ML23312A192

Issuance of Amendment Nos. 316 & 316 Regarding a Risk Informed Approach for Tornado Classification of the Fuel Handling Trolley Support Structure

ML23312A192
Person / Time
Site:	Surry
Issue date:	01/18/2024
From:	John Klos Plant Licensing Branch II
To:	Carr E Virginia Electric & Power Co (VEPCO)
Klos, J
References
EPID L-2022-LLA-0068
Download: ML23312A192 (31)

Text

January 18, 2024

Mr. Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 316 AND 316 REGARDING A RISK-INFORMED APPROACH FOR TORNADO CLASSIFICATION OF THE FUEL HANDLING TROLLEY SUPPORT STRUCTURE (EPID L-2022-LLA-0068)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 316 to Subsequent Renewed Facility Operating License No. DPR-32 and Amendment No. 316 to Subsequent Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2 (Surry), respectively. The amendments are in response to your application dated May 11, 2022, as supplemented by letters dated July 11, 2022, and April 6, 2023.

These amendments approve a risk-informed approach to demonstrate that the fuel handling trolley support structure, as designed, meets the intent of a tornado resistant structure under the current Surry licensing basis for a 360 miles per hour maximum tornado wind speed. The Surry Updated Final Safety Analysis Report is being updated to reflect the approved change.

E. Carr

A copy of the related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice. If you have any questions, please contact me at john.klos@nrc.gov, or 301-415-5136.

Sincerely,

/RA/

John Klos, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 316 to DPR-32

2. Amendment No. 316 to DPR-37

3. Safety Evaluation

cc: Listserv

VIRGINIA ELECTRIC AND POWER COMPANY

DOCKET NO. 50-280

SURRY POWER STATION, UNIT NO. 1

AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE

Amendment No. 316 Subsequent Renewed License No. DPR-32

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated May 11, 2022, as supplemented by letters dated July 11, 2022, and April 6, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, by Amendment No. 316, Subsequent Renewed Facility Operating License No. DPR-32 is hereby amended to authorize revision to the Updated Final Safety Analysis Report of the Surry Power Station, Unit No. 1, as set forth in the licensees application dated May 11, 2022, as supplemented by letters dated July 11, 2022, and April 6, 2023, and evaluated in the NRC staffs safety evaluation enclosed with this amendment.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-32 and the Updated Final Safety Analysis Report

Date of Issuance: January 18, 2024

VIRGINIA ELECTRIC AND POWER COMPANY

DOCKET NO. 50-281

SURRY POWER STATION, UNIT NO. 2

AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE

Amendment No. 316 Subsequent Renewed License No. DPR-37

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated May 11, 2022, as supplemented by letters dated July 11, 2022, and April 6, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, by Amendment No. 316, Subsequent Renewed Facility Operating License No. DPR-32 is hereby amended to authorize revision to the Updated Final Safety Analysis Report of the Surry Power Station, Unit No. 2, as set forth in the licensees application dated May 11, 2022, as supplemented by letters dated July 11, 2022, and April 6, 2023, and evaluated in the NRC staffs safety evaluation enclosed with this amendment.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes License No. DPR-37 and the Updated Final Safety Analysis Report

Date of Issuance: January 18, 2024

ATTACHMENT

SURRY POWER STATION, UNIT NOS. 1 AND 2

LICENSE AMENDMENT NO. 316 TO

SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32

DOCKET NO. 50-280

AND

LICENSE AMENDMENT NO. 316 TO

SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37

DOCKET NO. 50-281

Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages

License License

License No. DPR-32, page 3 License No. DPR-32, page 3 License No. DPR-37, page 3 License No. DPR-37, page 3

TSs TSs None None

3. This subsequent renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level

The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 316 are hereby incorporated in the subsequent renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports

The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records

The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 65

F. Deleted by Amendment 71

G. Deleted by Amendment 227

H. Deleted by Amendment 227

Surry - Unit 1 Subsequent Renewed License No. DPR-32 Amendment No. 316

3. This subsequent renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level

The licensee is authorized to operate the facility at steady state reactor core power Levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 316 are hereby incorporated in this subsequent renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports

The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records

The licensee shall keep facility operating records in accordance with the Requirements of the Technical Specifications.

E. Deleted by Amendment 54

F. Deleted by Amendment 59 and Amendment 65

G. Deleted by Amendment 227

H. Deleted by Amendment 227

Surry - Unit 2 Subsequent Renewed License No. DPR-37 Amendment No. 316 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO

AMENDMENT NO. 316 TO SUBSEQUENT RENEWED

FACILITY OPERATING LICENSE NO. DPR-32

AND

AMENDMENT NO. 316 TO SUBSEQUENT RENEWED

FACILITY OPERATING LICENSE NO. DPR-37

VIRGINIA ELECTRIC AND POWER COMPANY

SURRY POWER STATION, UNIT NOS. 1 AND 2

DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated May 11, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22131A351), as supplem ented by letters dated July 11, 2022, and April 6, 2023 (ML22192A075 and ML23096A197, respectively), Virginia Electric and Power Company (Dominion Energy Virginia, the licen see) submitted a license amendment request (LAR) for the Surry Power Station, Units 1 and 2 (Surry). The requested amendment would apply a risk-informed approach to demonstrate that the fuel handling trolley support structure (FHTSS), as designed, meets the intent of a tornado resistant structure under the current Surry licensing basis for a 360 miles per hour (mph) maximum tornado wind speed.

The supplements dated July 11, 2022, and April 6, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on September 6, 2022 (87 FR 54553).

2.0 REGULATORY EVALUATION

2.1 Description of the FHTSS

The fuel building (FB) is located between the auxiliary building, decontamination building, and the units 1 and 2 containment buildings. The lower portion of the FB is comprised of the spent

Enclosure 3

fuel pool (SFP) with reinforced concrete walls, and the upper portion of the FB is comprised of a steel structure that provides a weather cover for the spent fuel assemblies in the SFP. The steel structure includes the FHTSS that houses the FBs 125-ton fuel cask trolley. Attachment 2 of the LAR (ML22131A351) provides the steel framing drawings of the FB and FHTSS and attachment 1 of the LAR states that station pr ocedures require the FBs 125-ton fuel cask trolley to be moved to the crane enclosures north, outside of the FB, during abnormal weather conditions to prevent it from potentially falling into the SFPs cask loading area of the spent fuel pool (SFP) during a tornado.

