L-MT-18-010, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML18087A323
Person / Time
Site:	Monticello
Issue date:	03/28/2018
From:	Church C Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-18-010
Download: ML18087A323 (44)

Text

2807 West County Road 75 Monticello, MN 55362 800.895.4999 xcelenergy.com March 28, 2018 L-MT-18-010 10 CFR 50.90 10 CFR 50.69 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 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors In accordance with the provisions of 10 CFR 50.69 and 10 CFR 50.90, Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, is requesting an amendment to the Renewed Facility Operating License (FOL) of the Monticello Nuclear Generating Plant (MNGP).

The proposed amendment would modify the MNGP licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The enclosure to this letter provides the basis for the proposed change to the MNGP Renewed FOL. The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) Report NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005, which was endorsed by the NRC in Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, dated May 2006. Attachment 1 of the enclosure provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met.

Document Control Desk Page 2 The Probabilistic Risk Assessment (PRA) models described within this license amendment request (LAR) are the same as those described within NSPM submittal of the LAR dated December 19, 2017 to adopt TSTF-425 (ADAMS Accession No. ML17353A189), with routine updates applied. NSPM requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process. This would reduce the number of NSPM and NRC resources necessary to complete the review of the applications. This request should not be considered a linked request as the details of the PRA models in each LAR are complete, which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

NSPM requests approval of the proposed change by April 30, 2019, with an implementation period of 90 days.

In accordance with 10 CFR 50.91 (b)(1 ), a copy of this application, with attachments, is being provided to the designated Minnesota Official.

If there are any questions or if additional information is required, please contact Mr. Shane Jurek at (612) 330-5788.

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 March 28, 2018.

~1c.1/4 Christopher R. Church 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-18-010 NSPM Enclosure ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT Evaluation of the Proposed Change TABLE OF CONTENTS

1.

SUMMARY

DESCRIPTION ................................................................................................ 1

2. DETAILED DESCRIPTION ................................................................................................. 1 2.1 CURRENT REGULATORY REQUIREMENTS............................................................. 1 2.2 REASON FOR PROPOSED CHANGE ........................................................................ 1

2.3 DESCRIPTION

OF THE PROPOSED CHANGE ......................................................... 3

3. TECHNICAL EVALUATION ............................................................................................... 3 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i)) .................... 4 3.1.1 Overall Categorization Process ................................................................................. 4 3.1.2 Passive Categorization Process ................................................................................ 8 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii)) .............................. 9 3.2.1 Internal Events and Internal Flooding ........................................................................ 9 3.2.2 Fire Hazards .............................................................................................................. 9 3.2.3 Seismic Hazards ....................................................................................................... 9 3.2.4 Other External Hazards ........................................................................................... 10 3.2.5 Low Power and Shutdown ....................................................................................... 10 3.2.6 PRA Maintenance and Updates .............................................................................. 11 3.2.7 PRA Uncertainty Evaluations .................................................................................. 11 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii)) .................................. 12 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv)) .......................................................... 13

4. REGULATORY EVALUATION ......................................................................................... 13 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA .................................... 13 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS ................................... 14

4.3 CONCLUSION

S ......................................................................................................... 15

5. ENVIRONMENTAL CONSIDERATION ............................................................................ 16

6. REFERENCES ................................................................................................................. 16 i

L-MT-18-010 NSPM Enclosure LIST OF ATTACHMENTS : List of Categorization Prerequisites ................................................................ 18 : Description of PRA Models Used in Categorization ........................................ 19 : Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items .................................................................................................. 20 : External Hazards Screening ........................................................................... 32 : Progressive Screening Approach for Addressing External Hazards ............... 38 : Disposition of Key Assumptions/Sources of Uncertainty ................................ 39 ii

L-MT-18-010 NSPM Enclosure

1.

SUMMARY

DESCRIPTION The proposed amendment modifies the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment and evaluation). For equipment determined to be low safety significant (LSS),

alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be high safety significant (HSS), requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

2. DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS The Nuclear Regulatory Commission (NRC) has established a set of regulatory requirements for commercial nuclear reactors to ensure that a reactor facility does not impose an undue risk to the health and safety of the public, thereby providing reasonable assurance of adequate protection to public health and safety. The current body of NRC regulations and their implementation are largely based on a deterministic approach.

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DBEs to protect public health and safety. The Structures, Systems and Components (SSCs) necessary to defend against the DBEs are defined as safety-related. These SSCs are the subject of many regulatory requirements, herein referred to as special treatments, designed to ensure that they are of high quality and high reliability, and have the capability to perform during postulated design basis conditions. Treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations.

The distinction between treatment and special treatment is the degree of NRC specification as to what must be implemented for particular SSCs or for particular conditions. Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: safety-related, important to safety, or basic component. The terms safety-related and basic component are defined in the regulations, while important to safety, used principally in the general design criteria (GDC) of Appendix A to 10 CFR 50, is not explicitly defined.

2.2 REASON FOR PROPOSED CHANGE A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing Page 1 of 40

L-MT-18-010 NSPM Enclosure a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessments (PRAs) address credible initiating events by assessing the event frequency. Mitigating systems reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

To take advantage of the safety enhancements available through the use of PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment and evaluation). For equipment determined to be LSS, alternative treatment requirements can be implemented in accordance with the regulation. For equipment determined to be HSS, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The rule contains requirements on how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four risk-informed safety class (RISC) categories. The determination of safety significance is performed by an integrated decision-making process, as described by Nuclear Energy Institute (NEI) Report NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline (Reference 1), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSC is applied as necessary to maintain functionality and reliability, and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable requirements.

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, the rule enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained or enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50.69 will allow Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, to improve focus on equipment that has safety significance resulting in improved safety at the Monticello Nuclear Generating Plant (MNGP).

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L-MT-18-010 NSPM Enclosure

2.3 DESCRIPTION

OF THE PROPOSED CHANGE NSPM proposes the addition of the following condition to the Renewed Facility Operating License of the MNGP to document the NRCs approval of the use of 10 CFR 50.69.

NSPM is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3 and RISC-4 structures, systems and components specified in the license amendment request dated March 28, 2018.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

3. TECHNICAL EVALUATION 10 CFR 50.69 specifies the information to be provided by a licensee requesting adoption of the regulation. This request conforms to the requirements of 10 CFR 50.69(b)(2), which states:

A licensee voluntarily choosing to implement this section shall submit an application for license amendment under § 50.90 that contains the following information:

(i) A description of the process for categorization of RISC-1, RISC-2, RISC-3 and RISC-4 SSCs.

(ii) A description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

(iii) Results of the PRA review process conducted to meet § 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy § 50.69(c)(1)(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions).

Each of these submittal requirements are addressed in the following sections.

The PRA models described within this license amendment request (LAR) are the same as those described within NSPM submittal of the LAR dated December 19, 2017, to adopt Page 3 of 40

L-MT-18-010 NSPM Enclosure TSTF-425 (Reference 2), with routine maintenance updates applied. NSPM requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process. This would reduce the number of NSPM and NRC resources necessary to complete the review of the applications.

This request should not be considered a linked request as the details of the PRA models in each LAR are complete, which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process NSPM will implement the risk categorization process in accordance with NEI 00-04, Revision 0, as endorsed by Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance (Reference 3). NEI 00-04, Section 1.5, states, Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant. A separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00-04, as endorsed by RG 1.201. RG 1.201 states that the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv). However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed. Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and, as long as they are all completed, they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as LSS by all other elements. Similarly, NEI 00-04 only requires Item 4 to be completed for safety-related active components/functions categorized as LSS by all other elements.

