in both divisions

2. Manual Injection 2.k - LPCI System, Reactor Steam Dome Pressure - Time Delay Relay (Break Detection)

Actuate one LPCI system division Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Steam Dome Pressure Time Delay Relay (Break Detection)

2. Manual Injection 2.l - LPCI

System, Recirculation Pump Differential Pressure - Time Delay Relay (Break Detection)

Actuate both LPCI Divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Recirc Pump dP Time Delay Relay (Break Detection)

2. Manual Injection 2.m - LPCI

System, Recirculation Riser Differential Pressure - Time Delay Relay (Break Detection)

Actuate/de-actuate both LPCI divisions Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Recirc Riser dP Time Delay Relay (Break Detection)

2. Manual Injection 3.a - High Pressure Coolant Injection (HPCI)

System, Reactor Vessel Water Level - Low Low Actuate HPCI system Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Vessel Water Level -

Low Low,

b. Drywell Pressure High

2. Manual Initiation 3.b - HPCI System, Drywell Pressure - High Actuate HPCI system Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Drywell Pressure - High,

b. Reactor Vessel Water Level -

Low Low

2. Manual Initiation 3.c - HPCI System, Reactor Vessel Water Level - High Trip HPCI turbine None

L-MT-20-036 NSPM Enclosure Page 65 of 75 Table 24-4: Emergency Core Cooling System (ECCS) Instrumentation Diversity TS 3.3.5.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 3.d - HPCI

System, Condensate Storage Tank Level - Low Change HPCI suction path Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Condensate Storage Tank Level - Low

2. Manual Path Swap 3.e - HPCI

System, Suppression Pool Water Level -

High Change HPCI suction path Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Suppression Pool Water Level - High

2. Manual Path Swap 3.f - HPCI System, High Pressure Coolant Injection Pump Discharge Flow -

Low (Bypass)

Provide HPCI Pump Protection Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Pump Discharge Flow - Low (Bypass)

2. Manual Bypass 4.a - ADS Trip System A, Reactor Vessel Water Level -

Low Low Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Vessel Water Level - Low Low

2. Manual Initiation 4.b - ADS Trip System A, Automatic Depressurization System Initiation Timer Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. System Initiation Timer

2. Manual Initiation 4.c - ADS Trip System A, Core Spray Pump Discharge Pressure - High Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. CS Pump Discharge Pressure with contacts from both CS divisions

2. Manual Initiation 4.d - ADS Trip System A, Low Pressure Coolant Injection Pump Discharge Pressure - High Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. LPCI Pump Discharge Pressure with contacts from both CS divisions

2. Manual Initiation 5.a - ADS Trip System B, Reactor Vessel Water Level -

Low Low Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Vessel Water Level - Low Low

2. Manual Initiation

L-MT-20-036 NSPM Enclosure Page 66 of 75 Table 24-4: Emergency Core Cooling System (ECCS) Instrumentation Diversity TS 3.3.5.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 5.b - ADS Trip System B, Automatic Depressurization System Initiation Timer Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. System Initiation Timer

2. Manual Initiation 5.c - ADS Trip System B, Core Spray Pump Discharge Pressure - High Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. CS Pump Discharge Pressure with contacts from both CS divisions

2. Manual Initiation 5.d - ADS Trip System B, Low Pressure Coolant Injection Pump Discharge Pressure - High Actuate all ADS valves Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. LPCI Pump Discharge Pressure with contacts from both CS divisions

2. Manual Initiation Table 24-5: Reactor Core Isolation Cooling System (RCIC) Instrumentation Diversity TS 3.3.5.2 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 1 - Reactor Vessel Water Level - Low Low Actuate RCIC None

1. Automatic Initiation

a. Reactor Vessel Water Level -Low Low

2. Manual Initiation 2 - Reactor Vessel Water Level - High Trip RCIC turbine None

1. Automatic RCIC Stop

a. Reactor Water Level - High

2. Manually Secure 3 - Condensate Storage Tank Level - Low Change RCIC suction path None

1. Automatic Initiation

a. Condensate Storage Tank Level - Low

2. Manual Path Swap

L-MT-20-036 NSPM Enclosure Page 67 of 75 Table 24-6: Primary Containment Isolation (PCI) Instrumentation Diversity TS 3.3.6.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 1.a - Main Steam Line Isolation, Reactor Vessel Water Level - Low Low Main Steam Line Isolation Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Water Level - Low Low,

b. Main Steam Line Pressure - Low

c. Main Steam Line Flow - High,

d. Main Steam Line Tunnel Temperature - High

2. Manual Isolation Main Steam Line Break Accident (USAR 14.7.3)

1. Automatic Initiation

a. Reactor Water Level - Low

Low,

b. Main Steam Line Pressure -

Low,

c. Main Steam Line Flow - High,

d. Main Steam Line Tunnel Temperature - High

2. Manual Isolation 1.b - Main Steam Line Isolation, Main Steam Line Pressure - Low Main Steam Line Isolation Main Steam Line Break Accident (USAR 14.7.3)

1. Automatic Initiation

a. Main Steam Line Pressure -

Low,

b. Reactor Water Level - Low

Low,

c. Main Steam Line Flow - High,

d. Main Steam Line Tunnel Temperature - High

2. Manual Isolation 1.c - Main Steam Line Isolation, Main Steam Line Flow - High Main Steam Line Isolation Main Steam Line Break Accident (USAR 14.7.3)

1. Automatic Initiation

a. Main Steam Line Flow - High,

b. Reactor Water Level - Low Low,

c. Main Steam Line Pressure -

Low,

d. Main Steam Line Tunnel Temperature - High

2. Manual Isolation 1.d - Main Steam Line Isolation, Main Steam Line Tunnel Temperature -

High Main Steam Line Isolation Main Steam Line Break Accident (USAR 14.7.3)

1. Automatic Initiation

a. Main Steam Line Tunnel Temperature - High,

b. Reactor Water Level - Low

Low,

c. Main Steam Line Pressure -

Low,

d. Main Steam Line Flow - High

2. Manual Isolation

L-MT-20-036 NSPM Enclosure Page 68 of 75 Table 24-6: Primary Containment Isolation (PCI) Instrumentation Diversity TS 3.3.6.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 2.a - Primary Containment Isolation, Reactor Vessel Water Level - Low Primary Containment Isolation Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Reactor Vessel Water Level -

Low Low,

b. Drywell Pressure - High

2. Manual Isolation of Individual Valves 2.b - Primary Containment Isolation, Drywell Pressure - High Primary Containment Isolation None

1. Automatic Initiation

a. Drywell Pressure - High,

b. Reactor Vessel Water Level -

Low Low

2. Manual Isolation of Individual Valves 3.a - High Pressure Coolant Injection (HPCI)

System Isolation, HPCI Steam Line Flow - High HPCI Isolation Main Steam Line Break Accident (USAR 14.7.3) - refer to 7.6.3.3.2

1. Automatic Initiation

a. HPCI Steam Line Flow - High,

b. HPCI Steam Supply Line Pressure - Low,

c. HPCI Steam Line Area Temperature - High Manual Isolation of Individual Valves 3.b - HPCI System Isolation, HPCI Steam Supply Line Pressure - Low HPCI Isolation Main Steam Line Break Accident (USAR 14.7.3) - refer to 5.2-3b and 7.6-2

1. Automatic Initiation

a. HPCI Steam Supply Line Pressure - Low,

b. HPCI Steam Line Flow - High,

c. HPCI Steam Line Area Temperature - High Manual Isolation of Individual Valves 3.c - HPCI System Isolation, HPCI Steam Line Area Temperature -

High HPCI Isolation Main Steam Line Break Accident (USAR 14.7.3) - refer to 5.2-3b and 7.6-2

1. Automatic Initiation

a. HPCI Steam Line Area Temperature - High,

b. HPCI Steam Line Flow - High,

c. HPCI Steam Supply Line Pressure - Low Manual Isolation of Individual Valves 4.a - Reactor Core Isolation Cooling (RCIC) System Isolation, RCIC Steam Line Flow -

High RCIC Isolation Main Steam Line Break Accident (USAR 14.7.3) - refer to 7.6.3.3.2

1. Automatic Initiation

a.

RCIC Steam Line Flow - High,

b.

RCIC Steam Supply Line Pressure - Low,

c. RCIC Steam Line Area Temperature - High Manual Isolation of Individual Valves

L-MT-20-036 NSPM Enclosure Page 69 of 75 Table 24-6: Primary Containment Isolation (PCI) Instrumentation Diversity TS 3.3.6.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 4.b - RCIC System Isolation, RCIC Steam Supply Line Pressure - Low RCIC Isolation Main Steam Line Break Accident (USAR 14.7.3) - refer to 5.2-3b and 7.6-2

1. Automatic Initiation

a.

RCIC Steam Supply Line Pressure - Low,

b.

RCIC Steam Line Flow - High,

c. RCIC Steam Line Area Temperature - High Manual Isolation of Individual Valves 4.c - RCIC System Isolation, RCIC Steam Line Area Temperature

- High RCIC Isolation Main Steam Line Break Accident (USAR 14.7.3) - refer to 5.2-3b and 7.6-2

1. Automatic Initiation

a.

RCIC Steam Line Area Temperature - High,

b.

RCIC Steam Line Flow - High,

c. RCIC Steam Supply Line Pressure - Low Manual Isolation of Individual Valves 5.a - Reactor Water Cleanup (RWCU) System Isolation, RWCU Flow - High RWCU Isolation None

1. Automatic Initiation

a. RWCU Flow-High,

b. RWCU Room Temperature -

High,

c. Drywell Pressure - High,

d. Reactor Vessel Level - Low

2. Manual Isolation of Individual Valves 5.b - RWCU System Isolation, RWCU Room Temperature -

High RWCU Isolation None

1. Automatic Initiation

a. RWCU Room Temperature -

High,

b. RWCU Flow-High,

c. Drywell Pressure - High,

d. Reactor Vessel Level - Low

2. Manual Isolation of Individual Valves 5.c - RWCU System Isolation, Drywell Pressure -

High RWCU Isolation None

1. Automatic Initiation

a. Drywell Pressure - High,,

b. RWCU Flow-High,

c. RWCU Room Temperature -

High

d. Reactor Vessel Level - Low

2. Manual Isolation of Individual Valves 5.d - RWCU System Isolation, SLC System Initiation RWCU Isolation to Prevent Dilution None

1. Manually Initiated

L-MT-20-036 NSPM Enclosure Page 70 of 75 Table 24-6: Primary Containment Isolation (PCI) Instrumentation Diversity TS 3.3.6.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 5.e - RWCU System Isolation, Reactor Vessel Water Level - Low Low RWCU Isolation Main Steam Line Break Accident (USAR 14.7.3)

1. Automatic Initiation

a. Reactor Vessel Level - Low,

b. RWCU Flow-High,

c. RWCU Room Temperature -

High,

d. Drywell Pressure - High

2. Manual Isolation of Individual Valves 6.a - Shutdown Cooling System Isolation, Reactor Steam Dome Pressure - High Shutdown Cooling System Isolation None

1. Automatic Initiation

a. Reactor Pressure-High,

b. Reactor Vessel Level-Low

2. Manual Isolation of Individual Valves 6.b - Shutdown Cooling System Isolation, Reactor Vessel Water Level - Low Shutdown Cooling System Isolation None

1. Automatic Initiation

a. Reactor Vessel Level-Low,

b. Reactor Pressure-High

2. Manual Isolation of Individual Valves 7.a - Traversing Incore Probe System Isolation, Reactor Vessel Water Level - Low Traversing Incore Probe System Isolation None

1. Automatic Initiation

a. Reactor Vessel Water Level -

Low,

b. Drywell Pressure - High

2. Manual Isolation 7.b - Traversing Incore Probe System Isolation, Drywell Pressure - High Traversing Incore Probe System Isolation None

1. Automatic Initiation

a. Drywell Pressure - High,

b. Reactor Vessel Water Level - Low

2. Manual Isolation Table 24-7: Mechanical Vacuum Pump Isolation Instrumentation Diversity TS 3.3.7.2 Function Safety Function Plant Condition/Accident Diverse Reactor Trips Main Steam Line Tunnel Radiation

- High Mechanical Vacuum Pump Isolation Control Rod Drop Accident (USAR 14.7.1)

Automatic Initiation Main Steam Line Radiation - High Manual Initiation

L-MT-20-036 NSPM Enclosure Page 71 of 75 Table 24-8: Loss of Power (LOP) Instrumentation Diversity TS 3.3.8.1 Function Safety Function Plant Condition/Accident Diverse Reactor Trips 1 - 4.16 kV Essential Bus, Loss of Voltage Sense Essential Bus Loss of Voltage and Transfer to EDGs Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Essential Bus, Loss of Voltage

2. Manual EDG Initiation and Transfer 2.a - 4.16 kV Essential Bus, Degraded Voltage - Bus Undervoltage Sense Essential Bus Degraded Voltage and transfer to EDGs Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Essential Bus, Degraded Voltage - Bus Undervoltage

2. Manual EDG Initiation and Transfer 2.b - 4.16 kV Essential Bus, Degraded Voltage - Time Delay Sense Essential Bus Degraded Voltage and transfer to EDGs Loss of Coolant Accidents (USAR 14.7.2)

1. Automatic Initiation

a. Essential Bus, Degraded Voltage - Time Delay

2. Manual EDG Initiation and Transfer NSPM Response to RAI 24.b These "manual actuations" are defined in plant operation procedures to which operators are trained. Select manual actions are modelled in the PRA when it is determined that credit for the actions have a positive risk impact. Those actions that are modelled incorporate appropriate human error probabilities. For manual actions that arent modelled, failure of the automatic function results in a failure of the function in the fault tree.