2.2 Description of the Licensees Proposed Change

The licensee requested NRC approval to apply its proposed risk-informed approach to demonstrate that the FHTSS, as designed, meets the intent of a tornado resistant structure (i.e., tornado criterion T) under the current Surry licensing basis for a 360-mph maximum tornado wind speed. Surrys Updated Final Safety Analysis Report (UFSAR), Revision 54 (ML22283A015), Table 15.2-1, Structures, Systems, and Components Designed for Seismic and Tornado Criteria, states that tornado criterion T refers to structures that will not fail during the design tornado.

The proposed amendment is needed because the existing design-basis calculation for the FB does not consider a maximum tornado wind speed of 360 mph as stated in the UFSAR. The licensee stated that the UFSAR will be updated upon approval of the LAR to reflect the approved methodology change that demonstrates the FHTSS meets the intent of a tornado criterion T structure.

The licensee stated that no changes to the Surry technical specifications are required.

2.3 Description of Regulatory Requirements and Guidance

The General Design Criteria (GDC) included in Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50 became effective on May 21, 1971. The construction permits for Surry were issued prior to May 21, 1971; consequently, Surry is not subject to the current GDC requirements. Surrys UFSAR Section 1.4.2, Performance Standards; Section 1.4.40, Missile Protection; Section 2.2.2.1, Tornadoes; and Section 15.2.3, Tornado Criteria, meet the intent of GDC 2, Design bases for protection against natural phenomena, and GDC 4, Environmental and dynamic effects design bases.

Surry UFSAR, Section 1.4.2 states, in part, that:

Those systems and components of reactor facilities that are essential to the prevention of accidents that could affect the public health and safety or to the mitigation of their consequences are designed, fabricated, and erected in accordance with performance standards that enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established reflect (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area, and (b) an appropriate margin for withstanding forces greater than those recorded, in view of uncertainties about the historical data and their suitability as a basis for design.

Surry UFSAR, Appendix 14B.1.1 states, in part, that:

The analysis ensures that the Commissions General Design Criterion 4 is met, i.e., that all structures, systems, and components important to safety are designed to accommodate the effects of and are compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents (LOCAs).

These structures, systems, and components are protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids that may result in equipment failures and from events and conditions outside the nuclear power unit.

The regulatory requirements and guidance applicable to this LAR are listed below.

Regulations

The regulations in 10 CFR Part 50, Appendix A, GDC 2, Design bases for protection against natural phenomena, state, in part, that, structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.

The regulations in 10 CFR Part 50, Appendix A, GDC 4, Environmental and dynamic effects design bases, state, in part, that, structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant [LOCA]

accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

The regulations in 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, state, in part, that monitoring the performance or condition of structures, systems, or components, against licensee-established goals, shall be in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions.

The regulations in 10 CFR 50.67, Accident source term, state, in part, that a licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under § 50.90. The

application shall contain an evaluation of the consequences of applicable design basis accidents1 previously analyzed in the safety analysis report.

Guidance

Regulatory Guide (RG) 1.25, Revision 0, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors, March 1972 (ML083300022), provides guidance that is acce ptable to the NRC staff for acceptable assumptions that may be used in evaluating the radiological consequences of a fuel handling accident in the fuel handling and storage facility.

RG 1.76, Revision 1, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants, March 2007 (ML070360253), provides guidance that is acceptable to the NRC staff for use by licensees in selecting the design-basis tornado and design-basis tornado-generated missiles that a nuclear power plant should be designed to withstand to prevent undue risk to the health and safety of the public. RG 1.76 states that nuclear power plants must be designed so that they remain in a safe condition under severe meteorological events, including those that could result in the most severe tornado that could reasonably be predicted to occur at the site. The design-basis tornado wind speeds presented in this RG are based on NUREG/CR-4461, Revision 2, Tornado Climatology of the Contiguous United States (ML070810400), which examines the implications of switching from the Fujita Scale to the Enhanced Fujita Scale on design wind speed estimates for tornadoes.

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256), describes an approach that is acceptable to the NRC staff for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights.

RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ML003716792), provides guidance to licensees of operating power reactors on acceptable applications of Alternate Source Term (AST); the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. RG 1.183 establishes an acceptable AST and identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. RG 1.183 also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants, February 2001 (ML010430066), documents the NRC staffs evaluation of the potential accident risk in a SFP at decommissioning plants and use of sensitivity studies related to the acceptance guidelines in RG 1.174. In

1 The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of design analyses or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.

decommissioning, core damage frequency (CDF) is less relevant as all fuel would be offloaded from the core into the SFP such that large early release frequency (LERF) would be the more applicable risk measure for evaluation.

NUREG-2191, Revision 1, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report,Section X I.S6, Structures Monitoring (ML16274A389 -

Volume 1; ML16274A399 - Volume 2) provides guidance on the content of applications for the renewal concerning plant management programs and the technical basis for adequacy of those programs.

NUREG/CR-4461, Revision 2, Tornado Climatology of the Contiguous United States, February 2007 (ML070810400), provides wind speed estimates in this climatology based on the Enhanced Fujita Scale that correlates wind speeds with damage caused by tornadoes.

3.0 TECHNICAL EVALUATION

3.1 Engineering Evaluations

3.1.1 Evaluation of Tornado Wind Speeds

Section 15.2.3 of the UFSAR (ML22283A015), specifies that the existing licensing basis tornado characteristics used for design have a rotational velocity of 300 mph, translational velocity of 60 mph, pressure drop of 3 psi (pounds per square inch) in 3 seconds, overall diameter of 1200 feet, and radius of maximum winds of 200 feet. Therefore, applicable structures are designed to resist a maximum wind velocity associated with a tornado of 360 mph (300 mph rotational velocity + 60 mph translational velocity). This is commonly referred to as the maximum licensing basis tornado.

As part of its proposed risk-informed approach, the licensee calculated tornado initiating event frequencies using data from NUREG/CR-4461, Table 6-1, Tornado Wind Speed Estimates for United States Nuclear Power Plant Sites. The NRC staff verified that table 6-1 presents tornado wind speed estimates using the Fujita scale for events with a frequency of 1.0E-5, 1.0E-6, and 1.0E-7 events per year for each power plant. In its letter dated May 11, 2022, Section 3.2, Initiating Event Frequencies, the licensee stated that it used these Fujita tornado wind speed estimates for those specific frequencies since those values, at that frequency, are higher and more conservative than the wind speed estimates at the same corresponding frequencies stated at the enhanced Fujita scale level.

