L-PI-11-074, Redacted- Volume 2 to Prairie Island Independent Spent Fuel Storage Installation Safety Analysis Report, Rev. 14

Redacted- Volume 2 to Prairie Island Independent Spent Fuel Storage Installation Safety Analysis Report, Rev. 14

ML113560205
Person / Time
Site:	Prairie Island
Issue date:	10/20/2011
From:	Northern States Power Co, Xcel Energy
To:	NRC/NMSS/SFST
References
L-PI-11-074
Download: ML113560205 (766)

Text

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1.1-1 SECTION A1 INTRODUCTION AND GENERAL DESCRIPTION OF STORAGE SYSTEM A

1.1 INTRODUCTION

Section 1.1 provides historical information regarding the need to provide dry storage at the Prairie Island Independent Spent fuel Storage Installation. The information in Section 1.1 remains relevant and is applicable for storage utilizing the TN-40HT dry casks with the exception that the descriptions of the fuel to be stored and the description of the TN-40HT dry cask is contained within this Addendum. The TN-40HT cask is a modified version of the TN-40 dry cask and is designed for high enrichment and high burnup fuel.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1 .2-1 A1.2 GENERAL DESCRIPTION OF LOCATION The information in Section 1.2 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1 .3-1 A1.3 GENERAL STORAGE SYSTEM DESCRIPTION The description of the ISFSI site including the storage building contained in Section 1.3 is independent of the cask design. A general description of the TN-40HT casks is contained in the following sections.

A1 .3.1 TN-40HT General Description The TN-40HT cask is very similar in design to the TN-40 cask. The major difference is that the TN-40HT cask is designed to store higher enrichment and higher burnup fuel.

In order to accomplish this, the heat transfer capability of the basket design was enhanced. Additionally, to accommodate the enhanced basket, some minor changes were made to the cask body. The TN-40HT cask employs a slightly thinner lid, shield shell and cask bottom shield. However, the radial neutron shield thickness is increased to offset the higher neutron source of the high burnup fuel.

The TN-40HT cask accommodates up to 40, 14x14 PWR fuel assemblies with or without fuel inserts. It consists of the following components:

• A basket assembly which locates and supports the fuel assemblies, transfers heat to the cask body wall, and provides neutron absorption to satisfy nuclear criticality requirements.

• A containment vessel including a bolted closure lid and seals which provides radioactive material confinement and a cavity with an inert gas atmosphere.

• Gamma shielding surrounding the containment vessel.

• Radial neutron shielding surrounding the shield shell which provides additional radiation shielding. This neutron shielding is enclosed in an outer steel shell.

• A top neutron shield which is attached to the outer surface of the cask lid and provides additional neutron shielding.

• An overpressure system which monitors and maintains the pressure between the cask closure seals and provides a positive pressure differential across the inner seals so that any inner seal leak will result in inleakage to the cask cavity.

• A protective cover which provides weather protection for the closure lid, top neutron shield and overpressure system.

• Sets of upper and lower trunnions which provide the means for lifting and rotating the cask.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1 .3-2 The maximum allowable initial enrichment of the fuel to be stored in a TN-40HT cask is 5.0 wt% U 235. The maximum bundle average burnup, maximum decay heat, and minimum cooling time are 60 GWd/MTU, 0.80 kW/assembly, and 12 years, respectively.

The cask is designed for a maximum heat load of 32 kW.

A1 .3.2 TN-40HT cask Characteristics Each TN-40HT cask consists of a fuel basket, a cask body (shell, shell flange, bottom, and lid), a protective cover, an overpressure system, four trunnions, penetrations with bolted and sealed covers for leak detection and venting, closure bolts and locating pins.

A set of reference drawings is presented in Section A1 .5. The cask is a self-supporting cylindrical vessel. Dimensions and the estimated weight of the cask are shown in Table A1 .3-1. Where more than one material has been specified for a component, the most limiting properties are used in the analyses in the subsequent sections of this addendum.

The TN-40HT containment boundary components are shown in Figure A1 .3-1. The containment vessel for the TN-40HT cask consists of: an inner shell (1) which is a welded, low alloy steel cylinder with a low alloy steel bottom closure; a welded shell flange (3); a low alloy steel lid with bolts and metal seals (2); and vent and drain covers with bolts and metal seals (4 and 5). The overall containment vessel length, wall thickness, cask cavity diameter, and length are shown in Table A1 .3-1.

There are two penetrations through the lid. One is for draining and the other is for venting. A double-seal mechanical closure is provided for each penetration. The lid is fastened to the flange by 48 bolts. A double metal seal with interspace leakage monitoring is provided for the lid closure. To preclude air in-leakage, the cask cavity is pressurized above atmospheric pressure with helium.

The pressure monitoring system is the same as the TN-40 cask system. The interspace between the metal seals is connected to an overpressure tank and a pressure monitoring system. The overpressure tank and the interspace between the metal seals are pressurized with helium to a higher level than the cavity so that any seal leakage would result in flow into the cavity. A decrease in the pressure of the monitoring system would be signaled by a pressure transducer/switch wired to a monitoring/alarm panel.

For weather protection, a steel torispherical weather cover with an elastomeric seal is installed over the lid and bolted to the shell flange.

The materials used for neutron and gamma shielding are the same as those used for the TN-40 cask. A gamma shield is provided around the walls and bottom of the containment vessel by an independent shell and bottom shield of carbon steel which is welded to the shell flange. The gamma shield surrounds the containment vessel inner shell and bottom closure. A lid shield plate is also welded to the inside of the containment lid.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1 .3-3 Neutron shielding is provided by a borated polyester resin compound surrounding the cask body. The resin compound is cast into long, slender aluminum containers.

The array of resin-filled containers is enclosed within a smooth outer steel shell. In addition to serving as resin containers, the aluminum provides a heat conduction path from the cask body to the outer shell. A pressure relief valve is mounted on the top of the resin enclosure for venting pressure due to heating of the resin and entrapped air after fuel loading.

A 4.0 inch thick nominal disc of polypropylene enclosed by steel plates is attached to the cask lid to provide neutron shielding during storage.

The basket structure consists of an assembly of rectangular stainless steel tubes joined by a proprietary fusion welding process to 1.75 in. wide stainless steel bars. Above and below the bars are slotted Boral, borated aluminum or boron carbide/aluminum metal matrix composite plates (neutron poison plates) and aluminum plates which form an egg-crate structure. The poison plates provide the necessary criticality control and aluminum plates provide the heat conduction paths from the fuel assemblies to the cask cavity wall. This method of construction forms a very strong honeycomb-like structure of cell liners which provide compartments for 40 fuel assemblies. The nominal open dimension of each cell is 8.05 in. x 8.05 in. which provides clearance around the fuel assemblies. The overall basket length is less than the cask cavity length to allow for thermal expansion and fuel assembly handling.

The cask cavity surfaces and outer shell have a sprayed metallic coating of Zn/Al for corrosion protection similar to the TN-40 cask. The external surfaces of the cask are also painted for ease of decontamination.

A stainless steel overlay is applied to the cask and lid wherever sealing surfaces for the metallic seals are required.

Four trunnions are attached to the cask for lifting and rotation of the cask. Two of the trunnions are located near the top of the cask and two near the bottom. The upper trunnions are sized for lifting. The lower trunnions are used for rotating the cask between vertical and horizontal positions.

Threaded holes are provided in the lid and top neutron shield for attachment of component lifting devices.

During dry storage of the spent fuel, no active systems are required for the removal and dissipation of the decay heat from the fuel. The TN-40HT cask is designed to transfer the decay heat from the fuel to the basket, from the basket to the cask body and ultimately to the environment by radiation and natural convection. The cask is capable of removing 32 kW of decay heat without external fins, thus providing a smooth outer surface for ease of decontamination.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1 .4-1 A1 .4 IDENTIFICATION OF AGENTS AND CONTRACTORS The agents and contractors associated with the TN-40HT casks are the same as those described in Section 1.4 for the TN-40 casks.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1 .5-1 A1.5 SUPPLEMENTAL DATA The following TN Drawings are enclosed:

. TN-40HT Cask (Drawings TN40HT-72-1 through 10)

. TN-40HT Basket (Drawings TN40HT-72-21, sheets 1 through 7)

. TN-40HT Basket Rails (Drawings TN40HT-72-22, sheets 1 and 2)

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A1 .6-1 A

1.6 REFERENCES

None.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A1 .3-1 NOMINAL DIMENSIONS AND WEIGHT OF THE TN-40HT CASK Overall length (with protective cover, in) 199.6 Outside diameter (in) 101 Cavity diameter (in) 72.0 Cavity length (in) 163 Containment shell thickness (in) 1.5 Containment vessel length (in) 170 Cask wall thickness (in) 8.75 Containment Lid thickness (in) 4.5 Lid thickness (in) 10.0 Bottom thickness (in) 8.75 Resin and aluminum box thickness (in) 5.25 Outer shell thickness (in) 0.50 Overall basket length (in) 160 Top neutron shield thickness (in) 4 Protective cover thickness (in) 0.38 Nominal Cask weight:

Loaded on storage pad (tons) 121.2

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Figure not to scale. Features exaggerated for clarity.

Phantom line (--- - --- -)

indicates containment boundary.

Containment boundary components are listed below:

Cask body inner shell Lid assembly outer plate, closure bolts and inner metal seal Shell flange Vent port cover plate, bolts and seals Drain port cover plate, bolts and seals FIGURE A1.3-1 TN-40HT CONTAINMENT BOUNDARY COMPONENTS

Figure Withheld Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 HI IH IF IEI i: I ID

1 OHT

Figure Withheld Under 10 CFR 2.390 Figure Withheld Under 10 CFR 2.390 Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 I .: -

Ii HI IH IF IEI ID II 2-8

[ TF

-I STORAGE CASK OVERPRESSURE TANK A Tr*.I4URT- . 2NP-8 N.JDkIE 4 .3

Figure Withheld Under 10 CFR 2.390 Figure Withheld Under 10 CFR 2.390 Figure Withheld Under 10 CFR 2.390

Figure Withheld Under 10 CFR 2.390

Figure Withheld Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2.1-1 SECTION A2 SITE CHARACTERISTICS A2.1 GEOGRAPHY AND DEMOGRAPHY OF SITE SELECTED A2.1.1 SITE LOCATION The information in Section 2.1.1 is independent of cask design.

A2.1.2 SITE DESCRIPTION The information in Section 2.1.2 is independent of cask design.

A2.1.3 POPULATION DISTRIBUTION AND TRENDS The information in Section 2.1.3 is independent of cask design.

A2.1.4 USES OF NEARBY LAND AND WATERS The information in Section 2.1.4 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2.2-1 A2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES The information in Section 2.2 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2.3-1 A2.3 METEOROLOGY A2.3.1 REGIONAL CLIMATOLOGY The information in Section 2.3.1 is independent of cask design.

A2.3.2 LOCAL METEOROLOGY A2.3.2.1 DATA SOURCE The information in Section 2.3.2.1 is independent of cask design.

A2.3.2.2 TOPOGRAPHY The information in Section 2.3.2.2 is independent of cask design.

A2.3.3 ONSITE METEOROLOGICAL MEASUREMENT PROGRAM The information in Section 2.3.3 is independent of cask design.

A2.3.4 DIFFUSION ESTIMATES A2.3.4.1 BASIS The information in Section 2.3.4.1 is independent of cask design.

A2.3.4.2 CALCULATIONS The information in Section 2.3.4.2 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2.4-1 A2.4 HYDROLOGY The information in Section 2.4 is independent of cask design.

A2.4.1 SURFACE WATER The information in Section 2.4.1 is independent of cask design.

A2.4.2 GROUND WATER The information in Section 2.4.2 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2.5-1 A2.5 GEOLOGY AND SEISMOLOGY A2.5.1 BASIC GEOLOGIC AND SEISMIC INFORMATION The information in Section 2.5.1 is independent of cask design.

A2.5.1.1 STORAGE SITE GEOMORPHOLOGY The information in Section 2.5.1.1 is independent of cask design.

A2.5.1.2 GEOLOGIC HISTORY OF STORAGE SITE AND SURROUNDING REGION The information in Section 2.5.1.2 is independent of cask design.

A2.5.1.3 SPECIFIC STRUCTURAL FEATURES OF SIGNIFICANCE The information in Section 2.5.1.3 is independent of cask design.

A2.5.1.4 LARGE SCALE GEOLOGIC MAP The information in Section 2.5.1.4 is independent of cask design.

A2.5.1.5 PLOT PLAN AND SITE INVESTIGATIONS The information in Section 2.5.1.5 is independent of cask design.

A2.5.1.6 GEOLOGIC PROFILES The information in Section 2.5.1.6 is independent of cask design.

A2.5.1.7 PLAN AND PROFILE DRAWINGS The information in Section 2.5.1.7 is independent of cask design.

A2.5.1.8 LOCAL GEOLOGIC FEATURES AFFECTING SITE LOCATION The information in Section 2.5.1.8 is independent of cask design.

A2.5.1.9 SITE GROUNDWATER CONDITIONS The information in Section 2.5.1.9 is independent of cask design.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2.5-2 A2.5.1.10 GEOPHYSICAL SURVEYS AND STUDIES The information in Section 2.5.1.10 is independent of cask design.

A2.5.1 .11 SOIL AND ROCK PROPERTIES The information in Section 2.5.1.11 is independent of cask design.

A2.5.1.12 ANALYSIS TECHNIQUES The information in Section 2.5.1.12 is independent of cask design.

A2.5.2 VIBRATING GROUND MOTION The information in Section 2.5.2 is independent of cask design.

A2.5.3 SURFACE FAULTING The information in Section 2.5.3 is independent of cask design.

A2.5.4 STABILITY OF SUBSURFACE MATERIALS The information in Section 2.5.4 is independent of cask design.

A2.5.5 SLOPE STABILITY The information in Section 2.5.5 is independent of cask design.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2.6-1 A

2.6 REFERENCES

The references listed in Section 2.6 are independent of cask design and there are no additional references associated with Section A2.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2A-1 A2A BORING LOGS The information in Appendix 2A is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A2B-1 A2B GRAIN SIZE DISTRIBUTION TEST REPORTS The information in Appendix 2B is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.1-1 SECTION A3 PRINCIPAL TN-40HT CASK DESIGN CRITERIA A3.1 PURPOSES OF CASK A3.1 .1 SPENT FUEL TO BE STORED The TN-40HT cask is designed to store 40 Westinghouse and Exxon 14x14 Pressurized Water Reactor (PWR) spent fuel assemblies with or without fuel inserts. The maximum allowable initial enrichment is 5.0 wt % U-235. The maximum bundle average burnup, maximum decay heat, and minimum cooling time for the fuel assembly are 60 GWd/MTU, 0.80 kW/assembly (including heat from inserts), and 12 years, respectively.

The cask is designed for a maximum heat load of 32 kW.

Sensitivity analyses were performed to determine the fuel assembly type which was the most limiting for each of the analyses including shielding, criticality, heat load and confinement. The fuel assemblies considered are listed in Table A3.1-1. The design basis fuel for decay heat, shielding and confinement is the Westinghouse Standard 14x14 assembly with 179 fuel rods and with burnup, bundle average enrichment, and cooling time of 60 GWd/MTU, 3.4 wt% U-235, and 18 years, respectively. The fuel qualification screening which is developed based on this fuel type is conservative when applied to other fuel types with lower mass of uranium. For the criticality analysis, all fuel assembly types are analyzed with 5.0% lattice enrichment. The Westinghouse Standard 14x14 assembly is the most reactive and is evaluated for configurations which bound all normal, off-normal and accident conditions. The thermal and radiological characteristics for the PWR spent fuel were generated using the SAS2H/ORIGEN S modules of SCALE (Reference 14). These characteristics for the design basis Westinghouse Standard 14x14 assembly are shown in Table A3.1-2. The thermal analysis uses bounding fuel rod parameters for the fuel assembly model with the maximum allowable decay heat.

Fuel with various combinations of burnup, enrichment, and cooling time can be stored in the TN-40HT cask as long as the combination results in decay heat, surface dose rates, and radioactive sources for confinement that are bounded by the design basis fuel. As discussed in Section A7, an evaluation was performed to determine various combinations of burnup, enrichment, and cooling time that would result in acceptable dose rates for the TN-40HT. Table A3.1-3 provides information for determining the qualification of fuel assemblies for storage in the TN-40HT cask. This table also shows a parametric equation that can be utilized to qualify spent fuel assemblies for the allowable decay heat load. The decay heat load can be calculated based on a fuel assemblys burnup, cooling time, and initial enrichment parameters. This table ensures that the fuel assembly decay heat load is less than the maximum allowable. The development of this equation is provided in Section A3.3.2.2.8.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.1-2 Assemblies containing fuel inserts (TPAs and BPRAs) may be stored in the TN-40HT cask. Reconstituted assemblies, (natural uranium dioxide replacement rods, Zirconium inert rods, or stainless steel rods replacing fuel rods), may also be stored in the cask.

The decay heat of a reconstituted assembly with stainless steel rods is bounded by an intact assembly. However, the irradiated stainless steel rods do increase the gamma source term for a period of time after irradiation. This period is shorter than the 12 year minimum cooling time required and thus no additional cooling time is required for these reconstituted assemblies.

The maximum combined weight of any fuel assembly and insert is limited to 1,330 lbs and the total weight of all fuel assemblies and inserts is limited to 52,000 lbs.

A3.1.2 GENERAL OPERATING FUNCTIONS The fuel assemblies will be stored unconsolidated and dry in sealed storage casks. The casks will rest on a reinforced concrete pad, and provide safe storage by ensuring a reliable decay heat path from the spent fuel to the environment and by providing appropriate shielding and confinement of the fission product inventory. Storage of spent fuel in storage casks is a totally passive function, with no active systems required to function. Cooling of the casks is accomplished by radiant and convective cooling.

Each cask will be handled with a lifting yoke, the 125 ton capacity auxiliary building crane, a transport vehicle, or other appropriate equipment. The crane will lift the cask from the spent fuel pool, in the spent fuel pool enclosure, move the cask laterally through an access door, and lower the cask to ground level in the rail bay of the Auxiliary Building. The cask will then be picked up by the transport vehicle which will be pulled to the ISFSI by a tow vehicle. After the transport vehicle has been maneuvered to locate the cask in its storage position, the cask will be set down.

All the handling equipment to be used outside the Auxiliary Building will be designed according to appropriate commercial codes and standards, and will be operated, maintained, and inspected in accordance with the supplier's recommendations.

Documentation will be maintained to substantiate conformance with all applicable standards.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-1 A3.2 DESIGN CRITERIA FOR ENVIRONMENTAL CONDITIONS AND NATURAL PHENOMENA The storage casks are designed with the objectives of ensuring that fuel criticality is prevented, cask integrity is maintained, and fuel is not damaged so as to preclude its ultimate removal from the cask. The conditions under which these objectives are met are described below.

The safe storage of the spent fuel assemblies must depend only on the capability of the storage casks to fulfill their design functions. The casks are self-contained, independent, passive systems, which do not rely on any other systems or components for their operation. Therefore, the casks are important to safety. The criteria used in the design of the casks ensure that their exposure to credible site hazards will not impair their safety functions. Because the ISFSI is located on the plant site, all ISFSI design criteria for environmental conditions and natural phenomena are the same as the plant design basis, as found in the USAR (Reference 15).

The design criteria satisfy the requirements of 10 CFR Part 72 (Reference 1). They consider the effects of normal operation, natural phenomena and postulated man-made accidents. The criteria are defined in terms of loading conditions imposed on the storage cask. The loading conditions are evaluated to determine the type and magnitude of loads induced on the storage cask. The combinations of these loads are then established based on the number of conditions that can superimpose. The load combinations are then classified as Design and/or Service Conditions consistent with Section III of the ASME Boiler and Pressure Vessel Code (Reference 2). The stresses resulting from the application of these loads are then evaluated based on the rules of the ASME Code as defined herein.

A3.2.1 TORNADO AND WIND LOADINGS Tornado loadings specified in Section 12.2.1.3.2 of the Prairie Island USAR were used in the design of the TN-40HT cask. These loads consist of the following:

A differential pressure equal to 3 psi. This pressure is assumed to build up from normal atmospheric pressure in 3 seconds.