RAI 25 - Differences Between TSTF-505 and the LAR The LAR proposed to modify TS requirements to permit the use of Risk-Informed Completion Times in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b." The amendment requests to make changes to TS 3.6.1.3 "Primary Containment Isolation Valves (PCIVs)".

TSTF-505, Revision 2 TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)" pertaining to BWR/4 STS (page 591 of 850) contains the following:

Condition E requires additional justification. RAs A.2, C.2, E.2 and E.3 to specify the periodic performance of an action and Conditions F, G, and H are default Conditions. Therefore, they are excluded. Conditions B, C, and D are excluded.

The following differences were noted by the NRC staff between TSTF-505, Revision 2 and the LAR:

L-MT-20-036 NSPM Enclosure Page 72 of 75 The LAR "Completion Time" for C.2 has an AND statement that reads in entirety "Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if performed within the previous 92 days, for isolation devices inside primary containment." This AND statement does not appear in the corresponding TSTF-505 Revision 2.

Conversely, TSTF-505 Revision 2 Condition E.2 contains a similar (not identical) AND statement that is not replicated for the "Completion Time" of LAR D.2.

The LAR "Completion Time" for the "Required Action" of D.1 proposes to invoke "INSERT RICT 2." TSTF-505 Revision 2 for the "Required Action" of E.1 of the counterpart "Completion Time" proposes to invoke the equivalent of LAR "INSERT RICT 1." The difference between "INSERT RICT 2" versus "INSERT RICT 1" consists of the former including a NOTE to the effect "Not applicable when a loss of function occurs."

TSTF-505, Revision 2 has an E.3 "Required Action" of "Perform SR 3.6.1.3.7 for resilient seal purge valves closed to comply with Required Action E.1" with a "Completion Time" of "Once per [92] days [following isolation]." The staff notes that in the LAR there is no corresponding D.3 invoking SR 3.6.1.3.11.

The LAR does not address the above identified differences between the proposed changes of the LAR and TSTF-505, Revision 2.

Provide justification of the noted differences between the LAR and TSTF-505, Rev 2. Include justification for the conclusion that the proposed change to TS 3.6.1.3 does not represent a loss of function.

NSPM Response to RAI 25:

Regarding the LAR "Completion Time" for C.2 AND statement, during the MNGP Improved Technical Specifications (ITS) conversion, a new CT was added to TS 3.6.1.3 Required Action C.2 since these are single valve penetrations that can be isolated by a device inside primary containment. This CT is identical as that for ISTS 3.6.1.3 Required Action A.2. Attachment 3 to this enclosure contains a revised Table A4-1 that provides this additional item.

Regarding the TSTF-505 Condition E.2 AND statement that is not replicated in LAR D.2, the second Completion Time in ISTS 3.6.1.3 Required Action E.2 is deleted in the MNGP TS 3.6.1.3 Required Action D.2 because there are no isolation devices inside the containment associated with the primary containment vent and purge valve penetration flow paths. Attachment 3 to this enclosure contains a revised Table A4-1 that provides this additional item.

Regarding the LAR D.1 proposing to invoke "INSERT RICT 2" and the TSTF-505 Rev. 2 markup of the equivalent Required Action E.1 not including the loss of function note, TSTF-505 Rev. 2 does NOT include any such notes in any of the ISTS markups as the

L-MT-20-036 NSPM Enclosure Page 73 of 75 markup is done generically and because of plant design diversity. However, TSTF-505 Rev. 2 contains guidance at the end of Table 1 that Conditions of the form "one or more" should get completion times notes that restrict the RICT from a loss of function scenario.

Therefore, the NOTE in our markup is not inconsistent with TSTF-505. Attachment 3 to this enclosure contains a revised Table A4-1 that clarifies this item.

Regarding the LAR having no corresponding D.3 and SR 3.6.1.3.11, MNGP does not have a Required Action D.3 corresponding to ISTS E.3. This Required Action was not brought over during ITS conversion as this allowance was "consistent with requirements in place" at the time of the MNGP ITS conversion. Therefore, the TSTF-505 markup for Required Action E.3 is not applicable. Attachment 3 to this enclosure contains a revised Table A4-1 that provides this additional item.

3.0 REFERENCES

1.

Letter (L-MT-20-003) from NSPM to the NRC, "License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," dated March 30, 2020 (ADAMS Accession No. ML20090F820)

2.

Letter from the Technical Specification Task Force (TSTF) to the NRC, "TSTF Comments on Draft Safety Evaluation for Traveler TSTF-505, Provide Risk-Informed Extended Completion Times and Submittal of TSTF-505, Revision 2," Revision 2, dated July 2, 2018 (ADAMS Accession No. ML18183A493)

3.

Email from the NRC to NSPM, "Monticello Request for Additional Information RE:

TSTF-505 license amendment request," dated October 26, 2020 (ADAMS Accession No.ML20302A197)

4.

NRC Report NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," dated February 2007 (ADAMS Accession No. ML070650650)

5.

NEI 16-06, "Crediting Mitigating Strategies in Risk-Informed Decision Making," dated August 2016

6.

NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power Facilities" (EPRI 1011989),

dated September 2005

7.

NUREG-2178, "Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE -FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume," (ADAMS Accession No. ML16110A140)

L-MT-20-036 NSPM Enclosure Page 74 of 75

8.

NUREG-2180, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE),"

dated December 2016 (ADAMS Accession No. ML16343A058)

9.

NUREG/CR-5500, Vol 3, "Reliability Study: General Electric Reactor Protection System, 1984-1995," dated February 1999

10. NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b," Rev. 0-A, dated November 2006

11. NUREG/CR-6931, "Cable Response to Live Fire (CAROLFIRE) Volume 1: Test Descriptions and Analysis of Circuit Response Data," dated April 2008 (ADAMS Accession No. ML081190230)

12. ASME/ANS-Ra-Sa-2009 Standard Nonmandatory Appendix 1-A, "PRA Maintenance, PRA Upgrade, and the Advisability of Peer Review"

13. RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Rev. 2, dated March 2009

14. Letter (L-MT-13-055) from NSPM to the NRC, "License Amendment Request for Transition to AREVA ATRIUM 10XM Fuel and AREVA Safety Analysis Methodology,"

dated July 15, 2013 (ADAMS Accession No. ML13200A187

15. Letter (L-MT-19-033) from NSPM to the NRC, "Thirty-Day Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46," dated June 7, 2019 (ADAMS Accession No. ML19158A452)

16. Letter (L-MT-18-010) from NSPM to the NRC, "Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," dated March 28, 2018 (ADAMS Accession No. ML18087A323)

17. EPRI Technical Report TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," dated December 2008

18. Letter from the NRC to NSPM, "Monticello Nuclear Generating Plant - Issuance of Amendment to Transition to AREVA ATRIUM 10XM Fuel and AREVA Safety Analysis Methods (TAC No. MF2479)," dated June 5, 2015 (ADAMS Accession No. ML15072A141)

19. NRC NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, dated March 2017

20. Letter from the NRC to NSPM, "Monticello Nuclear Generating Plant (MNGP) - Issuance of Amendment Regarding Completion Time to Restore a Low-Pressure Emergency Core

L-MT-20-036 NSPM Enclosure Page 75 of 75 Cooling Subsystem to Operable Status (TAC No. MD9170)," dated July 10, 2009 (ADAMS Accession No. ML091480782)

21. Letter from the NRC to NSPM, "Monticello Nuclear Generating Plant - Issuance of Amendment to Revise Technical Specification 3.5.1, ECCS [Emergency Core Cooling System] - Operating (TAC No. MF3911)," dated November 3, 2014 (ADAMS Accession No. ML14246A449)

22. Letter from the NRC to NSPM, "Monticello Nuclear Generating Plant - Issuance of Amendment No. 176 to Renewed Facility Operating License Regarding Extended Power Uprate (TAC No. MD9990)," dated December 9, 2013 (ADAMS Accession No. ML13316B298)

23. AREVA Report ANP-3211(P), Revision 1, "Monticello EPU LOCA Break Spectrum Analysis for ATRIUM' 10XM Fuel," dated July 2013

24. Letter (L-MT-17-083) from NSPM to the NRC, "License Amendment Request: Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program," dated December 19, 2017 (ADAMS Accession No. ML17353A189)

25. NUREG-75/014 (WASH-1400), "Reactor Safety Study: An assessment of Accident Risks in US Commercial Nuclear Power Plants," October 1975

ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" UPDATED MNGP TSTF-505 LAR ENCLOSURE 4.0 INFORMATION SUPPORTING JUSTIFICATION OF EXCLUDING SOURCES OF RISK NOT ADDRESSED BY THE PRA MODELS Sections 3.0 and 6.0 (8 Pages Follow)

L-MT-20-036 NSPM Page 1 of 8 3.0 CONSERVATIVE SEISMIC ANALYSIS This section presents a conservative analysis of the potential seismic impact for inclusion in the decision-making process, as a seismic PRA (SPRA) is not available for MNGP. The process for analyzing an unscreened external hazard without the use of a full PRA involves the following three steps:

1. Estimate Conservative Seismic CDF (SCDF) Contribution

2. Evaluate Potential Risk Increases Due to Out of Service Equipment

3. Estimate Conservative Seismic LERF (SLERF) Contribution 3.1 Conservatively Estimate Seismic CDF A seismic PRA is not developed for MNGP. MNGP performed the equivalent of a reduced scope seismic margins assessment (SMA) for its Individual Plant Examination for External Events (IPEEE) (Reference 9), with an additional focus on a few components, in accordance with Supplement 5 of Generic Letter 88-20 (Reference 10). The seismic hazard for the MNGP site was re-evaluated in 2014 and provided to the NRC (Reference 11). The site safe shutdown earthquake (SSE) is documented in this report is 0.12 g. For screening purposes, a Ground Motion Response Spectrum (GMRS) was developed and a probabilistic seismic hazard analysis (PSHA) was completed using the Central and Eastern United States Seismic Source Characterization (CEUS-SSC) for nuclear facilities and the updated Electric Power Research Institute (EPRI) Ground-Motion Model (GMM). For both the 1 to 10 Hz response spectrum and portions of higher frequency (>10 Hz), the GMRS exceeds the SSE, which merited a seismic risk evaluation, spent fuel pool evaluation and a high frequency (HF) confirmation. However, based on the NRC staff's further comparison of the GMRS to the SSE and the review of additional hazard and risk information, the NRC staff concluded, as described in a letter dated October 27, 2015 (Reference 12), that a seismic risk evaluation was not merited for the MNGP. In addition, the staff concluded that the GMRS determined by the licensee adequately characterizes the reevaluated seismic hazard for the MNGP site. NSPM submitted a high frequency (HF) confirmation report (Reference 13) for the MNGP to the NRC and the NRC concluded that the licensee correctly implemented the guidance in conducting the HF confirmation for the MNGP (Reference 14).

NSPM also submitted a Mitigating Strategies Assessment (MSA) report (Reference 15) stating that the MNGP MSA was performed consistent with Appendix H of NEI 12-06, "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide," Revision 4 (Reference 16), which describes acceptable methods for demonstrating that the reevaluated seismic hazard is addressed within the MNGP mitigation strategies for beyond-design-basis external events.

Guidance document NEI 12-06, Revision 4, has been endorsed (with exceptions) by the NRC in JLD-ISG-2012-01, "Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events,"

Revision 2 (Reference 17). Therefore, the methodology used by the licensee is acceptable to perform an assessment of the mitigation strategies that addresses the reevaluated seismic hazard. The NRC completed its review of the seismic hazard MSA for the MNGP and concluded that sufficient information has been provided to demonstrate that the licensee's

L-MT-20-036 NSPM Page 2 of 8 plans for the development and implementation of guidance and strategies under Order EA-12-049 appropriately addressed the reevaluated seismic hazard information stemming from the 10 CFR 50.54(f) letter and that no further responses or regulatory actions associated with Phase 2 of Near-Term Task Force (NTTF) Recommendation 2.1 "Seismic" were required for MNGP (Reference 14).

Therefore, an alternative approach is taken to conservatively estimate seismic core damage frequency (SCDF) based on the current MNGP seismic hazard curve and assuming the seismic capacity of a component whose seismic failure would lead directly to core damage.

This approach to estimation of the SCDF uses the plant-level high confidence of low probability of failure (HCLPF) seismic capacity obtained from Table C-2 of Reference 18 and convolves the corresponding failure probabilities as a function of seismic hazard level with the site-specific hazard estimates for plants in the CEUS and spectral ratios developed from Reference 11. This is a commonly used approach to estimate SCDF when a seismic PRA is not available; see Section 10-B.9 of the ASME/ANS PRA Standard (Reference 5). This approach is consistent with approaches that have been used in other regulatory applications.