The licensee also stated that this data from NUREG/CR-4461, Table 6-1, was used with the process described in Nuclear Energy Institute (N EI) 17-02, Revision 1B, Tornado Missile Risk Evaluator (TMRE), as implemented and approved at the Shearon Harris Nuclear Power Plant (ML18347A385), to develop a site-specific hazard curve to describe the expected frequency of tornadoes that exceed wind speeds from 70 mph to 360 mph in 10 mph increments. The licensee selected this range because it encompasses th e range of high wind events considered in Surrys UFSAR.

The NRC staff evaluated the licensees calculation of the site-specific hazard curve and finds the licensees calculation of the site-specific hazard curve acceptable because (1) the licensee used the appropriate site-specific data from RG 1.76 and NUREG/CR-4461; (2) the licensee used the more conservative original Fujita scale; and (3) the licensees calculation utilizes

NEI 17-02, Revision 1B, dated September 19, 2018, which is incorporated by reference into the licensees UFSAR and the current licensing basis which serves as an update to NEI 17-02, Revision 1, reflective of a Surry site-specific TMRE methodology.

3.1.2 Evaluation of Structural Capacity of the FHTSS

Evaluation of the Median Wind Speed Capacity of the FHTSS

The licensee performed an analysis to determine the FHTSS probability of collapse. Utilizing the tornado wind exceedance frequencies from NUREG/CR 4461, Table 6-1, discussed above, the licensee developed FHTSS tornado wind fragility curves using the median wind speed capacity of the FHTSS.

In its letter dated, May 11, 2022, Section 3.1, Tornado Wind Fragility for the FHTSS, the licensee stated, in part, that,

The median tornado wind speed capacity is defined as the maximum tornado wind speed a structure can withstand based on its median capacity. The median capacity of each structural member of the FHTSS is defined as its unfactored (nominal) strengthusing median linear elastic material properties. Based on results from analysis of the FHTSS computer model, the wind speed that causes at least one (1) structural member to reach the maximum interaction ratio (i.e.,

demand-to-capacity ratio equal to 1.0) is considered to be the median wind speed capacity of the structure.

This was based on the results of a three-dimensional (3-D) FHTSS computer model, which was evaluated following the guidelines of American Society of Civil Engineers (ASCE) publication, Wind Load Design for Petrochemical and Other Industrial Facilities, 2 nd Edition, Appendix 5A.

The licensee further stated that,

the drag coefficients for the wind load evaluations used in the design basis calculations of the FB [fuel building] were based on ASCE 3269 [Wind Forces on Structures]. In the design basis calculations of the FB, the drag coefficients were conservatively selected from Table 3 of ASCE 3269 [7] for infinitely long structural members. Application of these drag coefficients to structural members comprising the 3-D frame structure of the FHTSS is conservative, since this approach ignores the shielding effect of closely spaced structural members.

In Section 3.1 of its letter dated May 11, 2022, the licensee stated, in part, that,

Based on results from the analysis of the FHTSS computer model, a median wind speed capacity of 210 mph was calculated. A comparative study showed the FHTSS will experience a much smaller base shear and moment due to a 150 mph wind acting on the FHTSS with siding intact, as compared to the base shear and moment due to a 210 mph wind applied on the bare structural steel framing of the FHTSS. Therefore, the calculated fragility with a median wind speed capacity of 210 mph and the corresponding risk will not be affected by the presence of siding during tornado wind speeds of 150 mph or lower.

In Section 3.2,Initiating Event Frequencies, of its letter dated May 11, 2022, the licensee stated that,

Initiating event frequencies for tornadoes were developed using the data in Table 6-1 of NUREG/CR 4461, Rev. 2 [4]. Tornado wind speed estimates from the Fujita Scale for events with a frequency of 1.0E-5, 1.0E-6, and 1.0E-7 events per year were used since these data are more conservative than the Enhanced Fujita Scale. This data was used with the process described in NEI 17-02, Rev. 1 B [12]

to develop a site-s pecific hazard curve to de scribe the expected frequency of tornadoes that exceed wind speeds from 70 to 360 mph in 10 mph increment s.

This range was selected be cause it encompasses the range of high wind events considered in the SPS UFSAR [3]. The estimate d initiating event frequencies range from 9.37E-5/yr for the smallest Fujita prime F'1 tornado inte nsity to 6.83E-10/yr for the largest F'6 tornado intensity.

In Section 3.3, Tornado Wind Fragility Estimates, of its letter dated May 11, 2022, the licensee stated that,

The lognormal tornado wind fragility curve described above was used to produce fragility estimates for each tornado wind speed interval in the scope of the analysis. These fragility estimates ranged from 2.01 E-9 at 70 mph to 0.445 at 200 mph. The fragility was conservatively assumed to be 1.0 at wind speeds above the calculated median wind speed capacity of 210 mph (i.e., the FHTSS was conservatively assumed to collaps e at wind speeds above the calculated median wind speed capacity of 210 mph).

In Section 3.6, Treatment of Uncertainty, of its letter dated May 11, 2022, the licensee stated, in part, that,

Conservatism was used throughout this analysis to ensure the estimates maintained a conservative bias to bound these uncertainties. For the event frequencies, credit was not taken for the reduced frequencies characterized by the Enhanced Fujita Scale data in NUREG/CR 4461, Rev. 2 [4]. For the fragilities, the structural failure rate was assumed to be 1.0, which means the FHTSS is assumed, conservatively, to collapse at wind speeds above the calculated median wind speed capacity of 210 mph. This is a conservative assumption because structures reaching their median capacity do not necessarily fail until they develop enough plastic hinges to initiate instability. With regard to impact to fuel, spent fuel damage was assumed to take place any time the analysis indicated a structural member would experience plastic deformation, and no credit was taken for post-yield capacity of structural members. The conservative bias resulting from the use of these conservative assumptions ensures the results can be considered reasonable despite the presence of uncertainties.

Additionally, a sensitivity study was performed by doubling the composite variability in the fragility curve from 0.175 to 0.35. The SFDF in the sensitivity study was estimated to be 7.56E-06/yr. This demonstrates that the risk assessment is sensitive to the composite variability, but the incremental SFDF would still be assessed as 'small' per the guidelines in RG 1.174, Rev. 3 [1] even if the variability was twice the typical value.

