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Flood Hazard Reevaluation Report
	ML16145A180
Person / Time
Site:	Monticello
Issue date:	04/21/2016
From:	Masopust P; Aterra Solutions
To:	; Office of Nuclear Reactor Regulation
References
	L-MT-16-024
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L-MT-16-024 ENCLOSURE 2 FLOOD HAZARD REEVALUATION REPORT IN RESPONSE TO THE 50.54(f) INFORMATION REQUEST REGARDING RECOMMENDATION 2.1: FLOODING OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAl-ICHI ACCIDENT for the MONTICELLO NUCLEAR GENERATING PLANT RENEWED LICENSE NO. DPR-22 (REDACTED VERSION) 39 pages follow FLOOD HAZARD REEVALUATION REPORT IN RESPONSE TO THE S0.54(f) INFORMATION REQUEST REGARDING RECOMMENDATION 2.1: FLOODING OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAl-ICHI ACCIDENT for the MONTICELLO NUCLEAR GENERATING PLANT RENEWED LICENSE No. DPR-22 Prepared by: Black & Veatch and Aterra Solutions Rev.O Printed Name Affiliation Signature Date Prepared by: Petr Masopust Aterra Solutions*

4/21/2016 Reviewed by: Adam Liebergen Black & Veatch :ffi "fl.1s-hiW-Approved by: Steven Thomas Black & Veatch t//trliu1' Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant Table of Contents List of Figures .....................................................................................................................................................

4 List of Tables ......................................................................................................................................................

5 List of Appendices

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6 Acronyms and Abbreviations

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7 1. Information Related to the Flood Hazard ..................................................................................................

9 1.1 Introduction

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9 1.2 Purpose ..............................................................................................................................................

9 1.3 Method ...............................................................................................................................................

9 1.4 Detailed Site Information

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10 1.4.1 Elevation Values .......................................................................................................................

10 1.4.2 Site Layout and Topography

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11 1.4.3 Elevation of Safety Related Structures, Systems and Components

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11 1.5 Current Design Basis Flood Elevations

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12 1.5.1 Flooding in Streams and Rivers ................................................................................................

12 1.5.2 Ice Induced Flooding ................................................................................................................

14 1.6 Flood-Related Changes to the Licensing Basis and Flood Protection and Mitigation Changes since License Issuance ...........................................................................................................................................

14 1.7 Watershed and Local Changes .........................................................................................................

14 1.8 Licensing Basis Flood Protection and Mitigation Features ..............................................................

15 2. Flood Hazard Reevaluation

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16 2.1 Local Intense Precipitation

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16 2.1.1 Methodology

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16 2.1.2 Results ......................................................................................................................................

19 2.1.3 Conclusions

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19 2.2 Flooding in Streams and Rivers ........................................................................................................

20 2.2.1 Maximum Stillwater Elevation

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20 2.2.2 Wind-Generated Waves ...........................................................................................................

22 2.3 Darn Breaches and Failures ..............................................................................................................

24 2.4 Storm Surge, including Wind-Wave Activity ....................................................................................

24 2.5 Seiche ................................................................................................................................................

24 2.6 Tsunami ............................................................................................................................................

25 2.7 2.7.1 2.7.2 Ice Induced Flooding ........................................................................................................................

25 Methodology

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25 Results ......................................................................................................................................

26 Page I 2 of 35 ,.:..._.

Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 2.7.3 Conclusions

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26 2.8 Channel Migration or Diversion

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27 2.9 Combined Effect Flood .....................................................................................................................

27 2.10 Interim Evaluations

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27 2.10.1 Evaluation of Internal Flooding during the LIP .........................................................................

27 2.10.2 Structural Evaluation of Doors for LIP Loads ...........................................................................

30 3. Comparison of Current Design Basis and Reevaluated Flood Hazard .....................................................

31 3.1 Comparison of Flood Hazard Elevations

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31 3.2 Comparison of Flood Parameters

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31 4. References

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34 Page I 3 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant List of Figures Figure 1-MNGP Site and Nearby Vicinity .......................................................................................................

10 Figure 2-MNGP Power Block Area .................................................................................................................

12 Figure 3 -Local Intense 6-Hr All-Season PMP Hyetograph

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17 Figure 4 -LIP Drainage Areas and Directions of Flow .....................................................................................

18 Figure 5 -Probable Maximum Flood Flow and Stage Hydrographs at MNGP Site (River Station 900.5) ....... 22 Figure 6 -Inundation Limits and Fetch Lines ..................................................................................................

23 Page I 4 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant List of Tables Table 1-All-Season LIP Calculations and Cumulative Depth ..........................................................................

16 Table 2-LIP Evaluation Results .......................................................................................................................

19 Table 3 -Summary of Ice Induced Flooding Evaluation

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26 Table 4 -Estimated Inflow Rates through Door Openings during the LIP Event ............................................

28 Table 5 -Summary of Current Design Basis and Reevaluated Flood Hazard Elevations

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31 Table 6-Local Intense Precipitation

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33 Page I 5 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant List of Appendices Appendix 1-50.54(f) Letter -Requested Information Cross-Reference Page I 6 of 35 Flood Hazard Reevaluation Report NSPM -Monticello Nuclear Generating Plant Acronyms and Abbreviations ACES ADAMS ANS ANSI APWA B COB CEM CFR cfs d/b/a DEM EC EFT EM FFE FIS FLEX ft ft3 ft/s H HEC-HMS HEC-RAS HHA HPCI HMR hr ISFSI ISG JLD km 2 LiDAR LIP m mi 2 MNGP Automated Coastal Engineering System Agencywide Documents Access and Management System American Nuclear Society American National Standards Institute American Public Works Association Bounded current design basis Coastal Engineering Manual Code of Federal Regulations cubic (foot) feet per second Doing Business As Digital Elevation Model Engineering Change Emergency Filtration Train Engineer Manual finished floor elevation Flood Insurance Study Diverse and Flexible Coping Strategies foot (feet) cubic foot (feet) feet per second horizontal Hydrologic Engineering Center Hydrologic Modeling System Hydrologic Engineering Center River Analysis System Hierarchical Hazard Assessment High-Pressure Coolant Injection Hydrometeorological Report hour Independent Spent Fuel Storage Installation Interim Staff Guidance Japan Lessons-Learned Directorate square kilometer(s)

Light Detection and Ranging Local Intense Precipitation meter(s) square mile(s) Monticello Nuclear Generating Plant Rev.a Page I 7 of35 Flood Hazard Reevaluation Report NSPM -Monticello Nuclear Generating Plant mph MSL N/A NAVD 88 NGDC NGP NB NGVD 29 NOAA NRC NSPM NTIF NU REG NU REG/CR PMF. PMP psf RHR scs SS Cs TBA TR USA CE USAR USGS v WSE yr miles per hour mean sea level not applicable North American Vertical Datum of 1988 National Geophysical Data Center Nuclear Generating Plant not bounded National Geodetic Vertical Datum of 1929 National Oceanic and Atmospheric Administration United States Nuclear Regulatory Commission Northern States Power Company, a Minnesota corporation Near Term Task Force NRC technical report designation NUREG contractor report probable maximum flood probable maximum precipitation pound(s) per square foot Residual Heat Removal Soil Conservation Service structures, systems, and components Turbine Building Addition Technical Release United States Army Corps of Engineers Updated Safety Analysis Report United States Geological Survey vertical water surface elevation year Rev.a Page I 8 of35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1. Information Related to the Flood Hazard 1.1 Introduction In response to the nuclear fuel damage at the Fukushima Dai-ichi power plant due to the March 11, 2011 earthquake and subsequent tsunami, the United States Nuclear Regulatory Commission (NRC) established the Near Term Task Force (NTIF) to conduct a systematic review of NRC processes and regulations, and to make recommendations to the Commission for its policy direction.

The NTIF reported a set of recommendations that were intended to clarify and strengthen the regulatory framework for protection against natural phenomena.

On March 12, 2012, the NRC issued an information request pursuant to Title 10 of the Code of Federal Regulations (CFR), Section 50.54 (f) (Reference

3) which included six (6) enclosures:

1. [NTIF] Recommendation 2.1: Seismic 2. [NTIF] Recommendation 2.1: Flooding 3. [NTIF] Recommendation 2.3: Seismic 4. [NTIF] Recommendation 2.3: Flooding 5. [NTIF] Recommendation 9.3: Emergency Preparedness

6. Licensees and Holders of Construction Permits In Enclosure 2 of Reference 3, the NRC requested that licensees reevaluate the flooding hazards at their sites against present-day regulatory guidance and methodologies being used for early site permits and combined license reviews. 1.2 Purpose This report provides the information requested in the March 12, 50.54(f) letter; specifically, the information listed under the "Requested Information" section of Enclosure 2 of Reference 3, paragraph 1 ("a" through "e") for Monticello Nuclear Generating Plant (MNGP). Evaluation of the eight flood-causing mechanisms and associated effects (when required), as well as the combined effect flood, identified in Attachment 1 to Enclosure 2 of the NRC information request (Reference

3) and the potential effects on the MNGP site is described in Section 2 of this report. 1.3 Method This flooding hazard reevaluation followed the Hierarchical Hazard Assessment (HHA) approach, as described in NUREG/CR-7046, "Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America" (Reference

2) and its supporting reference documents.

The HHA approach consists of a series of stepwise, progressively more refined analyses to evaluate the hazard resulting from phenomena at a given nuclear power plant site to structures, systems, and components (SSCs) important to safety with the most conservative plausible assumptions consistent with the available data. The HHA starts with the most conservative, simplifying assumptions that maximize the hazards from the maximum probable event. If the assessed hazards result in an adverse effect or exposure to any related SSC, a more site-specific hazard assessment is performed for the probable maximum event. The steps typically involved to estimate flood hazard include the following:

Page I 9 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1. Identify flood-causing phenomena or mechanisms by reviewing historical data and assessing the geohydrological, geoseismic, and structural failure phenomena in the vicinity of the site and region. 2. For each flood-causing phenomenon, develop a conservative estimate of the flood from the corresponding probable maximum event using conservative simplifying assumptions.

3. If any safety-related SSC is adversely affected by flood hazards, use site-specific data and/or more refined analyses to provide a more realistic condition and flood analysis, while ensuring that these conditions are consistent with those used by Federal agencies in similar design considerations.

Repeat Step 2; if all safety-related SSCs are unaffected by the estimated flood, or if all site-specific data have been used, specify design bases for each using the most severe hazards from the set of floods corresponding to the flood-causing phenomena.

Section 2 of this report provides additional HHA detail for each of the flood-causing mechanisms evaluated. 1.4 Detailed Site Information 1.4.1 Elevation Values Unless otherwise stated, all elevation values sited in this report are in feet above mean sea level (MSL), which is also referred to as National Geodetic Vertical Datum of 1929 (NGVD 1929). Figure 1-MNGP Site and Nearby Vicinity N W WI; 0 0.5 s 2 3 Page I 10 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1.4.2 Site Layout and Topography The site is located within the city limits of the Monticello , Minnesota on the right bank of the Mississippi River and about 70 river miles upstream from Minneapolis

-St. Paul. The plant site occupies an area of approximately 2 , 150 acres. The topography of the MNGP site is characterized by relat i vely level bluffs , which rise sharply above the river. Three distinct bluffs exist at the plant site at elevations 920 , 930, and 940 ft. Bluffs located approximately a mile north and south of the site rise to 950 ft. Further to the north, the terrain is relatively level with numerous lakes and wooded areas. To the south, west , and east , the terrain is hilly and dotted with numerous small l akes (Reference 10). The Mississippi River abuts the site to the north and northwe s t. The flow in the Mi ss i s sippi River in the vicinity of the plant is unregulated and subject to la r ge variations throughout the year. Normal river level is at elevation 905 ft and the maximum river flood stage was recorded in 1965 at elevation 916 ft. The 1 , 000-year projected river flood stage is at elevation 921 ft (Reference 10). The natural grade of the power block is at elevation 930 ft with elevations of the majority of critical structure open i ngs ranging from 931 ft to 935 ft (Refe r ence s 10 and 18). The MNGP site and the regional vicinity are shown in Figure 1. 1.4.3 Elevation of Safety Related Structures, Systems and Components Class I structures, which are vital to safe shutdown of the plant and removal of decay heat have been identified in the USAR (Reference 10), as following:

Primary Containment (Drywell , Vents, Torus, and Penetrations)

Reactor Building (up to Operating Floor -1027-foot 8-inch)

High Pressure Coolant Injection (HPCI) Building

Plant Control and Cable Spreading Structure

Spent Fuel Storage Pool

Off-gas Stack

Reactor Primary Vessel Biological Shield and Support Pedestal

Standby Diesel Generator Building

Diesel Fuel Oil Transfer House Containing Diesel Fuel Oil System

Emergency Filtration Train (EFT) Building

Intake Structure Pump Room Conta i ning Emergency Service W a te r and Re s idual Heat Removal (RHR) Service Water Pumps and Connecting Pipe Tunnel