1. PRA-based evaluations (e.g., the internal events, internal flooding, and fire PRAs)

2. Non-PRA approaches (e.g., fire safe shutdown equipment list (SSEL), seismic SSEL, other external events screening, and shutdown assessment)

3. Seven qualitative criteria in Section 9.2 of NEI 00-04

4. The defense in depth assessment

5. The passive categorization methodology Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various Page 4 of 40

L-MT-18-010 NSPM Enclosure elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., HSS or LSS) that is presented to the Integrated Decision-Making Panel (IDP). Note: the term preliminary HSS or LSS is synonymous with the NEI 00-04 term candidate HSS or LSS. A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1 below. The safety significance determination of each element, identified above, is independent of each other. Therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will only be preliminary until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final RISC category can be assigned.

Table 3 IDP Changes from Preliminary HSS to LSS Drives Categorization Step IDP Change Element Evaluation Level Associated (NEI 00-04 Section) HSS to LSS Functions Internal Events Base Case Not Allowed Yes (Section 5.1)

Fire, Seismic and Other Risk External Events Base Case Allowable No (PRA- (Sections 5.2, 5.3 and 5.4) Component Modeled) PRA Sensitivity Studies Allowable No Integral PRA Assessment Not Allowed Yes (Section 5.6)

Fire, Seismic and Other Risk External Hazards Component Not Allowed No (Non- (Sections 5.2, 5.3 and 5.4)

Modeled) Shutdown Function/Component Not Allowed No (Section 5.5)

Core Damage Function/Component Not Allowed Yes Defense in (Section 6.1)

Depth Containment Component Not Allowed Yes (Section 6.2)

Qualitative Considerations Function Allowable N/A Criteria (Section 9.2)

Passive Passive Segment/Component Not Allowed No (Section 4)

The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04, Section 10.2. The IDP may always elect to change a preliminary LSS component or function to HSS; however, the ability to change component categorization from preliminary HSS to LSS is limited. This ability is only available to the IDP for select process steps as described in NEI 00-04 and endorsed by RG 1.201. Table 3-1 summarizes these IDP Page 5 of 40

L-MT-18-010 NSPM Enclosure limitations in NEI 00-04. The steps of the process are performed at either the function level, component level, or both. This is also summarized in Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

The mapping of components to system function is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., Internal Events PRA or Integral PRA Assessment) or defense in depth evaluation will be initially treated as HSS. However, NEI 00-04, Section 10.2, allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS; and Section 4 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with an HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g., passive, non PRA-modeled hazards - see Table 3-1). These components from the component level assessments will remain HSS (i.e., IDP cannot override) regardless of the significance of the functions to which they are mapped. Therefore, if a HSS component is mapped to a LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven to HSS based on Table 3-1, or may remain LSS.

The following are clarifications to be applied to the NEI 00-04 categorization process:

The IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and PRA. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address, at a minimum, the purpose of the categorization; present treatment requirements for SSCs including requirements for design basis events; PRA fundamentals; details of the plant-specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense in depth philosophy and requirements to maintain this philosophy.

The decision criteria for the IDP for categorizing SSCs as HSS or LSS pursuant to 10 CFR 50.69(f)(1) will be documented in NSPM procedures. Decisions of the IDP will be arrived at by consensus. Differing opinions will be documented and resolved, if possible. If a resolution cannot be achieved concerning the safety significance of an SSC, then the SSC will be classified as HSS.

Passive categorization will be performed using the processes described in Section 3.1.2 of this enclosure. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

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L-MT-18-010 NSPM Enclosure An unreliability factor of 3 will be used for the sensitivity studies described in Section 8 of NEI 00-04. The factor of 3 was chosen as it is representative of the typical error factor of basic events used in the PRA model.

NEI 00-04, Section 7, requires assigning the safety significance of functions to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in Section 5, but does not require this for SSCs determined to be HSS from non PRA-based, deterministic assessments in Section 5. This requirement is further clarified in the NRC Safety Evaluation (SE) (Reference 4) approving the Vogtle license amendment to adopt 10 CFR 50.69, which states, if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense in depth assessment (Section 6), the associated system function(s) would be identified as HSS.

Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The IDP must intervene to assign any of these HSS function components to LSS.

With regard to the criteria that consider whether the active function is called out or relied upon in the plant Emergency/Abnormal Operation Procedures, NSPM will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

The risk analysis to be implemented for each hazard is described below:

Internal Event Risks: Internal events including internal flooding PRA model Revision 3.4.

The internal events PRA model described within this LAR is the same as the one described within NSPMs submittal of the LAR to adopt TSTF-425 (Reference 2).

Fire Risks: Fire PRA model Revision 4. This PRA model was credited in the MNGP LAR to extend the integrated leak rate testing (ILRT) interval, as accepted by the NRC in Reference 5. Furthermore, the Fire PRA model described within this LAR is the same as the one described within NSPMs submittal to adopt TSTF-425 (Reference 2).

Seismic Risks: SSEL referenced in the Individual Plant Examination of External Events (IPEEE) seismic analysis accepted by NRC Staff Evaluation Report dated April 14, 2000 (Reference 6).

Other External Risks (e.g., tornados, external floods): Using the IPEEE screening process as approved by the NRC in Reference 6. The other external hazards were determined to be insignificant contributors to plant risk.

Low Power and Shutdown Risks: Qualitative defense in depth shutdown model for shutdown configuration risk management based on the framework for defense in depth provided in Nuclear Management and Resource Council (NUMARC) Report Page 7 of 40

L-MT-18-010 NSPM Enclosure NUMARC 91-06, Guidance for Industry Actions to Assess Shutdown Management (Reference 7), which provided guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above (e.g.,

change from a seismic margins approach to a seismic PRA approach) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

2. System functions, identified and categorized, with the associated bases

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members 3.1.2 Passive Categorization Process For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Arkansas Nuclear One (ANO) Risk-Informed Repair/Replacement Activities (RI-RRA) methodology as approved by the NRC in Reference 8.

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., defense in depth, safety margins) in determining safety significance. Component supports are assigned the same safety significance as the highest passively ranked segment within the bounds of the associated analytical pipe stress model. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

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L-MT-18-010 NSPM Enclosure The use of this method was previously approved to be used for a 10 CFR 50.69 application by the NRC in Reference 4. The RI-RRA method, as approved for use at Vogtle for 10 CFR 50.69, does not have any plant specific aspects and is generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to change in treatment. The passive categorization process is intended to apply the same risk-informed process approved for use at ANO for passive categorization of Class 2, 3, and non-class components. Consistent with the ANO RI-RRA method, Class 1 pressure retaining SSCs in the scope of the system being categorized will be assigned HSS and cannot be changed by the IDP. Therefore, the RI-RRA methodology for passive categorization is acceptable and appropriate for use at MNGP for 10 CFR 50.69 SSC categorization.

3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii))

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed.

3.2.1 Internal Events and Internal Flooding The MNGP categorization process for internal events and flooding hazards will use the plant-specific PRA model. The NSPM risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant. Attachment 2 to this enclosure identifies the applicable internal events PRA model, which encompasses internal flooding.

3.2.2 Fire Hazards The MNGP categorization process for fire hazards will use a peer-reviewed, plant-specific fire PRA model. The internal fire PRA model was developed consistent with NUREG/CR-6850 (Reference 9) and only utilizes methods previously accepted by the NRC. The NSPM risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant. Attachment 2 to this enclosure identifies the applicable fire PRA model.