The NRC staff independently reviewed the licensee s calculation and confirmed the licensees assumptions were conservative. The NRC staff c oncludes that the licensees determination of a

median wind speed capacity of 210 mph on the bare structure steel frame bounds the 150-mph requirement of the structure with siding intact and is, therefore, acceptable. The NRC staff also noted that NUREG/CR-4461, Section 2.1, Tornado Intensities, states that, the maximum wind speed associated with a tornado occurs for only a small fraction of the footprint and at any location for only a short period of time. Depending on the translational velocity of the tornado, a location might experience wind speeds near the maximum for 5 to 10 seconds. This suggests that a tornado impacting the FHTSS structure directly with the maximum wind speed will be unlikely given that the FHTSS footprint is very small relative to the overall powerplant site, adding to the conservatism in the licensees determination of the FHTSS median wind speed capacity.

Evaluation of the Postulated Collapse of the FHTSS

In its letter dated July 11, 2022, the licensee addressed the likelihood of falling structural members of the FHTSS targeting the SFP with an angle of attack that could result in localized SFP damage. The licensee stated, in part, that, When structural members of the FHTSS fail under the applied tornado wind loads, structural members of the FHTSS could separate from the connecting members of the FHTSS and fall into the SFP. However, not all structural members of the FHTSS would cause the worst-case impact to the SFP for the following reasons: The licensee stated, in part, that,

1) Some structural members are not likely to fail under tornado winds because they are oriented parallel to the governing wind direction (i.e., north-south direction) and do not have a large sail area against the tornado wind.

2) Some structural members of the FHTSS (e.g., columns) are longer than the width of the pool. These members will not fit into the pool if they fail along the north-south direction.

3) Structural members may experience plastic deformations but not reach their rupture strain and will therefore remain connected to the FHTSS without falling into the SFP.

4) Members oriented along the east-west direction on the north and south ends of the FHTSS are located outside the SFP envelope. Even if these structural members fail and get separated from the FHTSS, they are not likely to fall into the pool.

5) Structural members that fall into the SFP can target the floor mat or walls with different angles of attack. An angle of attack that engages a large impact area (e.g., if members hit the SFP horizontally) will results in the impact load being distributed over a large area such that little or no localized damage will occur.

Therefore, while no quantitative analysis has been performed, the probability that structural members of the FHTSS will fail under the tornado wind load, break at both end connections, and target the SFP with the worst-case angle of attack is judged to be very low.

6) It is not credible for multiple members to fall and impact the same location of the SFP simultaneously.

7) The FHTSS is shielded against tornado winds in the west direction by the presence of the Unit 2 Containment Building. This significantly reduces the likelihood that falling members from the FHTSS will impact Region 2 of the SFP.

In its letter dated, May 11, 2022, Section 3.4, Probabilistic Assessment of Impact to Spent Fuel Damage Frequency, the licensee stated that its evaluation showed potential falling structural members from the FHTSS, due to a postulated collapse of the FHTSS under tornado winds, will not cause a perforation in the reinforced concrete walls or the floor mat of the SFP.

In its letter dated July 11, 2022, the licensee provided additional information on the analysis demonstrating the worst-case impact of a single falling member of the FHTSS will not result in perforation of the reinforced concrete walls or the floor mat of the SFP. The licensee explained its use of the Modified National Defense Research Committee method to estimate local effects (i.e., penetration, perforation, and back-scabbing) on the reinforced concrete members due to the heaviest structural member of the FHTSS falling into the SFP. Although this method is best suited for missiles with velocities greater than 500 ft/s, the licensee used it to estimate the upper bound of damage from falling structural members. The licensee identified conservative assumptions made in the analysis such as ignoring the effects of water on the velocity of the falling member and ignoring the presence of the steel liner plate. Even with these conservative assumptions, the licensee stated that the calculation demonstrated no perforation or back-scabbing of the concrete. The NRC staff independently reviewed the licensees calculation for the worst-case impact of a single falling member of the FHTSS onto the reinforced concrete walls or the floor mat of the SFP during t he audit. The NRC staff independently confirmed that the analysis demonstrated that a falling member of the FHTSS structure has an inconsequential impact on the SFP liner and reinforced concrete walls and floor mat. Based on the above, the NRC staff finds that the worst-case single falling member of the FHTSS will not result in perforation of the reinforced concrete walls or the floor mat of the SFP.

In Section 3.4 of the licensees letter dated, May 11, 2022, the licensee stated that,

A potential radioactive release resulting from spent fuel damage is expected to be bounded by the fuel accidents analyzed under the design basis, because it conservatively assumed in the cask-drop analysis that all 324 fuel assemblies stored in the first three rows of storage racks adjacent to the cask loading area, under the FHTSS, would fail.

The licensee also stated that,

Falling members from the FHTSS may cause local perforation of the SFP liner, especially if these members hit the SFP along a sharp edge. However, no significant leakage, i.e., more than 5 gallon per minute (gpm) as described in Section 9.12.4.13 of SPS UFSAR [3] for the cask-drop in the SFP, is expected due to a potential SFP liner perforation under the worst-case impact of a single falling member of the FHTSS because there will be no perforation (i.e., through cracking) in the reinforced concrete walls or the floor mat of the SFP.

Additionally, UFSAR, Section 9.12.4.13 states that the normal makeup capability from the primary-grade water system is 200 gpm and an emergency source of makeup is available from the fire main at a rate of up to 2000 gpm.

In its letter dated July 11, 2022, the licensee provided additional information on the comparison of potential falling structural members of the FHTSS by analogy to the cask-drop analysis. The licensee stated that,

The kinetic energy due to the free-fall of the heaviest structural member of the FHTSS is less than 20% of the kinetic energy due to the cask-drop analysis that is transferred to the floor mat or walls. Therefore, the overall effects of the impact from falling structural debris would be bounded by those due to the postulated cask-drop accident in the SPS safety analysis.

The NRC staff independently reviewed the licens ees analysis comparing potential falling structural members of the FHTSS to the cask-drop analysis during the audit and supplemental information provided in the licensees letter dated July 11, 2022, The NRC staff confirmed that the analysis demonstrated that the overall effects of the impact from falling structural debris would be bounded by those due to the postulated ca sk drop. Therefore, the NRC staff finds that the overall effects of the impact from potential falling structural members of the FHTSS are bounded by those due to the postulated cask-drop accident in the USFAR.

3.1.3 Evaluation of Spent Fuel Damage Frequency

In Section 3.4 of its letter dated May 11, 2022, the licensee calculated a spent fuel damage frequency (SFDF) based on the convolution of the site-specific tornado wind hazard curve and a tornado wind fragility curve with a median wind speed capacity of 210 mph and a generic composite variability of 0.175. The licensee also performed a sensitivity study by doubling the composite variability in the fragility curve from 0.175 to 0.35.