Parts of Turbine Building Housing Cla s s I Equipment

Underground Duct Bank-3 rd Floor, EFT to Reactor Building. Plant grade for Class I structures is at elevation 930 ft , except for the Intake Structure. These structures have been designed for a river flood stage up to this elevation. Certain components of the above mentioned Class I structures may be located below elevation 930 ft; however , flood protection features (e.g., steel plates, grout , or sandbags) are installed to close any openings below elevation 930 ft (Reference 10). The operating floor ofthe Intake Structure is located at elevation 919 ft with an opening between the Screenhouse and the Intake Structure at elevation 919.5 ft (References 10 and 18). The power block area is shown in Figu re 2. Page I 11 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant Figure 2-MNGP Power Block Area 1.5 Current Design Basis Flood Elevations The current design basis flood elevations for MNGP are described in the USAR (Reference

10) as well as the recent walkdown reports required as part of NRC's 10 CFR 50.54(f) letter (Reference 3). The only flood causing mechanism quantified as part of the current design basis was "Flooding in Streams and Rivers." A summary of this flood causing mechanism and the corresponding flood hazard is provided in the following section. A brief discussion of " Ice Induced Flooding" is also provided. 1.5.1 Flooding in Streams and Rivers The probable maximum flood (PMF) in the Mississippi River adjacent to the MNGP site was determined to be 364,900 cfs with a corresponding peak flood stage of 939.2 ft. A detailed description of the methodology , inputs, assumptions and results is provided in Appendix G of the USAR. The limiting flood resulted from a combination of meteorological conditions including snowmelt that could occur in the spring and could reach maximum river level in about 12 days. It was estimated that the flood elevation would remain above elevation 930 ft for approximately 11 days (Reference 10). From flow and stage hydrographs provided in Appendix G of the USAR , it can also be estimated that the flood stage would reach elevation 919 ft (Intake Structure operating floor) in less than 4 days and would remain above this elevation for approximately 20 days. Page I 12 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant The PMF study included a watershed area of 13,900 square miles upstream of MNGP. The watershed is characterized by level to rolling prairie land interspersed with areas of glacial moraines. The average elevation of the watershed is approximately 1,200 ft. The main hydrologic characteristic of the watershed are the numerous lakes, which significantly influence the streamflow characteristics of the Mississippi River and its tributaries. The long-term mean annual runoff for the basin is approximately 5.0 inches (Reference 10). The PMF at the MNGP site was determined by transposing actual storms to the watershed and maximizing the precipitation for potential moisture. The selected storms considered were the March 23-27, 1913 storm centered at Bellefontaine, Ohio (spring storm) and the August 28-31, 1941 storm centered at Hayward, Wisconsin (summer storm). The spring storm combined with a snowmelt occurring during the storm resulted in higher discharge than the summer storm. For the purpose of the PMF study, maximum snow water equivalent for a period between March 16 and March 31 having one percent probability was used. Methods developed by the USACE and described in EM 1110-2-1406 "Runoff from Snowmelt" were used to compute snowmelt (Reference 10). Initial retention losses were assumed to be zero and infiltration losses were assumed to be 0.02 inches per hour during the snowmelt period and 0.03 inches per hour during the period following the beginning of rainfall (Reference 10). The most critical sequence of event leading to a major flood in the watershed would be a combination of unusually heavy spring snowfall and low temperatures after a period of intermittent warm spells and freezing temperatures forming an impervious ground surface followed by a period of extremely high temperatures and a major storm event. For the purposes of the design basis study, snow water equivalent having a one percent probability was assumed to cover the watershed at the beginning of the simulation.

This was immediately followed by the maximum historical temperature sequence and after 5 days the probable maximum spring precipitation was initiated (Reference 10). The watershed was divided into four major sub-basins and synthetic hydrographs were developed for each using the Snyder's method. Unit hydrograph peaks were increased by 25 percent and basin lag time decreased by one-sixth.

Snowmelt and rainfall excess were applied to the unit hydrographs and the resulting hydrographs were determined for each sub-basin, which were then routed to the MNGP site using a computer model. The travel time for flood routing was based on the USACE recorded travel times for large floods. Baseflow of 5,000 cfs, based on long-term USGS records for the Elk River, Minnesota stream gage, was added to the routed flood hydrographs (Reference 10). A stage-discharge curve at the MNGP site was developed using a computer model. The inputs for the computer model included channel and overbank cross-sections based on 2-ft and 10-ft topographic maps; Manning's n-values for left overbank , right overbank , and channel; discharges of various magnitudes; and starting water elevations. Average Manning's n-values were determined to be 0.032 for the main channel, 0.050 for the left overbank , and 0.045 for the right overbank.

A higher value of 0.065 was used for the right overbank at MNGP and a value of 0.06 was used for the island immediately upstream of MNGP. The model was verified against the maximum flood of record (April 1965) for which records of high water marks exist at several points along the river (Reference 10). Based on the results of the study described in the USAR, the limiting design basis flood for the MNGP site is the PMF of the Mississippi River of 364,900 cfs, with a peak rive r flood stage of 939.2 ft. Dam breaches and wind-wave runup were not considered in the current design basis evaluation. Page I 13 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1.5.2 Ice Induced Flooding Appendix G of the USAR (Reference

10) briefly discusses backwater flooding caused by ice jams as one of the two types of flooding occurring the Mississippi River watershed. However, the ice induced flooding is not quantified and is not considered to result in flooding levels exceeding the precipitation

-driven flooding levels. 1.6 Flood-Related Changes to the Licensing Basis and Flood Protection and Mitigation Changes since License Issuance No changes to flood elevations have been made since the issuance of the original license. Procedural changes and enhancements to flood protection were implemented following the completion of the Recommendation 2.3: Flooding Walkdowns , during which several deficiencies and observations were identified. These deficiencies and observations were documented in the flooding walkdown report (Reference 11). A summary of the changes is provided below:

Enhancements to Procedure A.6 were implemented to streamline actions described in the procedure (Reference 12).

Engineering Change (EC) 21937 was p r epared to install permanent flood protection features on and around the Intake Structure and reduce the amount of field work required during a flooding event and ultimately improve response time (Reference 11). Th i s included bin wall/earthen levee design changes (Reference

15) and design of debris barrier to protect the flood barrier from impact by debris floating in the flooded river (Reference 16). 1.7 Watershed and local Changes The watershed contributory to the Mississippi River upstream of MNGP is approximately 13,900 square miles (Reference 10). There have not been significant changes to the watershed since the last license renewal and the land use changes are relatively minimal. Local area changes have also been minimal since plant operation began at the MNGP site. Changes consistent with most nuclear plant sites have been made at MNGP since operations began , including the addition of the following structures:

Administration Buildings

Independent Spent Fuel Storage Installation (ISFSI)

Security Buildings

Warehouses

FLEX Equipment Storage Building

Security barriers The changes also included the bin wall sections adjacent to the Intake Structure , which are now permanently installed. Location and configurat i on of current structures, as relevant, were i nputs to the Local Intense Precipitation (LIP) calculations as related to the flooding impacts on SSCs. Page I 14 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1.8 Licensing Basis Flood Protection and Mitigation Features Flood protection features utilized at MNGP in the event of the PMF include incorporated and temporary active and passive barriers. For flood protection below elevation 930 ft , installation of flood protection features (such as pumps and steel plates , grout, or sandbags to close openings up to elevation 930 ft) provides flood protection for Class I structures and Class II structures housing Class I equipment.

Suitable steel flood protection plates are stored at the plant to ensure that they are readily accessible (Reference 10). For flood protection above elevation 930 ft, a levee consisting of a bin wall and an earthen berm is constructed around Class I structures (e x cluding the Off-Gas Stack), Class II structures housing Class I equipment (excluding Off-Gas Storage Building), and Radwaste Building to protect them from the effects of a flood. The Off-Gas Stack is outside the boundary of the levee and is protected by sandbags.

The Off-Gas Storage Building is excluded because the only areas that house Class I components are the Fan and Foyer Rooms for Stand-By Gas Treatment and the components are located at an elevation above the PMF. Additional flood protection features (such as steel plates , grout , or sandbags) to close openings may be used as a defense in depth measure when river levels are expected to exceed elevation 930 ft (Reference 10). Procedure A.6, " Acts of Nature," (Section 5.0-External Flooding) (Reference

12) outlines actions to be taken in the event flood waters are predicted to exceed elevation 918 ft. Should the projected river level exceed 918 ft, an orderly plant shutdown would be commenced to place the reactor in a cold shutdown condition (Reference 10). Procedure 8300-02 (Reference

13) provides instruction for protection of MNGP from damage by floodwaters , including the construction of the bin wall/earthen levee and installation of the debris barrier to protect the bin wall and Intake Structure roof plates. The bin wall sections adjacent to the Intake Structure are permanently installed (Reference 15). The remaining bin wall/levee sections would be constructed ifthe river levels are forecast to exceed elevation 930 ft. The north-west portion of the levee and east and west temporary bin walls must be started prior to floodwaters reaching elevation 917 feet and remain ahead of the river. The remaining areas of the levee must be started prior to floodwaters reaching elevation 930 feet (Reference 12). Page I 15 of 35 Flood Hazard Reeva l uation Report Rev.O NSPM -Monticello Nuclear Generat i ng P l ant 2. Flood Hazard Reevaluation The flood i ng hazard reeva l uat i on for each of the eight flood causing mechanisms and the combined effect flood, i s described i n the follo wi ng subsect i ons. 2.1 Local Intense Precipitation The methodology and results presented i n th i s section a r e based on the evaluation of the LIP event performed i n Calcu l at i on 180999.51.1005 (Reference 18). 2.1.1 Methodology The LIP is a measure of the extreme precip i tat i on (high i ntens i ty/short durat i on) at a given locat i on. NUREG/CR-7046 (Reference

2) speci fi es that the LIP should be equivalent to the 1-hr, 2.56-km 2 (1-mi 2) PMP at the p l ant site. For the LIP evaluat i on at the MNGP site, the storm duration was extended to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , as shown in Appendix B of NUREG/CR-7046. Case 3 assumptions in Appendix B of NUREG/CR-7046 were a l so app li ed. Case 3 assumes that the design of the site grade and the pass i ve dra i nage channels are incapable of rout i ng flow from the i mmed i ate p l ant s i te and , therefore , overland flo w occurs o v er the entire p l ant s i te d uri ng the LIP event. Roof runoff w as accounted for i n the analys i s and rooftops a r e not pro vi d i ng any storage. The LIP eva l uat i on included an a ll-season storm event and a cool-season ra i n-on-sno w storm event. The all-season event was determ i ned to be the controlling event. Therefore, onl y the all-season event i s d i scussed in th i s section. Ra i nfall depths and temporal d i stribut i on fo r the LIP storm were developed using HMR 51 (Reference

24) and HMR 52 (Reference 25), respect i vely. The 1-hr 1-m i 2 ra i nfa ll depths and the correspond i ng percentages for the 5-, 15-, and 30-m i nute interva l s w ere determ i ned us i ng the approach descr i bed i n HMR 52. While HMR 52 does not specifically state that the time intervals be arranged i n th i s order, w i th the typical west-east flow across North America , the type of storm set-up that w ou l d provide an LIP event at the MNGP s i te would l i ke l y be a mesosca l e convective system (such as squall li ne for example). Us i ng the conceptual model of th i s type of s y stem , the i n i t i a l prec i pitat i on is associated wi th the matu r e ce ll s and a zone of con v ergence and as such will be very i ntense. The storm motion and nature of the system w ou l d then see a decrease in the prec i pitat i on after the in i t i a l burst as the rear tra ili ng strat i form region w ith the co l d poo l moves over the area. Th i s type of meteoro l ogica l s y stem fits w ith the front loaded d i str i but i o n. The 6-hr 10-mi 2 rainfa ll depth i s prov i ded i n HMR 51. The tempora l d i str i b u t i on of the LIP storm used i n the eva l uat i on i s provided i n Table 1. Table 1-All-Season LIP Calculat i ons and Cumulat i ve Depth Duration Area UP (minutes) (m12) Multiplier Applied to (Inches) 5 1 0.345 1-hr , 1-m i 2 PMP 5.78 15 1 0.55 1-hr , 1-mi 2 PMP 9.22 30 1 0.777 1-hr , 1-mi 2 PMP 13.02 60 1 0.71 6-hr , 10-mi 2 PMP 16.76 360 10 N/A N/A 23.60 Runoff losses w ere i gnored dur i ng the LIP event to max i mize runoff per NU REG/CR-7046. As a resu l t , i nfiltration (i.e., constant loss) w as not cons i dered and i n i t i a l abstraction w as set to ze r o. Page I 16 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant The Digital Elevation Model (DEM) was derived from bare earth LiDAR data combined with topographic survey data. For any buildings and structures not within the survey limits , a height of 10 feet is assumed for the shorter buildings and 100 feet for the tallest building. The top elevation of security barriers was defined by adding 3.3 feet (based on measurements) to the ground elevation reported by the LiDAR. Manning's

" n" roughness coefficients were based on land cover information and published guidance.