• A lateral force caused by a funnel of wind having a peripheral tangential velocity of 300 mph and a forward progression of 60 mph.

The design tornado driven missile was assumed equivalent to an airborne 4 x 12 x 12¢ 0 plank traveling end-on 300 mph, or a 4000 lb. automobile flying through the air at 50 mph and at not more than 25 ft. above ground level.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-2 An analysis of impact on the cask of tornado missiles in accordance with Table 12.2-9 of the Prairie Island USAR (Reference 15) is presented below. Wind loading is not significant in comparison to that due to tornados; therefore, the wind loading is conservatively taken to be the same as the tornado loading.

A3.2.1.1 APPLICABLE DESIGN PARAMETERS The external pressure drop of 3 psi associated with passing of the tornado is small and, when combined with the other internal pressure loads, is far exceeded by the design internal pressure (22 psig normal design rating, 100 psig analyzed) for the cask. Its effect is included in the effects of the other internal pressure loadings listed in Table A3.2-2.

A3.2.1.2 DETERMINATION OF FORCES ON STRUCTURES The 360 mph tornado wind loading is converted to a dynamic pressure (psf) acting on the cask by multiplying the square of the wind velocity (in mph) by a coefficient (0.002558 at ambient sea level condition) dependent on the air density, based on data presented in a paper by T.W. Singell (Reference 5). The result is a pressure of 332 psf (Figure A3.2-1).

The net force acting on the cask is obtained by multiplying this pressure by the product of the area of the cask projected onto a plane normal to the direction of wind times a drag coefficient. A drag coefficient of 1.0 is used based on the geometric proportions of the cask (i.e., length to diameter ratio of approximately 2) and the conservative assumption that the cask surface is rough.

An additional load on the structure is that created by the impact of tornado missiles on the cask. These impacts are analyzed for 2 types of missiles specified in the Prairie Island USAR (Reference 15):

Missile A: High energy deformable missile (4000 lb. automobile) impacting the cask a) horizontally at normal incidence at 50 mph.

Missile B: A 4 inch thick by 12 inch wide by 12 foot long wood plank weighing 200 lbs:

a) horizontally at normal incidence at 300 mph.

b) vertically at normal incidence at 80% of the horizontal velocity.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-3 The tornado missiles are summarized in the following table:

Mass Horizontal Velocity Vertical Velocity Missiles (lbs) (mph) (mph)

A Automobile 4,000 50 -

4" Thick Wood Plank B 12" Wide x 12' Long 200 300 240 A3.2.1 .2.1 STABILITY OF THE CASK IN THE VERTICAL POSITION UNDER WIND LOADING The cask stands in an upright position on a concrete pad. The coefficient of friction between the steel cask and the concrete is taken as 0.25 for dry concrete. This coefficient is conservatively low based on data in Marks Handbook (Reference 6) which gives a value of 0.29 for steel on sandstone. Steel on concrete would be similar. A wind velocity of 409 mph would then be required to cause the cask to slide.

Cask Sliding The wind loading on the cask body is:

2 q = 0.002558 V q = F/A F = force to overcome the friction force, F f A = projected area of cask, ft 2 and the projected area is:

A = (199.63 in. x 101.0 in.)/144 = 140.00 ft2 (conservatively use the overall dimension for the cask height)

The friction force at the cask base is:

F f = W cask x ji = 240,000 x 0.25 = 60,000 lbs.

(A conservatively low cask weight of 240,000 lbs is used for the stability analyses.)

60,000 Therefore: 0.002558 V2 =

140.0 V = 409 mph The velocity to produce cask sliding is 409 mph which is greater than the 360 mph tornado wind velocity. Thus, the tornado wind will not slide the cask.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-4 Cask Tipping If the cask does not slide the force of the wind may cause the cask to tip. The minimum wind velocity required to tip the cask is calculated below:

M c= the tipping moment about the bottom edge of the cask l

2 db M c= - wx- c W

where: 2 2 W = cask weight = 240,000 lb.

w = distributed load to tip cask, lb/in.

d b= cask diameter at base = 89.5 in.

lc = cask length = 199.63 in.

Therefore:

89.5 2 w=240,000 2 199.63 2

= 539 lb / in The corresponding pressure load q, is:

539 q= x144=768 psf 101.0 The corresponding wind speed, V, is:

768 V= /_

.002558

= 548 mph Therefore the design basis tornado wind velocity of 360 mph will neither slide nor tip the cask.

A3.2.1 .2.2 STABILITY OF THE CASK IN THE VERTICAL POSITION UNDER MISSILE IMPACT It is assumed that Missiles A and B impact inelastically on the cask as shown in Figure A3.2-2. Both Missile A (the automobile) and Missile B (the wood plank) are assumed to crush. The cask will tend to slide if the missile strikes it below the CG (unless it is blocked in position) or tilt if the missile strikes it above the CG.

Conservation of momentum is assumed for both sliding and tipping with a coefficient of restitution of zero. The energy transferred to the cask is dissipated by friction in the sliding case or transformed into potential energy as the cask CG lifts in the tipping case.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-5 When a missile strikes the side of the cask at an elevation near the CG, the following velocities are calculated:

V =mv o

In the sliding case M+m

- md vo In the tipping case: - m(d

)2

+

p Where:

V = cask translational velocity after impact V0= missile initial velocity m = mass of missile, lbf/386.4 M = cask mass, 240,000 lbf/386.4 wp = cask angular velocity about P after impact d CG

= distance from center of gravity to pivot point P

= (91 . 582 +44.75 2 ) 0.5 = 101.93 in.

Ip

= mass moment of inertia of cask about pivot point P 2 9

= (2,281,240 + 240,000 x 101 .93 )/386.4 = 2.496 x 10 lbf.in ./386.4 When the appropriate substitutions are performed for Missile A (automobile) impact, the cask translational velocity after impact, V, in the sliding case is found to be 14.4 in/sec (1.20 ft/sec). The rotational velocity, ^ p , about P is found to be 0.141 rad/sec for impact at the top. Missile B impact produces lower cask translational and rotational velocities because of its lower initial momentum. Vertical impact of Missile B has no effect on cask stability.

Cask Sliding The cask may slide if the missile strikes it below the CG. Assuming no rotation and ignoring friction, the cask velocity could reach 14.4 in/sec after the impact. Therefore, the final kinetic energy is:

4 KE = 1/2(W cask x V2 )/g = 1/2 (240,000 x 14 . 42 )/386 . 4 = 6.44 x 10 in-lb If the cask slides on the concrete pad, the cask kinetic energy after impact is absorbed by friction. The friction work can be equated to the kinetic energy. Assuming a coefficient of friction of 0.25:

F friction = m Wcask = 0.25 x 240,000 = 6.00 x 1 0 4 lbs

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-6 Where:

F friction = friction force m = coefficient of friction The sliding distance is determined by:

4 4 Sliding Distance L = KE/ F friction = 6.44 x 10 /6.00 x 10 >> 1.1 in.

Given the above assumptions, the cask will slide 1.1 in. if the missile strikes it below the CG of the cask.

Cask Tipping When the cask tips or pivots about point P after impact, the kinetic energy is transformed into potential energy as the center of gravity rises:

E tipping

increase in potential energy = kinetic energy 1 2 E ppig

Mgd [cos(b + a - p/2)- cos b]= Ip w 2

where:

E tipping

= the increase in potential energy of the cask since the center of gravity rises as the cask pivots about the corner.

o -1 13 = 90 -sin (B/d cg ) = 26.0° a and 13 are indicated in Figure A3.2-2 The angle a is determined to be 89.7° for impact at the top of the cask (the cask tilts 0.3° and the c.g. lifts about 0.3 in.). The cask is stable and will not tip over since it will return to the vertical position as long as a is greater than about 64.0°.

Even at this 89.7° angle, wind will not tip over the cask (if the wind force occurs simultaneously with Missile A impact). At an 89.7° angle, the tipping moment about point P due to the 360 mph wind is less than 44% of the restoring moment due to the weight. Therefore the cask is stable in the vertical orientation under simultaneous tornado wind and tornado missile loadings.

A3.2.1.3 LOCAL EFFECT ON CONTAINMENT OF MISSILE IMPACT Forces App lied to the Cask Due to Missile Impact The impact forces applied to the cask as it is struck by the missiles are determined as follows:

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-7

• Missile A (automobile) is assumed to crush 1.5 ft (18 in.) under a constant force during the impact. The loss of kinetic energy is assumed to be dissipated by crushing of the missile.

Average Impact Force, F a of missile A can be calculated from the relation:

(Average Impact Force) x (Crush Length) = (Initial Kinetic of Missile A)

The initial kinetic energy, KE, of the missile can be determined from the relation:

1 2 KE = - (Missile A Mass)x (Missile A Velocity) 2 1 2 KE = 4000 x (50

• 5280

• 12/3600) = 4.008 x1 0 6 in.lb .

2 32.2*12 Therefore, the Average Impact force (F a ) can be estimated as:

6

=KE 4.008x10 Fa = =222,682lb.

L 18 Frontal area of the automobile is assumed to be 3 ft. x 6 ft. = 18 ft. 2 , therefore the impact pressure on cask pa caused by Missile A is:

222,682 ----

1 pa = = 85.9 psi 18 144

• Missile B (Wood Plank) will crush under impact. The highest crush strength Scrush of any wood listed (Reference 6) is 9210 psi (hickory). The contact force required for the plank to start crushing is:

Fc = S h x A= 9210x(4x12)=442,080 lb .

The kinetic energy of the plank is dissipated as the plank crushes against the cask wall.

The kinetic energy, KE, of the plank is:

2 1 2 12 200 ^ 300x5280J KE= mv O =x---x 7215x10 6 inib.

2 32.2 ^ 3600 =

The crush length, L , of the plank, for a constant force, is evaluated as:

KE 7.215x 10 6 L= = = 16.3 in.

Fc 442,080

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-8 For the case of the plank striking the top of the cask, the velocity is 80% of the horizontal case. Hence the kinetic energy would be 4.618x10 6 in.lb . and the crush length of the plank is 10.4 in.

Missile A Missile A (automobile) deforms and is crushed during the impact. The local pressure on the cask structure is less than 1 % of the body yield strength. Therefore, no local penetration occurs. The shear stress in the cask wall is conservatively calculated below. It is assumed that the impact force is concentrated on a small curved section of the cask wall having rectangular dimensions. It is also assumed that only two edges are tending to shear (above and below the curved section). The impact force, F a , above is used in the calculation. It is assumed that only 3 foot sections are shearing. Then:

Shear Area = 2 x 36 x the thickness of the gamma shielding

= 2 x 36 x 7.25 = 522.0 in 2 The shear stress, t = Force/area = 222,682/522 = 426.6 psi.

The level D allowable shear stress for the gamma shielding is 0.42 S u = 0.42 x 70,000 =

29,400 psi. The shear stress is well below the allowable shear stress.

Assuming that the impact on the side of the cask is reacted by a 3 foot high section of shielding, Case 15, Table 9.2, of Reference 17 is used to model the impact as shown below.

^ M max = 3/2(wR 2 )

The 2itRw (from Table 9.2, case 15, Reference 17) is the side impact force = 222,682 lbs.

R = 41.125 inches (mean radius of gamma shield)

Therefore, w = F/2itR = 861.8 lbs/in.

And, ^ M max = 3/2(wR2 ) = 2.1863 x 106 in lbs.

The bending stress on the section is:

3 s b = M x c/ I = 2.1863 x 106 x (7.25/2)/(36 x 7.25 /12) = 6,932 psi Where: I denotes moment of inertia of cylindrical slab c is the maximum normal distance from neutral surface

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-9 A conservative estimate of the stress intensity is:

2 2 0.5

= (69322 + 4 x 426 . 62 )0.5 = 6985 psi (s + 4 x t )

The Level D allowable for membrane plus bending stress is S u . For SA-266 Cl. 2 shielding material at 300 °F, S u = 70,000 psi. Thus the allowable stress is 70,000 psi.

Therefore, the membrane plus bending stress is acceptable.

Missile B (wood plank, Horizontal Impact)

The shear stress is calculated using the contact force (F c above), the perimeter of the plank, and the thickness of the shield shell.

Shear Area = (plank perimeter) x (thickness of shield shell)

Shear Area = (4+12+4+12) x 7.25 = 232 in 2 Shear Stress, t = Force/area = 442,080/232 = 1,905 psi.

The Level D allowable shear stress for the gamma shielding is 0.42 S u = 0.42 x 70,000

= 29,400 psi. The shear stress is well below the allowable shear stress.

Assuming that the impact on the side of the cask is reacted by a 36 inch high section of shielding, the impact is modeled as in missile A above. The height of the section is a conservative estimate of the portion of the shield shell that reacts to the impact. In reality, the entire shell would be affected.

^ M max ^ = 3/2(wR 2 )

The 2p Rw (from Table 9.2, Case 15, Reference 17) is the side impact force = 442,080 lbs.

R = 41.125 inches (mean radius of gamma shield)

Therefore, w = F/2 p R = 1711 lbs/in.

And, ^ M max ^ = 3/2(wR2 ) = 4.34 x 106 in lbs.

The bending stress on the section is:

3 sb = M x c/I = 4.34 x 10 6 x (7.25/2)/(36 x 7.25 /12) = 13,762 psi Where: I denotes moment of inertia of cylindrical slab c is the maximum normal distance from neutral surface A conservative estimate of the stress intensity is:

2 2 0.5

= (13762 2 + 4 x 1905 2 ) 0.5 = 14280 psi (s + 4 x t )

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-10 From above, the Level D allowable stress is 70,000 psi. Therefore, the membrane plus bending stress is acceptable.

Missile B (wood plank, Vertical Impact)

The shear stress is calculated using the contact force, F c , above, the perimeter of the plank, and the thickness of the lid.

Shear Area = (plank perimeter) x (thickness of shield shell)

Shear Area = (4+12+4+12) x 4.5 = 144 in 2 Shear Stress, t = Force/area = 442,080/1 44 = 3,070 psi.

The Level D allowable shear stress for the gamma shielding is 0.42 S u = 0.42 x 70,000

= 29,400 psi. The shear stress is well below the allowable shear stress.

The bending stress of the lid outer plate is calculated ignoring the protective cover and top neutron shield and assuming that the lid is simply supported circular plate. The bending stress is estimated using the analytic relation from Reference 17, Table 11.2, Case 18: the flat circular plate of constant thickness T, with uniform load over a small eccentric circular area of radius R o , simply supported.

The equivalent radius of wood plank contact area is calculated as:

1/2 1/2 R o = ((contact area)/p) = ((4 x 12)/p) = 3.91 in.

The lid model assumes that the inner radius of shell flange, R=36 inch, is the effective radius of the lid circular support and its outer radius.

To maximize bending the center of load is assumed to be at the center of the circular plate (p = 0). The Poissons ratio is assumed to be v = 0.3.

Per Reference 17, the maximum bending moment for this case can be expressed as:

2 CrushForce ^

_Ro M= x [1 +(1 +n)* -n)* ( )I 2

4*p o 4x(R - p) ^^

2

______ 36-0

_____ 3.91

_________ ^

M= 442080x[1+ (1+ 0.3)x ln( 3.91 ) - (1 -0.3)*(4* (36-0) 2 )

4*p ^

5

= 1.37 x 10 in-lb.

The bending stress on the section is:

5 2 sb = 6 x M / T 2 = 6 x (1.37 x 10 ) / 4.5 = 40,593 psi

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-1 1 A conservative estimate of the stress intensity is:

(a 2 + 4 x t 2

)

0.5

= (40593 2 + 4 x 3070 2 ) 0.5 = 41,055 psi This stress is much below the Level D allowable stress of 70,000 psi.

A3.2.2 WATER LEVEL (FLOOD) DESIGN The design basis flood for the Prairie Island ISFSI is described in Section 2.4.1. The probable maximum flood level is calculated to occur at the ISFSI is 706.7 ft above mean sea level (msl) with a water velocity of 6.2 ft/sec. This includes wave run-up. The ISFSI is designed and sited such that the lowest point of potential leakage into the cask is above the level of the probable maximum flood.

The casks are designed to withstand loads from forces developed by the probable maximum flood including hydrostatic effects and dynamic phenomena such as momentum and drag.

The TN-40HT cask is designed for an external pressure of 25 psig which is equivalent to a static head of water of approximately 56 ft. This is greater than would be anticipated due to the probable maximum flood.

The drag force (F) exerted on the cask by the flowing water is:

77 2 F= C D Ar -'---

2g Where:

F = Drag force CD = Drag coefficient >>1 .0 A = Projected area, 140.0 ft 2 p = 62.4 lb/ft 3 V = 6.2 ft/sec g = 32.2 ft/sec 2 Therefore, F = 5,214 lbs The drag force is applied to the cask in the same manner as the wind force discussed in Section A3.2.1. Since the wind force did not move the cask and the drag force is less than 20% of the wind force (0.2

• 332 psf

• 140ft2 = 9,296 lbs), the postulated flood will not cause the cask to slide or tip.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-12 As mentioned previously, the maximum flood height including wave run-up does not exceed the height of the cask seals. Therefore no inleakage of water can occur. Also, the interspace between the containment seals and the containment vessel cavity are pressurized to further preclude any possibility of water inleakage.

A3.2.3 SEISMIC DESIGN Seismic design criteria are set forth in 10 CFR Part 72.102. The design earthquake for use in the design of the casks must be equivalent to the safe shutdown earthquake (SSE) for a collocated nuclear power plant, the site of which has been evaluated under the criteria of 1 0CFR1 00, Appendix A (Reference 7).

Section 2.6 of the Prairie Island USAR discusses site seismology and the development of the SSE. An SSE of 0.12g horizontal and 0.08g vertical has been established as the design criteria.

A3.2.3.1 SEISMIC ACCELERATION LEVEL The TN-40HT cask is a very stiff structure. The dominant structural frequency of vibration for a loaded cask in the lateral direction is determined as shown below:

3 1/2 f = 3.89/ (WL /8EI) (Reference 24, Page 369, Case #3)

Where:

W = Weight of Cask = 240,000 lbs L = Height of Cask = 199.63 in.

E = Modulus of Elasticity = 28.3 x 10 6 psi D o = Cask Body O.D. = 89.5 in.

D i = Cask I.D. = 72.0 in.

4 4 4

= (p/64)( D o - D i ) = (p/64)(89.5 - 72.0 ) = 1.83 x 10 6 in 4 4

I Substituting the values given above, f = 57 Hz The vertical structural frequency of cask will be still higher since the cask has higher axial stiffness than the lateral stiffness. Thus the cask standing vertically on its pad has dominant lateral and vertical frequencies higher than 33 Hz (corresponding to the maximum ground acceleration, NUREG 1.60 Reference 19).

Therefore, the cask can be treated as a rigid body and the maximum seismic load on the cask is the peak ground acceleration times the mass of the cask. The cask is, therefore, evaluated using an equivalent static seismic loading method, and there is no need to specify a design response spectrum or its associated time history. The equivalent static load used conservatively includes a factor of 1.5 times (per NUREG-0800, Reference 4) the basic g level specified earlier.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-13 A3.2.3.2 SEISMIC-SYSTEM ANALYSIS The only significant effects of seismic loading that might occur would be sliding and/or tipping (overturning) of the cask.

Cask Sliding If the cask is to slide due to seismic loading, the horizontal component of the seismic load must overcome the friction force between the cask base and concrete pad. The friction force is equal to the normal force due to gravity acting at the cask/ground interface multiplied by the coefficient of friction.

The vertical seismic force is applied upward (0.08W

• 1.5 = 0.12W) so as to decrease the normal force and hence the sliding resistance force. The downward load then becomes 0.88W and the friction force that must be overcome to slide the cask is 0.22W (0.25x 0.88W). The maximum side load is 0.12W x 1.5=0.18W; therefore, the cask will not slide.

Cask Tipping The cask will not tip over due to a seismic event if the stabilizing moment due to cask weight is higher than the seismic tipping moment. For a cylindrical cask, the horizontal g value necessary to tip the cask is calculated below:

M tip = g h W L v + (2/3) g h W L r Where:

M tip = Moment necessary to tip the cask, in-lbs g h = Horizontal acceleration value necessary to tip the cask W = Weight of cask on pad L v = Vertical distance to C.G. = 91.58 in.