As discussed in Section 3.1.1 of this safety evaluation, the licensee used the original Fujita scale for the tornado wind hazard. Although the original Fujita scale was replaced with the enhanced Fujita scale in 2007, it is more conservative than the enhanced Fujita scale. As discussed in Section 3.1.2 of this safety evaluation, the licensee conservatively assumed the fragility to be 1.0 for wind speeds above the median wind speed capacity of 210 mph. The NRC staff finds this approach to be acceptable because it is conservative to use original Fujita scale and assume a fragility of 1.0 for wind speeds above the median wind speed capacity of 210 mph.

The licensee calculated the SFDF to be 1.97E-6 per year, which was verified independently by the NRC staff. The NRC staff performed an additional sensitivity analysis using the enhanced Fujita scale and determined that the SFDF would be less than 1E-6 per year.

The NRC staff evaluated the licensees calculation of SFDF of 1.97E-6 per year and finds the licensees calculation of SFDF acceptable because (1) the licensee conservatively used the original Fujita scale; (2) the licensee conservatively assumed a fragility of 1.0 for wind speeds above the median wind speed capacity of 210 mph; and (3) the licensee selected a reasonable value of 0.175 for the composite variability and performed a sensitivity study using a composite variability of 0.35.

3.1.4 Evaluation of Consequence Analysis

To support the risk-informed approach, the licensee evaluated the radiological consequences of the FHTSS falling into the SFP and impacting decayed spent fuel using two methods. The first analysis is referred to as the design-basis analysis, which is intended to demonstrate to the

NRC staff reasonable assurance that safety-related structures, systems, and components will be maintained to mitigate the consequences of an accident and be compliant with the acceptance criteria of 10 CFR 50.67, Accident source term, in terms of total effective dose equivalent (TEDE). The second part of the analysis is a bounding Level 3 PRA calculation, as discussed in RAI APLC-1: Demonstration of RG 1.174 Acceptance Guidelines, is based on assumptions that are considered to provide a realistic, but still conservative, estimate of radiological consequences in terms of population-dose.

The licensee used the analyses to demonstrate that the potential radiological consequences of the FHTSS falling into the SFP are bounded by the NRC-approved design-basis accident analysis. The results of the analyses are compared to allowable risk increase values for individual early fatality risk and individual latent cancer fatality risk inferred from the RG 1.174 acceptance guidelines.

NUREG-1738 documents the NRC staffs evaluation of the potential accident risk in a SFP at decommissioning plants. Further, NUREG-1738 de scribes the acceptance guidelines in RG 1.174 pertaining to the CDF and LERF and the acceptable values for changes due to regulatory decisions as a function of the baseline frequencies.

Appendix 4C of NUREG-1738 states, in part, that,

For decommissioning plants, the risk is primarily because of the possibility of a zirconium fire associated with the spent fuel cladding. The consequences of such an event do not equate directly to either a core damage accident or a large early release as modeled for an operating reactor. Zirconium fires in SFPs have the potential for significant long-term consequences because: there may be multiple cores involved; the relevant clad/fuel degradation mechanisms could lead to increased releases of certain isotopes (e.g., short-lived isotopes such as iodine will have decayed, but the release of longer-lived isotopes such as ruthenium could be increased because of air-fuel reactions); and there is no containment surrounding the SFP to mitigate the consequences. On the other hand, they are different from a large early release because the postulated accidents progress more slowly, allowing time for protective actions to be taken to significantly reduce early fatalities (and to a lesser extent latent fatalities). In effect, an SFP fire would result in a large release, but this release would not generally be considered early because of the significant time delay before fission products are released.

NUREG-1738 uses the LERF acceptance guidelines from RG 1.174 to establish a pool performance guideline (PPG), which provides a threshold for controlling the risk from a plant SFP. Appendix 4C of NUREG-1738 states, in part, that, By maintaining the frequency of events leading to uncovery of the spent fuel at a value less than the recommended PPG value of 1x10 -5 per year, zirconium fires will remain highly unlikely, the risks will continue to meet the Commissions Quantitative Health Objectives, and changes to the plant licensing basis that result in very small increases in LERF may be permitted consistent with the logic in RG 1.174.

In addition, table 1 of Appendix 4D of NUREG-1738 i ndicates risk increase values of 8.7E-8 per year for individual early fatality risk at 1 mile, 6.9E -8 per year for individual latent cancer fatality risk at 10 miles, and population dose of 11 person-rem [Roentgen equivalent man] per year per year within 50 miles in accordance with RG 1.174 acceptance guidelines.

Evaluation of the Design-Basis Analysis

The licensees design-basis analysis concluded that the radiological consequences of falling members from the FHTSS on the spent fuel assemblies are bounded by the fuel handling accident in the SFP, as described in UFSAR Section 14.4.1.3, Fuel-Handling Accident in the Spent-Fuel Pool.

The NRC approved the current design-basis accident analysis in license amendment nos. 230 and 230 (ML020710159) for Surry, Units 1 and 2, respectively, using an AST that meets 10 CFR 50.67 and complies with the guidance stated in RG 1.183, Revision 0. The NRC staffs safety evaluation (SE) approving amendment nos. 230 and 230, in Table 3, Parameters and Assumptions Used in Radiological Consequence Calculations Fuel Handling Accident, states the licensees fuel handling accident analysis assumptions that:

The fuel handling accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown (i.e., the calculated fission product decay period), which is the minimum amount of time between the shutdown of a unit and initiation of fuel movement allowed by Technical Specification 3.10.

The radioactive inventory is based on maximum full power operation at the end of core life immediately preceding shutdown and a conservative radial peaking factor, which is applied to all fuel rods in the assembly, but varies slightly depending on whether the assembly is once-burned, twice-burned, or thrice burned in the core.

The fraction of fission products which migrate from the fuel matrix to the gap and plenum regions of the fuel rods during normal operation is available for immediate release into the water in the event of clad damage.

The core average activity per fuel assembly was multiplied by the peaking factor and then by the gap fractions prescribed PNNL-18212, Update of Gap Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard, Revision 1, as incorporated into Section 14.4.1.2.1 of the Surry UFSAR.

In amendment nos. 230 and 230, the NRC staffs SE states, in Section 2.2 Fuel Handling Accident, that the licensee assumed that all of the 204 fuel rods in one fuel assembly have failed instantaneously, releasing all fission products in the fuel gap to the water surrounding the fuel assemblies.