3.2.3 Seismic Hazards The MNGP categorization process will use the seismic margins analysis (SMA) performed for the IPEEE in response to Generic Letter (GL) 88-20, Supplement 4 (Reference 10), for evaluation of safety significance related to seismic hazards. No plant specific approaches were utilized in development of the SMA. The NEI 00-04 approved use of the SMA SSEL as a screening process results in the identification of all system functions and associated SSCs that are involved in the seismic margin success path as HSS. Since the analysis is being used as a Page 9 of 40

L-MT-18-010 NSPM Enclosure screening tool, importance measures are not used to determine safety significance. The NEI 00-04 approach using the SSEL identifies credited equipment as HSS regardless of their capacity, frequency of challenge, or level of functional diversity.

An evaluation was performed of the as-built, as-operated plant against the SMA SSEL. The evaluation compared the as-built, as-operated plant to the plant configuration originally assessed by the SMA. Differences were reviewed to identify any potential impacts to the equipment credited on the SSEL. Appropriate changes to the credited equipment were identified and documented. This documentation is available for audit. The NSPM risk management program will ensure that future changes to the plant will be evaluated to determine their impact on the SMA and risk categorization process.

3.2.4 Other External Hazards All other external hazards (i.e., not seismic or fire hazards) were screened from applicability to MNGP per a plant-specific evaluation in accordance with GL 88-20, Supplement 4, and updated to use the criteria in the ASME PRA Standard RA-Sa-2009 (Reference 11). to this enclosure provides a summary of the other external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

As part of the categorization assessment of other external hazard risk, an evaluation is performed to determine if there are components being categorized that participate in screened scenarios and whose failure would result in an unscreened scenario. Consistent with the flow chart in Figure 5-6 in Section 5.4 of NEI 00-04, these components would be considered HSS.

All remaining hazards were screened from applicability and considered insignificant for every SSC and, therefore, will not be considered during the categorization process.

3.2.5 Low Power and Shutdown Consistent with NEI 00-04, the MNGP categorization process will use the shutdown safety management plan described in NUMARC 91-06 for evaluation of safety significance related to low power and shutdown conditions. The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

NUMARC 91-06 specifies that a defense in depth approach should be used with respect to each defined shutdown key safety function. The key safety functions defined in NUMARC 91-06 are evaluated for categorization of SSCs.

SSCs that meet the two criteria (i.e., considered part of a primary shutdown safety system or a failure would initiate an event during shutdown conditions) described in Section 5.5 of NEI 00-04 will be considered preliminary HSS.

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L-MT-18-010 NSPM Enclosure 3.2.6 PRA Maintenance and Updates The NSPM risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant. The process delineates the responsibilities and guidelines for updating the PRA models, and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience), for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The process will assess the impact of these changes on the plant PRA model in a timely manner which is typically considered to be once every two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

In addition, NSPM will implement a process that addresses the requirements in NEI 00-04, Section 11, Program Documentation and Change Control. The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model upgrades will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.2.7 PRA Uncertainty Evaluations Uncertainty evaluations associated with any applicable baseline PRA models used in this application were evaluated during the assessment of PRA technical adequacy and confirmed through the self-assessment and peer review processes as discussed in Section 3.3 of this enclosure.

Uncertainty evaluations associated with the risk categorization process are addressed using the process discussed in Section 8 of NEI 00-04 and in the prescribed sensitivity studies discussed in Section 5.

In the overall risk sensitivity studies, NSPM will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference 4.

Consistent with the NEI 00-04 guidance, NSPM will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

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L-MT-18-010 NSPM Enclosure The detailed process of identifying, characterizing, and qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855 (Reference 12) and Section 3.1.1 of EPRI TR-1016737 (Reference 13). The process in these references was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

The list of assumptions and sources of uncertainty were reviewed to identify those which would be significant for the evaluation of this application. If the MNGP PRA model used a non-conservative treatment, or methods that are not commonly accepted, the underlying assumption or source of uncertainty was reviewed to determine its impact on this application.

Only those assumptions or sources of uncertainty that could significantly impact the risk calculations were considered key for this application.

Key MNGP PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned in Attachment 6 to this enclosure. The conclusion of this review is that no additional sensitivity analyses are required to address MNGP PRA model specific assumptions or sources of uncertainty.

3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))

The PRA models described in Section 3.2 have been assessed against RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 14), consistent with NRC Regulatory Issue Summary (RIS) 2007-06 (Reference 15).

The internal events PRA model, including internal flooding, was subject to a self-assessment and full-scope peer review conducted in accordance with RG 1.200, Revision 2, in April 2013.

Additionally, a focused scope peer review was conducted in April 2017 to review the convolution analysis portion of the PRA model.

The fire PRA model was subject to a self-assessment and full-scope peer review conducted in accordance with RG 1.200, Revision 2, in March 2015. Additionally, a focused scope peer review was conducted in December 2016, to account for enhanced fire modeling methods that were incorporated into the model.

Finding closure reviews were conducted on the internal events, including internal flooding, and fire PRA models in August and October 2017, respectively. Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-Out of Facts and Observations (Reference 16) as accepted by the NRC in Reference 17. The results of this review have been documented and are available for NRC audit. to this enclosure provides a summary of the remaining findings and open items, including:

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L-MT-18-010 NSPM Enclosure Open findings and disposition of the MNGP fire PRA model peer review. (Note: All internal events PRA model, including internal flooding, findings were closed during the finding closure review.)

Identification of and basis for any sensitivity analysis needed to address open findings.

The attachment identified above demonstrates that the PRA is of sufficient quality and level of detail to support the categorization process, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required in 10 CFR 50.69(c)(1)(i).

3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv))

The MNGP 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04.

The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common cause interactions, and meets the requirements of 10 CFR 50.69(b)(2)(iv). Sensitivity studies described in NEI 00-04, Section 8, will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF). The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

4. REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

The regulations in 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors.

RG 1.201, Guidelines for Categorizing Structures, Systems and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, May 2006 (Reference 3).

RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018 (Reference 18).

RG 1.200, An Approach for Determining Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009 (Reference 14).

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L-MT-18-010 NSPM Enclosure The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, proposes to modify the licensing basis of the Monticello Nuclear Generating Plant to allow for the voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

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

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

Response: No The proposed change will permit the use of a risk-informed categorization process to modify the scope of Structures, Systems and Components (SSCs) subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensure the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

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

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L-MT-18-010 NSPM Enclosure

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

Response: No The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC.

Under the proposed change, no additional plant equipment will be installed.

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

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

Response: No The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin.