L r = Radial distance to C.G. = 44.75 in.

M stab = W L r Where:

M stab = Stabilizing moment of the cask, in-lbs.

W = Weight of cask on pad L r= Radial distance to CG = 44.75 in Therefore, the g value necessary to tip the cask is found by equating Mtip to M stab:

g h W L v + (2/3) g h W L r = W L r

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-14 g h = 44.75/(91.58 + 0.667 x 44.75) = 0.37 The two horizontal components of seismic load are combined as indicated in Section 2.1 of Reference 8. At 45° to either horizontal component, the response due to a N-S earthquake is sin 45° x N-S response and likewise for an E-W earthquake is cos 45° x E-W response. If both components are equal, the combined response is:

(sin 2 45° + cos2 45°)1/2 x response = response in either axis.

Therefore, we only need to consider a single horizontal axis for the maximum seismic response.

For this evaluation the horizontal response is 0.1 2g x 1.5 = 0.1 8g which is less than the 0.37g required to tip the cask, the cask will not tip over.

A3.2.4 SNOW AND ICE LOADINGS The decay heat of the contained fuel will continue to heat the cask throughout its service life thus provide a heat source to melt snow or ice. However, to bound the possibility of unmelted snow or ice, a loading corresponding to 50 psf (0.35 psi) is bounded by the analyses.

The temperature of the protective cover attached to the top of the cask above the lid under certain conditions could fall below 32 °F and a layer of snow or ice might build up.

A 50 psf (0.35 psi) snow or ice load corresponds to approximately 6 ft of snow or 1 ft of ice. However, this load is insignificant on the TN-40HT since the cover is a 0.38 in.

thick torispherical steel head which can withstand an external load over 20 psi.

Therefore, the cover will maintain its intended protective function under snow or ice loading conditions.

A3.2.5 COMBINED LOAD CRITERIA A3.2.

5.1 INTRODUCTION

Sections A3.2.1 through A3.2.4, above, describe the most severe natural phenomena considered in the design of the TN-40HT. The forces and pressures applied to the cask due to these phenomena have been determined. These phenomena have been analyzed to show that the cask is stable. It will not tip over under any condition or slide on its pad more than approximately an inch.

All of the above phenomena are upper bound, low probability events. In most cases, however, there is a more regular or frequent similar phenomena of lower magnitude.

For instance, some small wind loading occurs often, but tornado winds are unlikely.

The forces and pressures determined for the severe phenomena can therefore be used as upper bound values for all similar events.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-15 It has been assumed that these bounding forces and pressures, with a single exception, can occur at any time and their effects are combined with those due to normal operations. The sole exception is the loading(s) due to the tornado missiles as described in Section A3.2.1 .3.

A3.2.5.2 TN-40HT CASK LOADING A brief explanation of the cask loads due to events that will occur or can be expected to occur in the course of normal operation at the ISFSI follows. The cask loads due to the severe natural phenomena and accidents are compared with those for similar but less severe normal events. Then loads equal to or higher than the upper bound values selected for design and analysis of the TN-40HT, defined as Service Loads, are described. Finally, the Service Loads are separated into two levels and superposition of simultaneous loading (combined loads) is discussed.

A3.2.5.2.1 NORMAL OPERATION During normal storage on the ISFSI pad, the cask is subjected to loading due to its own dead weight and that of its contents (fuel and basket), assembly stresses due to the bolt preload required to seat the double metal seals and react to the internal pressure, and internal pressure due to initial pressurization and any fuel clad failure resulting in fission gas release.

Additional normal loads include wind loading which produces a distributed lateral load on one side of the cask and can also result in slight external pressure drop on other portions of the cask.

Lifting loads due to cask dead weight, fuel and basket are applied to the cask through the trunnions.

An increased external pressure is applied to all surfaces of the cask during fuel loading when the cask is at the bottom of the spent fuel pool. Snow and ice loads apply local external pressure loading to the top of the protective cover. The cask will be subjected to the full range of thermal conditions produced by ambient variations (including insolation) and decay heat.

A3.2.5.2.2 LOADINGS DUE TO SEVERE NATURAL PHENOMENA AND ACCIDENTS The tornado wind loading described in Section A3.2.1 could produce higher lateral loading than any normal wind loading or flood water drag force. The external pressure drop due to the tornado wind is also more severe than due to any normal condition.

External pressure loading of the cask could occur due to flooding (see Section A3.2.2),

burial, or nearby explosion. The full range of thermal conditions due to ambient variations, decay heat and minor fires in the vicinity of the cask apply.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-16 A3.2.5.2.3 THERMAL CONDITIONS The TN-40HT component temperatures and thermal gradients are affected by the following thermal conditions:

• Fuel loading

• Decay heat

• Insolation

• Beginning of life unloading

• Ambient variations

• Lightning

• Minor fire

• Cask burial The thermal conditions which are of concern structurally are the temperature distributions in the cask and the differential thermal expansions of interfacing cask components.

A3.2.5.2.4 FUEL LOADING The cask is loaded in a spent fuel pool under water. The cask is cooled by pool water; therefore, the thermal gradients established during fuel loading are negligible.

A3.2.5.2.5 DECAY HEAT/SOLAR LOAD After the cask is loaded and removed from the pool, the cask body will gradually reach steady state conditions. Since the mass of the cask is large, the time to reach equilibrium will be approximately 1 to 2 days. The temperature gradients in the cask body have an insignificant effect on the structural integrity of the body.

Thermal analyses have been performed to determine the temperatures within the cask for different normal and accident conditions. The methods used to obtain these results are discussed in Section A3.3.2.2. The cask temperature distribution for the off-normal condition was used for the structural analysis.

A3.2.5.2.6 BEGINNING OF LIFE UNLOADING This condition would occur if it were necessary to place the cask back in the pool at the beginning of life after it had been loaded and reached thermal equilibrium. Prior to completely submerging the cask in the pool, the cask and fuel would have to be cooled by circulating water through the cask. Therefore, cold water would contact the hotter cask inside surfaces and fuel pins. Past experience/evaluations have shown that beginning of life unloading has little effect on the cask temperature and thus has an insignificant effect on the cask body and the fuel cladding loadings.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-17 A3.2.5.2.7 AMBIENT VARIATIONS Because the cask thermal inertia is large, the cask thermal response to changes in atmospheric conditions will be relatively slow. Ambient temperature variations due to changes in atmospheric conditions i.e., sun, ice, snow, rain and wind will not affect the performance of the cask. Snow or ice will melt as it contacts the cask because the outer surface will typically be above 32 °F. The cyclical variation of insolation during a day will also create insignificant thermal gradients.

A3.2.5.2.8 LIGHTNING Lightning will not cause a significant thermal effect. If struck by lightning on the lid, the electrical charge will be conducted by paths provided by the lid bolts to the body. It is not anticipated that lightning could cause significant increase in seal temperatures.

A3.2.5.2.9 FIRE The only real source of fuel for a fire in the vicinity of the cask is the fuel tank of the tow vehicle which transports the cask to the storage pad. An evaluation was made to determine the thermal response of the cask assuming this minor fire is an engulfing fire.

The results of this analysis are provided in Section A3.3.2.2.2.4. The cask will maintain its containment integrity during and after this bounding hypothetical fire accident.

A3.2.5.2.10 BURIED CASK An evaluation is made to determine the increase in cask temperature with time assuming that the cask is completely buried by dirt and debris with very low thermal conductivity. The results of this analysis are given in Section A3.3.2.2.2.4. The analysis shows that the cask will maintain its containment integrity for a maximum burial period of 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />.

A3.2.5.3 BOUNDING LOADS FOR DESIGN AND SERVICE CONDITIONS A3.2.5.3.1 DEAD (WEIGHT) LOADS The only dead loads (hereafter referred to as weights) on the cask are the cask weight including the contents. The calculated weights of the individual components of the cask and the total weights are given in Table A3.2-1. The weight of the cask assembly is reacted as a contact force between cask and storage pad except when the cask is supported (lifted) by the pair of trunnions at the top of the cask during handling prior to fuel loading.

A3.2.5.3.2 LIFTING LOADS The cask is provided with two trunnions at the top spaced 180 degrees apart for non-redundant lifting. The trunnions at the bottom of the cask are for rotation of the cask.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-18 The upper trunnions are designed to meet the requirements of NUREG-0612 (Reference 3) for non-redundant lifting fixture. This is accomplished by evaluating the trunnions to the stress design factors required by ANSI N 14.6 (Reference 9), i.e.

capable of lifting 6 times and 10 times the cask weight without exceeding the yield and ultimate strengths of the material, respectively. The trunnion loads are shown in Figure A3.2-3 and listed in Table A3.2-3.

The local region of the cask body is conservatively evaluated for a vertical load of 3 g (i.e., 3 times the weight of the cask) which is reacted at the trunnions involved in the handling operation. The factor of 3 provides ample allowance for sudden load application during lifting.

A3.2.5.3.3 INTERNAL PRESSURE The pressure inside the cavity of the storage cask results from several sources. Initially, the cavity is backfilled with helium to at least 19.5 psia . The purpose of pressurizing the cavity above atmospheric pressure is to prevent in-leakage of air. The initial pressure is determined on the basis that, at minimum, a 1 atm abs pressure must exist in the cavity on the coldest day at the end of life. Pressure variations due to daily and seasonal changes in ambient temperature conditions will be small due to the large thermal capacity of the cask. Fuel clad failure results in the release of fission gas which increases cavity pressure. Section A3.3.2.2.6.1 evaluates the increases in pressure due to off-normal and accident scenarios.

Another condition when internal pressure could increase is the cool down prior to unloading. This could occur at the beginning or end of life. Water will be gradually added to the cask during refilling to ensure that the cask pressure limits are not exceeded. See Section A3.3.2.2.5.2 for an evaluation of internal pressure during reflood i ng.

Table A3.2-2 presents a summary of internal pressures for the conditions identified. A pressure of 22 psig was chosen as the design internal pressure. This value bounds the normal and off-normal operating pressures.

A3.2.5.3.4 EXTERNAL PRESSURE There are several conditions which can result in external pressure on the cask. The external pressure due to a flood is less than the designed external pressure as discussed in Section A3.2.2.

During fuel loading or unloading the cask is at the bottom of the spent fuel pool, nominally 40 ft. deep. This results in an external hydrostatic pressure of approximately 20 psi.

An explosion on a barge in the vicinity of the Prairie Island plant has been shown to produce an overpressure of less than 2.25 psi at the ISFSI location.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-19 An earth pressure loading would occur if the cask were to be buried under dirt. This is similar to a hydrostatic pressure head of water. The density of loose dirt or earth is approximately 100 lb/ft 3 compared to 62.4 lb/ft3 for that of water. Therefore 36 ft. of earth is equivalent to a 56 ft. head of water and an external pressure of 25 psi.

The various external pressures are also summarized in Table A3.2-2. The cask is designed and evaluated for an external pressure of 25 psi. This value was selected because it exceeds the maximum external pressure which would be anticipated for any of the loading conditions considered above including floodwater discussed in Section A3.2.2 and snow and ice in Section A3.2.4.

A3.2.5.3.5 CASK BODY LOADS Globally distributed loads may be applied to the cask by wind (tornado is the upper bound case), flood water and seismic excitation. These loads are explained in detail and calculated in Sections A3.2.1 through A3.2.3. Table A3.2-4 lists the numerical values of these forces as calculated in the various sections. Note that bounding dynamic loads equal to the weight of the cask (1 g load) in each direction (lateral and vertical) applied as inertial loads for stress analysis purposes envelope all of these distributed loads with a substantial margin.

A3.2.5.4 DESIGN AND SERVICE LOADS The various cask loading conditions are listed in Table A3.2-5. These loading conditions include those described in 1 0CFR Part 72 (Reference 1), which are categorized as normal, man-made and natural phenomena. The applied loads acting on the different cask components due to these loading conditions have been determined and are discussed in the preceding sections and are listed in Table A3.2-2 through Table A3.2-4. This section describes the bases used to combine the loads for each cask component. The specific stress criteria against which each load combination will be evaluated are described in Section A4.2.3.

The bounding pressures and loads described above are used in the load combinations.

Certain combinations are evaluations of several events (e.g., one load combination represents stresses due to tornado wind, hurricane wind, normal wind, flood water, etc.).

Several loads are always present and are included in all evaluations. These are the assembly stresses due to bolt preload and metal seal compression. Lifting loads are always reacted by the cask weight (supported by trunnions - not the storage pad).

Lifting loads are not combined with those due to extreme natural phenomena since cask operations would be halted during a flood, tornado, etc. Dead weight loads are reacted at the bottom of the cask by the storage pad for all cases except the lifting cases.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-20 A3.2.5.4.1 CASK BODY The loading conditions for the cask body including the containment vessel and gamma shielding are categorized based on the rules of the ASME Boiler and Pressure Vessel Code Section III, Subsection NB, for a Class 1 nuclear component (Reference 2). The ASME code categorizes component loadings into five service loading conditions. They include Design Conditions (same as the Primary Service) and Levels A, B, C and D Service Loadings. The code provides different stress limits for each of these service loadings.

For each of these service loading conditions, there are several applied loads acting on the cask. The Design Loads are listed in Table A3.2-6. They include internal and external pressure; lid bolt preload including the effect of the gasket reactions; distributed loads due to weight, wind, and handling, and attachment loads applied through the trunnion to the cask body.

The inertia g loads are quasistatically applied loads which are multiples of the weight of the cask and/or contents. The magnitude of the Design Loads envelope the maximum Level A Service Loads. Thermal effects are excluded, except for their influence on the preload of the lid bolts (if any) because the ASME Code does not consider them Design (i.e., primary) Loads.

The Level A Service loads are listed in Table A3.2-7 and are basically the same as the Design Loadings except that the thermal effects on the containment vessel are included. The thermal effects consist of secondary (thermal) stresses caused by differential thermal expansion due to temperature differences caused by decay heat, solar insolation, ambient temperature variations and ambient conditions.

There are no Level B or C Service Loading Conditions. Events which occur infrequently which could be considered Level B or Level C Service Loadings are conservatively considered Level A loadings.

The loads due to Level D Service Loading Conditions, which are extremely unlikely conditions, are listed in Table A3.2-8.

Loading combinations for Normal Conditions (Design Conditions and Levels A Service Loadings) are given in Table A3.2-9. Loading combinations for Accident Conditions (Level D) are provided in Table A3.2-1 0. The loads are listed across the top of the table and the Load Combinations are designated in the first column of the table.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-21 There are eight normal (Design and Level A Service Loadings) load combinations listed, and five accident condition (Level D) combinations. The loads which are acting simultaneously for each of these combinations are denoted by an "X" under the load column heading. For example, for Normal Condition Load Combination N3, lid bolt preload, fabrication loads, distributed weight, internal pressure due to cavity pressurization and fission gas release, and heat due to maximum normal temperatures are acting simultaneously.

A3.2.5.4.2 BASKET Cask body internal and external pressures have no effect on the basket. External loads applied to the TN-40HT cask do not result in basket loads unless the cask actually moves. Therefore, tornado wind and flooding produce no basket loads. Seismic loading, however, is an inertial loading as discussed in Section A3.2.3, and is applied to the basket. The seismic acceleration loading (much less than 1g acceleration) is combined with dead weight loading since the two effects occur simultaneously.

Temperature effects due to snow, minor fire and ambient temperature variations which can cause thermal transients on the outside of the cask body will not cause similar transients in the basket. The high heat capacity of the body slows the temperature response and effectively eliminates transients at the wall of the cask cavity. The steady state temperature and temperature differences throughout the basket are, however, affected by decay heat, solar insolation and ambient temperature variations.

The basket is important for control of criticality of the fuel assemblies stored in the cask.

The bounding lateral and vertical inertial loadings on the basket are equal to 1g (in each direction) and have been shown to envelope the basket loadings. For the basket evaluation, an even more conservative 3g loading in the vertical direction is analyzed.

The stresses in the 304 stainless steel portions of the basket due to the primary loading, 1g in any lateral direction combined with 3g vertical (including dead weight), are determined by conservatively neglecting the tensile and bending strength of the plates (aluminum and/or poison) between fuel compartment boxes. The through thickness strength of the plates which separate the boxes is considered. Thus the aluminum and/or poison material is conservatively neglected in the primary load analysis where it can react some of the load. These primary stresses in the steel are evaluated at the maximum metal temperature occurring under extreme ambient conditions.

Clearance is provided between the aluminum/poison and stainless steel plates to provide for differential thermal expansion. The basket design criteria described in Section A4.2.3.3.3 is based on Section III Subsection NG and Appendix F of the ASME Code for stress limits.

The basket evaluation is provided in Appendix A4B.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.2-22 A3.2.5.4.3 UPPER TRUNNIONS The upper trunnions are designed to meet the requirements of NUREG-0612 (Reference 3) for non-redundant lifting fixture. This is accomplished by evaluating the trunnions to the stress design factors required by ANSI N 14.6 (Reference 9). During lifting, the trunnions are evaluated for vertical lifting reactions applied at the lifting shoulders required to support 6 times or 10 times the maximum weight of a fully loaded cask. When the load is equal to 6 times the weight, the maximum tensile stresses shall not exceed the minimum yield strength of the trunnion material. For the load equal to 10 times the weight, the maximum tensile stresses shall not exceed the minimum ultimate tensile strength of the trunnion material.

The loads acting on the trunnions are shown in Table A3.2-3. The structural analysis of the trunnions is presented in Section A4.2.3.

In a response to questions from the NRC Staff, NSP provided justification in Reference 10 for the adequacy of the load testing performed on the TN-40 cask trunnions during fabrication. In a Safety Evaluation dated May 11, 1995 (Reference 11) the NRC concluded that NSP had demonstrated that the trunnion-to-cask attachment welds have an acceptable level of quality and that load testing of the trunnions as described in Reference 10 provides an acceptable demonstration of the adequacy of the TN-40 cask trunnions. Since the trunnions for the TN-40HT cask are designed to the same requirements as the trunnions for the TN-40 cask, and the attachment welds have the same level of quality, the conclusions of Reference 11 are applicable to the TN-40HT cask.

A3.2.5.4.4 OUTER SHELL The outer shell is evaluated for the combined effects of inertia g loads due to lifting and internal pressure.

Out-gassing from the resin between the cask body and outer shell may cause a slight pressure on the inside of the outer shell. A pressure relief valve is provided in the outer shell to assure any pressure buildup is small. The outer shell is completely supported by the resin when subjected to an external pressure. An internal pressure of 3 psig will occur due to the reduced external pressure during a tornado. An internal pressure of 25 psig is conservatively used to evaluate the outer shell. The structural analysis of the outer shell is presented in Appendix A4A7. A summary of results and comparison with design criteria are given in Section A4.2.3.4.6.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-1 A3.3 SAFETY PROTECTION SYSTEMS A3.3.1 GENERAL The TN-40HT cask is designed to provide storage of spent fuel for at least 25 years.

The cask cavity pressure is always above ambient during the storage period as a precaution against the in-leakage of air which could be harmful to the fuel. Since the containment vessel consists of a steel cylinder with an integrally-welded bottom closure, the cavity gas can escape only through the lid closure system. In order to ensure cask leak tightness, two systems are employed. A double barrier system for all potential lid leakage paths consisting of covers with multiple seals is utilized. Additionally, pressurization of monitored seal interspaces provides a continuous positive pressure gradient which guards against a release of the cavity gas to the environment and the admission of air to the cavity.

A3.3.2 PROTECTION BY MULTIPLE CONFINEMENT BARRIERS AND SYSTEMS A3.3.2.1 CONFINEMENT BARRIERS AND SYSTEMS A combined cover-seal pressure monitoring system (Figure A3.3-1) always meets or exceeds the requirement of a double barrier closure which guarantees tight, permanent confinement. There are two lid penetrations, one for a drain pipe and one for venting and pressurization. When the cask is placed in storage, a pressure greater than that of the cavity is set up in the gaps (interspaces) between the double metallic seals of the lid and the lid penetrations. A decrease in the pressure of the monitoring system would be signaled by a pressure transmitter mounted at the side of the cask (Figure A3.3-1). The system is pressurized through a fill valve mounted near the overpressure tank. Lead shielding will be provided to reduce radiation exposure to the transmitter to acceptable levels.