Based on Section 5.1 and Tables B-1 and B-2 of the MNGP Fukushima Near-Term Task Force (NTTF) Report (Reference 29), the MNGP plant-level HCLPF is assessed to be 0.19 with Beta-C of 0.4. This is based on the Expedited Seismic Evaluation Process (ESEP) undertaken for MNGP. The intent of the ESEP was to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI 3002000704 (Reference 30). The RLGM (Review Level Ground Motion) acceleration for MNGP was determined by comparing the MNGP Safe Shutdown Earthquake (SSE) acceleration with the MNGP Ground Motion Response Spectrum (GMRS) acceleration between a frequency of 1 Hz to 10 Hz. The RLGM for all MNGP buildings (i.e., Reactor Building, Administrative Building, and Turbine Building) except the Emergency Filtration Building (EFT) was determined by linearly scaling the MNGP SSE by the maximum safety factor (SF = GMRS/SSE ratio) of 1.49 between the 1 and 10 Hz range. The resulting 5%

damped RLGM, based on scaling the horizontal SSE by the SF of 1.49, was determined to be 0.19g. A similar calculation was performed for the EFT Building using the SSE based on RG 1.60. Although the IPEEE (Reference 9) screens the EFT Building structure at a review level earthquake (RLE) of 0.3g, the RLGM for SSCs within the EFT Building was determined to be 0.134g. A review of Tables B-1 and B-2 of Reference 29 indicates that the HCLPFs for the majority of the SSCs are close to or exceed 0.19g. The EFT Building SSCs that were screened at 0.134g will not cause direct core damage or direct large early release. An NTTF plant-level HCLPF of 0.19g is applied to MNGP.

Convolving the MNGP plant-level HCLPF seismic capacity (0.19g), composite variability (c of 0.4), with the new site-specific hazard estimates for plants in the CEUS and spectral ratios developed from Reference 11, the corresponding SCDFs were calculated using CAFTA 6.0b and FRANX 4.2 for the peak ground acceleration (PGA) hazard, as well as the 10 Hz, 5 Hz and 1 Hz spectral frequency hazard curves. Based on these calculations, the PGA seismic hazard curve produces the highest SCDF and is controlling. The total MNGP SCDF is 6.42E-6

L-MT-20-036 NSPM Page 3 of 8 per year based on the MNGP PGA seismic hazard curve. This SCDF value will be used as the conservative estimate of instantaneous SCDF (ICDFseismic) for the TSTF-505 submittal RICT calculations.

3.2 Evaluate Potential Seismic Risk Increase Due to Out-of-Service Equipment The approach taken in the computation of SCDF assumes that the SCDF can be based on the likelihood that a single seismic-induced failure leads to core damage. This approach is conservative and implicitly relies on the assumption that seismic-induced failures of equipment show a high degree of correlation (i.e., if one SSC fails, all similar SSCs will also fail). Direct use of this assumption in evaluating the risk increase from out-of-service equipment could lead to an underestimation of the change in risk. However, if one were to assume no correlation at all in the seismic failures, then the seismic risk would be lower than the risk predicted by a fully correlated model. It should be noted, however, that the change in risk using the un-correlated model with a redundant piece of important equipment out-of-service would be equivalent to that predicted by the correlated model.

If the industry-accepted approach (Reference 18) of correlation is assumed, the conditional core damage frequency given a seismic event will remain unaltered whether equipment is out-of-service or not. Thus, the risk increase due to out of service equipment cannot be greater than the total SCDF calculated by the conservative method used in Reference 31. That is, for the MNGP site, the delta SCDF from equipment out of service cannot be greater than 6.42E-6 per year.

To summarize the above considerations:

The baseline seismic risk in this approach is assumed to be zero, whereas there will always be some level of baseline seismic risk for a zero-maintenance plant configuration. Therefore, the incremental seismic risk (configuration seismic risk -

baseline seismic risk) will always be overstated using a seismic penalty based on the total estimated seismic risk.

The limiting HCLPF approach assumes that a failure of a component with seismic capacity at that HCLPF leads directly to core damage (CD). However, even common failure of a given set of components (e.g., all emergency diesel generators (EDGs))

would not lead directly to CD, especially in light of the post-Fukushima FLEX mitigating strategies now in place. In reality, there are few SSCs whose failure would lead to seismic CD with any significant frequency. Examples could be important structures, or the reactor pressure vessel, or "distributed systems" such as all cable trays or all piping systems.

In a seismic PRA, seismic impacts to similar components (e.g., all the EDGs for a given unit) are typically assumed to be correlated unless there are reasons to justify not correlating. Correlation has the effect of introducing common cause impacts. So, if one train of emergency AC power fails seismically, both trains are modeled as likely to fail

L-MT-20-036 NSPM Page 4 of 8 given the same seismic event. So, in general, most seismic impacts would effectively be equivalent to TS loss of function.

Given the above, the use of a seismic penalty based on assuming seismic core damage given the plant level HCLPF is appropriate.

Note that there is another significant conservatism inherent in this approach in addition to the above considerations. In determining the SCDF to be used in the RICT calculations, the full annual seismic hazard has been used. Since the maximum RICT backstop is 30 days, accounting for the full annual seismic hazard introduces more than a factor of 10 increase in the calculated SCDF.

3.3 Conservatively Estimate SLERF Contribution The SLERF was conservatively estimated by including the containment fragility in the convolution calculations. The seismic capability of the containment for MNGP was evaluated in the IPEEE (Reference 9). The IPEEE concluded that there are no containment-related seismic vulnerabilities. The MNGP IPEEE confirmed that the containment SSCs required to respond to seismic events could be screened at the IPEEE screening level of 0.3g based on the walkdowns conducted. The corresponding SLERFs were calculated by convolving the same MNGP plant-level HCLPF seismic capacity (0.19g), composite variability (c of 0.4) and the plant limiting HCLPF for containment integrity (0.3g), with the new site-specific hazard estimates for plants in the CEUS and spectral ratios developed from Reference 11. The calculations were performed using CAFTA 6.0b and FRANX 4.2 for the PGA hazard, as well as the 10 Hz, 5 Hz and 1 Hz spectral frequency hazard curves. The non-seismic conditional large early release probability is calculated as the ratio of the internal events LERF and the internal events CDF. Based on Table E5-1, the internal events LERF is 6.1E-07/yr and the internal events CDF is 6.54E-06/yr. This yields a non-seismic conditional large early release probability of 9.33E-02. The SLERF for each seismic interval is the product of the seismic initiating event frequency, the CCDP for each seismic interval, and the sum of the seismic and non-seismic conditional large early release probability. Based on these calculations, the PGA seismic hazard curve produces the highest SLERF and is controlling. The total MNGP SLERF is 2.35E-06 per year based on the PGA seismic hazard curve. This SLERF value will be used as the conservative estimate of instantaneous SLERF (ILERFseismic) for the TSTF-505 submittal RICT calculations.

3.4 Conclusion The above analysis provides the technical basis for addressing the seismic-induced core damage risk for MNGP by reducing the ICDP/ILERP criteria to account for a conservative estimate of the configuration risks due to seismic events.

The RICT and RMAT calculations are based on the discussion provided above. The actual RICT and RMAT calculations performed by the MNGP Configuration Risk Management Tool are based on adding an incremental 6.42E-06 per year SCDF contribution and a corresponding 2.35E-06 per year SLERF contribution to the configuration-specific delta CDF

L-MT-20-036 NSPM Page 5 of 8 and delta LERF attributed to internal and fire events contributions. This is accomplished by adding these seismic contributions to the instantaneous CDF/LERF whenever a RICT is in effect. This method ensures that an incremental seismic CDF/LERF equal to the conservative SCDF/SLERF is added to internal and fire events incremental CDF/LERF contribution for every RICT occurrence.

6.0 REFERENCES

1.

NEI Topical Report NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)

2.

Letter from the NRC to NEI, "Final Safety Evaluation for Nuclear Energy Institute (NEI)

Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995)," dated May 17, 2007 (ADAMS Accession No. ML071200238)

3.

Letter (L-MT-18-010) from NSPM to the NRC, "Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors," dated March 28, 2018 (ADAMS Accession No. ML18087A323)

4.

Letter (L-MT-17-083) from NSPM to the NRC, "License Amendment Request: Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program," dated December 19, 2017 (ADAMS Accession No. ML17353A189)

5.

ASME Standard ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated February 2, 2009

6.

NRC NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466)

7.

NRC NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," September 1975 (ADAMS Accession No. ML081510817)

8.

NRC NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," dated June 1991 (ADAMS Accession No. ML063550238)

9.

NSPM Report NSPLMI-95001, "Monticello Individual Plant Examination of External Events (IPEEE)," Revision 1, dated November 17, 1995 (Legacy Accession No.

9511300255)

L-MT-20-036 NSPM Page 6 of 8

10. NRC Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54(f) (Generic Letter No. 88-20)," dated November 23, 1988 (ADAMS Accession No. ML031150465)

11. Letter (L-MT-14-045) from NSPM to the NRC, "MNGP Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated May 14, 2014 (ADAMS Accession No. ML14136A288)

12. Letter from the NRC to NSPM, "Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 Seismic of the Near-Term Task Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident," dated October 27, 2015 (ADAMS Accession No. ML15194A015)

13. Letter (L-MT-17-025) from NSPM to the NRC, "High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," dated April 11, 2017 (ADAMS Accession No. ML17101A598)

14. Letter from the NRC to NSPM, "Monticello Nuclear Generating Plant - Staff Review of Mitigating Strategies Assessment Report of The Impact of The Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter," dated October 10, 2017 (ADAMS Accession No. ML17277B007)

15. Letter (L-MT-17-055) from NSPM to the NRC, "Monticello Nuclear Generating Plant:

Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Revision 4, Appendix H, H.4.4, Path 4," dated July 26, 2017 (ADAMS Accession No. ML17208A015)

16. NEI Topical Report NEI 12-06, "Diverse and Flexible Coping Strategies (FLEX)

Implementation Guide," Revision 4, dated December 2016 (ADAMS Accession No. ML16354B421)

17. NRC Interim Staff Guidance JLD-ISG-2012-01, "Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," Revision 2, dated February 2017 (ADAMS Accession No. ML17005A188)

18. NRC Generic Issue 199 (GI-199) "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, Safety/Risk Assessment," dated August 2010 (ADAMS Accession No. ML100270639)

L-MT-20-036 NSPM Page 7 of 8

19. Letter from EPRI to NEI, "Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates," dated March 11, 2014 (ADAMS Accession No. ML14080A589)

20. NSPM PRA Document PRA-MT-L2, "Level 2 Accident Sequence Notebook," Revision 5.0, dated March 2019

21. NSPM PRA Document PRA-CALC-18-005, "MNGP Screening of External Hazards for 50.69," Revision 2, dated February 2020

22. NRC Generic Letter GL 89-13, "Service Water System Problems Affecting Safety-Related Equipment (Generic Letter 89-13)," dated July 18, 1989 (ADAMS Accession No. ML031150348)

23. NSPM Monticello Engineering Work Instruction EWI-08.22.01, "Generic Letter 89-013,"

Revision 14

24. Letter (L-MT-16-024) from NSPM to the NRC, "Monticello Nuclear Generating Plant:

Response to Post-Fukushima Near-Term Task Force (NTIF) Recommendation 2.1.

Flooding - Flood Hazard Reevaluation Report," dated May 12, 2016 (ADAMS Accession No. ML16145A179)

25. NRC Regulatory Guide 1.78, "Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," Revision 1, dated December 2001 (ADAMS Accession No. ML013100014)

26. NRC NUREG-0570, "Toxic Vapor Concentrations in the Control Room Following a Postulated Accidental Release," dated June 1979 (ADAMS Accession No. ML063480551)

27. NRC NUREG/CR-2300, Volume 2, "PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants," January 1983 (ADAMS Accession No. ML063560440)

28. NRC Handbook, "Risk Assessment of Operational Events Handbook, Volume 2 -

External Events, Internal Fires - Internal Flooding - Seismic - Other External Events -

Frequencies of Seismically-Induced LOOP Events," Revision 1.02, dated November 2017 (ADAMS Accession No. ML17349A301)

29. Letter (L-MT-14-093) from NSPM to the NRC, "Monticello Nuclear Generating Plant:

Expedited Seismic Evaluation Process (ESEP) - Augmented Approach to Post-Fukushima Near-Term Task Force (NTTF) 2.1," dated December 23, 2014 (ADAMS Accession No. ML14357A280)

L-MT-20-036 NSPM Page 8 of 8

30. EPRI 3002000704, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," Electric Power Research Institute, Final Report, May 2013

31. Letter from EPRI to NEI, "Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates," dated March 11, 2014 (ADAMS Accession No. ML14080A589)

ATTACHMENT 2 MONTICELLO NUCLEAR GENERATING PLANT Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" UPDATED TABLE E1-1 TS/LCO CONDITIONS TO CORESPONDING PRA FUNCTIONS (37 Pages Follow)

L-MT-20-036 NSPM Page 1 of 37 Revised Table E1-1 Table E1-1 provided with the LAR has been updated to reflect changes due to specific RAI responses in the preceding sections of the Enclosure. Changes from the version included in the LAR have been identified by change bars.

Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.1.7.B One [Standby Liquid Control (SLC)]

subsystem inoperable for reasons other than Condition A.

Two SLC subsystems (Mode 1 & 2)

Yes Provide a backup capability for bringing the reactor from full power to a cold, xenon free shutdown One of two SLC subsystems Same PRA also credits the control rod drive hydraulics system for reactivity control in non-anticipated transient without a SCRAM (ATWS) events.

3.3.1.1.A One or more required channels inoperable.

Intermediate Range Monitors (IRMs)

Function 1.a, eight Neutron Flux - High High channels (two IRM channels per Reactor Protection System (RPS) logic channel)

(Mode 2)

No Reactor Trip Initiation (SCRAM)

One Neutron Flux -

High High channel in each RPS trip system None (Notes 1 and 2)

Function 1.b, eight Inop. channels (two IRM channels per RPS logic channel)

(Mode 2)

No SCRAM One Inop. channel in each RPS trip system None (Notes 1 and 2)

L-MT-20-036 NSPM Page 2 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Average Power Range Monitors Function 2.a, four Neutron Flux - High (Setdown) channels (Mode 2)

No SCRAM Two Neutron Flux -

High (Setdown) channels None (Notes 1, 2, and 3)

Function 2.b, four Simulated Thermal Power - High channels (Mode 1)

No SCRAM Two Simulated Thermal Power -

High channels None (Notes 1, 2, and 3)

Function 2.c, four Neutron Flux - High channels (Mode 1)

No SCRAM Two Neutron Flux -

High channels None (Notes 1, 2, and 3)

Function 2.d, four Inop. channels (Mode 1)

No SCRAM Two Inop channels None (Notes 1, 2, and 3)

Function 2.e, four 2-Out-Of-4 Voter channels (Mode 1 & 2)

No SCRAM One 2-Out-Of-4 Voter channel in each RPS trip system None (Notes 1, 2, and 3)

Function 2.f, four

[Oscillation Power Range Monitor (OPRM)] Upscale channels

( 20% RTP)

No SCRAM Two Oscillation Power Range Monitor Upscale channels None (Notes 1, 2, and 3)

L-MT-20-036 NSPM Page 3 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 2.g, four Extended Flow Window Stability -

High channels (Within EFW boundary defined in COLR)

No SCRAM Two Extended Flow Window Stability -

High channels None (Notes 1, 2, and 3)

Function 3, four Reactor Vessel Steam Dome Pressure - High channels (Mode 1 & 2)

No SCRAM One Reactor Vessel Steam Dome Pressure -

High channel in each RPS trip system None (Notes 1 and 2)

Function 4, four Reactor Vessel Water Level -

Low channels (Mode 1 & 2)

No SCRAM One Reactor Vessel Water Level

- Low channel in each RPS trip system None (Notes 1 and 2)

Function 5, sixteen Main Steam Isolation Valve -

Closure channels (four Main Steam Isolation Valve - Closure channels per RPS logic channel)

(Mode 1; Mode 2 with reactor pressure 600 psig)

No SCRAM One of two Main Steam Isolation Valve - Closure channels in three of four steam lines)

None (Notes 1, 2, and 4)

L-MT-20-036 NSPM Page 4 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 6, four Drywell Pressure - High channels (Mode 1 & 2)

No SCRAM One Drywell Pressure - High channel in each of two trip systems None (Notes 1 and 2)

Scram Discharge Volume Water Level -

High Function 7.a, four Resistance Temperature Detector channels (Mode 1 & 2)

No SCRAM One Resistance Temperature Detector channel in each RPS trip system or one RTD channel in one trip system and one Float Switch channel in the other trip system None (Notes 1 and 2)

Function 7.b, four Float Switch channels (Mode 1 & 2)

No SCRAM One Float Switch channel in each RPS trip system or one Float Switch channel in one trip system and one RTD channel in the other trip system None (Notes 1 and 2)

L-MT-20-036 NSPM Page 5 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 8, eight Turbine Stop Valve -

Closure channels (two Turbine Stop Valve -

Closure channels per RPS logic channel)

(> 40% RTP)

No SCRAM Three Turbine Stop Valve - Closure channels in each of two trip systems None (Notes 1, 2, and 5)

Function 9, four Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure -

Low channels (two instruments per RPS logic channel)

(> 40% RTP)

No SCRAM One Turbine Control Valve Fast

Closure, Acceleration Relay Oil Pressure - Low channel in each RPS trip system None (Notes 1 and 2) 3.3.1.1.B ---NOTE---

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, 2.f or 2.g.

One or more Functions with one or more required channels inoperable in both trip systems.

See 3.3.1.1.A above, with the exception of the following Functions excluded by the Condition NOTE:

Average Power Range Monitors Function 2.a, Neutron Flux - High, (Setdown)

Function 2.b, Simulated Thermal Power - High Function 2.c, Neutron Flux - High Function 2.d, Inop.

Function 2.f, OPRM Upscale Function 2.g, Extended Flow Window Stability - High

L-MT-20-036 NSPM Page 6 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.3.2.2.A One or more feedwater pump and main turbine high water level trip channels inoperable.

Four Reactor Vessel Water Level - High channels (THERMAL POWER 25% RTP)

Yes Trip of Feedwater Pumps and Main Turbine One specific Reactor Vessel Water Level - High channel in each of two trip systems or both channels in a trip system Same (Note 15) 3.3.4.1.A One or more channels inoperable.

Function a, four Reactor Vessel Water Level -

Low Low channels (Mode 1)

Yes Trip both Recirculation Pumps Two Reactor Vessel Water Level

- Low Low channels in either of two trip systems Same Function b, four Reactor Vessel Steam Dome Pressure - High channels (Mode 1)

Yes Trip Both Recirculation Pumps Two Reactor Vessel Steam Dome Pressure -

High channels in either of two trip systems Same

L-MT-20-036 NSPM Page 7 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.3.5.1.B As required by Required Action A.1 and referenced in Table 3.3.5.1-

1.

Core Spray (CS)

System Function 1.a, four Reactor Vessel Water Level - Low Low channels (Mode 1 & 2)

Yes Actuate both CS system divisions and the associated EDG One specific Reactor Vessel Water Level - Low Low channel in each of two actuation systems or both channels in an actuation system for a given CS division Same (Note 15)

Function 1.b, four Drywell Pressure -

High channels (Mode 1 & 2)

Yes Actuate both CS system divisions and the associated EDG One specific Drywell Pressure -

High channel in each of two actuation systems or both channels in an actuation system for a given CS division Same (Note 15)

L-MT-20-036 NSPM Page 8 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Low Pressure Coolant Injection (LPCI) System Function 2.a, four Reactor Vessel Water Level - Low Low channels (Mode 1 & 2)

Yes Actuate both LPCI system divisions One specific Reactor Vessel Water Level - Low Low channel in each of two actuation systems or both channels in an actuation system for a given LPCI division Same (Note 15)

Function 2.b, four Drywell Pressure -

High channels (Mode 1 & 2)

Yes Actuate both LPCI system divisions One specific Drywell Pressure -

High channel in each of two actuation systems or both channels in an actuation system for a given LPCI division Same (Note 15)

L-MT-20-036 NSPM Page 9 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 2.f, sixteen Low Pressure Coolant Injection Pump Start -

Time Delay Relay channels (four relays in each of two logic channels in each LPCI actuation system)

(Mode 1 & 2)

Yes Actuate both LPCI system divisions Two Low Pressure Coolant Injection Pump Start - Time Delay Relays in one logic channel of each of two LPCI actuation systems Same Function 2.h, four Reactor Steam Dome Pressure - Low (Break Detection) channels (Mode 1 & 2)

Not explicitly Actuate both LPCI system divisions One specific Reactor Steam Dome Pressure -

Low (Break Detection) channel in each of two actuation systems or both channels in an actuation system for a given LPCI division Same (Notes 6 and 15)

Function 2.k, two Reactor Steam Dome Pressure - Time Delay Relay (Break Detection) channels (Mode 1 & 2)

Not explicitly Actuate one LPCI system division One Reactor Steam Dome Pressure -

Time Delay Relay (Break Detection) channel to actuate one LPCI division Same (Note 6)

L-MT-20-036 NSPM Page 10 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments HPCI System Function 3.a, four Reactor Vessel Water Level - Low Low channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Yes Actuate HPCI system One specific Reactor Vessel Water Level - Low Low channel in each of two LPCI actuation systems or both channels in an actuation system Same (Note 15)

Function 3.b, four Drywell Pressure -

High channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Yes Actuate HPCI system One Drywell Pressure - High channel in each of two CS actuation systems or both channels in an actuation system Same (Note 15) 3.3.5.1.C As required by Required Action A.1 and referenced in Table 3.3.5.1-

1.

CS System Function 1.c, two Reactor Steam Dome Pressure - Low (Injection Permissive) channels (Mode 1 & 2)

Yes Permit CS System Actuation One Reactor Steam Dome Pressure -

Low (Injection Permissive) channel Same

L-MT-20-036 NSPM Page 11 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 1.d, two Reactor Steam Dome Pressure Permissive

- Low (Pump Permissive) channels (Mode 1 & 2)

Yes Permit CS System Actuation One Reactor Steam Dome Pressure Permissive - Low (Pump Permissive) channel from either CS division Same Function 1.e, two Reactor Steam Dome Pressure Permissive

- Bypass Timer (Pump Permissive) channels (Mode 1 & 2)

Not explicitly Permit CS System Actuation One Reactor Steam Dome Pressure Permissive -

Bypass Timer (Pump Permissive) channel from either CS division Same (Note 7)

Function 1.f, two Core Spray Pump Start -

Time Delay Relay channels (Mode 1 & 2)

Yes Permit CS System Actuation One Core Spray Pump Start - Time Delay Relay channel per pump Same LPCI System Function 2.c, two Reactor Steam Dome Pressure - Low (Injection Permissive) channels (Mode 1 & 2)

Yes Permit actuation of both LPCI divisions One Reactor Steam Dome Pressure -

Low (Injection Permissive) channel Same

L-MT-20-036 NSPM Page 12 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 2.d, two Reactor Steam Dome Pressure Permissive

- Low (Pump Permissive) channels (Mode 1 & 2)

Yes Permit actuation of both LPCI divisions One Reactor Steam Dome Pressure Permissive - Low (Pump Permissive) channel Same Function 2.e, two Reactor Steam Dome Pressure Permissive

- Bypass Timer (Pump Permissive) channels (Mode 1 & 2)

Not explicitly Permit actuation of one LPCI division One Reactor Steam Dome Pressure Permissive -

Bypass Timer (Pump Permissive) channel Same (Note 7)

Function 2.i, eight Recirculation Pump Differential Pressure -

High (Break Detection) channels (Mode 1 & 2)

Not explicitly Actuate either LPCI pump in each LPCI division One of two channels of Recirculation Pump Differential Pressure - High (Break Detection) from each LPCI division Same (Note 6)

L-MT-20-036 NSPM Page 13 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 2.j, four Recirculation Riser Differential Pressure -

High (Break Detection) channels (Mode 1 & 2)

Not explicitly Actuate both LPCI Divisions One specific Recirculation Riser Differential Pressure - High (Break Detection) channel in each of two actuation systems or both channels in an actuation system Same (Notes 6 and 15)

Function 2.l, two Recirculation Pump Differential Pressure -

Time Delay Relay (Break Detection) channels (Mode 1 & 2)

Not explicitly Actuate one LPCI division One Recirculation Pump Differential Pressure - Time Delay Relay (Break Detection) channel Same (Note 6)

Function 2.m, two Recirculation Riser Differential Pressure -

Time Delay Relay (Break Detection) channels (Mode 1 & 2)

Not explicitly Actuate/de-actuate both LPCI divisions One Recirculation Riser Differential Pressure - Time Delay Relay (Break Detection) channel Same (Note 6)

L-MT-20-036 NSPM Page 14 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.3.5.1.D As required by Required Action A.1 and referenced in Table 3.3.5.1-

1.

HPCI System Function 3.d, two Condensate Storage Tank Level - Low channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Yes Change HPCI suction path One Condensate Storage Tank Level

- Low channel Same Function 3.e, two Suppression Pool Water Level - High channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Yes Change HPCI suction path One Suppression Pool Water Level -

High channel Same 3.3.5.1.E As required by Required Action A.1 and referenced in Table 3.3.5.1-

1.

LPCI System Function 2.g, four Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) channels (Mode 1 & 2)

Not explicitly Delay bypass flow for one LPCI pump on pump startup One Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) channel per LPCI pump Same (Note 8)

L-MT-20-036 NSPM Page 15 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.3.5.1.F As required by Required Action A.1 and referenced in Table 3.3.5.1-

1.