The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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

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

4.3 CONCLUSION

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

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L-MT-18-010 NSPM Enclosure

5. ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6. REFERENCES

1. NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005 (Agencywide Document Access and Management System (ADAMS) Accession No. ML052910035)

2. NSPM letter to 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)

3. NRC Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, dated May 2006 (ADAMS Accession No. ML061090627)

4. NRC letter to Southern Nuclear Operating Company, Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC Nos. ME9472 and ME9473), dated December 17, 2014 (ADAMS Accession No. ML14237A034)

5. NRC letter to NSPM, Issuance of Amendment Re: Technical Specification 5.5.11, Primary Containment Leakage Rate Testing Program (CAC No. MF7359), dated April 25, 2017 (ADAMS Accession No. ML17103A235)

6. NRC letter to NSP, Review of Monticello Individual Plant Examination of External Events (IPEEE) Submittal (TAC No. M83644), dated April 14, 2000

7. NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, dated December 1991 (ADAMS Accession No. ML14365A203)

8. NRC letter to Entergy Operations, Inc., Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC No. MD5250), dated April 22, 2009 (ADAMS Accession No. ML090930246)

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L-MT-18-010 NSPM Enclosure

9. NRC NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volumes 1 and 2, dated September 2005 (ADAMS Accession Nos.

ML15167A401 and ML15167A411)

10. NRC Generic Letter 88-20, Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), dated June 28, 1991 (ADAMS Accession No. ML031150485)

11. 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

12. NRC NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, dated March 2009 (ADAMS Accession No. ML090970525)

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

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

15. NRC Regulatory Issue Summary 2007-06, Regulatory Guide 1.200 Implementation, dated March 22, 2007 (ADAMS Accession No. ML070650428)

16. NEI letter to NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os), dated February 21, 2017 (ADAMS Accession No. ML17086A450)

17. NRC letter to NEI, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close out of Facts and Observations (F&Os), dated May 3, 2017 (ADAMS Accession No. ML17079A427)

18. NRC Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256)

19. NSPM letter to NRC, Monticello Nuclear Generating Plant: Response to Post-Fukushima Near-Term Task Force (NTTF) Recommendation 2.1, Flooding - Flood Hazard Reevaluation Report, dated May 12, 2016 (ADAMS Accession No. ML16145A179)

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L-MT-18-010 NSPM Enclosure Attachment 1: List of Categorization Prerequisites NSPM will establish procedure(s) prior to the use of the categorization process on a plant system. The procedure(s) will contain the elements/steps listed below.

IDP member qualification requirements.

Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary HSS or LSS based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.2 of this enclosure). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.

Component safety significance assessment. Safety significance of active components is assessed through a combination of PRA and non-PRA methods, covering all hazards.

Safety significance of passive components is assessed using a methodology for passive components.

Assessment of defense in depth and safety margin. Safety related components that are categorized as preliminary LSS are evaluated for their role in providing defense in depth and safety margin and, if appropriate, upgraded to HSS.

Review by the IDP. The categorization results are presented to the IDP for review and approval. The IDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.

Risk sensitivity study. For PRA-modeled components, an overall risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to CDF and LERF and meets the acceptance guidelines of RG 1.174.

Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.

Documentation requirements per Section 3.1.1 of this enclosure.

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L-MT-18-010 NSPM Enclosure Attachment 2: Description of PRA Models Used in Categorization Model Baseline CDF Baseline LERF Comments Internal Events PRA Model, This model represents including Internal Flooding, the current Internal Revision 3.4, dated June 1, Events, including Internal 2017. Flooding, PRA Model of 8.06E-06 6.95E-07 Record.

Full scope peer review against RG 1.200, Revision 2, in April 2013. Focused scope peer review in April 2017..

Fire PRA Model, Revision 4.0, This model represents dated January 30, 2017. the current Fire PRA Model of Record.

Full scope peer review against 5.24E-05 6.39E-06 RG 1.200, Revision 2, in March 2015. Focused scope peer review in December 2016.

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L-MT-18-010 NSPM Enclosure Attachment 3: Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 2-1 The requirement for this Supporting The fire PRA model was revised As recommended by the Requirement (SR) is to JUSTIFY credited spatial to evaluate spatially-separated Closure Review Team, the (PP-B3) separation. This section does not provide the fire areas in the multi- remaining few isolated cases (PP-C3) required justification. The documentation should compartment analysis with the where the barrier failure include an evaluation that establishes why the barrier failure probabilities set to probability is not set to 1.0 for separation provided by "space" will ensure that 1.0. The F&O finding closure spatially separated fire the adverse effects of fire will be substantially review team determined that the compartments will be contained in each of the adjacent Physical revised approach was corrected. These cases will be Analysis Units (PAUs). acceptable. However, several corrected by setting their isolated errors were identified barrier failure probability to 1.0 where barrier probabilities were in the Multi-Compartment set to lower values. A Analysis.

documentation discrepancy was also identified. These will be These changes are not corrected in a future update. expected to have a significant impact on CDF or LERF.

2-5 Section 5.6 discusses the apportionment of A comprehensive review of the These changes are not generic transient fire ignition frequencies and the influence factors was conducted expected to have any impact (IGN-A7) development of the influencing factors for each in response to this F&O including on CDF or LERF since the area. The influencing factors were assigned by a review by plant personnel. The recommendations are the Fire PRA analysts based on engineering maintenance influence factors associated with documentation judgment and a set of rules documented in were adjusted based on this changes to better explain Section 5.6.2 of the Ignition Frequency review. The F&O finding closure modeling rationale. If the Notebook. Assignment of these values resulted review determined that the "very low" factor cannot be in a comparatively low result. Based on the revised factors were acceptable. justified for compartments 8 information contained in the Fire Modeling However, better justification of and 33 then a different factor Database the influencing factors average as application of a very low factor will be appropriately applied follows: in two compartments needs to be with a justification.

provided.

Maintenance 1.7; Occupancy 2.2; and Storage 1.8. Its typically assumed that these factors will produce an average value, i.e., Medium or 3, by definition. Based on the values stated it appears Page 20 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 2-5 (cont.) that the influencing factors may have been underestimated. To increase the accuracy and reliability its suggested that these values be set or validated by plant operations and maintenance personnel.

For example, numerous fire zones were assigned LOW maintenance factors including H2 Seal Oil/Condensate Pump Area, Turbine Condenser Area, Air Ejector Room, Admin Bldg.

HVAC Room, Engineered Safety Features (ESF)

Motor Control Center, 13.8 kV Switchgear Rooms, Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI)

Rooms, Diesel Fuel Oil Pump House, etc. as these zones contain pumps, motors, electrical equipment that would require maintenance.

LOW storage factor was assigned in numerous fire zones including Lube Oil Storage Room, Contaminated Equipment Storage Area, etc.

which appear to be defined storage areas in the plant. Additionally, there are only 13 fire zones that are assigned storage factors greater than LOW.

3-6 The Fire PRA plant response model was not Documentation and model The Closure Review Team successfully modified to fail SSCs not selected changes were made to the fire Recommendation will be (PRM-B10) in the ES element. Representative examples of PRA to address this F&O. The addressed by including the this include: F&O Closure review determined specified basic events in the that the changes made largely fire failed events flag file.

1. In the PRM notebook, the basis for exclusion address the issues identified in of MSO 5j is that 'Monticello does not credit the F&O. However, there are still These changes are not operation of service water', however, service about ten component-level basic expected to have a significant water is not failed in the logic model. events that are not yet treated as impact on total CDF or LERF guaranteed failures in the fire since the current fire model

2. In section 3.2 of the PRM notebook Water, it model. A sensitivity study was contribution to total CDF and Page 21 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 3-6 (cont.) states that use of the Fire Water System as a performed which demonstrated LERF is 0.19% and 1.41%,

back-up to LPCI is not credited, however, this is that inclusion of these events respectively.

not failed in the model. result in a change in CDF and LERF of 0.19% and 1.41%,

3. In both the ES notebook and PRM notebook, respectively.

it is stated that CRDH and SLBC were not used in the fire PRA, however, these are not successfully failed in the model. They were failed by putting appropriate flags set to 1.0 in the model. However, basic events for SBLCS components (L) and HFE's appear in the results.

Basic events for CRD pump random failures (J) also appear in the model, with random failure probabilities. If the systems are correctly FLAGGED out, there should not be random failures of these systems. If the correct component is flagged, the logical 1.0 should propagate to the top of the tree, eliminating all other random failures. The fact that random events for these systems appear in cutsets indicate the correct basic event has not been chosen to be flagged. This particular example is not expected to be risk significant.