Connections to the overpressure tank are welded fittings. A quick connect coupling with a diaphragm valve is used to fill the tank.

The Helicoflex metal seals of the lid and lid penetrations possess long-term stability and have high corrosion resistance over the entire storage period. These high performance seals are comprised of two metal linings formed around a helically-wound spring. The sealing principle is based on plastically deforming the seal's outer lining. Permanent contact of the lining against the sealing surface is ensured by the outward force exerted by the helically-wound spring. Additionally, all metal seal seating areas are stainless steel overlay for improved surface control. The overlay technique has been previously used for the TN-40 casks.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-2 For protection against the environment, a torispherical protective cover equipped with an elastomer seal is provided above the lid. The lid and cover seals described above are contained in grooves. A high level of sealing over the storage period is assured by utilizing seals in a deformation-controlled design. The deformation of the seals is constant since bolt loads assure that the mating surfaces remain in contact. The seal deformation is set by its original diameter and the depth of the groove.

Metal gasket face seal fittings, diaphragm valves, and Helicoflex metal seals are all

-7 capable of limiting leak rates to less than 1 x 1 0 atm cm 3/sec of helium.

The initial operating pressure of the monitoring system's overpressure tank is set at 5.5 atm abs minimum. Over the storage period, the pressure decreases as a result of leakage from the system and as a result of temperature reduction of the gas in the system. Since the level of permeation through the containment vessel is negligible and leakage past the higher pressure of the monitoring system is physically impossible, a decrease in cavity pressure during the storage period occurs only as a result of a reduction in the cavity gas temperature with time. As long as the cavity pressure is greater than ambient pressure and the pressure in the monitoring system is greater than that of the cavity, no in-leakage of air or out-leakage of cavity gas is possible.

The analyses provided in Appendix A7A define the monitoring system helium test leakage rate which ensures that no cavity gas can be released to the environment nor air admitted to the casks for the 25 year storage period. All seals are considered collectively in the analysis as the monitoring system pressure boundary.

A3.3.2.2 HEAT TRANSFER DESIGN The TN-40HT cask is designed to passively reject decay heat under normal conditions of storage and hypothetical accident conditions while maintaining appropriate packaging temperatures and pressures within specified limits. An evaluation of the TN-40HT cask thermal performance is presented in this section. Objectives of the thermal analyses performed for this evaluation include:

• Determination of maximum and minimum temperatures with respect to material limits

• Determination of temperature distributions for analysis of thermal stresses

• Determination of temperatures for containment pressurization The TN-40HT basket consists of an assembly of 40 stainless steel fuel compartments with aluminum and neutron poison plates sandwiched between them. The compartments are joined by a fusion welding process to 1.75 in. wide stainless steel bars. Above and below the bars are slotted aluminum and neutron poison plates which form an egg-crate structure. Stainless steel basket rails including aluminum inserts are bolted to the basket periphery to provide a conduction path from the basket to the cask cavity wall. This thermal design feature of the basket allows the heat from the fuel assemblies to be conducted along the basket structure to the basket rails and dissipated to the cask cavity wall.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-3 The neutron shielding is provided by a resin compound cast into long slender aluminum containers placed around the cask shell and enclosed within a smooth outer shell. By butting against the adjacent shell surfaces, the aluminum containers provide a conduction path and allow decay heat to be conducted across the neutron shield.

To establish the heat removal capability, several thermal design criteria are established for the TN-40HT cask. These are:

• Confinement of radioactive material and gases is another design consideration.

Seal temperatures must be maintained below specified limits to satisfy the confinement function during normal and accident conditions. A maximum temperature of 536 °F (280 °C) is set for the double metallic seals in the containment vessel closure lid.

• To maintain the stability of the neutron shield resin during normal storage conditions, the neutron shield bulk average temperature cannot exceed 300 °F (149 °C).

• Maximum temperatures of the containment structural components must not adversely affect the confinement function.

• Maintaining fuel cladding integrity during storage is a major design requirement.

Fuel cladding temperature limits have been established per ISG-1 1 (Reference 25). For normal conditions of storage and short term loading operations such as vacuum drying, fuel cladding temperature limit is 400 °C (752 °F). For off-normal storage, and accident conditions, the fuel cladding temperature is limited to 570

°C (1058 °F).

In general, all thermal criteria are associated with maximum temperature limits and not with minimum temperatures. All materials can be subjected to the minimum environment temperature of -40 °F (-40 °C) without adverse effects.

The TN-40HT cask is analyzed based on a maximum heat load of 32 kW from 40 fuel assemblies with a maximum decay heat of 0.80 kW per fuel assembly (including inserts).

A3.3.2.2.1 THERMAL MODEL A3.3.2.2.1.1 THERMAL MODEL FOR NORMAL AND OFF-NORMAL STORAGE CONDITIONS To determine the maximum component temperatures, two finite element models are developed using the ANSYS computer code (Reference 36).

• Full-length cask model

• Sub-model of cask top

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-4 ANSYS is a comprehensive thermal, structural, and fluid flow analysis package. It is a finite element analysis code capable of solving steady-state and transient thermal analysis problems in one, two, and three dimensions. Heat transfer via a combination of conduction, radiation, and convection can be modeled by ANSYS.

A3.3.2.2.1.1.1 FULL LENGTH CASK MODEL The full length model is three dimensional and represents a 90° symmetric section of the TN-40HT cask. This model includes the geometry and the material properties of the basket, basket rails, cask shells, cask lid, protective cover, cask bottom plates, radial neutron shield, and top neutron shield, as well as concrete pad and supporting soil. The full-length model is used to determine the maximum temperatures for the fuel cladding and the other components (except the seals and the top neutron shield located within and above the cask lid).

The protective cover is modeled using SHELL57 elements (Reference 36). For conservatism, no heat transfer is considered between the protective cover and the upper surface of the cask lid in the full-length model to minimize the axial heat transfer.

Fuel assemblies are modeled as a homogenized material within the fuel compartments.

It is assumed that no convection heat transfer occurs within the cask cavity. The effective conductivity calculated for the homogenized fuel includes the radiation heat transfer. The conduction heat transfer through the basket and cask components is modeled using SOLID70 elements (Reference 36).

The aluminum basket rails are divided into four 40 in. sections. Axial gaps of 0.06 are considered between each two adjacent rail pieces. This model conservatively bounds the structural design of three-piece rails.

The following gaps are considered between components in the model at thermal equilibrium:

a) Air gaps:

• 0.01 radial gaps between the aluminum resin boxes and the adjacent shells

• 0.01 axial gap between the cask lid and the cask flange

• 0.01 axial gap between flange of the protective cover and the cask flange

• 0.01 axial gap between the cask lid and the shield plate

• 0.02 axial gap between the top neutron shield and the cask lid outer plate

• 0.125 axial gap between the bottom inner plate and the bottom shield plate

• 0.01 axial gap between the bottom shield plate and the shield shell

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-5 b) Helium gaps:

• 0.01 transverse gap between a poison plate or a stainless steel bar and the fuel compartment wall

• 0.02 transverse gap between an aluminum or a stainless steel bar and the fuel compartment wall This gap in combination with the above gap allows for using of multiple sheets and plates instead of using one pair of aluminum and poison plates.

• 0.06 axial gap between basket rail segments

• 0.125 axial gap between the basket and the inner bottom plate

• 0.13 radial gap between the rails and the cask inner shell

• 0.02 gap between the aluminum inserts and stainless steel plates of the basket rails at bolted joints

• 0.10 gap between the aluminum inserts and stainless steel plates of the basket rails at unbolted locations

• 0.01 gap between the basket plates and the basket rails or shims The axial cold gap of 0.07 between the stainless steel structural bars and the poison/aluminum plates is divided into a 0.01 axial gap at the bottom and a 0.060 axial gap at the top of the stainless steel bars.

The width of the cutout in basket aluminum plates or poison plates is considered to be 1 .11 which is larger than the nominal cold cutout width of 1.00.

A gap of 0.01 is used between the cask inner shell and the shield shell to simulate the thermal resistance of the shrink fit gap between these shells. Identical to Section 3.3.2.2.1, a contact conductance of 350 Btu/hr-ft 2 equal to an effective conductivity of 0.0243 Btu/hr-in-°F is used to model the shrink fit gap.

The geometry of the full-length model for TN-40HT cask is shown in Figure A3.3-2 and Figure A3.3-3. The locations of the gaps are shown in Figure A3.3-4 and Figure A3.3-5.

A3.3.2.2.1.1.2 SUB-MODEL FOR CASK TOP The second finite element model represents only the top portion of the TN-40HT cask.

This model is a sub-model of the full cask model in which radiation between the top neutron shield and protective cover are added to determine the maximum temperatures of the seals and the top neutron shield resin.

The cask top sub-model includes the cask flange, cask top shield plate, cask lid, protective cover, and the overpressure tank. The dimensions and the material properties for the sub-model are identical to those for the full-length model.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-6 The geometry of the sub-model for the cask top is shown in Figure A3.3-6.

A3.3.2.2.1 .1.3 STEADY STATE BOUNDARY CONDITIONS TH-40HT cask is designed for ambient temperatures between -40 °F and 120 °F. The maximum daily average temperature of 100°F is used for the maximum off-normal storage temperature.

Under the cold condition of -40 °F (-40 °C) ambient, the resulting packaging minimum component temperatures will approach -40 °F if no credit is taken for the decay heat load. Since the packaging materials, including containment structures and the seals, continue to function at this temperature, the minimum temperature condition has no adverse effect on the performance of the TN-40HT cask. Thermal analysis for minimum ambient temperature of -40 °F and no insolance is performed using the maximum decay heat load of 32 kW to determine the temperature distributions for structural evaluation.

Normal storage conditions (50 °F ambient) are bounded by the hot off-normal storage conditions (100 °F ambient). The temperature profiles for hot off-normal conditions are conservatively used for normal storage conditions.

The SOLID70 elements representing the homogenized fuel are given heat generation load throughout the 144 active fuel length. The active fuel length begins at about 4.0 above the bottom of the fuel assembly. The effective fuel properties are used for the entire basket length (160) within the homogenized fuel regions.

A typical axial decay heat profile from Reference 26 with a peaking factor of 1.1 is used to apply the heat in the active fuel region. The decay heat profile is smoothed and converted to a local average for each of the axial fuel segments defined for the homogenized fuel assemblies in the finite element model. Section A3.3.2.2.1.3 describes the conversion method and lists the resulting peaking factors used in the model.

Convection and radiation to the ambient are combined together to form a total heat transfer coefficient, which is defined as a temperature dependent material property in the model. The total heat transfer coefficients are used to apply to the outer surface of the cask. Section A3.3.2.2.4 describes the correlations to calculate these total heat transfer coefficients.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-7 Solar radiation is considered as a constant heat flux applied on the SURF152 elements overlaid on the outer surface of the cask. The outer surfaces of the cask are painted white. The insolance values from 1 0CFR71 (Reference 38) are considered as the maximum amount of solar radiation that is available for absorption on any surface.

These values are multiplied by the absorptivity factor of each surface to calculate the amount of solar heat flux that each surface absorbs. The resultant value is applied as a constant heat flux to the corresponding surface. The heat flux values used in the model are listed below.

Total solar heat Insolance flux averaged Solar heat flux (Reference 38) over 24 hrs. in the model Surface (gcal/cm2) (Btu/hr-in2) Absorptivity (Btu/hr-in2)

Cask Surface, Curved, Painted 400 0.427 0.3 0.128 Cask Surface, Flat Horizontal, Painted 800 0.853 0.3 0.256 Concrete, Flat, Horizontal 800 0.853 1.0 0.853 It is assumed that soil has a temperature of 70 °F at 10 feet below the cask bottom plate for hot conditions. The soil temperature for the cold condition (-40 °F) is assumed to be 45 °F. The concrete pad is 3 feet thick. Due to low conductivity of concrete and soil, the model is insensitive to the thickness of the pad / soil and the soil temperature.

Typical boundary conditions for the TN-40HT cask full-length model are shown in Figure A3.3-7.

The full-length model shows approximately 1.00 kW (3412.3 Btu/Hr) of heat dissipates from the top of the cask. To bound the analysis, a heat flow of 1.2 kW equal to 3.5% of the total heat load (decay heat plus insolance) is considered to be transferred to the cask top in the sub-model. This heat flow is applied as uniform heat flux at the bottom of the cask top sub-model.

qtop- 1.2x 3412.3 q top = 2 = 0.6509 Btu/hr-in 2 p/4 D - p/4* 89.5 2 The boundary conditions for the cask top sub-model are shown in Figure A3.3-8.

A3.3.2.2.1 .2 Thermal Models for Accident Conditions The TN-40HT casks will be stored on a concrete pad away from combustible material.

Therefore, no fires other than small electrical fires are considered credible at the ISFSI.

However, a hypothetical fire accident is evaluated for the TN-40HT cask based on a diesel fuel fire, resulting from a ruptured fuel tank of the cask transporter tow vehicle.

The bounding capacity of the fuel tank is 200 gallons and the bounding hypothetical fire is an engulfing fire around the cask.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-8 Another evaluation performed on the cask is the thermal response of the cask in the postulated event of it being completely buried by dirt and debris with very low thermal conductivity. The temperature-time history of the cask components during these events are reported in this section.

The purpose of the accident thermal analyses is to calculate the maximum fuel cladding and the maximum seal temperatures in order to demonstrate the containment integrity of the TN-40HT cask.

Two models, a cask cross section model and a lid-seal region model, are used for the evaluation. The cask cross-section model is created by selecting the nodes and elements at the midsection of the full-length cask model described in Section A3.3.2.2.1 .1.1. The selected segment contains all the nodes and elements of the full-length cask model from z = 37.57 to z = 67.57 in the axial direction, where the decay heat generation and the resultant temperatures are at their highest for the hot off-normal storage conditions. The height of this model is 30, which is twice the nominal height of one basket segment. The lid-seal region model is identical to the cask top sub-model described in Section A3.3.2.2.1 .1.2.

The geometry and the material properties of these sub-models are the same as those for the normal and off-normal cask models described above. The models for accident condition analysis are shown in Figure A3.3-9.

The initial temperatures for both cases are retrieved from the result file of the off-normal storage conditions using the sub-modeling ANSYS commands N WRITE and CBDOF (Reference 36). The boundary conditions for the sub-models are discussed in the following sections.

A3.3.2.2.1 .2.1 BOUNDARY CONDITIONS FOR THE HYPOTHETICAL FIRE ACCIDENT An average flame temperature of 1475 o F and flame emissivity of 0.9 from Reference 2 o 38 and a forced convection heat transfer coefficient of 4.5 Btu/hr-ft - F from Reference 39 were used for the fire accident case.

The "pool" of fuel is assumed to extend 1 meter beyond the cask surface (Reference 40). Based on an outer shell diameter of 101 inches, this gives a "pool" diameter of approximately 180 (with the cask standing upright) and a pool surface of 25,400 in 2 . A fuel consumption rate of 0.15 in/min. was selected from a Sandia Report (Reference 39) concerning gasoline/tractor kerosene experimental burning rates. This translates into a fuel consumption rate of approximately 16.5 gal/min. Therefore, the 200 gallons of fuel will sustain a fire for about 12 minutes and hence a 15 minute fire is evaluated.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-9 The convection and radiation heat transfer from the fire are combined together as a total heat transfer coefficient. The calculation of the total heat transfer coefficient for fire is discussed in Section A3.3.2.2.4. The total heat transfer coefficients are defined as temperature dependent material properties and used to apply the boundary conditions on the outer surface of the cask. The free convection and radiation during the post fire cool down period are also added together as a total heat transfer coefficient. An ambient temperature of 100 °F equal to the maximum hot off-normal storage temperature is assumed for the cool down period.

The fire accident can occur only during the transfer operation where combustible fuel is available. During transfer operation, the transferred cask is not surrounded by other casks. Nevertheless, a view factor of 0.8 is considered for the cask during cool down period for conservatism.

During the cool down period, the solar radiation is applied as a constant heat flux on the SURF1 52 elements (Reference 36) overlaid on the outer surface of the cask. It is assumed that the outer surfaces of the cask are covered with soot after the fire accident. A solar absorptivity of 1.0 (maximum value) is considered for these surfaces to bound the condition. The solar heat flux values applied in the model are listed below.

Total solar heat flux Insolance averaged over 24 Solar heat flux (Reference 38) hrs in the model Surface (gcal/cm2) (Btu/hr-in2) Absorptivity (Btu/hr-in2)

Cask surface, curved, painted 400 0.427 1.0 0.427 Cask surface, flat horizontal, painted 800 0.853 1.0 0.853 For the cross-section model, the heat generating boundary conditions are applied on the elements representing the homogenized fuel with a peaking factor of 1.1. No axial heat transfer is considered at the top and bottom ends of the cross-section model.

Heat generation rate = q= q x PF = 0.3218 Btu/hr-in 3 aLa where, q = Decay heat load per assembly = 2,730 Btu/hr (0.80 kW) a = Width of the modeled fuel assembly = 8.05 L a =Active fuel length = 144 PF = maximum peaking factor =1.1 Although the solid neutron shield resin might degenerate in the first few minutes of the fire accident, it is assumed that its conductivity remains unchanged throughout the fire.

After the fire, the conductivity of the neutron shield is replaced by air conductivity.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-10 To maximize the heat input from fire into the transfer cask, the 0.01 air gaps between the neutron shield aluminum boxes and the adjacent shells are removed and replaced by properties of Al-6063. For the same reason, the shrink fit gap between the inner shell and the shield is also removed during the burning time and replaced by properties of SA 516, Gr. 70 (cask shield shell). These gaps are restored immediately after the fire to maximize the thermal resistance, which consequently maximizes the fuel cladding temperature.

The maximum poison plate conductivity would be lower than that of pure aluminum.

Thus the conductivity of the poison plates was increased to the values for aluminum 1100 during the burning time and restored to the lower conductivity during the cool down period to bound the uncertainty about the conductivity of poison plates.

For the lid-seal region model, the heat flux from the cask calculated in Section A3.3.2.2.1 .1.3 (0.6509 Btu/hr-in 2 ) is applied at the bottom of the model.

Radiation heat transfer between the protective cover and the upper surfaces of the cask lid and the top neutron shield is considered in the lid-seal region model. This radiation heat transfer maximizes the heat input from the fire via the protective cover to the vent and port seals located in the cask lid.

Similar to the cask cross-section model, the 0.01 air gaps between the top neutron shield and the cask lid, the cask lid and the top shield plate, and the protective cover and cask flange in the cask top model are removed during the fire and replaced by properties of SA 516, Gr. 70 carbon steel. These gaps are restored in their original location during the cool down period.

The conductivity of neutron shield resin is considered for the elements within the top neutron shield case during the burning period. The conductivity of these elements is changed to air during the cool down period.

The resultant time-temperature histories of the fuel cladding, discussed in Section A3.3.2.2.2.4, show a continuous temperature increase without any extreme for the fire accident case. Therefore, the maximum fuel temperature for the post-fire accident case is calculated using a steady-state run of the cask cross-section model.

A3.3.2.2.1.2.2 BOUNDARY CONDITIONS FOR THE POSTULATED BURIED CASK The heat generation boundary conditions of the fuel are identical to those described for the hypothetical fire accident case. The heat flux toward the cask top applied at the bottom of the cask top sub-model is also identical to those described for the hypothetical fire accident case. Adiabatic boundary conditions are considered on the outermost nodes of the cask cross-section and the lid-seal models for buried cask case.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-1 1 A3.3.2.2.1.3 AXIAL DECAY HEAT PROFILE The normalized axial burnup profile for typical PWR fuels with burnups higher than 30 GWd/MTU is given in Reference 26 and is listed in Table A3.3-1. The active fuel length is 144.

Seventeen axial fuel regions are defined for the fuel assembly in the finite element model. Local peaking factors (PF z ) at the start and end of each region are calculated based on linear interpolation between data in Table A3.3-1. The average of the local peaking factors at the start and end of each region is assigned as PF m . The average peaking factors (PF m ) are smoothed to give an axial decay heat profile identical to the profile from Reference 26.