Automatic Depressurization System (ADS) Trip Systems A and B Functions 4.a and 5.a, four Reactor Vessel Water Level - Low Low channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Not explicitly Actuate all ADS valves Two Reactor Vessel Water Level

- Low Low channels in either of two ADS actuation systems Same (Note 9) 3.3.5.1.G As required by Required Action A.1 and referenced in Table 3.3.5.1-

1.

ADS Trip Systems A and B Functions 4.b and 5.b, two Automatic Depressurization System Initiation Timer channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Not explicitly Actuate all ADS valves One Automatic Depressurization System Initiation Timer channel on each ADS actuation system Same (Note 9)

Functions 4.c and 5.c, four Core Spray Pump Discharge Pressure -

High channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Not explicitly Actuate all ADS valves One Core Spray Pump Discharge Pressure - High channel from one of two CS pumps Same (Note 9)

L-MT-20-036 NSPM Page 16 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Functions 4.d and 5.d, eight Low Pressure Coolant Injection Pump Discharge Pressure - High channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Not explicitly Actuate all ADS valves One Low Pressure Coolant Injection Pump Discharge Pressure - High channel from one of two sets of LPCI pumps Same (Note 9) 3.3.5.2.B As required by Required Action A.1 and referenced in Table 3.3.5.2-

1.

Function 1, four Reactor Vessel Water Level -

Low Low channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Yes Actuate RCIC One specific Reactor Vessel Water Level - Low Low channel in each of two CS actuation systems or both channels in an actuation system Same (Note 15) 3.3.5.2.D As required by Required Action A.1 and referenced in Table 3.3.5.2-

1.

Function 3, two Condensate Storage Tank Level - Low channels (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Yes Change RCIC suction path One Condensate Storage Tank Level

- Low channels Same

L-MT-20-036 NSPM Page 17 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.3.6.1.A One or more required channels inoperable.

Main Steam Line Isolation Function 1.a, four Reactor Vessel Water Level - Low Low channels (Mode 1 & 2)

Not explicitly Main Steam Line Isolation Two Reactor Vessel Water Level

- Low Low channels in either of two trip systems Same (Note 10)

Function 1.b, four Main Steam Line Pressure - Low channels (Mode 1)

Not explicitly Main Steam Line Isolation Two Main Steam Line Pressure -

Low channels in either of two trip systems Same (Note 10)

Function 1.c, sixteen Main Steam Line Flow

- High channels (four instruments per Primary Containment Isolation logic channel)

(Mode 1 & 2)

Not explicitly Main Steam Line Isolation Two Main Steam Line Flow - High channels in either of two trip systems Same (Note 10)

Function 1.d, sixteen Main Steam Line Tunnel Temperature -

High channels (four instruments per Primary Containment Isolation logic channel)

(Mode 1 & 2)

Not explicitly Main Steam Line Isolation Two Main Steam Line Tunnel Temperature -

High channels in either of two trip systems Same (Note 10)

L-MT-20-036 NSPM Page 18 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Primary Containment Isolation Function 2.a, four Reactor Vessel Water Level - Low channels (Mode 1 & 2)

Not explicitly Primary Containment Isolation One Reactor Vessel Water Level

- Low channel in each of two trip systems Same (Note 10)

Function 2.b, four Drywell Pressure -

High channels (Mode 1 & 2)

Not explicitly Primary Containment Isolation One Drywell Pressure - High channel in each of two trip systems Same (Note 10)

HPCI System Isolation Function 3.a, two HPCI Steam Line Flow - High channels (Mode 1 & 2)

Not explicitly HPCI Isolation One HPCI Steam Line Flow - High channel in either of two logic systems causes isolation in two isolation systems Same (Note 10)

Function 3.b, four HPCI Steam Supply Line Pressure - Low channels (Mode 1 & 2)

Not explicitly HPCI Isolation Two specific HPCI Steam Supply Line Pressure - Low channels Same (Note 10)

L-MT-20-036 NSPM Page 19 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 3.c, sixteen HPCI Steam Line Area Temperature -

High (four instruments per HPCI logic channel)

(Mode 1 & 2)

Not explicitly HPCI Isolation Two specific HPCI Steam Line Area Temperature -

High channels in either of two logic channels causes isolation in two isolation systems Same (Note 10)

RCIC System Isolation Function 4.a, two RCIC Steam Line Flow - High channels (Mode 1 & 2)

Not explicitly RCIC Isolation One RCIC Steam Line Flow - High channel in either of two logic systems causes isolation in two isolation systems Same (Note 10)

Function 4.b, four RCIC Steam Supply Line Pressure - Low channels (Mode 1 & 2)

Not explicitly RCIC Isolation Two specific RCIC Steam Supply Line Pressure - Low channels cause isolation in two isolation systems Same (Note 10)

L-MT-20-036 NSPM Page 20 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 4.c, sixteen RCIC Steam Line Area Temperature -

High channels (Mode 1 & 2)

Not explicitly RCIC Isolation Two specific RCIC Steam Line Area Temperature -

High channels in either of two logic channels cause isolation in two isolation systems Same (Note 10)

Reactor Water Cleanup (RWCU) System Isolation Function 5.a, four RWCU Flow - High channels (Mode 1 & 2)

Not explicitly RWCU Isolation One RWCU Flow -

High channel in each of two trip systems Same (Note 10)

Function 5.b, four RWCU Room Temperature - High channels (Mode 1 & 2)

Not explicitly RWCU Isolation One RWCU Room Temperature -

High channel in each of two trip systems Same (Note 10)

Function 5.c, four Drywell Pressure -

High channels (Mode 1 & 2)

Not explicitly RWCU Isolation One Drywell Pressure - High channel in each of two trip systems Same (Note 10)

Function 5.d, two SLC System Initiation channels (Mode 1 & 2)

Not explicitly RWCU Isolation One SLC System Initiation channel in each of two trip systems Same (Note 10)

L-MT-20-036 NSPM Page 21 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 5.e, four Reactor Vessel Water Level - Low Low channels (Mode 1 & 2)

Not explicitly RWCU Isolation One Reactor Vessel Water Level

- Low Low channel in each of two trip systems Same (Note 10)

Shutdown Cooling System Isolation Function 6.a, four Reactor Steam Dome Pressure - High channels (Mode 1 & 2)

Not explicitly Shutdown Cooling System Isolation One Reactor Steam Dome Pressure -

High channel in each of two trip systems Same (Note 10)

Traversing Incore Probe System Isolation Function 7.a, four Reactor Vessel Water Level - Low channels (Mode 1 & 2)

Not explicitly Traversing Incore Probe System Isolation One Reactor Vessel Water Level

- Low channel in each of two trip systems Same (Note 10)

Function 7.b, four Drywell Pressure -

High channels (Mode 1 & 2)

Not explicitly Traversing Incore Probe System Isolation One Drywell Pressure - High channel in each of two trip systems Same (Note 10)

L-MT-20-036 NSPM Page 22 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.3.7.2.A One or more channels inoperable.

Four channels of Main Steam Line Tunnel Radiation - High instrumentation (Mode 1 & 2 with the mechanical vacuum pump in service and any main steam line not isolated)

No Mechanical Vacuum Pump Isolation in a Control Rod Drop Accident One Main Steam Line Tunnel Radiation - High channel in each of two trip systems Note 11 (Note 11) 3.3.8.1.A One or more channels inoperable.

Function 1, eight channels of 4.16 kV Essential Bus Loss of Voltage (four instruments per logic division)

(Mode 1 & 2)

Not explicitly Sense Essential Bus Loss of Voltage and Transfer to EDGs One 4.16 kV Essential Bus Loss of Voltage channel in each of two sets per bus Same (Note 12) 4.16 kV Essential Bus Degraded Voltage Function 2.a, six channels of Bus Undervoltage (three instruments per logic division)

(Mode 1 & 2)

Not explicitly Sense Essential Bus Degraded Voltage and transfer to EDGs Two Bus Undervoltage channels per bus Same (Note 12)

L-MT-20-036 NSPM Page 23 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments Function 2.b, six channels of Time Delay (three instruments per logic channel)

(Mode 1 & 2)

Not explicitly Sense Essential Bus Degraded Voltage and transfer to EDGs Two Time Delay channels per bus Same (Note 12) 3.4.3.A One or two required

[Safety/Relief Valves (S/RVs)]

inoperable.

Seven S/RVs (Mode 1 & 2)

Yes Reactor Pressure Vessel Overpressure Protection (RPV)

Five S/RVs Non-ATWS:

Two S/RVs ATWS:

Three S/RVs 3.5.1.B One [Low Pressure Coolant Injection (LPCI)]

subsystem inoperable for reasons other than Condition A.

OR One Core Spray subsystem inoperable.

Two LPCI subsystems and two Core Spray subsystems (Mode 1; Mode 2, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig)

Yes Low pressure injection into the RPV 1 CS pump and 2 LPCI pumps and 3 ADS valves One LPCI subsystem with one of two pumps injecting into the reactor vessel.

OR One CS subsystem injecting into the reactor vessel.

L-MT-20-036 NSPM Page 24 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.5.1.C One LPCI pump in both LPCI subsystems inoperable.

Four LPCI pumps (Mode 1; Mode 2, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig)

Yes Low pressure injection into the RPV 1 CS pump and 2 LPCI pumps and 3 ADS valves Same as PRA Success Criteria for TS 3.5.1.B 3.5.1.D Two LPCI subsystems inoperable for reasons other than Condition C or G.

Two LPCI subsystems (Mode 1; Mode 2, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig)

Yes Low pressure injection into the RPV 2 CS pumps and HPCI and 3 ADS valves Same as PRA Success Criteria for TS 3.5.1.B

L-MT-20-036 NSPM Page 25 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.5.1.E One Core Spray subsystem inoperable.

AND One LPCI subsystem inoperable.

OR One or two LPCI pump(s) inoperable.

Two CS subsystems and two LPCI subsystems including four LPCI pumps (two LPCI pumps per LPCI subsystem)

(Mode 1; Mode 2, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig)

Yes Low pressure injection into the RPV 1 CS pump and 2 LPCI pumps and 3 ADS valves Same as PRA Success Criteria for TS 3.5.1.B

L-MT-20-036 NSPM Page 26 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.5.1.I HPCI System inoperable.

One HPCI System (Mode 1; Mode 2, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig)

Yes High Pressure Injection into the RPV 3 ADS valves and 4 LPCI pumps and 2 CS pumps Feedwater subsystems OR One RCIC System OR CRDH System OR ADS in conjunction with one of four LCPI pumps or one of two CS pumps Based on thermal hydraulic calculations, feedwater, RCIC, or CRDH can provide adequate makeup for high pressure injection.

CRDH and RCIC are not adequate for a LOCA. For CRDH to be successful, operator actions are required to enhance flow.

One of the low pressure injection /

spray pumps (LPCI or CS) is adequate when depressurized.

L-MT-20-036 NSPM Page 27 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.5.1.J HPCI System inoperable.

AND Condition A, B, or C entered.

One HPCI System, two LPCI subsystems (containing four total LPCI pumps), and two Core Spray subsystems (Mode 1; Mode 2, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig)

Yes RPV Inventory Control and Decay Heat Removal (One of one HPCI System inoperable)

Three LPCI pumps (Condition A); or one LPCI subsystem or one Core Spray subsystem (Condition B); or one LPCI pump in each LPCI subsystem (Condition C), in conjunction with ADS Feedwater subsystems OR One RCIC System OR CRDH System OR ADS in conjunction with one of four LCPI pumps or one of two CS pumps Based on thermal hydraulic calculations, feedwater, RCIC, or CRDH can provide adequate makeup for high pressure injection.

CRDH and RCIC are not adequate for a LOCA. For CRDH to be successful, operator actions are required to enhance flow.

One of the low pressure injection/spray pumps (LPCI or CS) is adequate when depressurized

L-MT-20-036 NSPM Page 28 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.5.1.K One ADS valve inoperable.

Three ADS valves (Mode 1; Mode 2, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig)

Yes RPV Rapid Depressurization if HPCI is not available Three ADS valves OR HPCI Same See the response to RAI 23 3.5.3.A RCIC System inoperable.

One RCIC System (Mode 1; Mode 2 with reactor steam dome pressure > 150 psig)

Yes Supply High Pressure Makeup Water to the RPV (One of one RCIC System inoperable)

One HPCI System Feedwater subsystems OR One RCIC System OR CRDH System Based on thermal hydraulic calculations, feedwater, HPCI, or CRDH can provide adequate makeup for high pressure injection. CRDH and HPCI are not adequate for certain LOCA cases. For CRDH to be successful operator actions are required to enhance flow.

L-MT-20-036 NSPM Page 29 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.6.1.2.C Primary containment air lock inoperable for reasons other than Condition A or B.

One primary containment air lock (Mode 1 & 2)

Not explicitly Isolate Primary Containment during Personnel Entry and Exit One of two primary containment air lock doors closed with acceptable containment leakage per LCO 3.6.1.1 Same (Note 13) 3.6.1.3.A ---NOTE---

Only applicable to penetration flow paths with two PCIVs.