For calculation of time of concentration for hydrologic calculation the Manning's n-values ranged from 0.011 to 0.4. The MNGP site i s an industrial site with asphalt paving or gravel roads. Therefore, for on-site hydraulics a Manning's n-value of 0.02 was used to represent the roughness of the cross sections.

Travel times were calculated for each flow type (sheet flow , shallow concentrated flow , and channelized flow). The Technical Release 55 (TR-55) method (Reference

26) was used to determine the travel times for sheet flow and shallow concentrated flow. To determine the open channel flow time, the nomogram in American Public Works Association (APWA) 5600 (Figure 5602-3, Reference

28) representing typical hydraulic sections using Manning's equation, was used. To determine the time of concentration, the travel times for sheet flow, shallow concentrated flow and open channel flow were summed. The lag time for use in the HMS model was calculated as approximately 0.6 of the time of concentration (Part 630 Hydrology National Engineering Handbook , Reference 27). The USACE HEC-HMS model was used to evaluate runoff from the LIP event and generate inflow hydrographs for hydraulic analysis. Inputs consisted of the all season LIP hyetographs, drainage area size, and lag time. The 6-hr LIP hyetograph is provided in Figure 3. The Soil Conservation Service (SCS) Unit Hydrograph method was used to transform precipitation into runoff. Figure 3 -Local Intense 6-Hr All-Season PMP Hyetograph

,---7.0 -60 so -------------

0 lS GO 7S 90 105 120 HS 150 luS 180 l!IS HO us HO 2SS HO 285 no Tlm* lminut**I The drainage area of the MNGP site was divided in 10 sub-basins (Figure 4). Outputs from the HEC-HMS model are runoff hydrographs fo r each of the individual s ub-basins area s. Page I 17 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant An unsteady flow analysis was then performed using the USACE HEC-RAS model to evaluate dynamic hydraulic interactions between the different model components. Consistent with the NUREG/CR-7046 recommendations, it was assumed that all storm drains and roof drains are blocked. Based on the review of s ite conditions and site cover , it was concluded that all above-ground flow drainage paths would be available for conveyance and would not be blocked by debris. During the LIP event , precipitation that falls on the MNGP site flows away from the plant exiting either to the north into the Mississippi River or to the south through sections where concrete jersey barriers are not present. In areas where jersey barriers are present, they prevent outside runoff from flowing onto the site and limit the site runoff from flowing off-site.

Where considerable gaps exist in between rows of jersey barriers, water is allowed to leave the site and off-site downstream boundary condition is ta k en into account in the on-site hydraulic models. In order to characterize the various flow paths at and around the site , seven hydraulic models were created. Three hydraulic models were necessary to hydraulically describe runoff in the off-site subbasins (sub-basins 1 , 4, 5, 6, 7 and 8). Off-site basins were modeled to ensure that estimated flood levels would: a) not overtop protective jersey barriers; and b) would not impede onsite runoff. Additionally , four hydraulic models were necessary for characterizing the on-site runoff (runoff within the protective jersey barriers -sub-basins 2 , 2a , 3 , and 9). The downst r eam boundary condit i on wa s set to a known water level. Where the sub-ba si n discharges directly to the Mississippi River the downstream boundary condition was set to the 500-year water level in the Mississippi River , which is a conservative assumption. All models were run in the mixed flow reg i me, allowing for both supercritical and subcritical flows. Figure 4-LIP Drainage Areas and Directions of Flow l"9**d Dn l r\af*Bn in s ** ..... ._._ -t I -""'. -* ' 0 245 490 980 Feet -* . . -* Page I 18 of 35 Flood Hazard Ree v aluat i on Report Rev.O NSPM -Monticello Nuclear Generat i ng P l ant 2.1.2 Results The LIP evaluation determined the maximum water su r face e l evations at several critica l door openings.

The predicted flood elevat i ons vary spatially throughout the site, as shown in Table 2. An interim evaluation (Reference

18) was also performed to assess the impacts of flood levels on the doors and to evaluate possib l e l eakage through the doors. The results of th i s inter i m evaluation are presented in Sections 2.10.1and2.10

.2. T*ble 2 -LIP Ev*lu*tion Results Open i ng Invert/ Estimated Ma xi mum Water Openin g Locat i on S ill Level (FFE) MaximumWSE Depth at Opening feetNGVD29 feet I nta k e Stru ctu r e Doo r (D o or 2 0 9) -Int eri o r 9 1 9.50 920.62 1.12 betwee n Scree n h o use a nd I n t a k e S t r u ct u r e West Roll-Up Door (Door 119) -931.25 931.4 1 0.1 6 Turbine B uilding Additi o n E ast Roll-Up D oor (D o or 120) -931.2 5 9 3 1.53 0.28 Turbine B uilding A d diti o n Ra ilc a r E ntry (Do or 24) -T ur b ine B uil d ing 935.00 93 5.8 3 0.8 3 R ailcar En t ry (D oors 45 an d 46) -R eact o r B ui l ding 935.00 93 5.0 7 0.0 7 E me r gency D iesel G e ner ato r -931.00 9 31.41 0.4 1 East 3-ft wi d e M an Do or (D o o r 8) Emergency D iesel Gener at o r -931.00 931.4 1 0.41 West 3-ft wide Man Door (Door 7) EFT Room (Door 34 1) 932.83 933.20 0.37 13.8 KV R o om (D oor 1) 931.00 933.0 9 2.09 Off Gas Stack (Door 193) 932.50 934.63 2.13 Fuel Oil Transfer Pump House (Door 4 8 3) 931.00 931.2 9 0.29 2.1.3 Conclusions Based on the LIP eva l uat i on for MNGP , the maximum depth of flooding varies spatially and r anges from 0.07 ft at the Railcar Entry to the Reactor Building to 2.13 ft at the Off Gas Stack. The LIP evaluation was not addressed as part of the current design basis and, therefore , the reevaluated flood hazard is considered not bounded. Sign i ficant debris l oad i ng/transportation i s not a safety hazard due to the re l atively lo w velocity and depth of LIP flood w aters i n the v i cin i t y of SSCs , in addit i on to the lack of natura l debr i s sources wi th i n the on-s i te protected area. Page I 19 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant Regarding uncertainty as per NUREG/CR-7046, note that the LIP methodology herein incorporates conservatism which is anticipated to bound potential uncertainties in the analysis.

Specifically:

Rainfall loss rates (i.e., depression storage, infiltration) were conservatively not considered.

Roughness coefficients used in the hydraulic simulation are conservative.

HMR 52 PMP depths were used in the LIP evaluation.

These values are conservative compared to the site-specific LIP depths developed for MNGP in Calculation 180999.51.1008 (Reference 19).

Stormwater drainage systems and features were assumed to be blocked.

Rooftops were modeled as not providing any storage. 2.2 Flooding in Streams and Rivers The PMF is considered to be the most severe, reasonably possible flood resulting from a PMP across the watershed (Reference 2). The flows in the Mississippi River are affected by the presence of a large number of locks and dams. Some of these locks and dams were designed, constructed and are currently maintained by the USACE. To support Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy , development of the flood hazard reevaluation associated with the assessment of flood hazards due to flooding in streams and rivers, including potential flooding due to dam failure, NSPM requested NRC assistance in obtaining information related to the USACE dams, including all completed and pending dam failure analyses. The assistance was requested on March 5, 2014 (Reference 14). The USACE performed this portion of the flood hazard reevaluation. The results of the USACE analysis were transmitted to NSPM on November 18, 2015. The transmittal included a public letter (Reference

8) and non-public enclosures. The non-public enclosures contained security-related information. The USACE analysis provided in that letter determined the controlling PMF discharge for MNGP. The analysis was performed in accordance with JLD-ISG-2013-01, "Guidance for Assessment of Flooding Hazards Due to Dam Failure," (Reference

4) and was based on the USACE knowledge of the river system. The following sections provide a summary of the USACE analysis.

Additionally, the effects of waves induced by 2-year wind speed applied along the critical direction were evaluated in Calculation 180999.51.1002 (Reference

22) and are provided in Section 2.2.2. This calculation was performed using the maximum water surface elevation determined by the USACE in their PMF analysis. 2.2.1 Maximum Stillwater Elevation The information provided in this section is based on the results of the USACE PMF analysis transmitted to NSPM on November 18, 2015 (Reference 8), and the response to NSPM questions at the July 9, 2015 meeting with NRC and USACE (Reference 9). 2.2.1.1 Methodology The USACE developed a HEC-HMS model of the Mississippi River watershed upstream of MNGP to evaluate the PM F discharges. The initial and constant loss rate method was used to model precipitation losses. Loss parameters adopted to simulate the PMF event in HEC-HMS were selected based upon model calibration and the results of other hydrologic studies in the region. The model was calibrated to three large, historic events (1957, 1965 , and 2012) to replicate the conditions that would persist during the PMF event. The values derived via calibration are consistent with the initial and constant loss rates adopted for other studies performed by the USACE. HMR 51 was used to determine the precipitation depth-duration-area relationships for the all-season PMP event. HMR 52 was used to optimize the PMP storm orientation and size. The principles in HMR 52 were Page I 20 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant applied to determine the PMP rainfall hyetograph. To generate the spring PMP event, HMR 53 was used to determine the depth-duration-area relationships for a 10-mi 2 drainage area. The tables in "Probable Maximum Precipitation Estimates and Snowmelt Criteria for the Upper Mississippi River in Minnesota and the Fox-Wolf Rivers in Wisconsin" were used to translate this relationship to a wider range of drainage areas. The energy budget equations in USACE EM 1110-2-1406 were used to generate the snowmelt rates applied to model a 10-day melt prior to a spring rainfall event. The guidance in "Probable Maximum Precipitation Estimates and Snowmelt Criteria for the Upper Mississippi River in Minnesota and the Fox-Wolf Rivers in Wisconsin" was used to define the inputs to these equations and was used to define the 100-year Snow Water Equivalent (representative of the snowpack) prior to the melt. Unsteady flow modeling using the USACE HEC-RAS computer program was implemented for developing the PMF stage/flow hydrographs at the MNGP site. The downstream boundary condition was set to a stagedischarge rating curve at a USGS stream gage (USGS Gage 05288500), which is at a substantial distance downstream from MNGP. The rating curve was extended to PMF magnitude events. The HEC-RAS model extends upstream to a point immediately downstream of St. Cloud Dam , in St. Cloud, Minnesota.

As stated in Reference 9, The hydraulic model "was calibrated to the Anoka County and Sherburne County Flood Insurance Study (FIS) steady flow profiles for the 10-yr, 50-yr, 100-yr, and 500-yr events as well as published USGS rating curves at two locations. The model cross-sections were calibrated to water surface elevations within 0.5 ft of the 100-year FIS study water surface elevations and within 1.0 ft of the 500-yr flood water surface elevations, while ensuring that the 10-yr and 50-yr flood water surface elevations calibrated within a reasonable range." The primary parameters adjusted to achieve a better fit to the calibration data were Manning's n-values, which were consistent with the results of both the Anoka and Sherburne County FIS studies. 2.2.1.2 Results The representative HEC-RAS cross-section for the MNGP site is located at River Station 900.5. The peak flow rate corresponding to the combined-effect PMF is cfs. The peak stage resulting from the effect PMF is-ft NAVO 88 -ft NGVD 29). The average right overbank velocity is.ft/sand the average channel velocity is. ft/s. The combined-effect PMF flow and stage hydrographs for the governing precipitation driven discharge at the MNGP site are presented in Figure 5 .**************************

Page I 21 of 35 Flood Hazard Reevaluation Report Rev.O NSPM -Monticello Nuclear Generating Plant Flcure 5-Prob*ble Mnimum Flood Flow *nd St*c* Hydrocr*phs

t MNGP Site (River St*tion 900.5) -00 00 c --t -z £. 0 u: QI tlO nJ t;; Time (days) * * * * *

Mlsslnippi River Stage 2.2.1.3 Conclusions Based on the results of the PMF study performed by the USACE, the combined-effect PMF discharge of cfs corresponding to peak stage of-ft NAVO 88-ft NGVD 29) is bounded by the current design basis stillwater elevation of 939.2 ft. 2.2.2 Wind-Generated Waves The methodology and results presented in this section are based on the evaluation of wave prediction and wave runup performed in Calculation 180999.51.1002 (Reference 22). 2.2.2.1 Methodology Wave prediction and wave runup were calculated using the Automated Coastal Engineering System (ACES) model. The required inputs for the ACES model for wave pred i ction included wind speed and direction, depth, fetch, and wind duration.

For wave runup, additional information about the structure was required, including the depth of water at the structure, structure slope and roughness. Hourly wind data for the 10-m reference height were obtained from MNGP for years between 1991 and 2013, inclusive.