4. Individual components identified in Table D-1 of the ES notebook as not credited were not failed in the PRM [e.g., FPAP1AXXXR12-S -

CONDENSATE PUMP P-1A FAILS TO RUN (SHORT TERM)]

5. Conversely - Basic events that were not failed in the model, yet were not included in table C-1 as credited [e.g., ABSLPCIAXG - LPCI MCC FAULT (MCC-133A)]

4-11 An initial ambient temperature of 20°C was Validation studies of the three fire The Closure Review Team utilized in the fire modeling calculations for all modeling models used in the fire recommendation will be Page 22 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 4-11 (cont.) MNGP fire zones. This ambient temperature PRA were performed. In each addressed by revising the fire does not appear to be appropriate for areas that case, the model biases are models using expected plant (FSS-D4) are not temperature controlled such as the dispositioned as reasonable for ambient temperatures for each Turbine Building, Diesel Generator Building, and their use based upon Chapter 4 fire zone.

areas of the Reactor Building. of NUREG-1934. The F&O closure review team found the These changes are not validation studies to be expected to have a significant appropriate for cases in which impact on total CDF or LERF the ambient temperature is 20°C due to the conservative fire or less. Additional justification is modeling methods used.

required for plant areas which may have higher ambient temperatures.

4-20 Although the damage criteria for sensitive The fire PRA Single These changes are not electronics is defined in the Single Compartment Compartment Analyses have expected to have any impact (FSS-C5) Analysis Notebook 016015-RPT-06 and zones been updated to document the on CDF or LERF since the (FSS-D9) of influence (critical distances) are calculated in analysis associated with the recommendations are the Fire Modeling Database, there is no specific treatment of sensitive electronics. associated with documentation discussion of how specific sensitive electronics The F&O closure review team changes to better explain at Monticello are analyzed in the FPRA. determined that the methods modeling rationale.

used for all areas other than the main control room are either in agreement with FAQ 13-0004 or conservative in comparison to the FAQ. However, additional verification and documentation of the main control board configuration for sensitive electronics was determined to be required by the F&O closure team to fully resolve this F&O.

4-29 Appendix E of the Single Compartment Analysis For fire compartments resulting in These changes are not Notebook 016015-RPT-06 identifies that CDF greater than 1.0E-08/year, expected to have any impact (FSS-A1) scenarios for cable fires caused by welding and the process in FAQ 13-0005 was on CDF or LERF since the (FQ-A3) cutting and self-ignited cable fires result in high applied for cable fires caused by recommendations are total CDF contributions and further evaluation welding. The Single associated with documentation Page 23 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 4-29 (cont.) and refinement will be completed after risk Compartment Analysis notebook changes to better explain reduction activities are completed. These has been updated to document modeling rationale.

scenarios are not currently quantified in the the process used to treat cable FPRA model. fires due to hotwork, self-ignited cable fires, and junction box fires.

The F&O closure review team determined that the model changes were appropriate.

However, the F&O remains open since the documentation of the process used should be enhanced.

4-33 Wall and corner effects are not accounted for in Justification has been added to The Closure Review Team the FLASH-CAT modeling for heat release rate the fire PRA documentation to recommendation will be (FSS-A5) calculations that are used for the CFAST hot gas demonstrate that the FLASH- addressed by reviewing the layer models. CAT analyses would bound as- detailed modeled fire built conditions. However, the scenarios to determine which F&O finding closure review team ones meet the definition of determined that the results may wall or corner and revising not be bounding for cable trays in their model to address wall wall or wall-corner locations. and corner effects or provide Verification that FLASH-CAT justification that the approach results were not used for such was bounding for the expected configurations needs to be ignition sources in the fire performed. zone.

These changes are not expected to have a significant impact on total CDF or LERF due to the conservative fire modeling methods used.

6-3 Battery Chargers have been counted as either Many of the battery chargers The Closure Review Team Bin 10 or Bin 15 in the IGN development. have been re-assigned to Bin 10 recommendation will be (IGN-A1) for ignition frequency addressed by correctly It appears that well sealed low voltage panels determination, but some still assigning battery chargers (e.g. lighting panels) have been included in the need to be re-assigned to the D70, D80, and D90 to bin 15 Page 24 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 6-3 (cont.) bin 15 count that should be excluded. correct bin. A sensitivity study rather than bin 10.

shows that a change in CDF and LERF of -0.38% and -0.16%, These changes are not respectively will result once the expected to have a significant battery charger re-assignments impact on total CDF or LERF are completed. since their current fire model contribution to total CDF and The F&O closure team reviewed LERF is -0.38% and -0.16%,

the battery charger re-binning. respectively.

The team did not identify any specific issues with inappropriate counting of sealed low voltage panels. This F&O remains open since not all battery chargers have yet been reassigned.

6-9 Common cause and test and maintenance, and Common cause failure, test & The Closure Review Team pre-initiator human error basic events for core maintenance, and pre-initiator recommendation will be (SY-A19) spray and RHR are missing from the ASD Logic. basic events for core spray were addressed by correctly adding (PRM-B9) Additionally, the alternate shutdown modeling of added. RHR modeling under in the remaining one or two (SY-A1) core spray train B is also missing the failure multiple gates for the various basic events related to RHR (SY-A16) mode for: 'CS Pump P-208B to run after the first RHR functions were also added; B pump maintenance (SY-A14) hour.' Also, the review found that power supplies however, the F&O closure review unavailability that were not (SY-B9) for some of the active components were missing team identified that pump included under the alternate (SY-B5) from the alternate shutdown logic. maintenance unavailability shutdown logic.

(DA-E2) events needed to be added to some of the RHR fault tree logic. These changes are not expected to have a significant Power supplies were reviewed impact on total CDF or LERF and added to the model as since the few common cause appropriate. and maintenance events that will be included are not affected by a fire and are generally much less likely than the equipment failure probability.

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L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 6-11 Basic events added to the FPRA associated with Many of the identified issues The Closure Review Team the following failure modes have failure have been corrected with model recommendation will be (PRM-B9) probabilities set to zero: changes. However, a number of addressed by including the (SY-A1) UPS panel fault "breaker fail to remain open" specified basic events in the (SY-A2) Circuit breaker fails to remain open events and other miscellaneous fire failed events flag file.

(UNC-A2) Circuit breaker fails to open events that require data have not (DA-E2) Fused disconnect switch, fuse spuriously yet been updated in the model. These changes are not fails expected to have a significant Transformer fault impact on total CDF or LERF CS pump fails to start from the specified basic CS pump fails to run 1st hour events since their current fire model contribution to total MOV fails to remain open CDF and LERF is 0.00%.