One or more penetration flow paths with one PCIV inoperable for reasons other than Condition D or E.

Primary Containment Isolation Valves (Mode 1 & 2)

Yes Limit Fission Product Release during and following Postulated Design Basis Accidents (DBAs)

One of two Primary Containment Isolation Valves per penetration One isolation valve in each modeled penetration.

Lines less than 2 inches in diameter are screened from LERF and thus are not modeled and have no quantitative impact on LERF

L-MT-20-036 NSPM Page 30 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.6.1.3.D One or more penetration flow paths with one or more 18 inch primary containment purge and vent valves not within purge and vent valve leakage limits.

Seven 18 inch Primary Containment Purge and Vent Valves (Mode 1 & 2)

Yes Limit Fission Product Release during and following Postulated DBAs One or more penetration flow paths with one 18 inch primary containment purge or vent valve closed such that gross breach of primary containment does not exist One isolation valve in each modeled penetration.

Lines less than 2 inches in diameter are screened from LERF and thus are not modeled and have no quantitative impact on LERF 3.6.1.6.C One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

Four vacuum breakers, located two in series between two parallel lines (Mode 1 & 2)

Yes Relieve vacuum when primary containment depressurizes below reactor building pressure.

One line with two vacuum breakers OPERABLE for opening Same 3.6.1.7.A One required suppression chamber-to-drywell vacuum breaker inoperable for opening.

Eight suppression chamber-to-drywell vacuum breakers (Mode 1 & 2)

Yes Relieve vacuum in the drywell Six suppression chamber-to-drywell vacuum breakers OPERABLE for opening One suppression chamber-to-drywell vacuum breaker OPERABLE for opening There is an implementation item to reassess the PRA Success Criteria for the suppression chamber-to-drywell vacuum breakers.

L-MT-20-036 NSPM Page 31 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.6.1.8.A One RHR drywell spray subsystem inoperable.

Two RHR drywell spray subsystems (each containing two pumps)

(Mode 1 & 2)

Yes Lower Drywell Pressure and Temperature following a DBA One RHR drywell spray subsystem Same 3.6.2.3.A One RHR suppression pool cooling subsystem inoperable.

Two RHR suppression pool cooling subsystems (Mode 1 & 2)

Yes Removes Heat from the Suppression Pool following a DBA One RHR suppression pool cooling subsystem Same 3.7.1.A One [Residual Heat Removal Service Water (RHRSW)]

subsystem inoperable.

Two RHRSW subsystems (Mode 1 & 2)

Yes Provide cooling water for the RHR System heat exchangers, required for a safe shutdown following a DBA or transient One RHRSW subsystem Same 3.7.2.A One

[Emergency Service Water (ESW)]

subsystem inoperable.

Two ESW subsystems (Mode 1 & 2)

No Provide cooling water for the removal of heat from equipment required for a safe reactor shutdown following a DBA or transient One ESW subsystem None Thermal analysis has been performed to show that emergency service water (EFT-ESW) is not required to prevent CDF or LERF.

L-MT-20-036 NSPM Page 32 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.8.1.A One required offsite circuit inoperable.

Three qualified circuits consisting of all breakers, transformers, switches, interrupting devices, cabling, and controls to transmit power from the offsite transmission network to the Class 1E 4.16 kV essential bus (Mode 1 & 2)

Yes Provide power from offsite transmission network to onsite Class 1E 4.16 kV essential bus One qualified circuit to the grid for a Class 1E 4.16 kV essential bus Same when offsite power available 3.8.1.B One

[Emergency Diesel Generator (EDG)]

inoperable.

Two EDGs (Mode 1 & 2)

Yes Provide power to onsite Class 1E 4.16 kV essential bus when offsite power is lost One EDG Same when offsite power not available 3.8.1.C Two required offsite circuits inoperable.

Three qualified circuits consisting of all breakers, transformers, switches, interrupting devices, cabling, and controls to transmit power from the offsite transmission network to the Class 1E 4.16 kV essential bus (Mode 1 & 2)

Yes Provide power from offsite transmission network to onsite Class 1E 4.16 kV essential bus One qualified circuit to the grid for a Class 1E 4.16 kV essential bus Same when offsite power available

L-MT-20-036 NSPM Page 33 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.8.1.D One required offsite circuit inoperable.

AND One EDG inoperable.

Three qualified circuits consisting of all breakers, transformers, switches, interrupting devices, cabling, and controls to transmit power from the offsite transmission network to the Class 1E 4.16 kV essential bus and two EDGs (Mode 1 & 2)

Yes Provide power from offsite transmission network to onsite Class 1E 4.16 kV essential bus and provide power to onsite Class 1E 4.16 kV essential bus when offsite power is lost One qualified circuit to the grid and one EDG for a Class 1E 4.16 kV essential bus Offsite Power Available:

One offsite circuit Offsite Power Not Available:

One EDG for one essential bus 3.8.4.A One or more required battery chargers on Division 1 or Division 2 inoperable.

Six chargers in the 250 VDC electrical power subsystems; two normally inservice 125 VDC chargers and one spare 125 VDC charger per division.

Three chargers in the 125 VDC subsystems; one 125 VDC battery charger in each division plus one spare 125 VDC that can be used on either division.

(Mode 1 & 2)

Yes Ensure availability of required DC power to shut down the reactor and maintain it in a safe condition Two battery chargers for each 250 VDC electrical power subsystem One battery charger for each 125 VDC electrical power subsystem Same (Note 14)

L-MT-20-036 NSPM Page 34 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments 3.8.4.B One Division 1 or Division 2 DC electrical power subsystem inoperable for reasons other than Condition A.

Two 250 VDC and two 125 VDC electrical power subsystems (Mode 1 & 2)

Yes Ensure availability of required DC power to shut down the reactor and maintain it in a safe condition One of two electrical power subsystems Same 3.8.7.A One or more AC electrical power distribution subsystems inoperable.

Two AC electrical power distribution subsystems each consisting of one 4.16 kV essential bus, 480 VAC load centers, and transformers (Mode 1 & 2)

Yes Ensure availability of required AC power to shut down the reactor and maintain it in a safe condition One AC electrical power distribution subsystem capable of supporting minimum safety functions Same 3.8.7.B One or more DC electrical power distribution subsystems inoperable.

Two 125/250 VDC electrical power distribution systems each consisting of a 125/250 VDC distribution cabinet and 125 VDC distribution panel (Mode 1 & 2)

Yes Ensure availability of required DC power to shut down the reactor and maintain it in a safe condition One DC electrical power distribution subsystem capable of supporting minimum safety functions Same Table E1-1 Notes:

1.

The RPS is comprised of two independent trip systems (A and B) with three logic channels in each trip system (logic channels A1, A2, and A3, B1, B2, and B3) as described in Reference 3. The automatic trip logics of trip system A are logic channels A1 and A2; the manual trip logic of trip system A is logic channel A3. Similarly, the trip logics for trip system B are logic channels B1, B2, and B3.

L-MT-20-036 NSPM Page 35 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments The outputs of the automatic logic channels in a trip system are combined in a one-out-of-two logic so that either channel can trip the associated trip system. The tripping of both trip systems will produce a reactor scram. This logic arrangement is referred to as a one-out-of-two taken twice logic. The outputs of the manual logic channels in a trip system are combined in a one-out-of-one logic. The tripping of both manual logic channels will produce a scram. Each trip system can be reset by use of a reset switch. If a full scram occurs (both trip systems trip), a relay prevents reset of the trip systems for a short time delay after the full scram signal is received.

The short time delay on reset ensures that the scram function will be completed.

2.

The RPS is not modeled explicitly in PRA. This RPS is currently modeled by mechanical and electrical failures as point estimates.

These values were driven from quantifying the NUREG/CR 5500 Volume 3 (Reference 7) fault tree. For sample RICT calculations, RPS electrical failure probability was increased by quantification results from failing the modeled channels (functions 3 and 4) in the NUREG/CR 5500 Volume 3 fault tree. These two signals, along with others, are appropriate for several plant upset conditions, such as main steam line isolation valve or MSIV closure, loss of feedwater, and various losses of electrical loads. The PRA model will be updated to include the applicable RPS instrumentation prior to exercising the RICT program for this TS. The RPS modeling will replace the point estimate failure with detailed modeling in accordance with Regulatory Guide 1.200, Revision 2 (Reference 8).

3.

The APRM System is divided into four APRM channels and four 2-out-of-4 voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each; with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one un-bypassed APRM will result in a "half-trip" in all four of the voter channels, but no trip inputs to either RPS trip system. Because APRM trip Functions 2.a, 2.b, 2.c, 2.f, and 2.g are implemented in the same hardware, these trip Functions are combined with APRM lnop trip Function 2.d. Any Function 2.a, 2.b, 2.c, 2.d, or 2.g trip from any two un-bypassed APRM channels will result in a full trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip system logic channel (A1, A2, B1 and B2). Similarly, any Function 2.d, 2.f, or 2.g trip from any two un-bypassed APRM channels will result in a full trip from each of the four voter channels.

4.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve - Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve

- Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.

L-MT-20-036 NSPM Page 36 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments

5.

Turbine Stop Valve - Closure signals are initiated from position switches located on each of the four TSVs. One position switch and two independent contacts are associated with each stop valve. One of the two contacts provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram.

6.

LPCI loop select logic failure was used as a conservative surrogate. This basic event represents the probability that LPCI loop select fails in such a way that it causes LPCI injection to occur on the loop where the line break occurred.

7.

The ECCS auto start signals failed to actuate was used as a surrogate. These events are representative for reactor steam dome pressure permissive bypass timer relays.

8.

Failure of the minimum flow valve to open was used as a conservative surrogate for the RICT calculation. This surrogate is conservative as the valves failure to open makes the RHR pump unavailable.

9.

Failure of all ADS valves to open was used as a conservative surrogate for the RICT calculation. The model will be updated to include these SSCs prior to exercising the RICT program for this TS. The PRA Success Criteria will match the Design Success Criteria.

10.

Failure of primary containment isolation valves to close was used as a conservative surrogate as appropriate for each function evaluated in a RICT calculation.

11.

SSCs are not modeled. The PRA model will be updated to include the Mechanical Vacuum Pump system and Isolation instrumentation prior to exercising the RICT Program for this TS. The MVPI instrumentation system will be implemented in PRA to meet the ASME standard. Failure of the steam jet air ejectors was used as a conservative surrogate representation of the risk for the Table E1-2 sample calculations. This is a conservative surrogate, since the failure of steam jet air ejectors causes loss of condenser vacuum.

12.

Failure of loss of power relays to shift to the de-energized position was used as a conservative surrogate for the RICT calculation.

The failure to shift to the de-energized position was chosen for this TS, since the instrumentation function is to de-energize the relays on loss of power.

L-MT-20-036 NSPM Page 37 of 37 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions MNGP TS MNGP TS Description SSCs Covered by TS LCO Condition and Applicable Mode(s)

Modeled in PRA Function Covered by TS LCO Condition Design Success Criteria PRA Success Criteria Comments

13.

Since the containment airlock is not modeled, there are no explicit PRA Success Criteria. However, failure of the containment airlock function is modeled as a pre-existing leak probability in the PRA, which is a conservative surrogate in the PRA. This is a conservative surrogate, since the surrogate failure is equivalent to a break in containment in which an inoperable airlock may not be considered a break in the containment. Compliance with the remaining portions of LCO Condition 3.6.1.2.C ensure that at least one door is maintained closed in the air lock. Thus, the function is still maintained.

14.

While the spare Division 2 125 VDC battery charger can be used to supply either the Division 1 or Division 2 125 VDC subsystem, it can be used to meet the LCO requirements only for the Division 2 125 VDC subsystem. If it is supplying the Division 1 125 VDC subsystem, the Division 1 125 VDC subsystem is inoperable, but the function is maintained available.

15.

The logic is arranged such that it takes "specific" combinations of channels in each trip/actuation system to cause the logic to be made up. For example, there are two channel combinations where "one of two taken twice" will cause the logic to be made up and two channel combinations where one of two taken twice will not cause the logic to be made up. In addition, both channels in either trip system in "two of two taken once" will also cause the logic to be made up.

ATTACHMENT 3 MONTICELLO NUCLEAR GENERATING PLANT Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Revised Cross Reference of TSTF-505 and MNGP Technical Specifications (Provided for Information Only)

(15 Pages Follow)

L-MT-20-036 NSPM Page 1 of 15 Revised Table A4-1 Table A4-1 provided with the LAR has been updated to reflect changes due to specific RAI responses in the preceding sections of the Enclosure. Changes from the version included in the LAR have been identified by change bars.

Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments Completion Times 1.3 1.3 Example 1.3-8

[NEW TS]

1.3-8

[NEW TS]

1.3-8 No The MNGP TS do not currently contain this example. Example to be added to the TS to be consistent with TSTF-505. This is a new definition only (i.e., there is no RICT directly applicable to the TS).

Standby Liquid Control (SLC) System 3.1.7 3.1.7 One SLC subsystem inoperable [for reasons other than Condition A].