Only years with at least 80 percent complete records were used in the analysis.

Data for years 1991, 1993, 2000 through 2004, and 2006 through 2012 had sufficiently complete records to be used in the analysis.

The analysis was performed using a two-parameter Weibull distribution to determine the wind-speed with a 2-year recurrence interval.

The 2-year wind speed was determined to be 33.8 mph. Fetch length was determined based on the extent of the PMF inundation limits, as obtained from the USACE PMF study. As recommended in the Coastal Engineering Manual (CEM) (Reference 29), a straight line fetch method was used since it provides a more conservative value than the effective fetch method. Fetch lines Page I 22 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant were drawn in several directions (Figure 6) and average depth along the fetch line was calculated as a difference between the water surface elevation, as obtained from the USACE PMF study, and the terrain elevation. Based on the combination of fetch length and fetch average depth, wave prediction calculations were performed for fetch lines 4 and 5. Figure 6 -Inundation Limits and Fetch Lines It was conservatively assumed that the wind is oriented along the longest fetch. In accordance with the CEM, the final duration, which is the amount of time required for the waves to fully develop depending on the fetch length and wind speed, was calculated and used as input in the ACES model. The final durations for fetch lines 4 and 5 were 52 and 51 minutes, respectively. The aforementioned inputs were then used by the wave prediction module in the ACES model to calculate the significant wave height for fetch lines 4 and 5. The wave runup on the levee system that protects MNGP from the current design basis flooding was calculated only for the larger of the predicted significant wave heights. Since the levee system consists of an earthen levee and bin walls with vertical walls, wave runup both on a sloped berm (1 V: 2.5H) and a vertical wall was calculated. For this evaluation, it was conservatively assumed that the levee system is located at the river bank because the ACES model does not have the ability to simulate runup on a berm that is removed from the shoreline, as in the case of the MNGP site. The ACES model output was also checked using hand calculations, which showed very similar results. 2.2.2.2 Results The significant wave height for fetch lines 4 and 5 is predicted to be 1.65 ft and 1.61 ft, respectively.

The runup on the vertical bin wall and the sloped berm was calculated to be 1.9 ft and 1.3 ft, respectively. The corresponding water run up elevation for the vertical wall and the sloped berm is predicted to be-ft and-ft, respectively. 2.2.2.3 Conclusions The maximum wave runup elevation during the PMF event was estimated to be -ft. Note that the elevation is below the top elevation of the existing flood protection system Page I 23 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant (bin walls/levee).

While the current design basis did not specifically address wind-generated waves , this flood causing mechanism is bounded by the current design basis stillwater, elevation of 939.2 ft. 2.3 Dam Breaches and Failures Based on the information transmitted to NSPM on November 18, 2015 (Reference 8), potential dam breaches and failures were considered in the USACE PMF analysis: " The information contained in the enclosures was developed in accordance with Japan Lessons-Learned Directorate (JLD) interim staff guidance (ISG) JLD-ISG-2013-01, "Interim Staff Guidance For Assessment of Flooding Hazards Due to Dam Failure ," Revision 0, dated July 29 , 2013 (ADAMS Accession No. ML13151A153), and on the USACE knowledge of the river system." Furthermore , the USACE stated in its response to NSPM questions at the July 9, 2015 meeting between NSPM, the NRC , and the USACE: " All the dams upstream of the Monticello and Prairie Island Nuclear Generating Plants (NGP) were screened out with regard to dam failures that could potentially impact the NGPs. The Dam Screen process was conducted in accordance to the current version of the JLD-ISG guidance" (Reference 9). Furthermore, the USACE stated that " All dams were screened out in terms of flood risk to the NGP regardless of failure mode" (Reference 9). Based on the information provided by NRC and USACE , it can be concluded that potential upstream dam breaches and failures regardless of failure mode do not increase the flood hazard at MNGP. 2.4 Storm Surge, including Wind-Wave Activity The methodology and results presented in this section are based on the evaluation of storm surge performed in Evaluation 180999.50.2300-02 (Reference 20). JLD-ISG-2012-06 (Reference 5), " Guidance for Performing a Tsunami , Surge or Seiche Hazard Assessment," Section 3 , " Surge Hazard Assessment," states: "All coastal nuclear power plant sites and nuclear power plant sites located adjacent to cooling ponds or reservoirs subject to potential hurricanes , windstorms , and squall lines must cons i der the potential for inundation from storm surge and wind-waves

." As shown in Figure 1, MNGP is located on the Mississippi River and , therefore, not subject to a storm surge and wind-waves , as defined in the NRC guidance. Furthermore, an increase in water surface elevation on one bank of the river because of wind blowing across the river's water surface would be minor and negligible dur i ng non-flood conditions. The lowest elevation important to the plant is elevation 919 ft (Intake Structure operating floor) while the normal elevation for the Mississippi River at M NGP is 905 ft, which is approximately 14 ft lower. Therefore , flooding due to a storm surge is not applicable to the MNGP site (Reference 20). 2.5 Seiche The methodology and results presented in this section are based on the evaluation of seiche performed in Evaluation 180999.50.2300-02 (Reference 20). NUREG/CR-7046 (Reference 2), Section 3.6 , defines a seiche as follows: " A seiche is def i ned as an oscillation of the water surface in an enclosed or semi-enclosed body of water initiated by an external cause." As further described in Reference 2, seiches are considered for a lake or a reservoir.

Reference 5 , "Guidance for Perform i ng a Tsunami, Surge, or Seiche Hazard Assessment

," Enclosure 1 , Section 4, " Seiche Hazard Assessment," defines a seiche as follows: " Seiche i s an o s cillatory wave generated i n lakes , bays , or gulfs as a result of seismic or atmospheric d is turban c es and with a period ranging from a few m i nutes to a few hou rs." Page I 24 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant The MNGP site is located on the Mississippi River and not on a location susceptible to a seiche. Therefore, flooding due to a seiche is not applicable to the MNGP site (Reference 20). 2.6 Tsunami The methodology and results pre s ented in this section are based on the evaluat i on of tsunami performed in Evaluation 180999.50.2300-02

{Reference 20). Reference 5, "Guidance for Performing a Tsunami, Surge, or Seiche Hazard Assessment," Enclosure 2, Section 3, " Tsunami Hazard Assessment

," states: " All coastal nuclear power plant sites (including sites adjacent to o c eans , seas , lakes , rivers , and other inland bodies of water) must consider tsunami ha z ards." Reference 5 refers to NUREG/CR-6966 , "Tsunami Hazard Assessment at Nuclear Power Plant Sites in the United States of America ," (Reference

7) for performing a hierarchical hazard assessment to evaluate flooding potential due to a tsunami. Reference 7, Section 2 , describes the hierarchical hazard assessment approach. The first step is to determine the potential that a tsunami may pose a hazard to the site based on a regional survey of historical records in the region and an evaluation of potential tsunami generation sources. NOAA National Geophysical Data Center (NGOC) maintains a historical tsunami database which catalogs tsunami events. Reference 7, Section 4.1 refers to this database as the source for archived tsunami records in the United States. The data in the NGOC database was filtered to exclude invalid events. Based on this filtering, no tsunami events were identified in region of the M NGP site. Instances were identified of like waves on the Mississippi River caused by the New Madrid earthquakes in 1811 and 1812, but this is several hundred miles downriver from the MNGP site. The MNGP site is located in a relatively low seismic hazard area with a Design Basis Earthquake of 0.12g. The lowest elevation important to the plant is elevation 919 ft (Intake Structure operating floor) while the normal elevation for the Mississippi River at MNGP is 905 ft, which is approximately 14 ft lower. Therefore , based on historical records of tsunami events and the absence of significant tsunami generation sources in the region , flooding due to a tsunami is not applicable to the MNGP site (Reference 20). 2.7 Ice Induced Flooding The methodology and results presented in this section are based on the evaluation of ice induced flooding performed in Calculation 180999.51.1001 (Reference 21). 2.7.1 Methodology Potential ice induced flooding can either be the result of failure of an upstream ice jam or increased water elevation due to backwater conditions from a downstream ice jam. Consistent with the HHA approach described in NUREG/CR-7046 the following method was used to determine the maximum flood heights from an upstream ice jam failure. The most severe historical ice jam events on the Mississippi River in the vicinity of MNGP were determined using the Ice Jam Information Clearinghouse maintained by the USACE Cold Regions Research and Engineering Laboratory. Based on the information obtained from the Ice Jam Information Clearinghouse, two ice jam events caused a significant increase in the water surface elevation. The first was an ice jam of 15 ft in height about 1 mile upstream of Coon Rapids Dam holding approximately 10 ft of water on April 12, 1965. This location is approximately 35 miles downstream of the MNGP site. The second was an ice jam event at the same location in March 1984 , with water levels just 1.3 ft below the 1965 record. Page I 25 of 35 Flood Haza r d Ree v a l uation Report Re v.O N SPM -Mon t ice ll o Nuclea r Gene r at i ng P l a nt Ice jams w ere presumed t o form at four l ocations -at the closest upstream br i dge (C l ea rw ate r Br i dge at Route 24 , appro xi mate l y 13 m il es upstream of MNGP), at the i s l ands just upstream of MNGP, at the MNGP site , and at the closest downstream br i dge (H i gh w ay 25 B r idge). I t should be noted that there is no record of an i ce j am occurr i ng at the MNGP site. For l ocat i ons at and upstream of MNGP, conduci v e for the format i on o f ice jams , a s i mpl i fied/conservat i ve approach w as i n i t i ally app li ed by add i ng the 10-ft w ate r depth (equiva l ent to the i ncrease resu l t i ng from an i ce jam of record) directly to the winter base f l o w water surface elevation at the respective locat i ons and d i rectly transposing these elevat i ons to the s i te (w i thout attenuation). The 10-ft water depth was added to the w inter base flow water surface elevation at the MNGP s i te , the i s l ands upstream of the MNGP site , and the C l ea rw ater Br i dge upstream of the MNGP. The w inter base flow w ater surface e l e v at i on w as dete r m i ned by mode li ng ave r age flo w s during the month of Apr il i n HEC-RAS. Us i ng ave r age flo w rates i n Apr il is a conservative approach relative to average f l o w rates during the w i nter. Furthermore , NUREG/CR-7046 (Reference

2) does not require that s i mu l taneous precipitat i on-i nduced flood be considered as part of the ice i nduced f l ood i ng analys i s. If the resu l t i ng w ater surface e l evation e x ceeded the I nta k e Structure operating f l oor e l evation of 919 ft , a mo r e deta il ed analysis was performed using the HEC-RAS mode l to account for attenuat i on of the ice jam break flood wave. The HEC-RAS mode l w as a l so used to evaluate back w ater cond i tion from an ice jam do w nstream of MNGP (H i ghwa y 25 Bridge). 2.7.2 Results Us i ng the conservative approach, which d i d not i nclude HEC-RAS model i ng , the ma xi mum flood e l evat i on due to an i ce jam at the MNGP site and at the i s l a n ds just upstream of the MNGP site w as determined to be 916.6 ft and 917.3 ft , respecti v e l y. Th i s i s be l o w the e l evation of 919 ft (Intake Structure operat i ng floor); therefore , add i t i ona l ana l ys i s w as not w ar r anted fo r these two cases. Due to i ts upstream locat i on and distance from MNGP, the s i mp li fied ice jam e l evat i on (wi nter base flo w p l us 10 ft) upstream of the C l ear w ater B r idge is h i gher than the Intake Structure operat i ng floor e l evation of 919 ft. As such , the HEC-RAS mode l w as used to evalua t e the attenuation of the f l ood w a v e resu l t i ng from the breach of the ice jam. The result i ng w ate r surface ele v at i on at the MNGP s i te is 9 1 2.7 ft. The HEC-RAS est i mate d bac kw ater surface e l e v at i on at MNGP resu l t i ng from an ice jam at H i gh w a y 25 B ri dge w as determ i ned to be 910.1 ft. A summary of the r esu l ts i s prov i ded i n Tab l e 3. Table 3 -Summary of Ice Induced Flood i ng Evaluat i on Water Surface Margin to Intake Case Elevation at MNGP Structure at 919 ft (ft) (ft) Ice j a m a t H ighwa y 25 B r id ge 9 10.1 8.9 Ic e j am a t MNGP 9 16.6 2.4 Ice j a m a t the is la nds u pstr eam of M N GP 917.3* 1.7 Ice jam a t Clearwater B ri d ge upstre a m of M N G P (R oute 2 4) 9 1 2.7 6.3 *water surf a ce elevation a t the /acation of the ice jom, i.e. ot the islands upstream of M NG P. Ple a se note that t he m argin is based on a simplified analysis with significant conservatisms.