The average peaking factors used in the model are listed in Table A3.3-2.

Figure A3.3-10 compares the decay heat profile used in the model with the profile from Reference 26.

The area underneath the smoothed peaking factor profile is calculated as follows.

n A= PFmi, (z - - )

Where, A = total area underneath the axial heat profile PF m , i = average peaking factor in fuel region i z i = fuel region level at the end of region i The total area under the axial decay heat profile divided by the active fuel length must be equal to 1. As seen in Table A3.3-2, the resultant value is 0.993. A correction factor of 1/0.993 is therefore multiplied by the heat generating rate to avoid any degradation of the applied decay heat in the finite element model.

A3.3.2.2.2 RESULTS OF THE THERMAL ANALYSES A3.3.2.2.2.1 MAXIMUM TEMPERATURES FOR NORMAL/OFF-NORMAL STORAGE CONDITIONS Steady state thermal analyses are performed using a total decay heat load of 32 kW (0.80 kW/assembly), 100 °F daily average temperature, and the maximum insolance per 1 0CFR71 (Reference 38). Insolance is averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-12 The resultant temperature distributions for the cask models are shown in Figure A3.3-1 1, Figure A3.3-12 and Figure A3.3-1 3. Summaries of the maximum component temperatures are listed in Table A3.3-3. The temperatures for the normal storage conditions are bounded by those for hot off-normal conditions. Therefore, thermal stresses and cask maximum internal pressures are calculated based on the temperatures resulting from hot off-normal conditions.

The volumetric average temperatures for the basket, aluminum shim, cask inner shell, basket support bars, and basket aluminum plates at the hottest cross section are retrieved from the thermal models and are listed in Table A3.3-4.

A3.3.2.2.2.2 MINIMUM TEMPERATURES FOR NORMAL/OFF-NORMAL STORAGE CONDITIONS Temperature distributions under the minimum ambient temperature of -40 o F with no insolance and the maximum design heat load of 32 kW are determined under steady state conditions to maximize the temperature gradients in the cask. Table A3.3-5 summarizes the results of this analysis.

A3.3.2.2.2.3 EVALUATION OF THERMAL PERFORMANCE FOR NORMAL/OFF-NORMAL STORAGE CONDITIONS The thermal analysis for storage conditions demonstrates that the TN-40HT cask design meets all applicable requirements.

The maximum component temperatures calculated using conservative assumptions are lower than the allowable limits. The maximum cask lid seal and vent and port seal temperatures are below the long-term limit specified for continued seal function. The average resin temperature at the hottest cross section in the radial neutron shield and the top neutron shield are lower than the limit. Therefore, no degradation of the neutron shielding is expected. The calculated maximum fuel cladding temperature is well below the temperature limit considered for normal conditions of storage given in ISG-1 1 (Reference 25).

A3.3.2.2.2.4 MAXIMUM TEMPERATURES FOR ACCIDENT CONDITIONS The time-temperature histories of the cask components are shown in Figure A3.3-14 for the hypothetical fire accident analysis performed on the cask cross-section and the lid-seal models. As Figure A3.3-14 shows, the fuel cladding temperature increases during and after the fire without reaching a peak. Therefore, a steady-state run using post fire conditions is used to determine maximum fire accident temperature. Final temperature distributions resulting from the steady state run of the cask cross-section model are shown in Figure A3.3-1 5. The temperature distribution for the lid-seal model is shown in Figure A3.3-1 6 for fire accident conditions. Table A3.3-6 summarizes the maximum temperatures resulting from the transient or the steady state runs for the fire accident case.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-13 Due to adiabatic boundary conditions assumed at the outer surface of the buried cask, the cask component temperatures will continue to increase as a function of time. The times at which the component temperatures reach the allowable limits are listed in Table A3.3-7.

A3.3.2.2.2.5 EVALUATION OF THERMAL PERFORMANCE FOR ACCIDENT CONDITIONS The thermal analysis for the accident conditions demonstrates that the TN-40HT cask meets all applicable requirements.

The maximum fuel cladding temperature for the fire accident case is well below the allowable limit of 1058 °F recommended in Reference 25. The lid-seal temperature reaches the peak value at the end of fire. The vent and port seals reach their peak temperatures at 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the end of the fire. The seal temperatures remain well below the allowable limit of 536 °F. It is assumed that the solid neutron shield degenerates during the fire. Therefore, the maximum temperature of the solid neutron shield is irrelevant.

The analysis results of the buried cask accident show that if the cask is not uncovered within 1.85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br />, the neutron shield temperature exceeds the allowable limit of 300 °F.

The fuel temperature exceeds the allowable limit of 1058 °F about 95.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> after the cask is buried completely.

Seventy five (75) hours after the cask is buried, the average cavity gas temperature reaches 835 °F. The pressure based on this temperature is below a cavity internal pressure of 100 psig as concluded in Section A3.3.2.2.6. The seal temperature remains well below the allowable limit of 536 °F.

A3.3.2.2.3 MATERIAL PROPERTIES The temperature dependent thermal properties of material used in the thermal analyses are provided in Table A3.3-8. The tables are linearly interpolated to obtain values for temperature between table values.

A3.3.2.2.3.1 DETERMINATION OF NEUTRON ABSOBER (POISON) PLATES THERMAL PROPERTIES BoralTM , borated aluminum, and a metal matrix composite are possible poison materials to be used in the TN-40HT cask. BoralTM is the poison plate material conservatively analyzed below. The poison material consists of two cladding layers and one core. The core material conductivity is significantly lower than the cladding material conductivity.

For conservatism, no credit is taken for the cladding conductivity.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-14 Boral TMThermal conductivity of Core Thermal conductivity of Poison Plate (W/cm-K)

Temperature (°F) (Reference 33) (Btu/hr-in-°F) (Btu/hr-in-°F) 100 0.859 4.14 4.14 500 0.768 3.70 3.70 Based on data from Reference 33, the average density for BoralTM plate is calculated as follows:

Da =2.713(g / cm 3)x t a (cm); t a = cladding thickness 3

Dc = 2.48 1 ( g / cm ) x t c (cm) ; t c = core thickness Dt =( 2 x D a+ Dc)

Poison plate density = Dt / tt ; t = Boral TM plate thickness t

For a core thickness of 0.1 for a 0.125 thick BoralTM plate, the plate density is 3 3 0.091 lbm/in . A value of 0.0896 lbm/in is conservatively used for poison plates in the analysis.

Specific heat for BoralTM plate can be calculated as follows:

2 Cp ax D ac + Cp c xD Cp=

t (Reference 33)

Dt Specific heat Specific heat of Aluminum Cladding Specific heat of Core of Poison Plate Boral TM (Cp ) a (Cp )

c (Cp )

t Temperature (°F) (kJ/kg-K) (Btu/lbm-°F) (kJ/kg-K) (Btu/lbm-°F) (Btu/lbm-°F) 100 0.919 0.22 0.936 0.22 0.22 500 1.12 0.27 1.38 0.33 0.32*

• 0 . 33 Btu/lbm-°F is used in the model. Due to small differences between these values the thermal evaluation remains unaffected.

A3.3.2.2.3.2 DETERMINATION OF HELIUM THERMAL PROPERTIES The thermal properties for helium are calculated based on the following polynomial function from Reference 27.

k= CiTi for conductivity in (W/m-K) and T in (K)

For 300 < T < 500 K for 500< T < 1050 K c0 -7.76 1491 E-03 c0 -9.0656E-02 c1 8.66192033E-04 c1 9.37593087E-04 c2 -1 .5559338E-06 c2 -9.13347535E-07 c3 1.401 50565E-09 c3 5.55037072E-1 0 c4 0.0E+00 c4 -1.26457196E-13

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-15 A3.3.2.2.3.3 DETERMINATION OF AIR THERMAL PROPERTIES The thermal properties for air are calculated based on the following polynomial function from Reference 27.

k= CiTi for conductivity in (W/m-K) and T in (K)

For 250 < T < 1050 K c0 -2.2765010E-03 c1 1 .2598485E-04 c2 -1.4815235E-07 c3 1.7355064E-10 c4 -1.0666570E-13 c5 2.4766304E-1 7 A3.3.2.2.3.4 DETERMINATION OF CONCRETE AND SOIL THERMAL PROPERTIES The thermal conductivity of normal, saturated concrete varies from 1.2 to 2.0 Btu/ft-hr-°F at temperatures ranging from 50 to 150 °F (Reference 30). The conductivity of concrete decreases rapidly with the rise in temperature and assumes, at 750 °C (1382 °F) a conductivity value equal approximately to 50 percent of that of normal temperature (Reference 30). For the thermal analyses a thermal conductivity of 1.15 Btu/hr-ft-°F (0.0958 Btu/hr-in-°F) is used for concrete at 70 °F. This conductivity is reduced by half to a value of 0.0479 Btu/hr-in-°F at 1382 °F. A thermal conductivity of 0.3 W/m-K (0.0144 Btu/hr-in-°F) is considered for soil (Reference 31).

Since the concrete pad is not included in the transient runs such as the fire accident and the vacuum drying cases, the density and specific heat of concrete and soil are not provided.

A3.3.2.2.3.5 EMISSIVITIES AND ABSORPTIVITIES All outer surfaces of TN-40HT cask and all inner and outer surfaces of the protective cover are painted white. Reference 32 gives an emissivity between 0.92 and 0.96, and a solar absorptivity between 0.09 and 0.23 for white paints. To account for dust and dirt and to bound the problem, the thermal analysis uses an emissivity of 0.9 and a solar absorptivity of 0.3 for white painted surfaces.

Emissivity of concrete is between 0.9 and 0.94 (References 31 and 32). An emissivity of 0.90 is used for concrete surfaces. For conservatism a solar absorptivity of 1.0 is used for concrete surface to bound the effect of insolation.

The emissivity of the cask outer surface is set to 0.8 to be consistent with the requirements in Reference 38 during the accident fire. It is assumed that the cask surface is covered with soot after the fire. The solar absorptivity of soot is 0.95 (Reference 32). To bound the problem, the thermal analysis uses a solar absorptivity of 1.0 and an emissivity of 0.9 for cask surfaces after the fire.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-16 A3.3.2.2.3.6 EFFECTIVE FUEL THERMAL PROPERTIES A3.3.2.2.3.6.1 DISCUSSION The finite element models of the TN-40HT cask simulate the effective thermal properties of the fuel with a homogenized material occupying the volume within the basket where the fuel assemblies are stored. Effective values for density, specific heat, and conductivity are determined for this homogenized material to use in the finite element models.

The characteristics of the fuel assemblies to be stored in the TN-40HT cask are listed in Section A7, Table A7.2-1. Among these assemblies, the smallest cladding thickness and the largest gap between the fuel pellet and clad are chosen to determine the bounding thermal conductivities for the homogenized fuel assemblies.

A3.3.2.2.3.6.2

SUMMARY

OF FUEL MATERIAL PROPERTIES The properties used for UO 2 and Zircaloy for calculation of the effective fuel properties are given below. The conductivities of helium and air as backfill gases are listed in Table A3.3-8.

The air conductivity is used for vacuum drying conditions. The vacuum drying process generally does not reduce the pressure sufficiently to reduce the thermal conductivity of the water vapor and air in the cask cavity. Therefore, air conductivity is assumed for the backfill gas for vacuum drying conditions. For other conditions, the conductivity of helium is used for backfill gas.

A3.3.2.2.3.6.2.1 FUEL PELLET, UO 2 The conductivity, specific heat, and density for UO 2 fuel pellets are taken from Reference 14 and listed below.

Temperature (°C) k (cal/s-cm-°C) Temperature (°F) k (Btu/hr-in-°F) 25 0.025 77 0.503 100 0.021 212 0.423 200 0.018 392 0.362 300 0.015 572 0.302 500 0.01 32 932 0.266 700 0.0123 1292 0.248 800 0.0124 1472 0.250

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-17 Temperature (°C) c (cal/g-°C) p Temperature (°F) c (Btu/lbm-°F) p 0 0.056 32 0.056 100 0.063 212 0.063 200 0.0675 392 0.068 400 0.0722 752 0.072 1200 0.079 2192 0.079 3

The density of fuel pellets (UO 2 ) is 10.96 g/cc = 0.396 lbm/in .

The thermal conductivities shown above represent values for un-irradiated UO 2 pellets.

A study performed by Transnuclear (TN) and provided to the NRC in Reference 37 shows that the transverse effective fuel conductivity with irradiated UO 2 conductivity is approximately 3% lower than the one with un-irradiated UO 2 conductivity at a temperature of 700° F.

The sensitivity runs in the TN study showed that the fuel cladding temperature changes by approximately 1°F when using irradiated UO 2 conductivity. Since a cladding temperature change of 1 º F is negligible, the results of the study show that the fuel cladding temperature is not sensitive to the conductivity of UO 2 . Therefore, use of un-irradiated UO 2 fuel pellet conductivity from NUREG/CR-0200 (Reference 14) is reasonable for irradiated UO 2 .

A3.3.2.2.3.6.2.2 FUEL CLADDING, ZIRCALOY-4 / ZIRLO Table B-2.I of Reference 41 lists measured and calculated values of thermal conductivity for zircaloy at various temperatures. The measured values used in this calculation are listed below.

Temperature (K) k (W/m-K) Temperature (°F) k (Btu/hr-in-°F) 373.2 13.6 212 0.655 473.2 14.3 392 0.689 573.2 15.2 572 0.732 673.2 16.4 752 0.790 773.2 18.0 932 0.867 873.2 20.1 1112 0.968

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-18 Table B-1 .1 of Reference 41 lists specific heat values for zircaloy as a function of temperature.

3 The density of zircaloy is 6.56 g/cm 3 = 0.237 lbm/in , as defined in Reference 41.

Table B-3.1 1 of Reference 41 lists the measured emissivity values for fuel cladding. For ease of calculation a temperature independent emissivity of 0.8 is set for Zircaloy-4 in this calculation.

Note that Reference 42 states that ZIRLO and Zircaloy-4 alloys are very similar in terms of their thermal characteristics. Therefore zircaloy properties above are adequate for modeling ZIRLO.

A3.3.2.2.3.6.2.3 EMISSIVITY OF FUEL COMPARTMENTS, STAINLESS STEEL PLATES An emissivity of 0.3 for the stainless steel plates is used for the fuel compartments in calculating the transverse effective fuel conductivity. This value is conservative relative to the values provided in Reference 6.

A3.3.2.2.3.6.3 EFFECTIVE FUEL CONDUCTIVITY A3.3.2.2.3.6.3.1 TRANSVERSE EFFECTIVE CONDUCTIVITY The purpose of the effective conductivity in the transverse direction of a fuel assembly is to relate the temperature drop of a homogeneous heat generating square to the temperature drop across an actual assembly cross section for a given heat load. This relationship is established by the following equation obtained from Reference 43:

q q react k -

eff (0.29468) = (0.29468) 4L a (Tc - T) o (T - To )

Where:

keff = Effective thermal conductivity (Btu/hr-in-ºF) q = Assembly heat generation (Btu/hr) q react = Reaction solution retrieved from the quarter-symmetric 2D model (Btu/hr) q = 4xq react x La L a = Assembly active length (in)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-19 T c = Maximum temperature (°F)

T o = Compartment Wall temperature (°F)

A two dimensional, quarter-symmetric finite element model of a 14x14 fuel assembly is developed using the ANSYS computer code (Reference 36). All components are modeled using 2D PLANE55 thermal solid elements. This two-dimensional model simulates the heat transfer by radiation and conduction. No convection is considered within the fuel assembly for conservatism. Radiation between the fuel rods and the compartment walls is simulated using the /AUX12 processor in ANSYS. For this purpose, LINK32 elements are placed on the exteriors of the fuel rods for creation of the radiation super-element. All LINK32 elements are unselected prior to solution of the thermal problem.

The geometry of this model is shown in Figure A3.3-1 7.

A fuel assembly decay heat load of 0.80 kW is used for heat generation. An active length of 144 is assumed for the model. Several computational runs were made for the model using isothermal boundary temperatures ranging from 100 to 1000 °F. The isothermal boundary conditions are applied on the outermost nodes of the model, which represent the compartment walls. In determining the temperature dependent effective conductivities of the fuel assembly an average temperature, equal to (T o +T c )/2, is used for the fuel temperature. A typical temperature distribution for the fuel assembly model is shown in Figure A3.3-1 8.

The transverse effective conductivity is calculated with helium as backfill gas for all conditions except for vacuum drying. For vacuum drying conditions, the conductivity of helium is replaced by conductivity of air.

A3.3.2.2.3.6.3.2 AXIAL EFFECTIVE CONDUCTIVITY The backfill gas, fuel pellets, and fuel cladding behave like resistors in parallel.

However, due to the small conductivity of the fill gas and the axial gaps between fuel pellets, credit is only taken for the Zircaloy-4 in fuel cladding in the determination of the axial effective conductivities.

cladding area keffl =

xk (Btu/hr-in-°F) 4a 2 a = half of compartment width = 8.05/2 = 4.025 A3.3.2.2.3.6.4 EFFECTIVE FUEL DENSITY AND SPECIFIC HEAT Volume average density and weight average specific heat are calculated to determine the effective density and specific heat for the fuel assembly. The equations to determine the effective density and specific heat are shown below.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-20 ri V r

UO 2 VO +

r eff Vassembly

=

4a2L a mC=M O C

O2 +

=

p,eff mi M O2

+M Zr A3.3.2.2.3.

6.5 CONCLUSION

The transverse effective conductivity values for fuel assemblies in helium are listed in Table A3.3-9. The effective transverse conductivity used for fuel assemblies in thermal analyses for normal/off-normal (when in helium) and accident conditions is lower than the value calculated in this section. The applied values for transverse fuel conductivity are shown in Table A3.3-8 and are compared to the calculated value in Figure A3.3-1 9.

Note that the transverse effective conductivity values for fuel assemblies in air are also shown on Table A3.3-8 as are other calculated effective properties for a homogenized fuel assembly.

The transverse effective conductivities for fuel assemblies reported in Table A3.3-8 (Page 1 of 7) were calculated based on an incorrect Stefan-Boltzmann constant of 1 .983E-1 3 Btu/hr-in 2-°R4 . This value is 60 times lower than the correct value of 1 .190E-11 Btu/hr-in 2-°R4 . The transverse effective fuel conductivities calculated with correct parameters are reported in Table A3.3-9. As seen in Figure A3.3-1 9, the transverse effective fuel conductivities applied in the TN-40HT model are lower than those calculated with the correct parameters.

Since the transverse effective conductivities applied in the TN-40HT model are lower than the corrected values, the results for peak cladding temperatures under normal and off-normal storage conditions were not changed and are conservative.

As described in Section A3.3.2.2.2.4, the peak cladding temperature under fire accident conditions is determined using a steady-state run with the post fire conditions.

Therefore, using the lower transverse effective fuel conductivities is also conservative for peak cladding temperature under fire accident conditions.

The peak cladding temperatures will decrease for all transient conditions analyzed in the SAR as the transverse effective fuel conductivities increase. This is primarily due to the lower initial temperatures caused by higher conductivities at the initial conditions of the transient runs.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-21 A3.3.2.2.4 HEAT TRANSFER COEFFICIENTS A3.3.2.2.4.1 TOTAL HEAT TRANSFER COEFFICIENT TO AMBIENT (NORMAL/OFF-NORMAL)

The outer surfaces of the cask dissipate heat to the ambient via free convection and radiation. Total heat transfer coefficient is defined as:

Ht = h r + h c Where, h r = radiation heat transfer coefficient h c = free convection heat transfer coefficient The radiation heat transfer coefficient, h r , is given by the equation:

s (T 2- T 2) h =e F12 2

(Btu/hr-ft -°F)

T - T2 Where, e = surface emissivity F 12 = view factor from cask surface to ambient

-8 s = 0.1714 *10 Btu/hr-ft 2 -°R4 T 1 = cask surface temperature, °R T 2 = ambient temperature, °R

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-22 Since the TN-40HT casks can be modeled as being stored in a 2 x ¥ array with a nominal pitch of 18 ft. as shown in Figure A3.3-20, the radiation view factor from the radial cask surfaces to the environment is less than 1.0. A view factor of 0.8138 is calculated for in Section 3.3.2.2.1 between the TN-40 cask and the ambient for this same configuration. This value is multiplied by surface emissivity to calculate radiation heat transfer from cask radial surfaces.