MOV fails to open MOV fails to close 125 VDC distribution panel fault AOV fails to remain closed AOV fails to remain open Level transmitter spurious operation RHR Pump fails to run RHR Pump fails to start Solenoid valve fails to transfer 7-3 The purpose SR PRM-A4 is to confirm that the Corrections were made to the The Closure Review Team plant response model is constructed in such a Feedwater (FW) level control recommendation will be (PRM-A4) manner that it reflects the failure of identified logic. Corrections were also addressed by adding the HEP (PRM-C1) equipment due to the loss of the associated made to the FRANX mapping HPI-CNTRLY under and OR (SY-A1) equipment selected cables. and fault tree models to consider gate with L-RPV-INSTR: Loss (PRM-B9) dependency on both ADS of RPV Instrumentation in the Based on the review by peers, the following channels in the Alternate fault tree logic.

issues were identified. These are based on Shutdown logic.

limited time to review and are only examples. These changes are not Corrections were made to the expected to have a significant The fault tree modeling of essential cues for Main Steam Isolation Valve impact on total CDF or LERF HFE HPI-CNTRLY is not correct. The cues are (MSIV) pseudo-component logic. from the specified basic modeled under gate F-HEP-CNTRLML, and A review of the pseudo- events since their current fire ANDed with the medium LOCA initiator component and interlock model contribution to total Page 26 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 7-3 (cont.) IE_MLOCA 2.72E-4/yr (with no FPRA IE modeling was performed and CDF and LERF is 0.00%.

modeled there). The FPRA development team changes were incorporated into concurred that the cues modeling gate LRPV- the model. However, the F&O INSTR should be input into OR gate F020 finding closure review team (ORed with HPI-CNTRLY). identified some residual modeling issues with the HFE HPI-Equipment Selection report 016015-RPT-03 CNTRLY that is located in the Table B-2 identifies ADS-CHANNEL- FW level control logic which have A:Avail:Non-Spur and ADS-CHANNEL- not yet been corrected.

B:Avail:Non-Spur low level pseudo functions and equipment dependencies. It was determined during CS that the cables were properly mapped to the ADS pseudo component. Equipment SV271A, C, and D are dependent on both ADS A and B channel cables. However, review of the FRANX database FPRA CDF 2-2 determined improper Component to basic event mapping was made to the PRM. ADS channel A is mapped to SV271A and ADS channel B is mapped to the remaining SV271C and D.

A review of pseudo components MSIV-ISOL-A:Avail:Avail and LLS-DIV-B1:Avail:Non-Spur as identified in the ES procedure determined there was no modeling of the component to basic event relationship. From peer discussion it was determined that the noted pseudo-components were determined not required in the model following cut set review by the utility.

In addition, there was no evidence that the interlocks on the cable selection data worksheets were reviewed and properly incorporated into the PRM.

7-4 It is not clear if the internal events PRA initiating The fire PRA model has been The Closure Review Team events and accident sequences applicable to updated to use an adjusted Large recommendation will be Page 27 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR) 7-4 (cont.) two or more SRVs open similar to LLOCA have LOCA event tree for a spurious addressed by performing been correctly applied. If it is deemed that opening of 2 or more SRVs that thermal hydraulic MAAP (SY-A1) opening of two or more SRVs does not need to do not reclose due to an MSO analysis to determine the (PRM-B1) mimic LLOCA, then provisions for a new event. However, the F&O finding success criteria for the (PRM-B9) initiating event, success criteria and accident closure review team identified opening of 2 or more SRVs.

(PRM-C1) sequence is required. additional locations in the model The fault tree model will be (FQ-A2) where the revised logic model revised to reflect the PRM calculation 016015-RPT-05, MSO 3a and still needs to be added to fully determined success criteria.

3b for potential opening of two or more SRVs account for these fire-induced added additional logic to the PRM. Potential SRV opening scenarios.

opening of all SRVs mimics sequences similar to Large LOCA. Review of the PRM calculation noted in section 6.0 that the initiating events and accident sequences embodied in the MNGP internal events model are used as the basis for development of the FPRA model. Additional information received from the utility representative regarding the review of internal events initiating events determined that Large LOCA was deemed not applicable to the FPRA.

Review of the PRM CAFTA model located MSO failure of more than one SRV via gate F_SORV_2of8 with parents to gates different from LLOCA. If it is deemed that opening of all SRVs does not need to mimic LLOCA, then provisions for a new initiating event, success criteria and accident sequence is required.

FO-1 The analysis results of the thermal heat soak Documentation of the results of a The Closure Review Team method appear to credit ventilation limited sensitivity analysis that was recommendation will be (FSS-D3) burning in several PAUs without providing conducted to exclude credit of addressed by reviewing the (FSS-D4) sufficient basis. An example is the Group 1 ventilation-limited burning was cable heat soak fire modeling scenarios listed in Table J-6 of 016015-RPT-06. added. This analysis that credits ventilation limited Each of the four fire case CFAST results have demonstrates that the oxygen- burning. Credit for ventilation sensitivity cases due to the development of limited fire reaches the damage limited burning will be ventilation limited conditions. The baseline threshold for the cable at an removed.

CFAST results do not result in damage to a earlier time. This information was Page 28 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR)

FO-1 (cont.) generic target over a 60 min time interval. The used to justify the use of the These changes are not CFAST sensitivity cases that were originally run ventilation-limited cases for expected to have a significant with additional ventilation to verify constant estimating the time to hot gas impact on total CDF or LERF exposure damage times would likely result in layer formation. However, the due to the conservative fire damage to a generic TP target when assessed F&O finding closure review team modeling methods used.

in the heat soak model identified issues with the sensitivity case and its applicability in certain situations.

Additional justification concerning the treatment of the ventilation-limited modeling for those areas needs to be developed.

FO-2 A number of documentation issues have been The various documentation These changes are not identified. These include: issues identified in this F&O have expected to have any impact (FSS-D3) been addressed. However, the on CDF or LERF since the (FSS-D4) A. There are a number of scenarios that appear F&O finding closure review team recommendations are (FSS-H4) to credit the thermal heat soak method listed in determined that additional associated with documentation (FSS-H5) the FMDB but the HGL times do not match any information needs to be included changes to better explain (FSS-H9) scenario listed in Report 016015-RPT-06. An in the documentation concerning modeling rationale.

example is Equipment C-18 in the impacts of accumulation of tblIgnitionScenarios of the fire modeling damage at low temperature on database. Scenario 2 and the corresponding cables and on the impacts of comment indicates HGL time is 25 minutes cable size on the heat soak based on heat soak time. Table J-6 in Section J- methodology.

6 of Report 016015-RPT-06 does not list any damage times from any ignition source -

secondary combustible grouping of 25 minutes.

The database should be checked for additional examples and addressed as necessary.

B. There are a number of scenarios listed in the FMDB indicating HGL timings but there is inconsistent indication for when a scenario credits the thermal heat soak method. The only method to verify that the thermal heat soak method was applied in the FMDB was to query Page 29 of 40

L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR)

FO-2 (cont.) the results in TblIgnitionScenarios and match HGL timing to those reported in Table J-6 in Section J-6 of Report 016015-RPT-06, or to search for comment fields in IgScnComment.

C. Description of the method in which the results from the thermal heat soak analysis is incorporated in the MCA. It is not clear where the MCA heat soak calculations or their direct inputs are among the reviewed material. Section 5.4 of the MCA Report 016015-RPT-08 points to Table J-6 in the SCA Report 016015-RPT-06 for heat soak results. However, the compartments listed in Table J-6 do not completely match the compartments that were screened from the MCA using the heat soak method. This suggests that there may be other heat soak results that are not documented. For example, the MCA screens combinations involving compartments 19B and 32A, but the SCA does not indicate that thermal heat soak analysis was performed for these compartments. In addition it appears that the heat soak method was used to increase the HGL timing for combinations involving compartment 32A, but there is no documentation of the results used to justify this timing.

D. It is difficult to link the CFAST Group and the Fire Case as listed in Table J-6 in the SCA Report 016015-RPT-06 with the damage integral result listed in the database for the SCA and MCA where applied. There is no consolidated table which includes the CFAST Group and the Fire Case as applied to a given scenario in the FMDB.