2. 7 .3 Conclusions The reevaluated flood hazard due to ice induced flood i ng w as est i mated to be 917.3 ft, w hich is belo w the l o w est e l evat i on c ri tica l to the p l ant (919 ft) and s i gn i ficant l y be l o w the plant grade (930 ft). The r efore , ice i nduced flooding does not i mpact an y safety re l ated SSCs a n d i s comp l ete ly bounded by the " F l ood i ng i n St r eams and R i vers" m echan i sm. Page I 26 o f 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 2.8 Channel Migration or Diversion The methodology and results presented in this section are based on the evaluation of channel migration or diversion effects performed in Evaluation 180999.50.2300

-02 (Reference 20). Natural channels may migrate or divert either away from or toward the site. The relevant event flooding is diversion of water towards the site. The location of the MNGP site is adjacent to the natural channel of the Mississippi River. A review of USGS 7.5-minute topographic maps from 1961and2013 show 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. A review of national landslide hazards program information shows that the area in general is not susceptible to landslides and does not show any landslides of record. Based on the review of available data, there is no evidence of recent migration of the Mississippi River in the vicinity of the MNGP site and the possibility of a migration which would result in site flooding is considered extremely remote. Furthermore, the elevation of the lowest important structure to the plant is at 919 ft (Intake Structure operating floor) and the plant grade is at 930 ft. Normal water surface elevation in the Mississippi River at the site is 905 ft, which is approximately 14 ft below the critical elevation of the Intake Structure. Even in the highly unlikely event of channel diversion or migration, MNGP has a significant margin that would accommodate the potential increase in water surface elevation. There are no man-made channels , canals , diversions , or permanent levees used for conveyance of water and flood protection located near the site. Based on the above , channel migration or diversion is not considered an applicable flood hazard for the MNGP site. 2.9 Combined Effect Flood Combined effect floods as described in ANSl/ANS-2

.8-1992 (Reference

1) and Appendix H.1 Floods Caused by Precipitation Events of NUREG/CR-7046 (Reference

2) were considered as part of the flood hazard reevaluation.

The relevant combinations of flooding mechanisms are discussed in the previous sections under the individual flood causing mechanisms. 2.10 Interim Evaluations The following sections provide a description of interim evaluations that were performed as part of the flooding hazard reevaluation to assess the impact of increased flood levels at critical door openings.

No interim actions were deemed necessary in response to the reevaluated flood hazard. 2.10.1 Evaluation of Internal Flooding during the LIP As determined in Calculation 180999.51.1005 (Reference 18), the LIP flood levels exceed finished floor elevation at several critical openings. These critical door openings are either maintained closed or in some instances internal doors would prevent water from entering areas with safety-related SSCs. However , none of the critical door openings have been designed to be watertight and they can be classified into two categories:

1. Door openings with a visible gap between the bottom of the door and the door sill. 2. Door opening without a visible gap between the bottom of the door and the door sill. For the first door category, an engineering analysis was performed to calculate possible peak flow rates , total estimated inflow volumes , and total estimate inflow time through each door opening using the stage hydrograph obtained from the HEC-RAS model and th e standard orifi c e equation. The result s of thi s Page I 27 of 35 Flood Hazard Ree v a l ua t i on Report Rev.O N SP M -Mo n tice ll o Nu cl ear Gene r at i ng Plant eva l uation are presented i n Tab l e 4. Based on these resu l ts , the peak i n fl o w rate and tota l estimated i nflo w vo l ume w ere calculated at t he Ra il car Entry to the Turb i ne Building (Doors 45 and 46) at 6.9 cfs and 19 , 113 ft 3 , r espect i ve l y. The peak inflo w rate and total est i mated i nflow vo l ume were ca l cu l ated at the Emergency D i esel Generator Building at 0.23 cfs and 99 ft 3 , respect i ve l y. The calcu l ated peak i nf l o w rates and tota l est i mated i nflo w volumes into both the Emergency D i ese l Generato r Bu il ding and the Turb i ne Bu il d i ng are fess than the accepta n ce cr i ter i a. The maximum acceptab l e i nflow rate i nto the Emergency D i ese l Generato r Bu il d i ng w as determined to be less than 0.734 cfs. The max i mum acceptab l e inflow volume i nto the Turbine Bu i l d i ng was determ i ned to be l ess than 140 , 874 ft 3* Table 4 -Estimated Inflow Rates throuch Door Open i ncs durinc the LIP Event Opening E st i mated Max i mum Doo r Gap Total Tot a l In v ert/ Maximum W a ter Open i ng He i ght at Peak Inflo w Esti m ated Estimated Open i ng Location Sill Leve l Depth at Bottom Coefficient Rate Inflow Inflow Width (FFE) WSE Opening of Door of Discharge Volume Time f e etNGV029 feet in c he s ds ft3 m i nutes Railcar Entry -Turbine 1.00 0.71 6.9 19 , 113 Bu i ld i ng (Door 24 -see 935.00 935.83 0.83 16.00 note) 0.25 0.70 1.7 4 , 733 Emergency Di e sel Genera t or B uild i ng -93 1.00 93 1.41 0.4 1 3.00 0.25 0.70 0.23 99 East 3-ft wide Man D oo r (Door8) Emergency Diese l Generator Bldg -931.00 9 31.41 0.4 1 3.00 0.25 0.70 0.23 99 We s t 3-ft wide Man Door (Doo r 7) Note -the o verage gap between the bottom of the door and the doorsill at the Roi/c ar Entry to the Turbine Building i s 1.0 in c h. As a sensitivity analys i s and for i nformat i on purposes, inflow rate and inflow vo l ume were c alcu l ated for a 0.25-in c h gop. For t h e second door category , a qua li ta tiv e ana ly s i s w as performed to e v a l uate the potent i a l i mpact of il ea k age through the non-w atert i ght doors. The description of the ana l ys i s for each respect i ve door i s provided below:

Intake Structure Door from the Screenhouse (Door 209) T he door i s bet w een the I nta k e Structure and the Screenhouse. The door i s norma lly ma i nta i ned closed. Th e door s ill/in v ert i s at elevat i on 919.5 ft and the Sc r eenhouse fl oor is a t e l e v at i on 919 ft. The bottom of the door open i ng is s ix i nches abo v e the floor of the Screen h ouse. Therefore, w ater w ou l d need to poo l to a depth of s ix i nches i n the Sc r een ho u se before reach i ng the door s ill. Ho w ever, w ater enter i ng the Screenhouse w ou l d dra i n back i nto the I ntake , preclud i ng i t pool i ng and reach i ng the doo r s ill. Therefore, l eakage through th i s door i s not e x pected.

Tu r bine Building Addition (TBA) Roll-Up Doors (Doors 119 and 120) As the TBA ro ll-up doors can be either open or closed, the ro ll-up doors a r e not cred i ted wi th preclud i ng w ate r ingress. In l i eu of credit i ng the ro ll-up doors , the capab ili ty of i nternal Door 30 to preclude w ater i ng r ess w as eva l uated. When cl osed, the r o ll-up doors w ould provide redundancy.

A w a l kdo w n performed on Ma r ch 28 , 20 1 6 did not ident i fy any ob vi ous gaps at Doo r 30 that w ou l d a ll o w w ater i ng r ess. Ho w e v e r, th e door i s not des i gned as a w atert i ght door and some m i nor amount of w ate r l ea k age i s poss i b l e. Th e Turb i ne Bu i l d i ng can accommodate a l arge vo l ume of w ater (1 4 0 , 874 cub ic feet Page I 28 of 35 70 70 1 4 14 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant per Reference 17). Due to the minor amount of anticipated water ingress, the relatively short time period of the LIP, and the available volume in the Turbine Building there is reasonable assurance that this possible water leakage would not impact SSCs important to safety.

Rail Car Entry to Reactor Building (Doors 45 and 46) Doors 45 and 46 are normally maintained closed. A walkdown performed on March 28, 2016 did not identify any obvious gaps at Doors 45 that would allow water ingress. Door 45 was closed, thus, it was not possible to observe Door 46 without entering the Reactor Building.

However, these doors are not designed as watertight doors and some minor amount of water leakage is possible. The Reactor Building can accommodate a large volume of water (at least 300,000 gallons per Reference 17). Due to the minor amount of anticipated water ingress, the relatively short time period of the LIP, and the available volume in the Reactor Building there is reasonable assurance that this possible water leakage would not impact SSCs important to safety.

EFT Room Door (Door 341) Door 341 is normally maintained closed. A walkdown performed on March 28, 2016 did not identify any obvious gaps at Door 341 that would allow water ingress. However, the door is not designed as a watertight door and some minor amount of water leakage is possible.

Water leakage past the EFT door would flow down the nearby stairway into the Administrative Building; which (based on the walkdown) can accommodate a relatively large volume of water. Due to the minor amount of anticipated water ingress, the relatively short time period of the LIP event, and the available volume in the Administrative Building there is reasonable assurance that this possible water leakage would not impact SSCs important to safety.

13.8 KV Room Door (Door 1) Door 1 is normally maintained closed. A walkdown of performed on March 28, 2016 did not identify any obvious gaps at Door 1 that would allow water ingress. However , the door is not designed as a watertight door and some minor amount of water leakage is possible.

Leakage past Door 1 could possibly leak past other door(s) from the 13.8 KV Room into the Turbine Building.

The Turbine Building can accommodate a large volume of water (140,874 cubic feet per Reference 17). Due to the minor amount of anticipated water ingress, the relatively short time period of the LIP event, and the available volume in the Turbine Building there is reasonable assurance that this possible water leakage would not impact SSCs important to safety.

Off Gas Stack Door (Door 193) Door 193 is maintained closed. A walkdown performed on March 28, 2016 did not identify any obvious gaps around the doors. However, the door is not designed as a watertight door and some minor amount of water leakage is possible. Leakage into the Off Gas Stack could pool and, if the level builds up sufficiently, could leak into the Reactor Building.

The Reactor Building can accommodate a large volume of water (at least 300,000 gallons per Reference 17). Due to the minor amount of anticipated water ingress, the relatively short time period of the LIP event, and the available volume in the Reactor Building there is reasonable assurance that this possible water leakage would not impact SSCs important to safety. Page I 29 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant

Fuel Oil Transfer Pump House Door (Door 483) Door 483 is maintained closed. A walkdown performed on March 28, 2016 did not identify any obvious gaps at Door 483 that would allow water ingress. However, the door is not designed as a watertight door and some minor amount of water leakage is possible.

Leakage into the Fuel Oil Transfer Pump House could pool in the building, which does not have a large available volume to accommodate leakage. Calculation 180999.51.1005 (Reference

18) indicates that the water elevation would be above 931 ft for a short period of time (less than 16 minutes at the nearby Emergency Diesel Generator Room doors). Due to the minor amount of anticipated water ingress and the relatively short time period that the water elevation would be above the door sill, there is reasonable assurance that this possible water leakage would not impact SSCs important to safety. 2.10.2 Structural Evaluation of Doors for LIP Loads Consideration was also given to hydrodynamic and debris impacts during the LIP event. The maximum flood level predicted during the LIP event is 935.83 ft. The LIP event will not include any debris impact or any appreciable hydrodynamic effects due to the direction of all flow being away from the building. As stated in the previous section, several critical door openings will be subjected to water loading without flood protection.

An interim structural evaluation was performed in Calculation 180999.51.1010 (Reference

23) by comparing existing allowable pressure, differential pressure , or capacity qualifications for each door to resultant LIP loading. The results of the evaluation indicate that the existing allowable pressure, differential pressure , or capacity qualifications bound the resultant LIP loading. Therefore, no re-analysis of the critical door openings is necessary as part of the flood hazard reevaluation.