Cask gray body exchange factor = F 12 x c = 0.8138 x 0.9 = 0.7324 For conservatism, a value of 0.72 is used in the determination of the total heat transfer coefficient that is applied in the detailed TN-40HT cask model described in Section A3.3.2.2.1 .1 for calculating of radiation heat transfer.

A3.3.2.2.4.1 .1 STORAGE ARRAY RADIANT HEAT TRANSFER EVALUATION The view factor between the cask surface and the concrete pad and service road surrounding the concrete pad were not considered explicitly in the calculation of the view factor for the TN-40 in Section 3.3.2.2.1.

The surface temperature of the concrete pad and service roads maybe higher than ambient temperature due to solar radiation and thermal radiation from the casks, but it is significantly lower than the cask surface temperature. Therefore, the radiation exchange between the casks and the concrete pad/service road is significant, but it maybe lower than the radiation exchange between the casks and the ambient.

A three dimensional model of TN-40HT casks in a storage array is developed using ANSYS (Reference 36) to investigate the effects of the cask view factor on the thermal performance. The model includes 18 casks and considers a cask pitch of 18 ft. The storage pad width is extended three times the cask pitch (72 ft.) to minimize any uncertainty regarding the radiation exchange between the casks and the surroundings.

This model is shown in Figure A3.3-21.

Effective conductivities for cask shells and cask body in axial and radial directions are calculated using the detailed model of TN-40HT cask described in Section A3.3.2.2.1 .1.1. The TN-40HT cask is divided into five sections as shown in Figure A3.3-22 to calculate the effective conductivities. The methodologies and results for effective conductivities of cask sections are described in below. These effective conductivities are used in the storage array model to reduce the number of elements and create manageable input and output files.

The dimensions of the TN-40HT cask in the storage array model are identical to those used in Section A3.3.2.2.1 .1.1. The decay heat load is applied as a uniform heat flux on the cask inner surface over the length of basket (160 in.). This heat flux is calculated as follows.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-23 Decay heat = 32 kW = 109,193.6 Btu/hr I D cask = 72 I

Height basket = 160 Decay heat flux = Decay heat /(p ID cask x Height basket ) = 3.017 Btu/hr-in 2 The solar heat flux on the cask outer surface and the fixed temperatures at 10 ft. below the cask bottom plate are identical to those described in Section A3.3.2.2.1 .1.3. Free convection boundary conditions are considered at the cask outer surface using correlations described in Section A3.3.2.2.4.3. Radiation exchange between cask and surroundings is simulated using radiosity methodology in ANSYS (Reference 36). An ambient temperature of 100 °F is considered for this model.

The resultant temperature distributions are shown in Figure A3.3-23 for cask outer surfaces, in Figure A3.3-24 for the storage pad surface temperature, and in Figure A3.3-25 for the cask inner shell.

As seen in Figure A3.3-21, the storage array model considers complete casks without symmetry planes. The detail cask model described in Section A3.3.2.2.1.1.1 considers a 90 degree, quarter symmetric segment of the cask. In order to compare the temperature profiles of these two models, average temperatures at the hottest cross sections are retrieved form each model.

The average and maximum temperatures for storage array model and TN-40HT detailed cask model are compared in Table A3.3-10.

As seen in Table A3.3-1 0, the average temperatures for the cask inner shell in the array storage and the cask detail models are within 1 °F of each other. Therefore, the temperatures for the basket and its contents would not be affected significantly if the view factor from the cask to surrounding storage pads were included in the thermal model.

The average cask surface temperature in the detailed cask model is 242 °F while the average cask surface temperature in the storage array model is 260 °F. This shows a temperature increase of 18 °F due to the cask view factors.

The calculated average resin temperature at the hottest cross section of the detailed model is 267 °F. The addition of the storage pad to the cask view factor calculation increases the average resin temperature at the hottest cross section by at most 18 °F to 285 °F. This temperature remains below the allowable limit of 300 °F for the radial resin.

Since the basket temperatures calculated in the TN-40HT detailed model remain unaffected, these temperatures can be used for calculation of the thermal stress, thermal expansion, and cask cavity pressure for the structural evaluation.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-24 The view factors from the emitting cask (shown in Figure A3.3-20) to other casks and to the storage pad are calculated from the storage array model. These view factors are compared in Table A3.3-11 to the values calculated in Section 3.3.2.2.1 for information purposes only.

Effective Conductivity for Cask Sections To calculate the axial and the radial effective conductivities of the cask sections shown in Figure A3.3-22, corresponding nodes and elements for each section are selected from the detailed finite element model of the TN-40HT cask.

To calculate the axial effective conductivity of each cask section, constant temperature boundary conditions are applied at the top and bottom of that section. The value of reaction solution for the colder surface is retrieved from the model using PRNDL command (Reference 36) after solution was completed. Since the model is quarter symmetric, the amount of heat leaving the colder surface is four times the reaction solution resulted from PRNLD command. The axial effective conductivity is calculated using the following equation:

QxL keff,axl -

Ax D T Q = Amount of heat leaving the colder face of the cask section = Qreaction x 4 L = Length of the cask section A = Surface area of the colder face (upper face) of the cask section 2

= p ( r2 - r 12) r 1 = Cask cavity radius = 36 for cask sections 1 through 4, 0 for cask section 5 r2 = Outer radius of the cask section = 44.75 for cask sections 1, 2, and 4

= 50.5 for cask section 3

= 36 for cask section 5 DT = (T2 - T 1 ) = Temperature difference between lower and upper faces of the section

(°F)

T 1 = Constant temperature applied on the upper face of the model (°F)

T 2 = Constant temperature applied on the lower face of the model (°F)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-25 To calculate the radial effective conductivity of the cask sections, constant temperature boundary conditions are applied on the outermost and innermost nodes of the sections.

The value of reaction solution on the colder surface is then retrieved from the model using PRNDL command and multiplied by four to give the amount of heat leaving the colder surface. The radial effective conductivity is calculated using the following equation with the same terms defined for axial effective conductivity:

ln(r2/r1) k=Qx e11,rad 2p L D T For Section 5, the lowest conductivity between the cask lid and cask top shield (SA203, Gr. E, Table 3.3-8) is used for radial conductivity.

In determining the temperature dependent effective conductivities an average temperature, equal to (T 2 + T 1 )/2, is used. The resulting effective conductivities are listed in Table 3.3-12.

A3.3.2.2.4.2 TOTAL HEAT TRANSFER COEFFICIENT TO AMBIENT FOR FIRE The radiation and forced convection from the fire toward the cask surface are combined together as a total heat transfer coefficient. Total heat transfer coefficient is defined as:

H t,1ire = h r ,1ire + h 1 Where, h r,fire = radiation heat transfer coefficient from fire h f = forced convection heat transfer coefficient A forced convection value of 4.5 Btu/hr-ft2-°F (0.03125 Btu/hr-in 2-°F) is considered during the burning time from Reference 39.

The radiation heat transfer coefficient during the burning period of the hypothetical fire accident, hr,fire, is given by the following equation:

s(e 1 T1 4 -4) 2 F1 [

(Btu/hr-ft -°F)

(T1-T)

L Where, e s = surface emissivity = 0.8 (Reference 38) ef = fire emissivity = 0.9 (Reference 38)

F sf = view factor from surface to fire = 1

-8 s = 0.1714 *10 Btu/hr-ft 2 -°R4 T f = fire flame temperature, 1475°F = 1935°R (Reference 38)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-26 T s = cask surface temperature, °R A3.3.2.2.4.3 FREE CONVECTION COEFFICIENTS The free convection coefficients are calculated based on the shape and position of the convective surface using correlations from Reference 27. The convection correlations are described in the following sections.

A3.3.2.2.4.3.1 VERTICAL CYLINDER The following equations from Reference 27 are used to calculate the free convection coefficients for vertical cylindrical surfaces.

h=Nuk L

With L = height of the vertical cylinder D= diameter of vertical cylinder k = air conductivity g b (T - T¥ ) L3 Ra=Gr Pr ; Gr=

n2 Nu ate =

Cl Ra 1/4 , Cl = 0.5 15 for gases (Reference 27) 2.0 Nu l Plate=

ln(1 + 2 / Nu ;late) 1.8 L / D T

x= Nu Plate x

Nu l= Nu ;late Nusselt number for fully laminar heat transfer ln(1+)

V 0. 13 Pr 0.22 Ct =

(1+0.61Pr 081 ) 042 f=1+0.078 -1

¥ 1/

Nu t = CfRa T /(1+1.4x10Pr/ Ra ) Nusselt number for fully turbulent heat transfer Nu = [(Nu l ) m + (Nu t ) m ]

/3 with m=6 The correlations to calculate the total heat transfer coefficient are incorporated in the ANSYS model via a macro. Air properties are taken from Reference 27 and listed in Table A3.3-8.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-27 A3.3.2.2.4.3.2 HORIZONTAL FLAT SURFACES FACING DOWNWARDS The following equations from Reference 27 are used to calculate the free convection coefficients for horizontal flat surfaces facing downwards.

h=Nuk L

With L = A/P A=surface area of heated surface P= perimeter of the heated surface k = air conductivity b (T - T ) L3 Ra = Gr Pr ; Gr = g n2 1/5 0.527 Ra Nu l =

[1+ (1.9/Pr) 9/10 ) 2/9 Nu = Nu l The above correlations are incorporated in ANSYS model via a macro. Air properties are taken from Reference 27 and listed in Table 3.3-8.

A3.3.2.2.4.3.3 HORIZONTAL FLAT PLATE FACING UPWARDS The following equations from Reference 27 are used to calculate the free convection coefficients for horizontal flat surfaces facing upwards.

h=Nuk L

With L = A/P A=surface area of heated surface P= perimeter of the heated surface k = air conductivity g b (T - T¥ ) L3 Ra=Gr Pr ; Gr=

n2 Nu T = 0.835 Cl Ra 1/4 Cl = 0.5 15 for gases (Reference 27) 1.4 Nu l = Nusselt number for fully laminar heat transfer ln(1 +1.4 / NuT)

H C t >> 0.14 for Pr < 100 (Reference 27)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-28 H 1/3 Nu t = C t Ra Nusselt number for fully turbulent heat transfer Nu=[(Nu l ) m + (Nu t ) mmwith m =10 for Ra> 1 The above correlations are incorporated in ANSYS model via a macro. Air properties are taken from Reference 27 and listed in Table 3.3-8.

A3.3.2.2.5 THERMAL EVLUATION OF LOADING AND UNLOADING Fuel loading and unloading operations occur in the fuel handling building. During loading operation fuel assemblies are submerged in pool water permitting heat dissipation. After fuel loading is complete, the cask is removed from the pool, drained, sealed, dried, and backfilled with helium.

A3.3.2.2.5.1 VACUUM DRYING The vacuum drying operation evaluated is the heatup of the cask before helium is introduced into the cask cavity. The time for this operation (t loading) is defined as the interval from the start of water being drained out of the cask cavity to the beginning of helium backfilling.

The duration of vacuum drying operation being evaluated is limited by the amount of time required to reach the maximum fuel cladding temperature of 752 º F (400 °C) allowed by ISG-1 1 (Reference 25).

To determine the time-temperature histories of the cask components, a transient analysis is performed using the cross-section model of the cask described in Section A3.3.2.2.1 .2. Although cask cavity is full of water at t loading = 0, the following changes are considered to occur immediately at the start of the transient run and remain unchanged.

• Effective conductivity values in a vacuum are considered for elements representing homogenized fuel assemblies.

• Air conductivity is given to the elements representing the gas within the cask cavity.

The conductivity of air or helium is independent of the pressure for the conditions considered during the loading and vacuum drying operations. All other material properties of the cask cross-section model remain unchanged. The effective fuel conductivity values used for vacuum conditions are listed in Table A3.3-8.

The decay heat is applied as heat generation load on the elements representing the homogenized fuel assemblies with a peaking factor of 1.1. The applied value for the heat generation is:

q= q x PF = 0.3218 Btu/hr-in 3

Heat generation rate =

aLa

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-29 where, q = Decay heat load per assembly = 2,730 Btu/hr (0.80 kW) a = Width of the modeled fuel assembly = 8.05 in.

L a =Active fuel length = 144 in.

PF = maximum peaking factor = 1.1 (Reference 26)

Adiabatic boundary conditions are applied on the top and bottom faces of the cross-section model for conservatism.

An average, initial temperature of 215 °F is assumed for the cask at the start of water draining. This temperature is higher than boiling point of water.

It is assumed that the cask resides in the pool for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the water draining starts.

Maximum pool temperature is 150 °F. Conservatively, a constant temperature of 215 °F is applied on the cask surface during this period.

After leaving the pool, the cask dissipates heat to ambient inside of the fuel handling building. The free convection and radiation are combined together to calculate the total heat transfer coefficient from the cask outer surface to the ambient. An ambient temperature of 120 °F is considered conservatively for the fuel handing building.

The volumetric average temperatures of the basket, rails, inner shells, basket support bars and basket aluminum plates are retrieved from the thermal model to calculate the thermal expansion. A zero hot gap at thermal equilibrium is assumed between the rails and the cask inner shell in retrieving the average temperatures. These temperatures are listed in Table A3.3-1 3.

The time-temperature history of the maximum fuel cladding temperature is illustrated in Figure A3.3-26 for loading operations with 32 kW decay heat load. The time period for this operation was defined as the time from the beginning of cask draining to when helium backfilling begins. A time limit of 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> was chosen for this operation to ensure that the fuel cladding temperature limit of 752 º F was not exceeded.

The temperature distribution for cask cross-section model at t loading = 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> is shown in Figure A3.3-27.

The temperature limits for fuel cladding, radial neutron shielding material, and seals are not exceeded at 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> as shown in Table A3.3-14.

Once helium is introduced, subsequent evacuation and backfilling with helium will not change the conductivity of the cavity gas, and therefore no repeated cycling of the fuel cladding temperature occurs.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-30 A3.3.2.2.5.2 REFLOODING For unloading operations, the cask will be slowly filled with borated water to gradually cool the fuel in the cask.

As pool water is added to the cask cavity containing hot fuel and basket components, some of the water will flash to steam causing the internal cavity pressure to rise. This steam pressure is released through the vent port. The reflooding procedures will require that the pressure be monitored and the reflood flow controlled such that the pressure does not exceed the analyzed internal pressure of 100 psig. To provide margin to the analyzed limit and to account for any pressure drop between the monitoring location and the cask internal pressure, the procedure shall limit the monitored pressure to less than 75 psig.

A3.3.2.2.6 INTERNAL CASK PRESSURE DETERMINATION A3.3.2.2.6.1 MAXIMUM INTERNAL PRESSURE The following methodology is used to determine the maximum pressures within the TN-40HT cask cavity for normal, off-normal, and accident conditions:

• Average cavity gas temperatures are derived from the TN-40HT cask thermal models.

• The amount of helium present within the cask cavity after the initial backfilling is determined via the ideal gas law.

• The total amount of free gas within the fuel assemblies, including both fill and fission gases, are calculated based on NUREG 1536 (Reference 35) guidelines.

• The amount of released gas from the fuel rods into the cask cavity is determined based on the maximum fraction of the ruptured fuel rods considered in NUREG 1536 (Reference 35).

• The amount of helium backfill gas is added to the amount of released gases to make the total amount of gases in the cask cavity.

• Finally, the maximum cask internal pressures are determined via the ideal gas law.

To bound the maximum internal pressure, the maximum daily average temperature of 100 °F is considered for both normal and off-normal conditions.

As it is assumed in NUREG 1536 (Reference 35), the maximum fractions of the fuel rods (fB) that can rupture and release their free gases to the cask cavity for normal, off-normal, and accident conditions are 1, 10, and 100%, respectively.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-31 A3.3.2.2.6.1.1 AVERAGE GAS TEMPERATURE To determine the average cavity gas temperature, volume average temperatures of the elements representing the helium gaps (Tavg,void) and the homogenized fuel assemblies

( Tavg , fuel) are retrieved from the thermal models using the ETABLE commands in ANSYS (Reference 36). Although the average temperature of the homogenized fuel elements includes the fuel rods and the helium gas between them, this average temperature is conservatively used as the average gas temperature within the fuel compartments.

The volume of helium gaps in the model is (Vvoid) is retrieved from the model using ETABLE commands (Reference 36) and is equal to 28,272 in 3 .

The approximate volume of the gas in the fuel compartments (Vgas , comp) is determined as follows (note this is only an approximation and not intended to represent the physical configuration of the fuel, e.g. actual fuel pin length, open guide tubes, and fuel assembly hardware).

Vgas,comp = N c x (Volume of one fuel compartment - Volume of fuel rods in one assembly )

N c = Number of fuel compartment in finite element model (1/4 of cask is modeled) = 10 Volume of one fuel compartment = W x W x H 2

Volume of fuel rods in one assembly = N x p/4 OD x H W = compartment width = 8.05 H = compartment height = 160 N = number of fuel rods and tubes = 196 OD = the smallest outer diameter of fuel rods and tubes = 0.40 in.

Volume of one fuel compartment = 10,368 in 3 Volume of fuel rods in one assembly = 3,941 in 3 Vgas,comp = 64270 in 3 The average gas temperature in the cask cavity is calculated as follows.

T - Tavg ,fuel xVgas ,comp +T . xV.

avg ,vcnd vcnd

' Cavity -

(Vgas,comp + Vvoid The results are summarized below.

Operating Condition T Cavity (°F)

Normal (Off-Normal) Storage Conditions 456 Fire Accident 592 Buried Cask Accident, 75 hrs after occurrence 835 Buried Cask Accident, 95.75 hrs after occurrence 929 A3.3.2.2.6.1.2 AMOUNT OF INITIAL HELIUM BACKFILL A maximum cavity pressure of 1.43 atm abs (21 psia) is considered for the initial backfill pressure of helium. The amount of the helium backfill gas at this moment is calculated based on the ideal gas law.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-32 The cavity gas temperature for the cask located in the fuel handling building is calculated considering an ambient temperature of 70 °F, a view factor of 1.0, and no insolance. The full-length cask model described in Section A3.3.2.2.1.1.1 is used for this purpose with steady state conditions. The average cavity gas temperature is retrieved from the model using the methodology described in Section A3.3.2.2.6.1 .1.

The retrieved average cavity gas temperature when cask is in the handling building is 426 °F (886 °R).

For the maximum gas backfill, it is assumed that helium does not instantaneously reach 426 °F, but is conservatively assumed to be the average of the ambient (70 °F) and the steady state cavity temperature of 426 °F.

A displacement of 480 in 3 is used for each BPRA.

From the backfill pressure and the backfill gas temperature, the amount of helium backfill gas is calculated as follows.

n back = (PV)/(RT) = 0.580 lb-moles without BPRAs

= 0.549 lb-moles with BPRAs P = maximum initial backfill pressure = 1.43 atm = 21 psia V = cask cavity free volume without BPRAs (loaded) = 362,440 in 3 = 209.7 ft 3

= cask cavity free volume with BPRAs = cavity volume - BPRA volume

= (362,440 - 40*480)/1 2 3 = 198.6 ft 3 T = initial backfill temperature = 0.5 (70+426) = 248 °F = 708 °R R = universal gas constant = 10.730 psia-ft 3/lb-moles-°R (Reference 29)

A3.3.2.2.6.1.3 FREE GAS WITHIN FUEL ASSEMBLIES As indicated in Section A7.2, Table A7.2-1, the maximum volume of free gas per assembly is 0.226 m 3 at standard temperature and pressure (0 °C and 760 mmHg).