In its letter dated July 11, 2022, the licensee stated that the worst result for the fuel handling accident in the SFP occurs at the exclusion area boundary and is 3.2 rem TEDE over a 2-hour period assessed over a 30-day period. The regulatory acceptance criterion for the exclusion area boundary is 25 rem.

The results of the fuel handling accident bound the results of the cask-drop analysis, as discussed in UFSAR Section 9 and Appendix 9B. The cask-drop analysis includes the following list of inputs and assumptions, as stated in the licensees letter dated July 11, 2022:

The spent fuel in Region 1 of the SPS SFP was treated as having been discharged for 150 days per SPS Technical Specification (TS) 5.3, Fuel Storage, during spent fuel cask handling. Administrative controls associated with TS 5.3

are in place to preclude storage of fuel with <150 days of decay time in Region 1 at all times.

It was assumed that all damaged fuel assemblies had an initial enrichment of up to 5% by weight U-235, burnup of 60,000 MWD/MTU, core average power of 2546 MWt, and a core average relative power defect (RPD) of 1.2. (Note: The SPS current rated thermal power is 2587 MWt. The 1.2 RPD previously applied ensures the increased MWt rating and calorimetric uncertainty remain bounded.)

It was assumed that 400 fuel assemblies were damaged. This would account for all the fuel assemblies in Region 1 (324) with an allowance for up to 76 additional fuel assemblies in the cask being damaged.

Gap Fractions are obtained from RG 1.25.

Dose conversion factors are obtained from RG 1.109.

A breathing rate of 3.47E-4 m3/sec was assumed.

An EAB X/Q of 2.10E-3 sec/m3 was used. (Note: the EAB X/Q for the FHA is 1.02E-3 sec/m3 as documented in UFSAR Table 14.4-2.)

A pool Decontamination Factor (DF) of 100 was assumed for the removal of iodine.

The fuel assembly source term was derived using ORIGEN2, and the power, burnup, enrichment and decay assumptions described above included contributions from the following table:

Source Term (Ci/fuel assembly)

Kr-85 7.445E+3 Xe-131m 5.084E+1 Xe-133 2.300E+0 I-129 2.090E-2 I-131 1.246E+0 I-132 4.272E-4

In its letter dated July 11, 2022, the licensee also stated that the whole body immersion dose for the cask-drop accident in the FB at the exclusion area boundary is 0.97 rem whole body and 0.57 rem inhalation thyroid. The regulatory acceptance criteria are 25 rem whole body and 300 rem thyroid.

The NRC staff independently confirmed the current analysis of record was unchanged for this LAR and finds the use of the current analysis of record to evaluate the radiological consequence for this LAR is reasonable and appropriate as all methods, inputs, and assumptions remain unchanged. The current analysis of record is sufficient to address uncertainties in accident progression, fission product transport, and atmospheric dispersion. The current analysis of record also includes reasonable assumptions of a maximized fission product release major fuel rod damage, maximum allowable leak rates, a postulated single failure of any of the fission

product cleanup systems, adverse site meteorological dispersion characteristics, and the dose-receptor presumed to be an individual located at the exclusion area boundary and low population zone without protective personal equipment or emergency response actions. The NRC staff reviewed the licensees calculation of the radiological consequences concerning FHTSS falling members onto the spent fuel assemblies and finds that the licensees calculation is acceptable because the results are bounded by the fuel handling accident, analysis of record, for the SFP and would continue to meet control room dose limits.

Evaluation of Early Fatality Risk and Prompt Fatality Risk

The licensee compared the bounding result for the fuel handling accident in the SFP of 3.2 rem TEDE over a two-hour period assessed over a 30-day period to the 200-rem dose required for an early fatality. The licensee estimated the early fatality risk to be zero as the area around the facility is uninhabited, which is lower than the inferred limit of 8.7E-8 early fatalities per year in NUREG-1738 from the RG 1.174 acceptance guidelines.

The NRC staff compared the early fatality risk estimate to the NUREG-1738, Appendix 4C, risk increase value of 8.7E-8 per year for individual early fatality risk and finds that the proposed change meets the RG 1.174 acceptance guidelines.

Evaluation of Population Dose and Latent Cancer Risk

The licensees analysis applied the bounding fuel handling accident source term to evaluate population dose, and subsequently latent cancer risk, using Level 3 PRA techniques. In its letter dated April 6, 2023, the licensee stated that,

The year 2030 projected population was used from the previous Level 3 PRA analysis that was performed to support the 40-to-60-year license extension.

Population trends over the last thirty years were reviewed, and exponential growth was modeled to project the population forward to year 2100. This projected population was used to assess potential dose to the population in the 50 miles around the plant.

The source term was combined with the population to estimate the integrated population dose in every direction around the plant. This resulted in a total of 1.65E5 person-rem as the bounding consequence of the high-wind induced failure of the FHTSS, which equates to an average of 0.03 rem per person.

Multiplying the population dose with the SFDF frequency of 1.97E-6 gives a risk impact associated with the proposed change in terms of increase in dose of 0.325 p[person]-rem I yr. This value compares favorably with the RG 1.17 4 equivalent limit of 11 p-rem I yr. used in NUREG-1738.

Based on the above, the NRC staff finds that the proposed change meets both guidelines of risk increase values of 11 person-rem per year and 6.9E-8 per year for individual latent cancer fatality risk inferred from RG 1.174.

3.

1.5 NRC Staff Conclusion

on Engineering Evaluations

The NRC staff evaluated the licensees evaluations of tornado wind speeds, structural capacity of the FHTSS, SFDF, and radiological consequence analysis. The NRC staff determined that the FHTSS is designed to (1) withstand the effects of natural phenomena such as tornadoes

and (2) accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. Based on the discussion above, the NRC staff finds that the proposed change meets the plant design criteria in the UFSAR and, thus, the intent of GDC 2 and GDC 4 for protection against natural phenomena.

3.2 Risk-Informed Evaluation

RG 1.174 describes an approach that is acceptable to the NRC staff for developing risk-informed applications for a licensing basis change and provides five key principles that licensing basis changes are ex pected to meet. The NRC staff evaluated the LAR against these five key principles.

3.2.1 Principle 1 - Conformance with Current Regulations

The first key principle of risk-informed decision-making is that the proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12).