Page I 30 of 35 F l oo d Haza r d Ree v a l uat i on Report Re v.O NSP M -Mont i ce ll o N uclear Gene r at ing P l ant 3. Comparison of Current Design Basis and Reevaluated Flood Hazard 3.1 Comparison of Flood Hazard Elevations T ab le 5 prov i des a compar i son of t h e cur r ent des i gn ba sis and r eeva lu ated fl ood h azard e l e v at i ons and an assessment w he t her t h e r ee v a l uated f l ood hazard e l e v at i on i s bounde d by the cu rr ent des i gn bas i s flood e l e v at i on. Tab l e 5 -Summa ry of Cu rre nt De si g n B asis and ReeVll l uated F l ood H a za rd E le Vllt i ons Flood Caus i ng Current Des i gn Bas i s Flood Hazard Current Design Basis Bounds Reevaluat i on Mechani s m Flood Hazard Elevation Reeva l uat i on E l evation F l ood Hazard E l evation? L oca l Intense Not specifica lly 935.83 ft Not Bou n ded Pre ci p i ta ti on add r essed i n the US A R F l ood i ng in S tr ea ms 9 3 9.2 ft -ft Bo un ded and R iv er s Da m Breac h es a n d Not spe cifi ca lly Sc r ee n ed O u t Bo un ded 1 Fa il u r es addressed in the USAR S torm Surge Not spe cific a lly Screened O u t Bounde d 1 addressed i n the USAR Se ic he Not specifica lly Screened Ou t Bo u nded 1 addressed in t h e US A R T su nam i Not spec ifi ca lly Scree n ed O ut Bo u nded 1 addre ss ed in the U S A R Ic e I nd u ced Fl ood i ng Not quant ifi ed in t h e 91 7.3 ft Bounded 2 USAR C h anne l Mi grat i on o r Not spe cifi ca lly Screened Out Bounded 1 D iv e r s i on addressed i n the US A R Comb i ned Effects Not spec ifi ca lly F l ood -P MF wit h -ft Bou n de d 3 a dd r essed i n th e U S AR w a v e run up 1 These flood-cau si ng mechan i sms w ere not s pecifi ca lly addr ess ed i n the USAR; ho w e v er , a s c re e n i ng l e v e l analys i s s ho w ed that these mechan i sms w ere not app li cable and are c ompl e tely bounded b y other mec h an is ms. Sin c e the s e mechan is ms w ere screened out as part of the flood hazard ree v aluat i on , they w ere a l so considered bounded by the current des i gn bas i s. 2 Wh i le the i ce induced flooding hazard was not quantified i n th e USAR , the re s ultant flood el e vat i on is be l ow elevation s that w ould impact SS Cs i mportant to s afety. Furthermor e, thi s hazard i s fully bounded by the flood i ng in s tream and rivers hazard. 3 Wh i le the w ave run-up flood i ng hazard w as not specifi c ally addre s sed in the USAR , the re s ultant wa v e runup elevat i on is fully bounded by the current de si gn bas i s (CDS) still w ater ele v at i on. 3.2 Comparison of Flood Parameters Th e M a r ch 1 2 , 2012 , 5 0.5 4(f) l e tt e r (R e f e r e nce 3) r e qu e sted th a t a n int eg r a t e d ass e ssment o f t h e plant's resp o ns e to the reevaluated fl ood h a z a rd be pe rformed if the re e v a lu at ed fl ood h a zard e l e v at i o n is n o t Page I 31 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant bounded by the current design basis. If the reevaluated flood hazard elevation is not bounded , the NRC requested that the licensee define the applicable flood parameters and perform an integrated assessment. The applicable flood parameters include the following per JLD-ISG-2012-05 (Reference 6): 1. Flood height and associated effects a. Stillwater elevation;

b. Wind waves and run-up effects; c. Hydrodynamic loading , including debris; d. Effects caused by sediment deposition and erosion (e.g., flow velocities, scour); e. Concurrent site conditions , includ i ng adverse weather conditions; and f. Groundwater ingress. 2. Flood event duration parameters

a. Warning time (may include information from relevant forecasting methods (e.g., products from local , regional, or national weather forecasting centers) and ascension time of the flood hydrograph to a point (e.g. intermediate water surface elevations) triggering entry into flood procedures and actions by plant personnel);

b. Period of site preparation (after entry into flood procedures and before flood waters reach site grade); c. Period of inundation; and d. Period of recession (when flood waters completely recede from site and plant is in safe and stable state that can be maintained). 3. Plant mode(s) of operation during the flood event duration 4. Other relevant plant-specific factors (e.g. waterborne projectiles)

Since the reevaluated flood hazard elevation for the LIP event is not bounded , the applicable flood parameters for this flood causing mechanism were defined and are provided in Table 6. Page I 32 of 35 Flood Hazard Reeva l uation Report Rev.O NSPM -Monticello Nuclea r Gen e rating P l ant Tab l e 6 -Local Intense Precipitat i on COB Reevaluated Flood Bounded (B) or Flood Scenario Parameter Flood Hazard Not Bounded Hazard (NB) 1. Maximum Stillwater El e vat i on (ft) 935.83 NB .., a::: 2. Max i mum Wave Run-up E l evation (ft) 4: See Note 2 N/A .. Vl ::> "' 3. Max i mum Hydrodynamic/Deb ri s Loading (psf) See Note 3 N/A ;{ QJ ..c .., v ...., c .. :e 4. Effects of Sediment Deposition

/Eros i on c See Note 4 N/A Gi ... *-"C _, 5. Concurrent Site Conditions QJ See Note 5 N/A .., VI 8 VI QJ ..: .._ 6. Effects on Groundwater "C See Note 6 N/A "C ro 7. Warn i ng T i me (hours) £ See Note 7 N/A ro u .. 8. Per i od of Site Preparation (hours) ;;;;: See Note 8 N/A c c *u .2 ..... QJ '8 9. Per i od of Inundat i on (hours) a. "'1.2 (see Note 9) NB VI .2 0 ...., u.. 0 10. Peri o d of Recession (hours) c "'5.3 (see Note 10) NB VI ro 11. Plant Mode of Operations

!: See Note 11 N/A ] c... ::::; 0 12. Other Factors See Note 12 N/A Additi o na l notes, "N/A" justifications (why a particular parameter is judged not to affect the s i te), and explanations regarding the boun d e d/n on-bo un d ed d etermination.

1. No n e 2. Con s ide ra tion of wind-gener a ted wave a cti o n for the LIP event is not explicitly required i n NUREG/CR-7046, ANSI/ ANS-2.8-1992 or t he 50.54(f) letter. Furt h er m ore, wave runup is considered negligible due to lim i ted fl ood d epths and fetch. 3. Hydrodynamic l oading was not considered plausible due to surface water flow d i rect i on is not towards the buildings. Debris im pact loading was not considered plausible due to limited velocities and flood depths (Reference 18}. 4. Due to limited ve l ocities, and short du r ation of flooding (Reference 18}, sediment deposition a n d erosion is not considered to have a n effect on t he LIP floo d levels. 5. High w i nds a nd ha i l c o ul d coinci de w i th the LIP event. I n general, n o manual actions are required to be p e rformed o uts i de. Pers o nnel may be , however , exp o se d t o the elements while moving between locations.

Environmental c o nditions woul d be considered p rior to pers o nnel being d i r ected to move between locations. 6. D ue t o relatively s ho rt duratio n of the LIP event (Reference 18}, surch a rge t o groundwater is not considered.

7. Warn i ng time is not credited in the flood protection strategy (s i nce only permanent/passive measure s are used for the LIP floo d) and, therefore, was not consi d ered as part of the analysis. 8. SSCs imp o rtant to safety are protecte d by means of permanent/passive measures and, the r efore, site preparation was no t considered as part of the analysis. 9. T he p eri od of inun d ation va r ies t hro u ghout t h e site; however , at the critical d oor location with the highest water surface elev at io n , it was estimated that w a ter level woul d remain above fin i shed floor elevat i on for 70 minutes (Reference 18}. 1 0. O nc e t he flood waters rece d e below finished floor elevation, i t woul d take approximately 5.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months for flood waters to com pl e t e l y rece d e fr o m a reas near the critical doors, which is approximate l y within 30 minu t es after the en d of t h e 6-hr st o rm LIP event (Reference 18). 11. T her e are no l im i tations on p l ant mo d es of operation prior to, or during, the LIP event. 1 2. T here a re no o ther fact o rs, incl ud ing w a ter bo rne projectiles, ap p lica b le to t his floo d ca u sing mecha n ism. Page I 33 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 4. References

1. American Nuclear Society, "Determining Design Basis Flooding at Power Reactor Sites," ANSl/ANS-2.8-1992, 1992. 2. U.S. Nuclear Regulatory Commission, "Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America," NUREG/CR-7046, November 2011, ADAMS Accession No. ML11321A195.

3. U.S. Nuclear Regulatory Commission, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f} regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 12, 2012, ADAMS Accession No. ML12053A340.

4. U.S. Nuclear Regulatory Commission, "Guidance for Assessment of Flooding Hazard due to Dam Failures," JLD-ISG-2013-01, July 29, 2013, ADAMS Accession No. ML13151A153.

5. U.S. Nuclear Regulatory Commission, "Guidance for Performing a Tsunami, Surge, or Seiche Hazard Assessment," JLD-ISG-2012-06, January 4, 2013, ADAMS Accession No. ML12314A412.

6. U.S. Nuclear Regulatory Commission, "Guidance for Performing the Integrated Assessment for External Flooding," JLD-ISG-2012-05, November 30, 2012, ADAMS Accession No. ML12311A214.

7. U.S. Nuclear Regulatory Commission, "Tsunami Hazard Assessment at Nuclear Power Plant Sites in the United States of America," NUREG/CR-6966, March 2009, ADAMS Accession No. ML091590193.

8. U.S. Nuclear Regulatory Commission, "Monticello Nuclear Generating Plant -Transmittal of U.S. Army Corps of Engineers Flood Hazard Reevaluation Information (TAC No. MF3696}," November 18, 2015, ADAMS Accession Nos.: ML15296A365 (Package}, ML15324A383 (PUBLIC Letter} and ML 15296A274 (NON PUBLIC Letter}. 9. U.S. Nuclear Regulatory Commission, "Summary of July 9, 2015 closed meeting between representatives of the U.S. Army Corps of Engineers, U.S. Nuclear Regulatory Commission, and Northern States Power Company-Minnesota, to discuss flood analysis associated with Monticello Nuclear Generating Plant and Prairie Island Nuclear Generating Plant, Units 1 and 2 (TAC Nos. MF3696, MF3697, and MF3698}," October 2, 2015, ADAMS Accession No. ML15271A207.

10. NSPM, "Monticello Updated Safety Analysis Report," Revision 32. 11. NSPM, "Revision to MNGP Final Response to NRC Request for Information Pursuant to 10 CFR 50.54(f} Regarding the Flooding Aspects of Recommendation 2.3 ft of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," July 31, 2013. 12. NSPM, Procedure A.6 "Acts of Nature," Rev. 53. 13. NSPM, Procedure 8300-02 "External Flooding Protection Implementation to Support A.6 Acts of Nature," Rev 7. 14. NSPM, "Request for NRC Assistance to Obtain Information on Dams from the U.S. Army Corps of Engineers (USACE}," March 5, 2014, ADAMS Accession No. ML14064A291.

15. NSPM, "External Flooding Implementation Timeline," Calculation 14-080, Rev 2. 16. NSPM, "Debris Barrier Design for Impact Protection of External Flood Barrier," ECN 23429. Page I 34 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 17. NSPM, "Reactor Bldg, Turbine Building & Intake Structure Water Height -Internal Flooding," Calculation No. CA-07-021, Rev. 0. 18. Black & Veatch, "Local Intense PMP Hydrology and Hydraulics," Calculation 180999.51.1005, Rev. 2. 19. Black & Veatch, "Site Specific PMP and Ancillary Meteorological Analysis," Calculation 180999.51.1008, Rev. 1. 20. Black & Veatch, "MNGP Flood Scenario Evaluations," 180999.50.2300-02, Rev. 0. 21. Black & Veatch, "Ice Induced Flooding," Calculation 180999.51.1001, Rev. 1. 22. Black & Veatch, "MNGP Wave Prediction and Wave Runup," Calculation 180999.51.1002, Rev. 0. 23. Black & Veatch, "Evaluation of Structural Elements -Flood," Calculation 180999.51.1010, Rev. 0. 24. National Oceanic and Atmospheric Administration, "Probable Maximum Precipitation Estimates, United States East of the 105th Meridian," Hydrometeorological Report No. 51, June 1978 25. National Oceanic and Atmospheric Administration, "Application of Probable Maximum Precipitation Estimates

-United States East of the 105th Meridian," Hydrometeorological Report No. 52, August 1952. 26. U.S. Department of Agriculture, "Urban Hydrology for Small Watersheds -Technical Release 55," June 1986. 27. U.S. Department of Agriculture, "Part 630 Hydrology National Engineering Handbook," September 1997. 28. Kansas City Metropolitan Chapter, American Public Works Association Standard Specifications

& Design Criteria, Section 5600 Storm Drainage Systems & Facilities, February 16, 2011. 29. U.S. Army Corps of Engineers, "Coastal Engineering Manual," Engineer Manual 1110-2-1100.

Part II Coastal Hydrodynamics, 2015 and Part VI Design of Coastal Project Elements, 2011. Page I 35 of 35 Appendix 1 50.54{f) Letter -Requested Information Cross-Reference Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant This appendix provides a list of each item requested in Enclosure 2 ofthe NRC Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident; dated March 12, 2012 and the corresponding section(s) in the main body of the MNGP FHRR where the requested information is provided.

a. Site information related to the flood hazard. Relevant SSCs important to safety and the UHS are included in the scope of this reevaluation, and pertinent data concerning these SSCs should be included.