Total amount of free gases within the fuel assemblies is:

P V XSt 760x0.226x10 3 n fuel

=N = 40x = 403.3 g-mole = 0.889 lb-mole R TSd 62.361x 273.15 With N = number of fuel assemblies in the cask = 40 P Std = standard pressure = 760 mmHg Vfree = maximum volume of free gas per assembly = 0.226 m 3 T Std = standard temperature = 0 °C = 273.15 K

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-33 R = universal gas constant = 62.361 (mmHg-lit/g-mole-K) (Reference 29)

A bounding amount of 2.0E-4 lb-mole is considered for free gas in each BPRA rod. The maximum amount of free gases from the BPRA rods is:

n BPRA = Nx n Rod x (2.0E -4 lb - mole / rod) = 40x16x2.0E-4 = 0.128 lb-mole With N = number of fuel assemblies in the cask = 40 n Rod = maximum number of BPRA rods in one assembly = 16 Total amount of free gases within fuel assemblies is:

n free = n fuel = 0.889 lb-mole without BPRAs

= n fuel + n BPRA = 1.017 lb-mole with BPRAs A3.3.2.2.6.1 .4 TOTAL AMOUNT OF GASES WITHIN CASK CAVITY The total amount of gases within the cask cavity is equal to the amount of the initial helium backfill gas plus any free gases that are released to the cask cavity from ruptured fuel and BPRA rods.

Total amount of gases in the cask cavity is:

n total = n back + fB ( n free) n back = total amount of backfill gas n free = total amount of free gases fB = fraction of the ruptured fuel rods from NUREG-1 536 (Reference 35)

A3.3.2.2.6.1.5 MAXIMUM CASK INTERNAL PRESSURE The maximum cask internal pressure (Pcavity) is determined via the ideal gas law:

P cavity = ( n total R T cavity )/V V = cask cavity free volume without BPRAs = 209.7 ft 3

= cask cavity free volume with BPRAs = 198.6 ft3 R = universal gas constant = 10.73 (psia-ft 3/lb-mol-°R)

The results are summarized in Table A3.3-1 5 and Table A3.3-1 6.

The internal cask cavity pressures are at or below the design limit of 100 psig.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-34 A3.3.2.2.6.2 INTERNAL PRESSURE AT END OF SERVICE LIFE A minimum helium backfill pressure of 19.5 psia was determined on the basis that a minimum of 1 atm pressure must exist on the coldest day at the end of life.

The full-length cask model was run with steady state conditions in the handling building to determine the average cavity gas temperature after completion of the helium backfilling. An ambient temperature of 70 °F is considered for this run. The average gas cavity temperature of 426 °F (886 °R) was retrieved from the model using the methodology described in Section A3.3.2.2.6.1 .1.

The determination of the end of life cavity pressure was based on the average gas backfill temperature of 426°F (886°R) at the time of backfill and an average gas temperature of 216°F (676°R) after 25 years of storage an external ambient temperature of -40°F.

The initial pressure of 1 9.5psia assures that at the end of 25 years, on the coldest day

(-40 F ambient), the internal pressure of the cask is:

P cavity = 19.5 psia x (676°R/886°R) = 14.9 psi Therefore, the internal pressure of the cask is above the 1 atm minimum.

A3.3.2.2.7 RADIAL HOT GAP BETWEEN THE BASKET RAILS AND THE CASK INNER SHELL A nominal diametrical cold gap of 0.30 in. is considered between the basket and the cask cavity wall for the TN-40HT cask.

A radial, hot gap of 0.13 at thermal equilibrium is assumed in the ANSYS model for normal storage conditions. To verify this assumption, the hot dimensions of the cask inner diameter and basket outer diameter are calculated at thermal equilibrium as follows.

The outer diameter of the hot basket is:

OD B,hot = OD B + [ L SS,B x a SS ( T avg , B - T ref) + L Al x a Al ( T avg,Al - T ref)]

Where:

OD B,hot = Hot outer diameter of the basket OD B = Cold outer diameter of the basket = 72 - 0.30 = 71.70 L SS,B = Length of basket at 90-270 direction = O DB - 2x0.46 = 70.78 L Al = Length of aluminum shim = 2x0.46 = 0.92 a SS = Stainless steel axial coefficient of thermal expansion

= 9.66E-6 in/in-°F @ 479 °F (Reference 28)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-35 a Al = Aluminum coefficient of thermal expansion

= 13.43E-6 in/in-°F @ 358 °F (Reference 28)

T avg,B = Average basket temperature at the hottest cross section = 479 °F (Table A3.3-4)

T avg,Al = Average shim temperature at the hottest cross section = 358° F (Table A3.3-4)

T ref = reference temperature for stainless steel and aluminum alloys = 70°F (Reference 28)

The inner diameter of the hot cask is:

ID c,hot = ID c [1 + a LS ( T avg,C - T ref)]

Where:

ID C,hot = Hot inner diameter of cask cavity ID C = Cold inner diameter cask cavity = 72 a LS = Coefficient of thermal expansion for low alloy steel

= 6.91E-6 in/in-°F @ 302 °F (Reference 28)

T avg,C = Average cask inner shell and gamma shield temperature at hottest x-section = 302 °F (Table A3.3-4)

T ref = reference temperature for low alloy steel = 70°F (Reference 28)

The hot gap between the basket outer diameter and cask inner diameter is:

G hot = ID C,hot - OD B,hot =0.132 (diametrical)

Radial hot gap = 0.066 The assumed radial hot gap of 0.13 is larger than the above calculated hot gap. This assumption is therefore conservative.

A3.3.2.2.8 HEAT GENERATION RATE AS A FUNCTION OF SPENT FUEL PARAMETERS A3.3.2.2.

8.1 INTRODUCTION

The spent fuel loading into the TN-40HT Cask is based on a uniform loading of 800 watts per fuel assembly. The objective of this section is to describe a mathematical function that determines the heat generation rate for high burnup fuel as a function of initial enrichment, assembly burnup and cooling time. This attachment also provides a sample fuel qualification table for heat generation rate, based on the mathematical function. This function applies only to 14x14 high burnup fuel assemblies.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-36 A3.3.2.2.8.2 CALCULATIONAL METHODOLOGY AND INPUT MODELS The SAS2H and the ORIGEN-S modules of SCALE4.4 (Reference 14) computer code, with the 44 Group ENDF-V cross section library, are used to determine the thermal source terms.

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 A3.3.2.2.8.3 MATHEMATICAL FUNCTION TO DETERMINE HEAT GENERATION PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 The functional form is expressed below with decay heat (DH) in watts as:

H I DH = F1

• Exp({G*[1 -(1 2.0/X3)]}*[(X3/X1 ) ]*[(X2/X1 ) ])

Where:

F1 = A + B*X1 + C*X2 + D*X1 2 + E*X1*X2 + F*X2 2 and, F1 Intermediate Function, basically the Thermal source at 12 year cooling X1 Assembly Burnup in GWd/MTU (45 to 60)

X2 Initial Enrichment in wt % U235 (3.4 - 5.0)

X3 Cooling Time in Years (12 min)

A 18.76 B 11.27

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-37 C 6.506 D 0.163 E -1.826 F 6.617 G -0.309 H 0.431 I -0.374 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 A3.3.2.2.8.4 DECAY HEAT FOR LOW BURNUP ASSEMBLIES For assemblies with burnups less than 44 Gwd/Mtu, the SASH2 analyses shows that decay heat load is less than 800 watts with a minimum cooling time of 12 years.

A3.3.2.2.8.5 HEAT GENERATION FOR INSERTS The decay heat load for a thimble plug device with a minimum of 16 years cooling was determined to be 0.42 watts. This value was based on a maximum cumulative host assembly(s) burnup of 125,000 Mwd/Mtu.

The decay heat load for a burnable poison rod assembly with a minimum of 18 years cooling was determined to be 0.90 watts. This value was based on a maximum cumulative host assembly(s) burnup of 30,000 Mwd/Mtu.

If the fuel assembly to be stored contains an insert, these heat loads are added to the fuel assembly heat load to determine a combined heat load for comparison to the 800 watt criterion.

A3.3.3 PROTECTION BY EQUIPMENT AND INSTRUMENTATION SELECTION A3.3.3.1 EQUIPMENT The design criteria for the TN-40HT casks are described in Section A3.4 and summarized in Table A3.4-1.

A3.3.3.2 INSTRUMENTATION Due to the totally passive and inherently safe nature of the storage system, important to safety instrumentation is not necessary. Instrumentation to monitor cask pressure is furnished. Appropriate capabilities to check and recalibrate these monitors are also provided. The pressure monitoring system is further described in Section A3.3.2.1.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-38 A3.3.4 NUCLEAR CRITICALITY SAFETY A3.3.4.1 CONTROL METHODS FOR PREVENTION OF CRITICALITY The design criterion for criticality is that an upper subcritical limit (USL) of 0.95 minus benchmarking bias and modeling bias will be maintained for all postulated arrangements of fuel within the cask. The fuel assemblies are assumed to stay within their basket compartment based on the cask and basket geometry.

The control methods used to prevent criticality are:

Incorporation of neutron absorbing material (boron) in the basket material.

2. Loading of the irradiated fuel assemblies in the fuel pool containing at least 2450 ppm boron.

3. Prevention of fresh water entering the loaded cask.

The basket has been designed to assure an ample margin of safety against criticality under the conditions of fresh fuel in a cask flooded with borated pool water. The method of criticality control is in keeping with the requirements of 10 CFR72. 124.

The TN-40HT cask system's criticality safety is ensured by fixed neutron absorbers in the basket, soluble boron in the pool. The cask basket uses a Borated-Aluminum alloy, Aluminum/B4C metal matrix composite or Boral as the fixed neutron poison material.

The collective term B-Al refers to the Borated-Aluminum alloy and Aluminum/B4C metal matrix composite materials where the analysis uses a 90% credit for the neutron poison.

A credit of 75% is taken for the presence of neutron poison for Boral plates.

A single fixed poison loading is utilized in the TN-40HT basket. Table A3.3-17 lists the minimum B10 poison loading required for the various poison materials and the corresponding poison content modeled in the analyses.

The minimum soluble boron concentration in the pool credited in the analysis is 2450 ppm.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-39 A3.3.4.1.1 DISCUSSION AND RESULTS Figure A3.3-28 shows the cross section of the TN-40HT cask. The TN-40HT cask stainless steel basket consists of an egg-crate plate design. The fuel assemblies are housed in 40 stainless steel fuel compartment tubes. The basket structure, including the fuel compartment tubes, is held together with stainless steel insert bars (note that the input files in Appendix A3A refer to these bars as plates) and the poison and aluminum plates that form the egg-crate structure. The fuel compartment tube structure is connected to the perimeter rail assemblies forming the cylindrical outer geometry required to be housed within the cask. The poison/aluminum plates are located between the fuel compartment tubes.

The analysis presented herein is performed for a TN-40HT Cask during normal, off-normal and accident loading conditions. This analysis also bounds all conditions of storage, including loading and transfer. The cask consists of an inner steel shell, a steel gamma shield shell and a hydrogenous neutron shield.

Table A3.3-1 8 lists the fuel assemblies considered as authorized contents of the TN-40HT cask system. The criticality analysis begins by determining the most reactive assembly type for the Westinghouse 14x14 fuel assembly (WE 14x14) class identified in Table A3.3-1 8. Then the most reactive configuration for the basket (including rail configuration) and fuel assembly position is determined. Next, criticality calculations are performed using the maximum allowable initial fuel enrichment for the TN-40HT cask shown in Table A3.3-1 7. These calculations determine keff with the CSAS25 control module of SCALE-4.4 (Reference 14) including all uncertainties to assure criticality safety under all credible conditions.

The Control Components (CCs) are also authorized for storage in the TN-40HT casks.

The authorized CCs are Burnable Poison Rod Assemblies (BPRAs) and Thimble Plug Devices (TPDs).

The results of the evaluation demonstrate that the maximum expected keff, including all applicable biases and uncertainties, is less than the Upper Subcritical Limit (USL) determined from a statistical analysis of benchmark criticality experiments. The statistical analysis procedure includes a confidence band with an administrative safety margin of 0.05.

A3.3.4.1.2 PACKAGE FUEL LOADING The TN-40HT Cask is capable of transferring and storing a maximum 40 intact WE 14x14 class PWR fuel assemblies. The reactivity of a cask loaded with less than 40 PWR fuel assemblies is lower than that calculated here since the more absorbing borated water replaces the fuel in the empty locations. Reconstituted fuel assemblies, where the fuel pins are replaced by lower enriched fuel pins or non-fuel pins that displace the same amount of borated water, would also lower the reactivity of the cask.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-40 Table A3.3-1 9 lists the fuel parameters for the PWR fuel assemblies considered in this evaluation. Reload fuel from other manufacturers with the same parameters are also considered as authorized contents.

For the WE 14x14 class assemblies, CCs are also included as authorized contents.

The only change to the package fuel loading to evaluate the addition of these CCs is replacing the borated water in the water holes conservatively with 11 B4C. Since these CCs displace borated moderator in the assembly guide tubes, an evaluation is performed to determine the potential impact of storage of CCs that extend into the active fuel region on the system reactivity. For CCs such as BPRAs no credit is taken for the cladding and absorbers; rather the CCs are modeled as 11 B4C in the entire tube of the respective design. Thus, the highly borated moderator in the tube is conservatively modeled as 11 B4C. The inclusion of more Boron-11 and carbon enhances neutron scattering causing the neutron population in the fuel assembly to be slightly increased which increases reactivity. Therefore, these calculations bound any CC design that is compatible with the WE 14x14 class assemblies. CCs that do not extend into the active fuel region of the assembly (e.g., TPD) do not have any effect on the reactivity of the system as evaluated because only the active fuel region is modeled in this evaluation with periodic boundary conditions making the model infinite in the axial direction. The dimensions of the fuel assemblies reported in Table A3.3-19 remain unchanged for the CC cases. The models that include CCs only differ in that the region inside the guide tubes are modeled as 11 B4C instead of moderator.

A3.3.4.1 .3 MODEL SPECIFICATION The following subsections describe the physical models and materials of the TN-40HT cask used for input to the CSAS25 module of SCALE-4.4 (Reference 14) to perform the criticality evaluation. The reactivity of cask under storage conditions is bounded by the analysis with zero (or near zero) internal moderator density case during loading and transfer.

A3.3.4.1.

3.1 DESCRIPTION

OF CALCULATIONAL MODEL The TN-40HT cask is explicitly modeled using the appropriate geometry options in KENO V.a of the CSAS25 module in SCALE-4.4. Several models are developed to evaluate the fabrication tolerances of the basket/cask, fuel assembly locations, fuel assembly design and storage of fuel with CCs like BPRAs and TPDs.

The criticality evaluation is performed using an egg-crate section length of 14.49 inches in the basket. The actual egg-crate length is 15.0 inches in the active fuel region of the assembly. This represents more reactive design than the actual basket because of the shorter egg-crate section length. Utilizing a shorter section length in the calculational model ensures that the model is conservative since the amount of poison per unit length is minimized. The key basket dimensions utilized in the calculation are shown in Table A3.3-20.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-41 The fixed poison modeled in the calculation is based on a poison plate thickness of 0.075 inches for initial sensitivity calculations and 0.125 inches for the final design basis calculations. The important parameter is the minimum B-1 0 areal density; therefore the modeled thickness of the poison plate does not affect the results of the calculation.

The basic calculational KENO model, as discussed above, is a 14.49-inch axial section and full-radial cross section of the basket within the cask with periodic boundary conditions at the axial boundaries (top and bottom) and reflective boundary conditions at the radial boundaries (sides). This axial section essentially models one building block of the egg crate basket structure. Periodic boundary conditions ensure that the resulting KENO model is essentially infinite in the axial direction. The model does not explicitly include the solid neutron shield (polyester resin); however the infinite array of casks without the neutron shield does contain unborated water between the casks. For the purpose of storage, the TN-40HT cask configuration is not expected to encounter any regions containing unborated water once the fuel assemblies are loaded. Therefore, this hypothetical configuration that models an infinite array of casks in close reflection is conservative.

The fuel assemblies within the basket are modeled as arrays of fuel pins and guide/instrument tubes. Spacer grids and sub-components like oversleeves (when present) are not modeled since their effect on reactivity is insignificant. The fuel compartment tubes surround each fuel assembly that is in-turn bounded by the basket plates consisting of 0.4375" aluminum/poison plates. These plates are arranged to represent an egg-crate design with the 0.3625"- Aluminum and a 0.075"-poison plate.

The thermal expansion and egg-crate slot gaps are not modeled assuming plate continuity, thus replacing the more absorbing borated water (internal moderator) with basket material (aluminum/poison). KENO model plots in 2D for the various views of the basket compartment are shown in Figure A3.3-29 through Figure A3.3-35.

There are a total of 13 poison plates in the TN-40HT basket. They are located at all the faces where at least five fuel assemblies are lined up. Thus, all the interior 30 fuel assemblies are surrounded by poison plates on all four faces and the outer 10 fuel assemblies do not have poison plates on the radially outward face. The fuel assembly and poison plate positions (and the aluminum plate positions) in the KENO model of the basket is shown in Figure A3.3-36. Even though the poison and aluminum plates have been shown as discrete plates around the fuel compartment, they are all continuous running from one end of the basket to the other.

The basket structure is connected to the cask shell by perimeter rail assemblies. The rail material is a combination of aluminum and SS304. The rails provide a structural function as well as provide a heat conduction path from the basket to the cask shell.

Due to limitations in the geometry options available in KENO, it is not possible to exactly model the rails. However, bounding evaluations are performed to determine the effect of rail material / geometry modeling on the reactivity of the system.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-42 A list of all the geometry units used in the basic KENO model is shown in Table A3.3-21. Figure A3.3-37 shows the various radial cylinders utilized in the KENO model surrounding the fuel assemblies. Basically, this shows the cask details.

The first model developed uses nominal dimensions for the fuel compartments, fuel compartment thickness, poison plate thickness and the fuel assemblies centered in the fuel compartment. The rails are modeled simply using horizontal and vertical aluminum plates around the periphery of the basket. The cavity between the fuel compartments and the cask inner shell is modeled with internal moderator.

This basic KENO model is used to determine the most reactive fuel assembly design, the most reactive assembly-to-assembly pitch, and to determine the most reactive cask configuration accounting for manufacturing tolerances. The second model is of the most reactive configuration identified above. This model is used to determine the maximum allowable initial enrichment for the TN-40HT cask. Plots of the various KENO models utilized in these calculations are shown in Figure A3.3-38 through Figure A3.3-41.

A3.3.4.1.3.2 PACKAGE REGIONAL DENSITIES The Oak Ridge National Laboratory (ORNL) SCALE code package (Reference 14) contains a standard material data library for common elements, compounds, and mixtures. All the materials used for the TN-40HT cask analysis are available in this data library. Table A3.3-22 provides a complete list of all the relevant materials used for the criticality evaluation. The material density for the B1 0 in the poison plates includes a 10% reduction for B-Al poison and a 25% reduction for Boral poison.

A3.3.4.1 .4 CRITICALITY CALCULATIONS This section describes the analysis methodology utilized for the criticality analysis. The analyses are performed with the CSAS25 module of the SCALE system. A series of calculations are performed to determine the relative reactivity of the various fuel assembly designs to determine the most reactive assembly type without CCs. The most reactive intact fuel design, as demonstrated by the analyses, is the WE 14x14 Standard assembly. The most reactive credible configuration is an infinite array of flooded casks, each containing 40 fuel assemblies, with minimum fuel compartment tube ID, maximum fuel compartment tube thickness, minimum stainless steel plate thickness, and minimum assembly-to-assembly pitch.

Finally, using the most reactive credible configuration determined above, keff is determined for the maximum initial enrichment for the WE 14x14 assembly class at a soluble boron concentration of 2450 ppm.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-43 A3.3.4.1.4.1 CALCULATIONAL METHOD A3.3.4.1.4.1.1 COMPUTER CODES The CSAS25 control module of SCALE-4.4 (Reference 14) is used to calculate the effective multiplication factor ( k eff) of the fuel in the TN-40HT Cask. The CSAS25 control module allows simplified data input to the functional modules BONAMI-S, NITAWL-II, and KENO V.a. These modules process the required cross sections and calculate the keff of the system. BONAMI-S performs resonance self-shielding calculations for nuclides that have Bondarenko data associated with their cross sections. NITAWL-II applies a Nordheim resonance self-shielding correction to nuclides having resonance parameters. Finally, KENO V.a calculates the keff of a three-dimensional system. A sufficiently large number of neutron histories are run so that the standard deviation is below 0.0015 for all calculations.