In its letters dated July 11, 2022, and April 6, 2023, the licensee addressed the five key principles of risk-informed licensing basis changes. The licensee also stated that, under the proposed change, the current regulations continue to be met and the FHTSS would continue to meet design and construction requirements.

As described in Sections 3.1.4 and 3.1.5 of this SE, the NRC staff determined that the design of the FHTSS continues to meet control room dose criteria in 10 CFR 50.67 and plant design criteria in the UFSAR and, thus, the intent of GDC 2 and GDC 4 for protection against natural phenomena. Based on the above, the NRC staff finds that the proposed change meets the first key principle of risk-informed decision-making.

3.2.2 Principle 2 - Defense in Depth

The second key principle of risk-informed decision-making is that the proposed licensing basis change is consistent with the defense-in-depth philosophy. Defense-in-depth is an approach to designing and operating nuclear facilities involving multiple independent and redundant layers of defense to compensate for human and system failures. Section C.2.1.1.2, Considerations for Evaluating the Impact of the Proposed Licensing Basis Change on Defense in Depth, of RG 1.174 provides the following seven considerations that should be used to evaluate the impact of the proposed change on defense-in-depth:

1. Preserve a reasonable balance among the layers of defense.

2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

3. Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

4. Preserve adequate defense against potential common-cause failures (CCFs).

5. Maintain multiple fission product barriers.

6. Preserve sufficient defense against human errors.

7. Continue to meet the intent of the plants design criteria.

In its letter dated April 6, 2023, the licensee addressed each of the seven considerations and how its risk-informed assessment is consistent with the philosophy of defense-in-depth. The following sections provide the NRC staffs eval uation of each of the seven considerations.

1. Preserve a reasonable balance among the layers of defense

The licensee discussed the reasonable balance among the layers of defense. The licensee provided a list of layers of defense to protect the public and the environment from a radioactive release from the SFP, including structures of fuel rod, fuel assembly, and SFP; systems of fission products and decay heat removal; and availability of water sources for SFP cooling. The NRC staff reviewed the licensees analysis and finds that the proposed change does not alter the defense-in-depth design function or operation of any plant structure, system, or component and it continues to preserve a reasonable balance among the layers of defense to prevent radioactive materials release.

2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures

The licensee discussed the primary function of the FHTSS and sufficiency of the design under normal loading conditions. The proposed change will not allow plant operation in a configuration outside of the design basis. The NRC staff review ed the licensees analysis and finds that the proposed change does not rely on compensatory measures to support the proposed change.

Based on its review, the NRC staff finds that the proposed change preserves adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

3. Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty

The licensee discussed the redundancy of the fuel cooling, emergency power, and beyond-design-basis systems; independence of multiple layers of defense; and diversity of SFP cooling.

The NRC staff reviewed the licensees analysis and finds that the proposed change does not introduce new system dependencies. Based on its review, the NRC staff finds that the proposed change preserves system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including considerations of uncertainty.

4. Preserve adequate defense against potential CCFs

The licensee stated that there are no CCFs that could cause a total loss of safety function of SFP inventory or cooling. The NRC staff reviewed the licensees analysis for the potential introduction of new CCF mechanisms and finds that the licensee adequately assessed that the proposed change does not degrade defenses against potential CCFs and directly considers the

impact of the common-cause initiator. Based on its review, the NRC staff finds that the proposed change preserves adequate defense against potential CCFs.

5. Maintain multiple fission product barriers

The licensee stated that the proposed change demonstrates that it does not modify the fission product barriers, including the existing spent fuel cladding and SFP water. As discussed in Section 3.1.3 of this safety evaluation, the licensee conservatively demonstrated that the likelihood of a tornado that results in the failure of multiple fission product barriers is low. The NRC staff reviewed the licensees analysis and fi nds that the proposed change neither removes a fission product barrier nor affects the independence of the fission product barriers. Based on its review, the NRC staff finds that the proposed change maintains multiple fission product barriers.

6. Preserve sufficient defense against human errors

The licensee stated that the proposed change has no new human actions required to preserve the layers of defense, and the probability of any existing potential human error is unchanged.

The NRC staff reviewed the proposed changes and finds that the proposed change preserves sufficient defense against human error because no new human errors are introduced.

7. Continue to meet the intent of the plants design criteria

The licensee stated that the capacity of the structure meets the intent of the design-basis criteria during normal and abnormal operations on site. As discussed in Section 3.1.5, the NRC staff finds that the proposed change continues to meet the intent of the plants design criteria.

Based on the discussion above, the NRC staff finds that the proposed change meets the second key principle of risk-informed decision-making.

3.2.3 Principle 3 - Safety Margins

The third key principle of risk-informed decision-making is that the proposed licensing basis change maintains sufficient safety margins. Section C.2.1.2, Safety Margin, of RG 1.174 states that the engineering evaluation should assess whether the impact of the proposed change is consistent with the principle that sufficient safety margins are maintained. With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met or proposed revisions provide suffi cient margin to account for uncertainty in the analysis and data.

In the letter dated April 6, 2023, the licensee addressed safety margins. The licensee stated that safety margins are preserved by the presence of significant conservative assumptions that are not discretely factored into the risk-informed approach. The licensee stated that these conservative assumptions ensure that the conclusions are well supported, including consideration of uncertainty in parameters used to assess the risk. The assumptions for the consequence analysis are discussed in more detail in Section 3.1.4 of this SE

In its letter dated April 6, 2023, the following list of conservative assumptions for the structural analysis:

In the estimation of the median wind speed capacity for the fragility analysis of the FHTSS, it is assumed that structural members of the FHTSS would collapse once they reach their yield strengths. This is a conservative assumption because structural steel members would typically undergo plastic deformations and exhibit strain hardening behavior and resistance to the applied loads before their failure at ultimate strength.

In the estimation of the potential damage to the SFP due to the impact of potential falling structural debris, the heaviest members (i.e., trolley girders) of the FHTSS with the worst-case angle of attack are conservatively considered.

Also, conservatively, the kinetic energy of the heaviest members is calculated assuming a maximum travel distance from their installed location to the surface of the SFP floor mat.

No credit is taken to account for energy dissipation due to travel of the postulated falling structural members through borated water inventory of the SFP before their potential impact to the reinforced concrete walls or floor mat of the SFP.

No credit is taken to account for energy dissipation and load distribution due to impact of the postulated falling structural members to the spent fuel storage racks and spent fuel assemblies. The majority of the SFP area under the footprint of the FHTSS is occupied by the spent fuel racks that provide a suitable means to distribute the impact loads to a larger area, thus minimizing, if not fully eliminating, any possible local damage to the reinforced concrete walls and floor mat and steel liner of the SFP.