Other relevant site data includes the following:

i. Detailed site information (both designed and as-built), including present-day site layout, elevation of pertinent SSCs important to safety, site topography, as well as pertinent spatial and temporal data sets: Response:

See Section 1.4 for detailed site information.

ii. Current design basis flood elevations for all flood causing mechanisms:

Response:

See Section 1.5 which describes current design basis flood hazards. iii. Flood-related changes to the licensing basis and any flood protection changes (including mitigation) since license issuance:

Response:

See Section 1.6 for description offlood-related changes to the licensing basis and any flood protection changes (including mitigation) since license issuance.

iv. Changes to watershed and local area since license issuance:

Response:

See Section 1. 7 for any changes to watershed and local area since license issuance.

v. Current licensing basis flood protection and pertinent flood mitigation features at the site: Response:

See Section 1.8 for current licensing basis flood protection and mitigation features.

vi. Additional site details, as necessary, to assess the flood hazard (i.e. bathymetry, walkdown results, etc.): Response:

No additional information beyond the information provided in the above-mentioned sections was required to assess the flood hazard.

The walkdown reports are referenced, as relevant, in Sections 1.5 and 1.6. b. Evaluation of the flood hazard for each flood causing mechanism, based on present-day methodologies and regulatory guidance.

Provide an analysis of each flood causing mechanism that may impact the site including local intense precipitation and site drainage, flooding in streams and rivers, dam breaches and failures, storm surge and seiche, tsunami, channel migration or diversion, Appendix 1 Page I 2 of4 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant and combined effects. Mechanisms that are not applicable at the site may be screened-out; however, a justification should be provided.

Provide a basis for inputs and assumptions, methodologies and models used, including input and output files, and other pertinent data: Response:

A description of the flood hazard reevaluation for each flood causing mechanism is provided in the FHRR as referenced below:

Local Intense Precipitation (LIP} and Site Drainage:

See Section 2.1;

Flooding in Streams and Rivers: See Section 2.2;

Dam Breaches and Failures:

See Section 2.3;

Storm Surge including Wind-Wave Activity:

See Section 2.4;

Seiche: See Section 2.5;

Tsunami: See Section 2.6;

Ice Induced Flooding:

See Section 2.7;

Channel Migration and Diversion:

See Section 2.8;

Combined Effects (including wind-waves and runup effects):

See Section 2.9 -the relevant combinations of flooding mechanisms are discussed under the individual flood causing mechanisms;

Other Associated Effects (i.e., hydrodynamic/debris loading, effects caused by sediment deposition and erosion, concurrent site conditions, and groundwater ingress):

See Table 6. Note that other associated effects are only applicable to the LIP since LIP was the only bounded flood causing mechanism for MNGP;

Flood Event Duration Parameters (i.e., warning time, period of site preparation, period of inundation, and period of recession):

See Table 6. Note that flood duration parameters are only applicable to the LIP since LIP was'the only non-bounded flood causing mechanism for MNGP. c. Comparison of current and reevaluated flood causing mechanisms at the site. Provide an assessment of the current design basis flood elevation to the reevaluated flood elevation for each flood causing mechanism.

Include how the findings from Enclosure 4 of the 50.54(f) letter (i.e., Recommendation 2.3 flooding walkdowns) support this determination.

If the current design basis flood bounds the reevaluated hazard for all flood causing mechanisms, include how this finding was determined.

Response:

A comparison of the current design basis and reevaluated flood hazard elevations for each flood causing mechanism is provided in Section 3.1 and Table 5. It was determined that the current design basis flood bounds the reevaluated hazard for all applicable flood causing mechanisms, including combined-effects flooding, with the exception of the LIP flood hazard. The following provides additional detail for each reevaluated flood causing mechanism:

Appendix 1 Page I 3 of4 Flood .Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant i. Local Intense Precipitation (LIP): Since the LIP flood hazard is not addressed in the current design basis, the reevaluated LIP hazard is considered to be non-bounded.

See Section 2.1 for the LIP analysis and Section 3.2 and Table 6 for the LIP Flood Scenario Parameters.

ii. Flooding in Stream and Rivers: Based on the USACE analysis performed for the NRC, the current design basis bounds the reevaluated flood hazard. See Section 2.2 and Table 5. iii. Dam Breaches and Failures:

Based on the USACE analysis performed for the NRC, it was determined that potential upstream dam breaches and failures, regardless of failure mode, do not increase the flood hazard at MNGP. See Section 2.3 and Table 5. iv. Storm surge, seiche and tsunami: These hazards were screened out as not applicable/not plausible at the MNGP site. See Sections 2.4, 2.5, and 2.6 and Table 5. v. Ice Induced Flooding:

The reevaluated ice-induced flooding hazard was determined to be fully bounded by the combined-effects flooding and, therefore, considered to be bounded. See Section 2.7 and Table 5. vi. Channel Migration and Diversion:

This hazard was found to not be applicable/not plausible at the MNGP site. See Section 2.8 and Table 5. vii. Combined-Effect Flood: Based on the combination of the PMF (Section 2.2.1) and the generated wave analysis (Section 2.2.2), the current design basis bounds the reevaluated flood hazard. See Table 5. d. Interim evaluation and actions taken or planned to address any higher flooding hazards relative to the design basis, prior to completion of the integrated assessment described below, if necessary:

Response:

An interim evaluation was performed to assess the potential impact of the reevaluated LIP flood hazard on the plant. The result of the interim evaluation for the LIP is that there is no adverse impact to safety-related SSCs. See Section 2.10. e. Additional actions beyond Requested Information item 1.d taken or planned to address flood hazards, if any: Response:

None required.

Appendix 1 Page I 4 of4 L-MT-16-024 ENCLOSURE 2 FLOOD HAZARD REEVALUATION REPORT IN RESPONSE TO THE 50.54(f) INFORMATION REQUEST REGARDING RECOMMENDATION 2.1: FLOODING OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAl-ICHI ACCIDENT for the MONTICELLO NUCLEAR GENERATING PLANT RENEWED LICENSE NO. DPR-22 (REDACTED VERSION) 39 pages follow FLOOD HAZARD REEVALUATION REPORT IN RESPONSE TO THE S0.54(f) INFORMATION REQUEST REGARDING RECOMMENDATION 2.1: FLOODING OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAl-ICHI ACCIDENT for the MONTICELLO NUCLEAR GENERATING PLANT RENEWED LICENSE No. DPR-22 Prepared by: Black & Veatch and Aterra Solutions Rev.O Printed Name Affiliation Signature Date Prepared by: Petr Masopust Aterra Solutions*

4/21/2016 Reviewed by: Adam Liebergen Black & Veatch :ffi "fl.1s-hiW-Approved by: Steven Thomas Black & Veatch t//trliu1' Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant Table of Contents List of Figures .....................................................................................................................................................

4 List of Tables ......................................................................................................................................................

5 List of Appendices

..............................................................................................................................................

6 Acronyms and Abbreviations

.............................................................................................................................

7 1. Information Related to the Flood Hazard ..................................................................................................

9 1.1 Introduction

.......................................................................................................................................

9 1.2 Purpose ..............................................................................................................................................

9 1.3 Method ...............................................................................................................................................

9 1.4 Detailed Site Information

.................................................................................................................

10 1.4.1 Elevation Values .......................................................................................................................

10 1.4.2 Site Layout and Topography

.....................................................................................................

11 1.4.3 Elevation of Safety Related Structures, Systems and Components

.........................................

11 1.5 Current Design Basis Flood Elevations

.............................................................................................

12 1.5.1 Flooding in Streams and Rivers ................................................................................................

12 1.5.2 Ice Induced Flooding ................................................................................................................

14 1.6 Flood-Related Changes to the Licensing Basis and Flood Protection and Mitigation Changes since License Issuance ...........................................................................................................................................

14 1.7 Watershed and Local Changes .........................................................................................................

14 1.8 Licensing Basis Flood Protection and Mitigation Features ..............................................................

15 2. Flood Hazard Reevaluation

......................................................................................................................

16 2.1 Local Intense Precipitation

...............................................................................................................

16 2.1.1 Methodology

.............................................................................. , .............................................

16 2.1.2 Results ......................................................................................................................................

19 2.1.3 Conclusions

..............................................................................................................................

19 2.2 Flooding in Streams and Rivers ........................................................................................................

20 2.2.1 Maximum Stillwater Elevation

.................................................................................................

20 2.2.2 Wind-Generated Waves ...........................................................................................................

22 2.3 Darn Breaches and Failures ..............................................................................................................

24 2.4 Storm Surge, including Wind-Wave Activity ....................................................................................

24 2.5 Seiche ................................................................................................................................................

24 2.6 Tsunami ............................................................................................................................................

25 2.7 2.7.1 2.7.2 Ice Induced Flooding ........................................................................................................................

25 Methodology

............................................................................................................................

25 Results ......................................................................................................................................

26 Page I 2 of 35 ,.:..._.

Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 2.7.3 Conclusions

..............................................................................................................................

26 2.8 Channel Migration or Diversion

.......................................................................................................

27 2.9 Combined Effect Flood .....................................................................................................................

27 2.10 Interim Evaluations

..........................................................................................................................

27 2.10.1 Evaluation of Internal Flooding during the LIP .........................................................................

27 2.10.2 Structural Evaluation of Doors for LIP Loads ...........................................................................

30 3. Comparison of Current Design Basis and Reevaluated Flood Hazard .....................................................

31 3.1 Comparison of Flood Hazard Elevations

.......................... ...............................................................

31 3.2 Comparison of Flood Parameters

....................................................................................................

31 4. References

...............................................................................................................................................

34 Page I 3 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant List of Figures Figure 1-MNGP Site and Nearby Vicinity .......................................................................................................

10 Figure 2-MNGP Power Block Area .................................................................................................................

12 Figure 3 -Local Intense 6-Hr All-Season PMP Hyetograph

.............................................................................

17 Figure 4 -LIP Drainage Areas and Directions of Flow .....................................................................................

18 Figure 5 -Probable Maximum Flood Flow and Stage Hydrographs at MNGP Site (River Station 900.5) ....... 22 Figure 6 -Inundation Limits and Fetch Lines ..................................................................................................

23 Page I 4 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant List of Tables Table 1-All-Season LIP Calculations and Cumulative Depth ..........................................................................

16 Table 2-LIP Evaluation Results .......................................................................................................................

19 Table 3 -Summary of Ice Induced Flooding Evaluation

..................................................................................

26 Table 4 -Estimated Inflow Rates through Door Openings during the LIP Event ............................................

28 Table 5 -Summary of Current Design Basis and Reevaluated Flood Hazard Elevations

................................

31 Table 6-Local Intense Precipitation

...............................................................................................................

33 Page I 5 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant List of Appendices Appendix 1-50.54(f) Letter -Requested Information Cross-Reference Page I 6 of 35 Flood Hazard Reevaluation Report NSPM -Monticello Nuclear Generating Plant Acronyms and Abbreviations ACES ADAMS ANS ANSI APWA B COB CEM CFR cfs d/b/a DEM EC EFT EM FFE FIS FLEX ft ft3 ft/s H HEC-HMS HEC-RAS HHA HPCI HMR hr ISFSI ISG JLD km 2 LiDAR LIP m mi 2 MNGP Automated Coastal Engineering System Agencywide Documents Access and Management System American Nuclear Society American National Standards Institute American Public Works Association Bounded current design basis Coastal Engineering Manual Code of Federal Regulations cubic (foot) feet per second Doing Business As Digital Elevation Model Engineering Change Emergency Filtration Train Engineer Manual finished floor elevation Flood Insurance Study Diverse and Flexible Coping Strategies foot (feet) cubic foot (feet) feet per second horizontal Hydrologic Engineering Center Hydrologic Modeling System Hydrologic Engineering Center River Analysis System Hierarchical Hazard Assessment High-Pressure Coolant Injection Hydrometeorological Report hour Independent Spent Fuel Storage Installation Interim Staff Guidance Japan Lessons-Learned Directorate square kilometer(s)

Light Detection and Ranging Local Intense Precipitation meter(s) square mile(s) Monticello Nuclear Generating Plant Rev.a Page I 7 of35 Flood Hazard Reevaluation Report NSPM -Monticello Nuclear Generating Plant mph MSL N/A NAVD 88 NGDC NGP NB NGVD 29 NOAA NRC NSPM NTIF NU REG NU REG/CR PMF. PMP psf RHR scs SS Cs TBA TR USA CE USAR USGS v WSE yr miles per hour mean sea level not applicable North American Vertical Datum of 1988 National Geophysical Data Center Nuclear Generating Plant not bounded National Geodetic Vertical Datum of 1929 National Oceanic and Atmospheric Administration United States Nuclear Regulatory Commission Northern States Power Company, a Minnesota corporation Near Term Task Force NRC technical report designation NUREG contractor report probable maximum flood probable maximum precipitation pound(s) per square foot Residual Heat Removal Soil Conservation Service structures, systems, and components Turbine Building Addition Technical Release United States Army Corps of Engineers Updated Safety Analysis Report United States Geological Survey vertical water surface elevation year Rev.a Page I 8 of35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1. Information Related to the Flood Hazard 1.1 Introduction In response to the nuclear fuel damage at the Fukushima Dai-ichi power plant due to the March 11, 2011 earthquake and subsequent tsunami, the United States Nuclear Regulatory Commission (NRC) established the Near Term Task Force (NTIF) to conduct a systematic review of NRC processes and regulations, and to make recommendations to the Commission for its policy direction.