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L-MT-18-010 NSPM Enclosure F&O Number Description Resolution Disposition for 10 CFR 50.69 (SR)

FO-2 (cont.) E. The thermal heat soak method does not fully document the approach for target damage accumulation at low temperatures. No technical deficiencies were noted in the method review; however, the treatment of the low temperature damage accumulation can have a significant influence on the overall result and should be clearly discussed.

F. Additional documentation of the limits of applicability for the thermal heat soak method is needed in Report 016022-RPT-01. For example, is there a maximum exposure temperature or maximum/minimum cable size over which the results can be used?

G. Documentation of sources of model uncertainty and its treatment in the analysis is needed to achieve a Cat II for FSS-H5 and FSS-H9. Since the heat soak method is an interpolation of the generic cable damage times listed in NUREG/CR-6850, there is no new uncertainty introduced with the heat soak method, except for the damage accrual estimates at temperatures below the damage threshold.

H. Reports 016015-RPT-06, Rev. 4 and 016015-RPT-08, Rev. 4 were draft at the time of the review. They will need to be finalized and signed.

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L-MT-18-010 NSPM Enclosure Attachment 4: External Hazards Screening Screening Result External Screening Screened?

Hazard Criterion Comment (Y/N)

(Note a)

Aircraft Impact Y PS4 The nearest major airport is Minneapolis-St. Paul International (MSP) which is located approximately 45 miles from the site. The IPEEE reports results from bounding assessments to demonstrate that the risk due to this hazard is less than 1.0E-06 per year.

Avalanche Y C3 The topography surrounding MNGP precludes the possibility of a snow avalanche.

Biological Y C5 Actions committed to and completed by MNGP in Event response to Generic Letter 89-13 (Service Water System Problems Affecting Safety-Related Equipment) provide on-going control of biological hazards. These controls are described in MNGP procedure EWI-08.22.01, Generic Letter 89-013. Additionally, actions taken in response to INPO SOER 07-2 (Intake Structure Blockage) provide an additional layer of biological hazard management.

Based on these actions, the hazard is slow to develop and can be identified via monitoring and managed via standard maintenance processes.

Coastal Y C3 The mid-western location of MNGP precludes the Erosion possibility of coastal erosion.

Drought Y C5 These effects would take place slowly allowing time for orderly plant reductions including shutdowns.

External Y C1 The external flooding hazard at MNGP was Flooding recently updated as a result of the post-Fukushima 50.54(f) Request for Information and the flood hazard reevaluation report (FHRR) was submitted to the NRC for review on May 12, 2016 (Reference 19). The results indicate that flooding from rivers and streams are bounded by the current licensing basis and do not pose a challenge to the plant.

Flooding from local intense precipitation was evaluated and will not challenge any safety functions at MNGP.

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L-MT-18-010 NSPM Enclosure Screening Result External Screening Screened?

Hazard Criterion Comment (Y/N)

(Note a)

Extreme Wind Y C1 Wind damage is bounded by tornadoes, and the or Tornado tornado wind speed corresponding to the 1.0E-06 PS4 per year exceedance frequency is less than the MNGP design value. Therefore, damage due to the forces associated with extreme winds or tornadoes can be screened.

For tornado missiles, those areas housing critical equipment required to assure safe shutdown were designed to prevent penetration of exterior walls by two types of missiles that could be generated by a tornado: A) A utility pole 35-feet long by 14-inches in diameter and a unit weight of 35 pounds per cubic foot having a velocity of 200 mph; and B) A one ton missile, such as a compact type automobile, traveling at 100 mph at a maximum height of 25-feet above grade and with a contact area of 25 square feet.

The CDF associated with tornado missiles is less than 1.1 E-07 per year.

Fog Y C1 The principal effects of such events (such as freezing fog) would be to cause a loss of offsite C4 power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for MNGP.

Forest or Y C1 The site landscaping and lack of forestation Range Fire prevent such fires from posing a threat to MNGP.

C3 Furthermore, the principal effects of such events would be to cause a loss of offsite power and are C4 addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for MNGP.

Frost Y C1 The principal effects of such events would be to cause a loss of offsite power and are addressed C4 in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for MNGP.

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L-MT-18-010 NSPM Enclosure Screening Result External Screening Screened?

Hazard Criterion Comment (Y/N)

(Note a)

Hail Y C1 Hail is bounded by other events for which the plant is designed. The principal effects of such C4 events would be to cause a loss of offsite power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for MNGP.

High Summer Y C1 The principal effects of such events would result Temperature in elevated river temperatures which are C4 monitored by station personnel. Should the ultimate heat sink temperature exceed the C5 Technical Specification limit, an orderly shutdown would be initiated. These effects would take place slowly allowing time for orderly plant reductions, including shutdowns. Another potential consequence would be to cause a loss of offsite power. This consequence is addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for MNGP.

High Tide, Y C3 High tide or lake level not applicable to the site Lake Level, or because of location. Impact of High River Stage River Stage C4 is included as an impact to external flooding.

Hurricane Y C3 The mid-western location of MNGP precludes the possibility of a hurricane. Additionally, hurricanes C4 would be covered under Extreme Wind or Tornado and Intense Precipitation.

Ice Cover Y C1 The principal effects of such events would be to cause a loss of offsite power and are addressed C4 in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for MNGP. Furthermore, ice induced flooding was evaluated in the FHRR and determined to be bounded by External Flooding.

Industrial or Y C3 There are no military facilities within the proximity Military Facility to the plant (the closest is a National Guard Accident C4 facility at MSP airport, ~45 miles away). The hazards associated with an industrial facility accident are screened elsewhere in this table (e.g., transportation accident, pipeline accident).

Internal N N/A The MNGP internal events PRA addresses risk Flooding from internal flooding events.

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L-MT-18-010 NSPM Enclosure Screening Result External Screening Screened?

Hazard Criterion Comment (Y/N)

(Note a)

Internal Fire N N/A The MNGP internal fire PRA addresses risk from internal fire events.

Landslide Y C3 Not applicable to the site because of topography.

Lightning Y C1 Lightning strikes can result in losses of offsite power or surges in instrumentation output if C4 grounding is not fully effective. The latter events often lead to reactor trips. Both results are incorporated into the MNGP internal events PRA model through the incorporation of generic and plant-specific data.

Low Lake Y C5 These effects would take place slowly over time Level or River allowing for orderly plant reductions, including Stage shutdowns.

Low Winter Y C1 The principal effects of such events would be to Temperature cause a loss of offsite power. These effects C4 would take place slowly allowing time for orderly plant reductions, including shutdowns. At worst, C5 the loss of offsite power events would be subsumed into the base PRA model results.

Meteorite or Y PS4 The frequency of a meteorite or satellite strike is Satellite judged to be very low such that the risk impact Impact from such events is insignificant.

Pipeline Y C1 The nearest hazardous material or natural gas Accident pipeline is located more than one mile south of the plant. The effects on plant structures due to a pipeline explosion located approximately one mile from the site are bounded by tornado loadings.

Release of Y C4 Chlorination of water systems is performed using Chemicals in a hypochlorite system. No chlorine gas is stored Onsite Storage PS1 on-site. The control room envelope habitability program, as described in Technical Specification 5.5.13 ensures NSPM retains the ability to safely operate the plant in the event of a chemical release onsite or in the vicinity of the plant.

Onsite and offsite chemical hazards are periodically re-evaluated to ensure continued compliance with this specification.

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L-MT-18-010 NSPM Enclosure Screening Result External Screening Screened?