3.1.7.B 3.1.7.B Yes TSTF-505 changes are incorporated.

L-MT-20-036 NSPM Page 2 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments Reactor Protection System (RPS)

Instrumentation 3.3.1.1 3.3.1.1 One or more required channels inoperable.

3.3.1.1.A.1 3.3.1.1.A.2 3.3.1.1.A.1 3.3.1.1.A.2 Yes Yes MNGP TS contains a NOTE which is not contained in NUREG-1433 which limits Required Action A.2 from being applied to MNGP TS 3.3.1.1 Functions 2.a, 2.b, 2.c, 2.d, 2.f, and 2.g.

The MNGP Function 2, "Average Power Range Monitors," design differs from that assumed in NUREG-1433. See Enclosure 1 of this submittal for further detail.

Under certain circumstances, with more than one required channel inoperable, a loss of function can occur. Condition C would be entered with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time and no RICT. Therefore, TSTF-505 changes are incorporated.

One or more Functions with one or more required channels inoperable in both trip systems.

3.3.1.1.B.1 3.3.1.1.B.2 3.3.1.1.B.1 3.3.1.1.B.2 Yes Yes MNGP TS contains a NOTE which is not contained in NUREG-1433 which limits Condition B from being applied to MNGP TS 3.3.1.1 Functions 2.a, 2.b, 2.c, 2.d, 2.f, and 2.g. The MNGP Function 2, "Average Power Range Monitors," design differs from that assumed in NUREG-1433. See Enclosure 1 of this submittal for further detail.

Under certain circumstances, with more than one required channel inoperable, a loss of function can occur. Condition C would be entered with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time and no RICT. Therefore, TSTF-505 changes are incorporated.

L-MT-20-036 NSPM Page 3 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments Source Range Monitor (SRM)

Instrumentation 3.3.1.2 3.3.1.2 One or more required SRMs inoperable in MODE 2 with intermediate range monitors (IRMs) on Range 2 or below.

3.3.1.2.A.1 3.3.1.2.A.1 No NSPM has determined there is negligible benefit to applying a RICT to this Condition in the MNGP TS. Therefore, TSTF-505 changes are not incorporated.

Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 3.3.2.2 One feedwater and main turbine high water level trip channel inoperable.

3.3.2.2.A.1 3.3.2.2.A.1 Yes Under certain circumstances, with more than one required channel inoperable, a loss of function can occur. Condition B would be entered with a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time and no RICT. Therefore, TSTF-505 changes are incorporated.

Two or more feedwater and main turbine high water level trip channels inoperable.

3.3.2.2.B.1 3.3.2.2.B.1 No The wording of the MNGP TS differs from that in NUREG-1433. The MNGP TS wording is for level trip capability not maintained, which represents loss of function. Therefore, TSTF-505 changes are not incorporated.

End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation 3.3.4.1 One or more required channels inoperable.

3.3.4.1.A.1 3.3.4.1.A.2 No No The MNGP TS do not contain this TS. Therefore, TSTF-505 changes are not incorporated.

L-MT-20-036 NSPM Page 4 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)

Instrumentation 3.3.4.2 3.3.4.1 One or more channels inoperable.

3.3.4.2.A.1 3.3.4.2.A.2 3.3.4.1.A.1 3.3.4.1.A.2 Yes Yes Under certain circumstances, with more than one required channel inoperable, a loss of function can occur. Condition B would be entered with a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time and no RICT.

Therefore, TSTF-505 changes are incorporated.

Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.1 3.3.5.1 As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

3.3.5.1.B.3 3.3.5.1.B.3 Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

3.3.5.1.C.2 3.3.5.1.C.2 Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

3.3.5.1.D.2.1 3.3.5.1.D.2.1 Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

L-MT-20-036 NSPM Page 5 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

3.3.5.1.E.2 3.3.5.1.E.2 Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

3.3.5.1.F.2 3.3.5.1.F.2 Yes The RICT insert format is modified from TSTF-505, Revision 2, to align with MNGP TS 1.2, "Logical Connectors," direction to only use first level logic for Completion Times.

Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

3.3.5.1.G.2 3.3.5.1.G.2 Yes The RICT insert format is modified from TSTF-505, Revision 2, to align with MNGP TS 1.2, "Logical Connectors," direction to only use first level logic for Completion Times.

Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

L-MT-20-036 NSPM Page 6 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments Reactor Core Isolation Cooling (RCIC)

System Instrumentation 3.3.5.2 3.3.5.2 As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

3.3.5.2.B.2 3.3.5.2.B.2 Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

As required by Required Action A.1 and referenced in Table 3.3.5.2-1.

3.3.5.2.D.2.1 3.3.5.2.D.2.1 Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

Primary Containment Isolation Instrumentation 3.3.6.1 3.3.6.1 One or more required channels inoperable.

3.3.6.1.A.1 3.3.6.1.A.1 Yes The RICT insert format is modified from TSTF-505, Revision 2, to align with MNGP TS 1.2, "Logical Connectors," direction to only use first level logic for Completion Times.

Under certain circumstances, with more than one required channel inoperable, a loss of function can occur. Condition B would be entered with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time and no RICT. TSTF-505 changes are incorporated.

L-MT-20-036 NSPM Page 7 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments Low-Low Set (LLS) Instrumentation 3.3.6.3 3.3.6.3 One or more LLS valves with one or more channels inoperable.

3.3.6.3.A.1 3.3.6.3.A.2 No Consistent with a RICT not being applied to MNGP TS 3.6.1.5, "Low-Low Set (LLS) Valves,"

NSPM does not propose to add a RICT to the LLS valve instrumentation. Therefore, TSTF-505 changes are not incorporated.

Mechanical Vacuum Pump Isolation Instrumentation 3.3.7.2

[MNGP TS Condition]

One or more channels inoperable.

3.3.7.2.A.1 3.3.7.2.A.2 Yes Yes This LCO is MNGP plant-specific and therefore not in NUREG-1433 or TSTF-505 Revision 2.

Under certain circumstances, with more than one required channel inoperable, a loss of function can occur. Condition B would be entered with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time and no RICT.

Therefore, changes consistent with TSTF-505 are incorporated.

Loss of Power (LOP) Instrumentation 3.3.8.1 3.3.8.1 One or more channels inoperable.

3.3.8.1.A.1 3.3.8.1.A.1 Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur.

Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function has occurred. TSTF-505 changes are incorporated.

Safety/Relief Valves (S/RVs) 3.4.3 3.4.3

[ One [or two] [required] S/RV[s]

inoperable.

3.4.3.A.1 3.4.3.A.1 Yes TSTF-505 changes are incorporated.

L-MT-20-036 NSPM Page 8 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments ECCS - Operating 3.5.1 3.5.1 One low pressure ECCS injection/spray subsystem inoperable.

OR One LPCI pump in both LPCI subsystems inoperable.

3.5.1.A.1 No The MNGP TS do not contain this TS. Therefore, TSTF-505 changes are not incorporated.

[MNGP TS Condition]

One LPCI subsystem inoperable for reasons other than Condition A.

OR One Core Spray subsystem inoperable.

3.5.1.B.1 Yes This is a MNGP plant-specific Condition. MNGP TS Condition B does not involve loss of function as the remaining OPERABLE ECCS subsystems provide adequate core cooling during a loss of coolant accident (LOCA). Therefore, changes consistent with TSTF-505 are incorporated.

[MNGP TS Condition]

One LPCI pump in both LPCI subsystems inoperable.

3.5.1.C.1 Yes This is a MNGP plant-specific Condition. MNGP TS Condition C does not involve loss of function as the remaining OPERABLE ECCS subsystems provide adequate core cooling during a LOCA.

Therefore, changes consistent with TSTF-505 are incorporated.

[MNGP TS Condition]

Two LPCI subsystems inoperable for reasons other than Condition C or G.

3.5.1.D.1 Yes This is a MNGP plant-specific Condition. MNGP TS Condition D does not involve a loss of function as the remaining OPERABLE ECCS subsystems provide adequate core cooling during a LOCA. Therefore, changes consistent with TSTF-505 are incorporated.

L-MT-20-036 NSPM Page 9 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments

[MNGP TS Condition]

One Core Spray subsystem inoperable.

AND One LPCI subsystem inoperable.

OR One or two LPCI pump(s) inoperable.

3.5.1.E.1 3.5.1.E.2 3.5.1.E.3 Yes Yes Yes This is a MNGP plant-specific Condition. MNGP Condition E does not involve a loss of function as the remaining ECCS subsystems provide adequate core cooling during a LOCA. Therefore, changes consistent with TSTF-505 are incorporated.

HPCI System inoperable.

3.5.1.C.2 3.5.1.I.2 Yes TSTF-505 changes are incorporated.

HPCI System inoperable.

AND Condition A entered.

3.5.1.D.1 3.5.1.D.2 3.5.1.J.1 3.5.1.J.2 Yes Yes The wording of MNGP TS Condition J differs from NUREG-1433 Condition D in that it applies to "HPCI System inoperable" and "Condition A, B, or C entered". Changes consistent with TSTF-505 are incorporated.

One ADS valve inoperable.

3.5.1.E.1 3.5.1.K.1 Yes TSTF-505 changes are incorporated.

One ADS valve inoperable.

AND Condition A entered.

3.5.1.F.1 3.5.1.F.2 No No The MNGP TS do not contain this Condition.

Therefore, TSTF-505 changes are not incorporated.

[Reactor Core Isolation Cooling (RCIC)] System 3.5.3 3.5.3 RCIC System inoperable.

3.5.3.A.2 3.5.3.A.2 Yes TSTF-505 changes are incorporated.

Primary Containment Air Lock 3.6.1.2 3.6.1.2

L-MT-20-036 NSPM Page 10 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments Primary containment air lock inoperable for reasons other than Condition A or B.

3.6.1.2.C.3 3.6.1.2.C.3 Yes TSTF-505 changes are incorporated.

Primary Containment Isolation Valves (PCIVs) 3.6.1.3 3.6.1.3

---

NOTE-----------------------

Only applicable to penetration flow paths with two [or more] PCIVs.

One or more penetration flow paths with one PCIV inoperable [for reasons other than Condition[s] D [and E)).

3.6.1.3.A.1 3.6.1.3.A.1 Yes TSTF-505 changes are incorporated.

---

NOTE------------------------

Only applicable to penetration flow paths with only one PCIV.

One or more penetration flow paths with one PCIV inoperable [for reasons other than Condition[s] D [and E)).

3.6.1.3 C.2 3.6.1.3 C.2 No During the MNGP ITS conversion, a new Completion Time was added to TS 3.6.1.3 Required Action C.2 since these are single valve penetrations that can be isolated by a device inside primary containment. This Completion Time is identical as that for ISTS 3.6.1.3 Required Action A.2.

[ One or more penetration flow paths with one or more containment purge valves not within purge valve leakage limits.

3.6.1.3.E.1 3.6.1.3.D.1 Yes Wording of MNGP TS differs from TSTF-505 (i.e.,

MNGP TS uses "18 inch primary containment purge and vent valves" and "purge and vent valve"). TSTF-505 changes are incorporated.

L-MT-20-036 NSPM Page 11 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments One or more penetration flow paths with one or more containment purge valves not within purge valve leakage limits.

3.6.1.3 E.2 3.6.1.3 D.2 No The second Completion Time in NUREG-1433 TS 3.6.1.3 Required Action E.2 has been deleted because there are no isolation devices inside the containment associated with the primary containment vent and purge valve penetration flow paths.

One or more penetration flow paths with one or more containment purge valves not within purge valve leakage limits.

3.6.1.3 E.3 No This Required Action was not brought over during ITS conversion as this allowance was "consistent with requirements in place" at the time of the MNGP ITS conversion. Therefore, the TSTF-505 markup for Required Action E.3 is not applicable.

Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 3.6.1.6 One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

3.6.1.7.C.1 3.6.1.6.C.1 Yes TSTF-505 changes are incorporated.

Two [or more] lines with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

3.6.1.7.D.1 3.6.1.6.D.1 No The MNGP design only includes two lines.

Consequently, the MNGP TS Condition does not contain the "or more" wording from NUREG-1433. Therefore, for MNGP, two lines with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening constitutes a loss of function. Therefore, TSTF-505 changes are not incorporated.

Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 3.6.1.7

L-MT-20-036 NSPM Page 12 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments One required suppression chamber-to-drywell vacuum breaker inoperable for opening.

3.6.1.8.A.1 3.6.1.7.A.1 Yes TSTF-505 changes are incorporated.

Residual Heat Removal (RHR) Drywell Spray 3.6.1.8

[MNGP TS Condition]

One RHR drywell spray subsystem inoperable.

3.6.1.8.A.1 Yes This is a MNGP plant-specific Condition. The MNGP safety analyses take credit for the operation of the drywell spray function, not the suppression pool spray function. MNGP TS Condition A does not involve a loss of function.

Therefore, TSTF-505 changes are incorporated.

Residual Heat Removal (RHR)

Suppression Pool Cooling 3.6.2.3 3.6.2.3 One RHR suppression pool cooling subsystem inoperable.