The NTIF reported a set of recommendations that were intended to clarify and strengthen the regulatory framework for protection against natural phenomena.

On March 12, 2012, the NRC issued an information request pursuant to Title 10 of the Code of Federal Regulations (CFR), Section 50.54 (f) (Reference

3) which included six (6) enclosures:

1. [NTIF] Recommendation 2.1: Seismic 2. [NTIF] Recommendation 2.1: Flooding 3. [NTIF] Recommendation 2.3: Seismic 4. [NTIF] Recommendation 2.3: Flooding 5. [NTIF] Recommendation 9.3: Emergency Preparedness

6. Licensees and Holders of Construction Permits In Enclosure 2 of Reference 3, the NRC requested that licensees reevaluate the flooding hazards at their sites against present-day regulatory guidance and methodologies being used for early site permits and combined license reviews. 1.2 Purpose This report provides the information requested in the March 12, 50.54(f) letter; specifically, the information listed under the "Requested Information" section of Enclosure 2 of Reference 3, paragraph 1 ("a" through "e") for Monticello Nuclear Generating Plant (MNGP). Evaluation of the eight flood-causing mechanisms and associated effects (when required), as well as the combined effect flood, identified in Attachment 1 to Enclosure 2 of the NRC information request (Reference

3) and the potential effects on the MNGP site is described in Section 2 of this report. 1.3 Method This flooding hazard reevaluation followed the Hierarchical Hazard Assessment (HHA) approach, as described in NUREG/CR-7046, "Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America" (Reference

2) and its supporting reference documents.

The HHA approach consists of a series of stepwise, progressively more refined analyses to evaluate the hazard resulting from phenomena at a given nuclear power plant site to structures, systems, and components (SSCs) important to safety with the most conservative plausible assumptions consistent with the available data. The HHA starts with the most conservative, simplifying assumptions that maximize the hazards from the maximum probable event. If the assessed hazards result in an adverse effect or exposure to any related SSC, a more site-specific hazard assessment is performed for the probable maximum event. The steps typically involved to estimate flood hazard include the following:

Page I 9 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1. Identify flood-causing phenomena or mechanisms by reviewing historical data and assessing the geohydrological, geoseismic, and structural failure phenomena in the vicinity of the site and region. 2. For each flood-causing phenomenon, develop a conservative estimate of the flood from the corresponding probable maximum event using conservative simplifying assumptions.

3. If any safety-related SSC is adversely affected by flood hazards, use site-specific data and/or more refined analyses to provide a more realistic condition and flood analysis, while ensuring that these conditions are consistent with those used by Federal agencies in similar design considerations.

Repeat Step 2; if all safety-related SSCs are unaffected by the estimated flood, or if all site-specific data have been used, specify design bases for each using the most severe hazards from the set of floods corresponding to the flood-causing phenomena.

Section 2 of this report provides additional HHA detail for each of the flood-causing mechanisms evaluated. 1.4 Detailed Site Information 1.4.1 Elevation Values Unless otherwise stated, all elevation values sited in this report are in feet above mean sea level (MSL), which is also referred to as National Geodetic Vertical Datum of 1929 (NGVD 1929). Figure 1-MNGP Site and Nearby Vicinity N W WI; 0 0.5 s 2 3 Page I 10 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1.4.2 Site Layout and Topography The site is located within the city limits of the Monticello , Minnesota on the right bank of the Mississippi River and about 70 river miles upstream from Minneapolis

-St. Paul. The plant site occupies an area of approximately 2 , 150 acres. The topography of the MNGP site is characterized by relat i vely level bluffs , which rise sharply above the river. Three distinct bluffs exist at the plant site at elevations 920 , 930, and 940 ft. Bluffs located approximately a mile north and south of the site rise to 950 ft. Further to the north, the terrain is relatively level with numerous lakes and wooded areas. To the south, west , and east , the terrain is hilly and dotted with numerous small l akes (Reference 10). The Mississippi River abuts the site to the north and northwe s t. The flow in the Mi ss i s sippi River in the vicinity of the plant is unregulated and subject to la r ge variations throughout the year. Normal river level is at elevation 905 ft and the maximum river flood stage was recorded in 1965 at elevation 916 ft. The 1 , 000-year projected river flood stage is at elevation 921 ft (Reference 10). The natural grade of the power block is at elevation 930 ft with elevations of the majority of critical structure open i ngs ranging from 931 ft to 935 ft (Refe r ence s 10 and 18). The MNGP site and the regional vicinity are shown in Figure 1. 1.4.3 Elevation of Safety Related Structures, Systems and Components Class I structures, which are vital to safe shutdown of the plant and removal of decay heat have been identified in the USAR (Reference 10), as following:

Primary Containment (Drywell , Vents, Torus, and Penetrations)

Reactor Building (up to Operating Floor -1027-foot 8-inch)

High Pressure Coolant Injection (HPCI) Building

Plant Control and Cable Spreading Structure

Spent Fuel Storage Pool

Off-gas Stack

Reactor Primary Vessel Biological Shield and Support Pedestal

Standby Diesel Generator Building

Diesel Fuel Oil Transfer House Containing Diesel Fuel Oil System

Emergency Filtration Train (EFT) Building

Intake Structure Pump Room Conta i ning Emergency Service W a te r and Re s idual Heat Removal (RHR) Service Water Pumps and Connecting Pipe Tunnel

Parts of Turbine Building Housing Cla s s I Equipment

Underground Duct Bank-3 rd Floor, EFT to Reactor Building. Plant grade for Class I structures is at elevation 930 ft , except for the Intake Structure. These structures have been designed for a river flood stage up to this elevation. Certain components of the above mentioned Class I structures may be located below elevation 930 ft; however , flood protection features (e.g., steel plates, grout , or sandbags) are installed to close any openings below elevation 930 ft (Reference 10). The operating floor ofthe Intake Structure is located at elevation 919 ft with an opening between the Screenhouse and the Intake Structure at elevation 919.5 ft (References 10 and 18). The power block area is shown in Figu re 2. Page I 11 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant Figure 2-MNGP Power Block Area 1.5 Current Design Basis Flood Elevations The current design basis flood elevations for MNGP are described in the USAR (Reference

10) as well as the recent walkdown reports required as part of NRC's 10 CFR 50.54(f) letter (Reference 3). The only flood causing mechanism quantified as part of the current design basis was "Flooding in Streams and Rivers." A summary of this flood causing mechanism and the corresponding flood hazard is provided in the following section. A brief discussion of " Ice Induced Flooding" is also provided. 1.5.1 Flooding in Streams and Rivers The probable maximum flood (PMF) in the Mississippi River adjacent to the MNGP site was determined to be 364,900 cfs with a corresponding peak flood stage of 939.2 ft. A detailed description of the methodology , inputs, assumptions and results is provided in Appendix G of the USAR. The limiting flood resulted from a combination of meteorological conditions including snowmelt that could occur in the spring and could reach maximum river level in about 12 days. It was estimated that the flood elevation would remain above elevation 930 ft for approximately 11 days (Reference 10). From flow and stage hydrographs provided in Appendix G of the USAR , it can also be estimated that the flood stage would reach elevation 919 ft (Intake Structure operating floor) in less than 4 days and would remain above this elevation for approximately 20 days. Page I 12 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant The PMF study included a watershed area of 13,900 square miles upstream of MNGP. The watershed is characterized by level to rolling prairie land interspersed with areas of glacial moraines. The average elevation of the watershed is approximately 1,200 ft. The main hydrologic characteristic of the watershed are the numerous lakes, which significantly influence the streamflow characteristics of the Mississippi River and its tributaries. The long-term mean annual runoff for the basin is approximately 5.0 inches (Reference 10). The PMF at the MNGP site was determined by transposing actual storms to the watershed and maximizing the precipitation for potential moisture. The selected storms considered were the March 23-27, 1913 storm centered at Bellefontaine, Ohio (spring storm) and the August 28-31, 1941 storm centered at Hayward, Wisconsin (summer storm). The spring storm combined with a snowmelt occurring during the storm resulted in higher discharge than the summer storm. For the purpose of the PMF study, maximum snow water equivalent for a period between March 16 and March 31 having one percent probability was used. Methods developed by the USACE and described in EM 1110-2-1406 "Runoff from Snowmelt" were used to compute snowmelt (Reference 10). Initial retention losses were assumed to be zero and infiltration losses were assumed to be 0.02 inches per hour during the snowmelt period and 0.03 inches per hour during the period following the beginning of rainfall (Reference 10). The most critical sequence of event leading to a major flood in the watershed would be a combination of unusually heavy spring snowfall and low temperatures after a period of intermittent warm spells and freezing temperatures forming an impervious ground surface followed by a period of extremely high temperatures and a major storm event. For the purposes of the design basis study, snow water equivalent having a one percent probability was assumed to cover the watershed at the beginning of the simulation.

This was immediately followed by the maximum historical temperature sequence and after 5 days the probable maximum spring precipitation was initiated (Reference 10). The watershed was divided into four major sub-basins and synthetic hydrographs were developed for each using the Snyder's method. Unit hydrograph peaks were increased by 25 percent and basin lag time decreased by one-sixth.

Snowmelt and rainfall excess were applied to the unit hydrographs and the resulting hydrographs were determined for each sub-basin, which were then routed to the MNGP site using a computer model. The travel time for flood routing was based on the USACE recorded travel times for large floods. Baseflow of 5,000 cfs, based on long-term USGS records for the Elk River, Minnesota stream gage, was added to the routed flood hydrographs (Reference 10). A stage-discharge curve at the MNGP site was developed using a computer model. The inputs for the computer model included channel and overbank cross-sections based on 2-ft and 10-ft topographic maps; Manning's n-values for left overbank , right overbank , and channel; discharges of various magnitudes; and starting water elevations. Average Manning's n-values were determined to be 0.032 for the main channel, 0.050 for the left overbank , and 0.045 for the right overbank.

A higher value of 0.065 was used for the right overbank at MNGP and a value of 0.06 was used for the island immediately upstream of MNGP. The model was verified against the maximum flood of record (April 1965) for which records of high water marks exist at several points along the river (Reference 10). Based on the results of the study described in the USAR, the limiting design basis flood for the MNGP site is the PMF of the Mississippi River of 364,900 cfs, with a peak rive r flood stage of 939.2 ft. Dam breaches and wind-wave runup were not considered in the current design basis evaluation. Page I 13 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1.5.2 Ice Induced Flooding Appendix G of the USAR (Reference

10) briefly discusses backwater flooding caused by ice jams as one of the two types of flooding occurring the Mississippi River watershed. However, the ice induced flooding is not quantified and is not considered to result in flooding levels exceeding the precipitation

-driven flooding levels. 1.6 Flood-Related Changes to the Licensing Basis and Flood Protection and Mitigation Changes since License Issuance No changes to flood elevations have been made since the issuance of the original license. Procedural changes and enhancements to flood protection were implemented following the completion of the Recommendation 2.3: Flooding Walkdowns , during which several deficiencies and observations were identified. These deficiencies and observations were documented in the flooding walkdown report (Reference 11). A summary of the changes is provided below:

Enhancements to Procedure A.6 were implemented to streamline actions described in the procedure (Reference 12).

Engineering Change (EC) 21937 was p r epared to install permanent flood protection features on and around the Intake Structure and reduce the amount of field work required during a flooding event and ultimately improve response time (Reference 11). Th i s included bin wall/earthen levee design changes (Reference

15) and design of debris barrier to protect the flood barrier from impact by debris floating in the flooded river (Reference 16). 1.7 Watershed and local Changes The watershed contributory to the Mississippi River upstream of MNGP is approximately 13,900 square miles (Reference 10). There have not been significant changes to the watershed since the last license renewal and the land use changes are relatively minimal. Local area changes have also been minimal since plant operation began at the MNGP site. Changes consistent with most nuclear plant sites have been made at MNGP since operations began , including the addition of the following structures:

Administration Buildings

Independent Spent Fuel Storage Installation (ISFSI)

Security Buildings

Warehouses

FLEX Equipment Storage Building

Security barriers The changes also included the bin wall sections adjacent to the Intake Structure , which are now permanently installed. Location and configurat i on of current structures, as relevant, were i nputs to the Local Intense Precipitation (LIP) calculations as related to the flooding impacts on SSCs. Page I 14 of 35 Flood Hazard Reevaluation Report Rev.a NSPM -Monticello Nuclear Generating Plant 1.8 Licensing Basis Flood Protection and Mitigation Features Flood protection features utilized at MNGP in the event of the PMF include incorporated and temporary active and passive barriers. For flood protection below elevation 930 ft , installation of flood protection features (such as pumps and steel plates , grout, or sandbags to close openings up to elevation 930 ft) provides flood protection for Class I structures and Class II structures housing Class I equipment.