Hazard Criterion Comment (Y/N)

(Note a)

River Y C3 Diversion of the Mississippi River was reviewed Diversion under the Fukushima 10 CFR 50.54(f) Request C5 for Information and the FHRR was submitted to the NRC for review in Reference 19. The location of the MNGP site is adjacent to the natural channel of the Mississippi River. A review of the United States Geological Survey from 1961 and 2013 showed no change in the course of the Mississippi River channels in the site vicinity. The river channel in the area of the site does not include prominent bluffs or other features that could be susceptible to landslide which could potentially result in migration of the channel more directly towards the site. There are no man-made channels, canals, diversions, or permanent levees used for conveyance of water and flood protection near the site.

Sand or Dust Y C1 The frequency of a loss of offsite power accounts Storm for severe weather, including sand or dust C3 storms.

C4 Seiche Y C3 The MNGP site is located on the Mississippi River and is not susceptible to a seiche.

Seismic N N/A See information in Section 3.2.3 of the enclosure.

Activity Snow Y C1 Snow cover is included as an input to the probable maximum flood. Potential impacts are C4 covered under external flooding. Snow loading on buildings was considered during the design of the plant and maintenance procedures exist to ensure snow loading remains within the design load.

Soil Shrink- Y C3 Excluded based on structures founded on Swell bedrock and/or engineered fill.

Consolidation Storm Surge Y C3 The mid-western location of MNGP precludes the possibility of a sea level driven storm surge.

Toxic Gas Y C4 The hazards associated with toxic gas are screened elsewhere in this table (e.g., Release of Chemicals in Onsite Storage)

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L-MT-18-010 NSPM Enclosure Screening Result External Screening Screened?

Hazard Criterion Comment (Y/N)

(Note a)

Transportation Y C1 Land Transportation - Based on the proximity of Accident the nearest major roadways, potential impacts C3 are covered by Extreme Wind or Tornado as well as Release of Chemicals in Onsite Storage.

C4 Rail Transportation - Based on the proximity of the nearest commercial railroad line, potential impacts are covered by Extreme Wind or Tornado as well as Release of Chemicals in Onsite Storage.

Water Transportation - MNGP is located near the headwaters of the Mississippi River. Therefore, the river is shallow near the plant limiting the activity primarily to pleasure craft.

Tsunami Y C3 The mid-western location of MNGP precludes the possibility of a tsunami.

Turbine- Y PS4 Turbine-Generated Missiles are evaluated in Generated USAR Section 12.2.3. The probability of Missiles unacceptable damage from turbine missile has been calculated to be 1.76E-08 per year.

Volcanic Y C3 Not applicable to MNGP.

Activity Waves Y C3 Waves associated with adjacent large bodies of water are not applicable to MNGP. The potential C4 impacts of waves associated with External Flooding was evaluated and wave runup was determined to be below flood protection barriers.

Note a - See Attachment 5 for descriptions of the screening criteria.

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L-MT-18-010 NSPM Enclosure Attachment 5: Progressive Screening Approach for Addressing External Hazards Event Criterion Source Comments Analysis Initial C1 Event damage potential NUREG/CR-2300 Preliminary is less than events for Screening which plant is designed ASME/ANS PRA Standard RA-Sa-2009 C2 Event has lower mean NUREG/CR-2300 frequency and no worse consequences than ASME/ANS PRA Standard other events analyzed RA-Sa-2009 C3 Event cannot occur NUREG/CR-2300 close enough to the plant to affect it ASME/ANS PRA Standard RA-Sa-2009 C4 Event is included in the NUREG/CR-2300 Not used to definition of another screen. Used only event ASME/ANS PRA Standard to include within RA-Sa-2009 another event.

C5 Event develops slowly ASME/ANS PRA Standard allowing adequate time RA-Sa-2009 to eliminate or mitigate the threat Progressive PS1 Design basis hazard ASME/ANS PRA Standard Screening cannot cause a core RA-Sa-2009 damage accident PS2 Design basis for the NUREG-1407 event meets the criteria in the NRC 1975 ASME/ANS PRA Standard Standard Review Plan RA-Sa-2009 PS3 Design basis event NUREG-1407 mean frequency is

< 1.0E-05 per year and ASME/ANS PRA Standard the mean conditional RA-Sa-2009 core damage probability is < 0.1 PS4 Bounding mean CDF is NUREG-1407

< 1.0E-06 per year ASME/ANS PRA Standard RA-Sa-2009 Detailed Screening not NUREG-1407 PRA successful. PRA needs to meet requirements in ASME/ANS PRA Standard the ASME/ANS PRA RA-Sa-2009 Standard.

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L-MT-18-010 NSPM Enclosure Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Assumption/Uncertainty Discussion Disposition Ignition source counting is an The conservatism in the This uncertainty has the area with inherent uncertainty; ignition frequency data, which potential to push more however, the results are not is also linked to conservatism components into the HSS particularly sensitive to in non-suppression probability categorization than LSS.

changes in ignition source data specified in NUREG/CR- Future model enhancement counts. The primary source of 6850, appears to introduce a incorporating NUREG-2169 uncertainty for this task is significant conservatism. will reduce this uncertainty, associated with the frequency but not eliminate it, due to values from NUREG/CR-6850 larger data set developing Supplement 1 which result in frequencies.

uncertainty due to variability among plants along with some significant conservatism in defining the frequencies, and their associated heat release rates, based on limited fire events and fire test data.

Fire PRA Cable Selection is Cable Selection for PRA This uncertainty has the based on risk significance of components was limited potential to push more PRA component. based on risk significance of components into the HSS the Internal Events categorization than LSS.

components impact to Fire Further cable selection to PRA risk. credit PRA equipment in the plant will be prioritized by the risk significance.

Common cause failures are This uncertainty potentially As directed by NEI 00-04, developed using available affects all SSCs evaluated common cause basic events industry data. during 50.69 categorization. are increased to their 95th percentile and also decreased to their 5th percentile values as part of the required 50.69 PRA categorization sensitivity cases. These results are capable of driving a component and respective functions HSS and, therefore, the uncertainty of the common cause failure probabilities are accounted for in the categorization process.

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L-MT-18-010 NSPM Enclosure Assumption/Uncertainty Discussion Disposition The approach taken for the The employment of generic This uncertainty has the Detailed Fire Modeling was: fire modeling solutions did not potential to push more

1) the use of generic fire introduce any significant components into the HSS modeling treatments in lieu of conservatism. Detailed fire categorization than LSS.

conservative scoping analysis modeling was only applied Future model enhancement techniques and 2) limited where the reduction in incorporating NUREG-2178 detailed fire modeling was conservatism was likely to will reduce this uncertainty, performed to refine the have a measurable impact. but not eliminate it, due to scenarios developed using The NUREG/CR 6850 heat larger data set developing the generic fire modeling release rates introduce frequencies.

solutions. The primary significant conservatism given conservatism introduced by the limited fire test data this task is associated with available to define the heat the heat release rates release rates and the specified in NUREG/CR- associated fire development 6850. timeline.

Human Reliability Analysis Human Reliability Analysis As directed by NEI 00-04, (HRA) is a continually human error basic events are evolving discipline. The increased to their 95th human error probabilities percentile and also decreased were obtained using the to their 5th percentile values current EPRI HRA calculator as part of the required 50.69 consistent with the Fire HRA PRA categorization sensitivity Methodology described in cases. These results are NUREG-1921. The Internal capable of driving a Events human error component and respective probabilities were obtained functions HSS and therefore using guidance from the uncertainty of the PRA NUREG/CR-1278 and modeled Human Error NUREG/CR-4772. Probabilities are accounted for in the 50.69 application.

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