3.6.2.3.A.1 3.6.2.3.A.1 Yes TSTF-505 changes are incorporated.

Residual Heat Removal (RHR)

Suppression Pool Spray 3.6.2.4 One RHR suppression pool spray subsystem inoperable.

3.6.2.4.A.1 No See MNGP TS 3.6.1.8 - the MNGP TS do not contain this TS. Therefore, TSTF-505 changes are not incorporated.

[Drywell Cooling System Fans]

3.6.3.1 Two [required] [drywell cooling system fans] inoperable.

3.6.3.1.B.2 No The MNGP TS do not contain this TS. Therefore, TSTF-505 changes are not incorporated.

Residual Heat Removal Service Water (RHRSW) System 3.7.1 3.7.1

L-MT-20-036 NSPM Page 13 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments One RHRSW pump in each subsystem inoperable.

3.7.1.B.1 No The MNGP TS do not contain this Condition.

Therefore, TSTF-505 changes are not incorporated.

One RHRSW subsystem inoperable for reasons other than Condition A.

3.7.1.C.1 3.7.1.A.1 Yes Wording of MNGP TS differs from TSTF-505 (i.e.,

MNGP TS does not have Condition A from NUREG-1433/TSTF-505). TSTF-505 changes are incorporated.

[Plant Service Water (PSW)] System and [Ultimate Heat Sink (UHS)]

3.7.2 3.7.2 One [PSW] pump in each subsystem inoperable.

3.7.2.B.1 No The MNGP TS do not contain this Condition.

Therefore, TSTF-505 changes are not incorporated.

[ One or more cooling towers with one cooling tower fan inoperable.

3.7.2.C.1 No The MNGP TS do not contain this Condition.

Therefore, TSTF-505 changes are not incorporated.

One [PSW] subsystem inoperable for reasons other than Condition[s] A [and C].

3.7.2.E.1 3.7.2.A.1 Yes Wording of MNGP TS differs from TSTF-505 (i.e.,

MNGP TS does not have Condition A or C from NUREG-1433/TSTF-505 and uses "Emergency Service Water (ESW)" instead of PSW).

TSTF-505 changes are incorporated.

The Main Turbine Bypass System 3.7.7 3.7.7

[Requirements of the LCO not met or Main Turbine Bypass System inoperable].

3.7.7.A.1 3.7.7.A.1 No The MNGP Main Turbine Bypass System design only includes two bypass valves. Therefore, one bypass valve inoperable results in a loss of function. Therefore, TSTF-505 changes are not incorporated.

AC Sources - Operating 3.8.1 3.8.1

L-MT-20-036 NSPM Page 14 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments One [required] offsite circuit inoperable.

3.8.1.A.3 3.8.1.A.3 Yes TSTF-505 changes are incorporated.

One [required] DG inoperable.

3.8.1.B.4 3.8.1.B.4 Yes TSTF-505 changes are incorporated.

Two [required] offsite circuits inoperable.

3.8.1.C.2 3.8.1.C.2 Yes TSTF-505 changes are incorporated.

One [required] offsite circuit inoperable.

AND One [required] DG inoperable.

3.8.1.D.1 3.8.1.D.2 3.8.1.D.1 3.8.1.D.2 Yes Yes TSTF-505 changes are incorporated.

[ One [required] [automatic load sequencer] inoperable.

3.8.1.F.1 No The MNGP TS do not contain this Condition.

Therefore, TSTF-505 changes are not incorporated.

DC Sources - Operating 3.8.4 3.8.4 One [or two] battery charger[s on one division] inoperable.

3.8.4.A.3 3.8.4.A.3 Yes Wording of MNGP TS differs from TSTF-505 (i.e.,

MNGP TS added the term "required" since each 250 VDC subsystem has two battery chargers and a spare, but only two are required to be OPERABLE in each 250 VDC subsystem. In addition, each 125 VDC subsystem has one battery charger, with a spare battery charger that is common to both 125 VDC subsystems. Lastly, MNGP Condition A contains "Division 1 or Division 2" specific to plant nomenclature).

TSTF-505 changes are incorporated.

One [or two] batter[y][ies on one division]

inoperable.

3.8.4.B.1 No The MNGP TS do not contain this Condition.

Therefore, TSTF-505 changes are not incorporated.

L-MT-20-036 NSPM Page 15 of 15 Table A4-1: Cross-Reference of TSTF-505 and MNGP Technical Specifications TSTF-505 Tech Spec Section Title /

Condition Description TSTF-505 TS MNGP TS Apply RICT?

Comments One DC electrical power subsystem inoperable for reasons other than Condition A [or B].

3.8.4.C.1 3.8.4.B.1 Yes Wording of MNGP TS differs from TSTF-505 (i.e.,

MNGP TS added the term "Division 1 or Division 2" specific to plant nomenclature). TSTF-505 changes are incorporated.

Inverters - Operating 3.8.7 One [required] inverter inoperable.

3.8.7.A.1 No The MNGP TS do not contain this TS. Therefore, TSTF-505 changes are not incorporated.

Distribution Systems - Operating 3.8.9 3.8.7 One or more AC electrical power distribution subsystems inoperable.

3.8.9.A.1 3.8.7.A.1 Yes TSTF-505 changes are incorporated.

[ One or more AC vital buses inoperable.

3.8.9.B.1 No The MNGP TS do not contain this Condition.

Therefore, TSTF-505 changes are not incorporated.

One or more [station service] DC electrical power distribution subsystems inoperable.

3.8.9.C.1 3.8.7.B.1 Yes TSTF-505 changes are incorporated.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 5.5.16 No The MNGP TS do not currently contain this program. The new RICT Program will be added to the MNGP TS 5.5.18 consistent with TSTF-505.

ATTACHMENT 4 MONTICELLO NUCLEAR GENERATING PLANT Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Revised RICT Program PRA Implementation Items (1 Page Follows)

L-MT-20-036 NSPM Page 1 of 1 Revised RICT Program PRA Implementation Items Revised Table A5-1 Table A5-1 provided with the LAR has been updated to reflect changes due to specific RAI responses in the preceding sections of the Enclosure. Changes from the version included in the LAR have been identified by change bars.

Table A5-1: RICT Program PRA Implementation Items No.

Implementation Items

1.

NSPM shall ensure that Reactor Protection System RPS Instrumentation is modeled in the MNGP PRA with sufficient detail to accurately calculate a RICT prior to implementation of the RICT Program.

2.

NSPM shall ensure that Mechanical Vacuum Pump system and isolation instrumentation are modeled in the MNGP PRA with sufficient detail to accurately calculate a RICT prior to implementation of the RICT Program.

3.

NSPM shall ensure that the Automatic Depressurization System (ADS) and instrumentation is modeled in the MNGP PRA with sufficient detail to accurately calculate a RICT prior to implementation of the RICT Program.

4.

NSPM shall ensure that the L-41 AC panel is modeled in the MNGP Fire PRA with sufficient detail to accurately calculate a RICT prior to implementation of the RICT Program.

5.

NSPM shall ensure that the Standby Liquid Control (SBLC) System is modeled with sufficient detail in the MNGP Fire PRA to accurately calculate a RICT prior to implementation of the RICT Program.

6.

NSPM shall ensure the PRA success criteria for Drywell vacuum breakers are clear in the MNGP PRA to accurately calculate a RICT prior to implementation of the RICT Program.

7.

NSPM shall ensure that appropriate joint HEP is used in the MNGP PRA to accurately calculate a RICT prior to implementation of the RICT Program.

ATTACHMENT 5 MONTICELLO NUCLEAR GENERATING PLANT Response to Request for Additional Information Related to License Amendment Request to Implement Technical Specification Task Force Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Revised Table E1-2: In-Scope TS/LCO Conditions RICT Estimate (3 Pages Follow)

L-MT-20-036 NSPM Page 1 of 3 Revised Table E1-2 Table E1-2 provided with the LAR has been updated to reflect changes due to specific RAI responses in the preceding sections of the Enclosure. Changes from the version included in the LAR have been identified by change bars.

Table E1-2: In-Scope TS/LCO Conditions RICT Estimate Tech Spec LCO Condition RICT Estimate 3.1.7.B One [Standby Liquid Control (SLC)] subsystem inoperable for reasons other than Condition A.

30 Days 3.3.1.1.A One or more required channels inoperable.

30 Days 3.3.1.1.B

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NOTE---------------------------------------------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, 2.f or 2.g.

One or more Functions with one or more required channels inoperable in both trip systems.

30 Days 3.3.2.2.A One or more feedwater pump and main turbine high water level trip channels inoperable.

30 Days 3.3.4.1.A One or more channels inoperable.

30 Days 3.3.5.1.B As required by Required Action A.1 and referenced in Table 3.3.5.1-1. 17 Days 3.3.5.1.C As required by Required Action A.1 and referenced in Table 3.3.5.1-1. 30 Days 3.3.5.1.D As required by Required Action A.1 and referenced in Table 3.3.5.1-1. 30 Days 3.3.5.1.E As required by Required Action A.1 and referenced in Table 3.3.5.1-1. 30 Days 3.3.5.1.F As required by Required Action A.1 and referenced in Table 3.3.5.1-1. 30 Days 3.3.5.1.G As required by Required Action A.1 and referenced in Table 3.3.5.1-1. 30 Days 3.3.5.2.B As required by Required Action A.1 and referenced in Table 3.3.5.2-1. 16 Days 3.3.5.2.D As required by Required Action A.1 and referenced in Table 3.3.5.2-1. 30 Days 3.3.6.1.A One or more required channels inoperable.

30 Days 3.3.7.2.A One or more channels inoperable.

30 Days 3.3.8.1.A One or more channels inoperable.

30 Days 3.4.3.A One or two required [Safety/Relief Valves (S/RVs)] inoperable.

30 Days 3.5.1.B One LPCI subsystem inoperable for reasons other than Condition A.

OR One Core Spray subsystem inoperable.

18 Days 3.5.1.C One LPCI pump in both LPCI subsystems inoperable.

30 Days

L-MT-20-036 NSPM Page 2 of 3 Table E1-2: In-Scope TS/LCO Conditions RICT Estimate Tech Spec LCO Condition RICT Estimate 3.5.1.D Two LPCI subsystems inoperable for reasons other than Condition C or G.

15 Days(1) 3.5.1.E One Core Spray subsystem inoperable.

AND One LPCI subsystem inoperable.

OR One or two LPCI pump(s) inoperable.

No voluntary entry (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Front Stop)(1) 3.5.1.I HPCI System inoperable.

30 Days 3.5.1.J HPCI System inoperable.

AND Condition A, B, or C entered.

16 Days 3.5.1.K One ADS valve inoperable.

30 Days 3.5.3.A RCIC System inoperable.

18 Days 3.6.1.2.C Primary containment air lock inoperable for reasons other than Condition A or B.

30 Days 3.6.1.3.A

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NOTE------

Only applicable to penetration flow paths with two PCIVs.

One or more penetration flow paths with one PCIV inoperable for reasons other than Condition D or E.

30 Days 3.6.1.3.D One or more penetration flow paths with one or more 18 inch primary containment purge and vent valves not within purge and vent valve leakage limits.

30 Days 3.6.1.6.C One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

30 Days 3.6.1.7.A One required suppression chamber-to-drywell vacuum breaker inoperable for opening.

30 Days 3.6.1.8.A One RHR drywell spray subsystem inoperable.

30 Days 3.6.2.3.A One RHR suppression pool cooling subsystem inoperable.

24 Days 3.7.1.A One RHRSW subsystem inoperable.

24 Days 3.7.2.A One ESW subsystem inoperable.

30 Days(2) 3.8.1.A One required offsite circuit inoperable.

7 Days

L-MT-20-036 NSPM Page 3 of 3 Table E1-2: In-Scope TS/LCO Conditions RICT Estimate Tech Spec LCO Condition RICT Estimate 3.8.1.B One EDG inoperable.

30 Days 3.8.1.C Two required offsite circuits inoperable.

No Entry(1) 3.8.1.D One required offsite circuit inoperable.

AND One EDG inoperable.

6 Days 3.8.4.A One or more required battery chargers on Division 1 or Division 2 inoperable.

30 Days 3.8.4.B One Division 1 or Division 2 DC electrical power subsystem inoperable for reasons other than Condition A.

No voluntary entry (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Front Stop)(1) 3.8.7.A One or more AC electrical power distribution subsystems inoperable.

No voluntary entry (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Front Stop)(1) 3.8.7.B One or more DC electrical power distribution subsystems inoperable. No voluntary entry (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Front Stop)(1)

Table E1-2 Notes:

1.

Several quantification results exceed the risk cap level of 1E-03 (CDF) or 1E-04 (LERF). Those LCOs are listed as "No Entry" given the quantified risk. However, it is possible that the LCO could be entered for a partial failure and would result in lower quantified risk. In a lower risk condition, entry into the RICT program would be allowed.

2.

The ESW subsystem was not required to be credited in thermal hydraulic analysis to mitigate core damage or LERF. Therefore, there is no risk impact by removing the subsystem from service.