Energy dissipation through impact to the crash pads, located in the cask loading area of the SFP, is not considered, conservatively.

The NRC staff reviewed the conservative assumptions mentioned above and finds that the proposed change meets the third key principle of risk-informed decision-making.

3.2.4 Principle 4 - Evaluation of Risk Increase

The fourth key principle of risk-informed decision-making is that when the proposed licensing basis change results in an increase in risk, the increase should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants. Section C.2.4, Acceptance Guidelines, of RG 1.174 provides acceptance guidelines for CDF and LERF. These guidelines are intended to provide assurance that proposed increases in CDF and LERF are small and are consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.

As described in Section 3.1.4, NUREG-1738 documents the NRC staffs evaluation of the potential accident risk in an SFP at decommissioning plants. NUREG-1738 used the LERF acceptance guidelines from RG 1.174 to establish a PPG of 1E-5 per year. In addition, NUREG-1738 infers risk increase values of 8.7E-8 per year for individual early fatality risk and 6.9E-8 per year for individual latent cancer fatality risk from the RG 1.174 acceptance guidelines. The NRC staff notes that the calculations in NUREG-1738 were based on the Surry site and are, therefore, directly applicable to this LAR.

In its letter dated April 6, 2023, the licensee addressed the risk increase associated with the proposed change. As described in Section 3.1.4 of this SE, the NRC staff found that the licensees increase in dose and increase in latent cancer risk values met the risk increase values of 11 person-rem per year for population dose and 6.9E-8 per year for individual latent cancer fatality risk inferred from RG 1.174.

Based on the above, the NRC staff finds that the proposed change meets both guidelines of risk increase values of 11 person-rem per year and 6.9E-8 per year for individual latent cancer fatality risk from RG 1.174. Therefore, the NRC staff finds that the proposed licensing change results in an increase in risk that is small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plant. Based on the discussion above, the NRC staff finds that the proposed change meets the fourth key principle of risk-informed decision-making.

3.2.5 Principle 5 - Performance Measurement

The fifth key principle of risk-informed decision-making is that the impact of the proposed licensing basis change should be monitored us ing performance measurement strategies.

Section C3, Element 3: Define Implementation and Monitoring Progra m, of RG 1.174 states that licensees should integrate, or at least coordinate, their monitoring for risk-informed changes with existing programs that monitor equipment performance and other operating experience on their site and industrywide. This section also states that monitoring under 10 CFR 50.65 can be used when the monitoring performed under the Maintenance Rule is sufficient for the structures, systems, and components affected by the risk-informed application.

In the letter dated April 6, 2023, the licensee addressed performance measurement. The licensee stated that monitoring to assure the structural performance of the FHTSS remains consistent with the current as-built design and is accomplished via programmatic activities for aging management, which are incorporated into the sites structural monitoring program. The licensee stated that the structural monitoring program, which includes the FHTSS, implements the requirements of 10 CFR 50.65 and has been enhanced to be consistent with the requirements of NUREG-2191.

In its letter dated April 6, 2023, the licensee stated,

Inspections performed under this program cover the entire Fuel Building structure, including the steel superstructure over the SFP and the attached FHTSS, as well as the fuel cask trolley crane. Qualified structural engineers are responsible for performing the inspections of plant structures, for the evaluating inspection results, for determining corrective actions as needed, and for ensuring structures inspected can meet their intended design functions.

In general, inspections include assessment of structural and support steel, as well as concrete and masonry element s. Inspections are performed on a frequency not to exceed 5 years. Inspections are documented in accordance with procedural program requirements and retained as plant records. Any deficiencies identified during inspections are assessed to determine whether corrective actions are warranted - if so, they are initiated as appropriate through the plant's corrective action program.

In conclusion, the FHTSS is monitored for aging as part of the Surry Structures Monitoring Program to ensure the structural performance remains consistent with the current as-built design. Additionally, the significance of this structure with respect to high wind events will be incorporated into the sites licensing basis (via UFSAR update), as well as design-basis documentation. This will ensure that any future plant changes take into consideration the importance of this structure, and that any design changes maintain the current structural capacity and preserve risk assessment results.

The NRC staff reviewed the licensees plans for monitoring performance and finds that the impact of the proposed licensing change will be monitored using performance measurement strategies. Based on the above, the NRC staff finds that the proposed change meets the fifth key principle of risk-informed decision-making.

3.

2.6 NRC Staff Conclusion

on Risk-Informed Evaluation

The NRC staff evaluated the proposed change against each of the five key principles of risk-informed decision-making in RG 1.174, in conjunction with the conclusions on engineering evaluations provided in Sections 3.1.4 and 3. 1.5 of this safety evaluation. The NRC staff concludes that the proposed change satisfies th e key principles of risk-informed decision-making and, therefore, the request to apply a risk-informed approach to demonstrate that the FHTSS, as designed, meets the intent of a tornado resistant structure under the current Surry licensing basis for a 360-mph maximum tornado wind speed is acceptable. The staff has reviewed the licensees submittal and finds that it is consistent with the guidance in RG 1.174, and that the proposed change is acceptable and meets the applicable requirements of 10 CFR 50.67, 10 CFR 50.65, and the plant design criteria in the UFSAR and, thus, GDC 2 and GDC 4.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth Virginia State official was notified of the proposed issuance of the amendments on November 8, 2023. On November 8, 2023, the State official confirmed that the Commonwealth of Virginia had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increa se in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on September 6, 2022 (87 FR 54553) and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributors: Steven Alferink, NRR Elijah Dickson, NRR Shao Lai, NRR Ata Istar, NRR Jason White, NRR De Wu, NRR

Date: January 18, 2024

ML23312A192 OFFICE DORL/LPL2-1/PM DRA/APLC DSS/SCPB/BC DEX/EXHB/BC NAME JKlos SAlferink BWittick BHayes DATE 11/8/23 10/16/23 10/25/23 by email 10/19/23 by email OFFICE DEX/ESEB/BC DRA/ARCB/BC DRA/APLC/BC OGC NAME ITseng KHsueh SVasavada ANaber DATE 10/23/23 by email 10/30/23 by email 10/30/23 by email 12/7/23 OFFICE DORL/LPL2-1/LA DORL/LPL2-1/BC DORL/LPL2-1/PM NAME KGoldstein MMarkley JKlos DATE 01/18/24 1/18/24 1/18/24