A3.3.4.1 .4.1.2 PHYSICAL AND NUCLEAR DATA The physical and nuclear data required for the criticality analysis include the fuel assembly data and cross-section data as described below.

Table A3.3-1 9 provides the pertinent data for criticality analysis for each fuel assembly evaluated in the TN-40HT cask. The criticality analysis used the 44-group cross-section library built into the SCALE system. ORNL used ENDF/B-V data to develop this broad-group library specifically for criticality analysis of a wide variety of thermal systems.

A3.3.4.1.4.1.3 BASES AND ASSUMPTIONS The analytical results reported in Section A4 demonstrate that the cask containment boundary, basket structure and fuel cladding do not experience any significant distortion under hypothetical accident conditions. Therefore, for both normal and hypothetical accident conditions the cask geometry is identical except for the neutron shield and neutron shield jacket (outer skin). As discussed above, the neutron shield and neutron shield jacket (outer skin) are removed and the interstitial space modeled as water.

The TN-40HT cask is modeled with KENO V.a using the available geometry input. This option allows a model to be constructed that uses regular geometric shapes to define the material boundaries. The following conservative assumptions are also incorporated into the criticality calculations for intact fuel:

1. No credit is taken for the presence of burnable poisons such as Gadolinia, Erbia or any other absorber in the fuel.

2. CCs that extend into the active fuel region, such as BPRAs are conservatively assumed to exhibit neutronic properties of 11 B 4 C.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-44

3. Unirradiated fuel - no credit taken for fissile depletion due to burnup or fission product poisoning.

4. For intact fuel, the lattice average fuel enrichment is modeled as uniform everywhere throughout the assembly. Natural Uranium blankets and axial or radial enrichment zones are modeled as enriched uranium at the lattice average enrichment.

5. All fuel rods are filled with full density fresh water in the pellet/cladding gap.

6. Only a 14.49-inch section of the basket (actual is 15.0-inches) with fuel assemblies is explicitly modeled with periodic axial boundary conditions, therefore the model is effectively infinitely long and the actual poison height for each section is conservatively modeled about 0.5 inches shorter.

7. The neutron shield material is modeled using water and is not expected to result in any significant change in the system reactivity since it is located in a relatively unimportant region for criticality.

8. Only 90% credit is taken for the B1 0 in the B-Al poison plates and 75% credit for Boral in the KENO models.

9. The fuel rods are modeled assuming a stack density of 96.5% theoretical density with no allowance for dishing or chamfer. This assumption conservatively increases the total fuel content in the model.

1 0. All calculations are carried out at a temperature of 20 °C (293 K).

11 . All basket stainless steel materials are modeled as SS304. The cask steel materials are modeled as SS304 and carbon-steel. The small differences in the composition of the various stainless steels have no effect on results of the calculation.

12. All zirconium based materials in the fuel (including ZIRLO) are modeled as Zircalloy-4. The small differences in the composition of the various clad / guide tube materials have no effect on the results of the calculation.

13. The thermal expansion and egg-crate gaps are replaced with the basket material wherever present. This modeling assumption results in replacing the soluble boron moderator in the gap regions with basket material (aluminum/poison),

thereby, decreasing the neutron absorption around the fuel.

14. The transition rails between the basket and the cask shell are modeled as solid aluminum with cuboid holes of borated water in the aluminum.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-45 A3.3.4.1.4.1.4 DETERMINATION OF K EFF The Monte Carlo calculations performed with CSAS25 (KENO V.a) use a flat neutron starting distribution. The total number of histories traced for each calculation is approximately 800,000. This number of histories is sufficient to achieve source convergence and produce standard deviations of less than 0.0015 in keff. The maximum keff for the calculation is determined with the following formula:

k eff = k KENO + 2cJ KENO A3.3.4.1.4.2 FUEL LOADING OPTIMIZATION A. Determination of the Most Reactive Fuel Type All fuel assemblies listed in Table A3.3-1 9 are evaluated to determine the most reactive fuel assembly type with initial enrichments of 4.50 wt. % U-235. The fuel types are analyzed with fresh water in the fuel pellet cladding annulus, a soluble boron concentration of 2400 ppm and a fixed borated aluminum poison loading of 18.7 mg 2

B1 0/cm . The parameters utilized for these sensitivity evaluations are not critical since it is intended to determine the most reactive configuration that estimates a relative reactivity and not absolute reactivity of the system. Nominal basket dimensions are utilized in the KENO model.

These evaluations are carried out for two fuel assembly positions within the fuel compartment - centered and inward. The centered position is when the fuel assemblies are centered within the fuel compartment while the inward position is when the fuel assemblies are positions closest relative to the center of the basket. The inward position results in the fuel assemblies being clustered closer together, thereby resulting in an increase in the reactivity. The internal moderator density (IMD) is varied from 80% to 100% of full density to determine any sensitivity of the fuel assembly design reactivity to moderator density.

In all other respects, the model is the same as the basic model described above. A simple representation of the peripheral rails as shown in Figure A3.3-38 is utilized for this evaluation.

The Cask model for this evaluation differs from the actual design in the following ways:

• The neutron shield of the cask is conservatively replaced with water between the casks.

• The stainless steel and aluminum basket rails, which provide support to the fuel compartment tube grid, are modeled as aluminum plates located in the periphery.

• The egg-crate section length is modeled as 14.49" high (12.67" basket section

+ 1 .75" steel insert bar + 0.07" gap). The actual design for the TN-40HT has an egg-crate section length of 1 5.0" (13.18" basket section + 1 .75" steel insert bar

+ 0.07" gap).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-46 The results of this evaluation are provided in Table A3.3-23. The most reactive design is the Westinghouse 14x14 Standard fuel assembly. The results also indicate that the inward position of the fuel assemblies results in a more reactive configuration and that the relative reactivity ranking of the fuel assemblies remains unchanged with IMD variations. The case corresponding to the highest keff in this evaluation is highlighted in Table A3.3-23.

B. Determination of the Most Reactive Configuration The fuel loading configuration of the cask affects the reactivity of the package. Several series of analyses determined the most reactive configuration for the TN-40HT cask.

For this analysis, the most reactive fuel type is used to determine the most reactive configuration. The cask is modeled with the WESTD assembly, over a 14.49-inch axial section with periodic axial boundary conditions and reflective radial boundary conditions. This represents and infinite array in the x-y direction of the cask that is infinite in length, which is conservative for criticality analysis. The starting model is identical to the model used above.

The cask model for this evaluation differs from the actual design in the following ways:

• The neutron shield of the cask is conservatively replaced with water between the casks.

• The stainless steel and aluminum basket rails, which provide support to the fuel compartment tube grid, are modeled as aluminum plates located in the periphery.

Further, an evaluation is performed to determine an acceptable and conservative representation of the rails in the basket periphery.

• The egg-crate section length is modeled as 14.49" high (12.67" basket section

+ 1.75" steel insert bar + 0.07" gap). The actual design for the TN-40HT has an egg-crate section length of 15.0" (13.18" basket section + 1.75" steel insert bar

+ 0.07" gap).

Each evaluation is performed with the fuel assemblies in the inward position at various IMD values to determine the optimum moderator density where the reactivity is maximized.

The first set of analyses evaluates the effect of fuel compartment tube internal width on the system reactivity. The model starts with the nominal poison plate thickness modeled as above. For this evaluation the fuel compartment tube internal width is varied from 8.00 to 8.10 inches square. The results of the evaluation are given in Table A3.3-24. The results show that the most reactive configuration is with the minimum fuel compartment tube size. The balance of this evaluation uses the minimum fuel compartment tube width because it represents the most reactive configuration. An example input file for the most reactive fuel evaluation is included in Appendix A3A.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-47 The second set of analyses evaluates the effect of fuel compartment tube and the stainless steel tie bar thicknesses on reactivity. The model starts with the most reactive configuration basket model determined above. The compartment tube thickness is varied from 0.1775" to 0.2325". The stainless steel tie bar thickness is varied from 0.4275" to 0.4925". Varying the thickness of the tie bar also serves the purpose of varying the aluminum plate thickness as the thickness of the poison plate is maintained constant. The results in Table A3.3-25 show that the system reactivity is not very sensitive to the compartment and stainless steel bar thicknesses.

However, in order to obtain a design basis configuration for criticality analyses, a few of the more reactive combinations are evaluated for variations with other parameters.

Therefore, the next sets of analyses are performed with bounding configurations that are highlighted in Table A3.3-25. Note that there are four bounding configurations and they are statistically similar.

The third set of analyses evaluates the effect of poison plate thickness and modeling on the system reactivity. Three of the bounding configurations identified above are utilized in this evaluation. For this evaluation Boral R is utilized as the poison material with a core thickness of 0.050" and a total (core + clad) thickness of 0.075". The effect of poison plate thickness (and hence the absorber thickness) variation on the system reactivity is shown to be statistically insignificant based on the results in Table A3.3-26.

These results also indicate that poison plates of higher thicknesses can be used as long as the minimum absorber loading is maintained. These results also demonstrate that there is no effect on the reactivity due to the aluminum panels when Boral poison is used. These results further indicate that effect of a change in the thickness of the egg-crate plates (poison and aluminum) is statistically insignificant provided the total thickness remains constant.

The fourth set of analyses evaluates the effect of rail structure modeling on the system reactivity. Due to the limitations in the geometry capabilities of the CSAS25 code, it is not possible to exactly model the rail structure. However, due to relatively low importance of the rail structure modeling to the criticality of the system, an exact model is not essential. In addition to the simplistic rail model utilized so far in these evaluations, three additional variations of the rails are also evaluated. The first variation is based on a modeling the rails as internal moderator ( rail1 option). The second variation is based on rail1 option with cuboids of aluminum placed within the periphery using hole modeling ( rail2 option). The third variation is different from the rail2 option where the periphery is modeled with solid aluminum with cuboids of internal moderator

( rail3 option where the aluminum and internal moderator from rail2 are interchanged).

The modeling of the cuboids as holes at the periphery ( rail2 and rail3 options) is shown in Figure A3.3-40.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-48 Note that the models - rail1, rail2 and rail3 represent a reduction in the volume fraction of internal moderator in the basket periphery. The results of the evaluation are given in Table A3.3-27. These results indicate that the rail3 option results in a bounding configuration. The results also indicate that the reactivity of the basket increases with a reduction in the volume fraction of the internal moderator at the basket periphery. It is clear based on a comparison of the rail configurations from Figure A3.3-28 and Figure A3.3-40 that the rail3 option conservatively models a lower volume fraction of internal moderator. Therefore, modeling the basket periphery with the rail3 option is appropriate and conservative. All the four bounding configurations identified above are utilized in this evaluation. These calculations also result in the determination of the most reactive configuration for the four bounding configurations evaluated.

Based on these evaluations the most reactive cask configuration is for:

• Fuel assemblies pushed toward the center of the basket (inward arrangement),

• Minimum fuel compartment tube internal width,

• Maximum fuel compartment tube wall thickness,

• Nominal poison plate thickness,

• Minimum stainless steel bar thickness and

• Basket periphery modeled using the rail3 option.

C. Maximum Initial Enrichment for the TN-40HT Cask The analysis performed in this section is performed using the most reactive configuration as determined in Section B above. The internal moderator density is varied to determine the peak reactivity for the specific configuration. The maximum initial enrichment (5.00 wt. % U-235) is also shown in Table A3.3-17.

The cask model for this evaluation differs from the actual design in the following ways:

• The neutron shield of the cask is conservatively replaced with water between the casks.

• The stainless steel and aluminum basket rails, which provide support to the fuel compartment tube grid, are modeled conservatively using the Rail3 option.

• The worst case material conditions, as determined in the previous Section above, are modeled,

• The egg-crate section length is modeled as 14.49" high (12.67" basket section

+ 1.75" steel insert bar + 0.07" gap). The actual design for the TN-40HT has an egg-crate section length of 1 5.0" (13.18" basket section + 1 .75" steel insert bar

+ 0.07" gap).

2 A fixed poison loading of 33.7 mg B-1 0/cm is utilized in the criticality calculations as described in Table A3.3-1 7. The soluble boron concentration utilized for these calculations is 2450 ppm. A maximum initial enrichment of 5.0 wt. % U-235 is utilized in these calculations. An example input file is included in Appendix A3A.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-49 The most reactive WE 14x14 class assembly is the WE 14x14 Standard fuel assembly as demonstrated in Table A3.3-23. The results for the WE 14x14 class assembly calculations without CCs are listed in Table A3.3-28.

The results for the WE 14x14 class assembly calculations with CCs are listed in Table A3.3-29. The results demonstrate that the no reduction in the initial enrichment is required due to the presence of CCs.

The maximum calculated keff corresponds to the configuration with an initial enrichment of 5.0 wt. % U235 with 2450 ppm borated water with CCs. The maximum calculated k eff is 0.9357 +/- 0.0008 or 0.9373 which is below the USL (0.9419) calculated in Section A3.3.4.3.2. The maximum calculated dry keff (normal condition for storage) is 0.5787 +/- 0.0004 or 0.5795.

A3.3.4.2 ERROR CONTINGENCY CRITERIA Provision for error contingency is built into the criterion used in Section A3.3.4.1 above.

The criterion, used in conjunction with the KENO-Va and NITAWL codes, is common practice for licensing submittals. Because conservative assumptions are made in modeling, it is not necessary to introduce additional contingency for error.

A3.3.4.3 VERIFICATION ANALYSIS - BENCHMARKING The computer codes described in Section A3.3.4.1.4.1.1 were used to benchmark 121 experiments. The results of these benchmarks were used to determine the Upper Subcritical Limit (USL-1).

The benchmark problems used to perform this verification are representative of benchmark arrays of commercial light water reactor (LWR) fuels with the following characteristics:

(1) water moderation, (2) boron neutron absorbers, (3) unirradiated light water reactor type fuel (no fission products or burnup credit) near room temperature (vs. reactor operating temperature),

(4) close reflection, and (5) uranium oxide.

The 121 uranium oxide experiments were chosen to model a wide range of uranium enrichments, fuel pin pitches, assembly separation, soluble boron concentration and control elements in order to test the codes ability to accurately calculate keff. These experiments are discussed in detail in NUREG/CR-6361 (Reference 18).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-50 A3.3.4.3.1 BENCHMARK EXPERIMENTS AND APPLICABILITY A summary of all of the pertinent parameters for each experiment is included in Table A3.3-30 along with the results of each run. The best correlation is observed for fuel assembly separation distance with a correlation of 0.66. All other parameters show much lower correlation ratios indicating no real correlation. All parameters were evaluated for trends and to determine the most conservative USL.

The USL is calculated in accordance to NUREG/CR-6361 (Reference 18). USL Method 1 (USL-1) applies a statistical calculation of the bias and its uncertainty plus an administrative margin (0.05) to the linear fit of results of the experimental benchmark data. The basis for the administrative margin is from NUREG/CR-5661 (Reference 20).

Results from the USL evaluation are presented in Table A3.3-31.

The criticality evaluation used the same cross section set, fuel materials and similar material/geometry options that were used in the 121 benchmark calculations as shown in Table A3.3-30.

The modeling techniques and the applicable parameters listed in Table A3.3-32 for the actual criticality evaluations fall within the range of those addressed by the benchmarks in Table A3.3-30.

A3.3.4.3.2 RESULTS OF THE BENCHMARK CALCULATIONS The results from the comparisons of physical parameters of each of the fuel assembly types to the applicable USL value are presented in Table A3.3-32. The minimum value of the USL is determined to be 0.9419 based on comparisons to the limiting assembly parameters as shown in Table A3.3-32.

A3.3.5 RADIOLOGICAL PROTECTION Provisions for radiological protection by confinement barriers and systems are described in Section A3.3.2.1.

A3.3.5.1 ACCESS CONTROL The ISFSI does not require the continuous presence of operators or maintenance personnel. In addition, it is located within a fenced-in area shared only with the Equipment Storage Building and Security Building which will be used for storage of cask handling and security related equipment and will not be continuously manned. Access to the fenced-in area is limited to personnel needed during operations at the ISFSI.

Activities will include periodic inspections of these facilities, emplacement of storage casks, and security checks. These activities will be defined and controlled by the Radiation Protection and Security procedures manuals covering the ISFSI.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-51 A3.3.5.2 SHIELDING The storage casks provide sufficient radiation shielding to allow handling of the loaded casks with as low as reasonably achievable (ALARA) doses to the operators and to comply with the radiation limits in 10 CFR72. For specific dose estimates, see Section A7.

A3.3.5.3 RADIOLOGICAL ALARM SYSTEMS There are no credible events which result in releases of radioactive products or unacceptable increases in direct radiation. In addition, the releases postulated as the result of the hypothetical accidents described in Section A8 are of a very small magnitude. Therefore, radiological alarm systems are not necessary. However, as described in Section A3.3.2.1, not important to safety pressure monitors are provided.

Procedures to be followed when these alarms are activated will be specified in the ISFSI operating procedures.

A3.3.6 FIRE AND EXPLOSION PROTECTION No hydrocarbon fuel of any sort will be stored in the ISFSI. The quantity of fuel carried in the tow vehicle will be limited so that only a small fire of short duration would be possible. There are no other significant combustible sources within the ISFSI security fence. Due to the large thermal mass of the casks any minor fires in the vicinity of the ISFSI would raise the cask temperature by only a few degrees and are not expected to affect cask integrity.

As indicated in Section 2.2, overpressures of 2.25 psi can be conservatively postulated to occur at the ISFSI as a result of accidents involving explosive materials which are stored or transported near the site. This impact is less than that postulated to result from the tornado wind loading and missile impact analysis, as described in Section A3.2.1, and is well within the design basis of the cask.

A3.3.7 MATERIAL HANDLING AND STORAGE A3.3.7.1 SPENT FUEL HANDLING AND STORAGE The handling of spent fuel within the Prairie Island Nuclear Generating Plant will be conducted in accordance with existing fuel handling procedures. Only fuel that is not a DAMAGED FUEL ASSEMBLY will be considered for storage in the TN-40HT casks.

In the TN-40HT casks, a DAMAGED FUEL ASSEMBLY is a Spent Nuclear Fuel Assembly that:

a. has visible deformation of the rods in the spent nuclear fuel assembly. Note:

This is not referring to the uniform bowing that occurs in the reactor. This refers to bowing that significantly opens up the lattice spacing;

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.3-52

b. has individual fuel rods missing from the assembly. Note: The assembly is not a DAMAGED FUEL ASSEMBLY if a dummy rod that displaces a volume equal to, or greater than, the original fuel rod, is placed in the empty rod location;

c. has missing, displaced, or damaged structural components such that radiological and/or criticality safety is adversely affected (e.g., significantly changed rod pitch);

d. has missing, displaced, or damaged structural components such that the assembly cannot be handled by normal means (i.e., crane and grapple);

e. has reactor operating records (or other records) indicating that the spent nuclear fuel assembly contains cladding breaches; or

f. is no longer in the form of an intact fuel bundle (e.g., consists of, or contains, debris such as loose fuel pellets or rod segments).

Handling of the sealed casks outside of the Auxiliary Building in the process of emplacing them at the ISFSI will be done according to procedures that ensure that their safety functions and the power station capability for safe shutdown are not impaired.

These operations for the TN-40HT casks are the same as for a TN-40 cask and are described in Section 5.4.

A3.3.7.2 RADIOACTIVE WASTE TREATMENT The ISFSI will not generate radioactive waste. However, cask loading and decontamination operations, while in the Auxiliary Building, may generate small amounts of waste. This waste is disposed of in accordance with the radioactive waste handling procedures described in Section A6, and is part of the 10 CFR50 licensed activities. Waste storage facilities are neither required nor provided for the ISFSI.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.4-1 A3.4

SUMMARY

OF STORAGE CASK DESIGN CRITERIA The principal design criteria for the TN-40HT cask are presented in Table A3.4-1. The TN-40HT cask is designed to store 40, 14 x 14 PWR spent fuel assemblies with or without fuel inserts. The maximum allowable initial enrichment is 5.0 wt% U-235. The maximum bundle average burnup, maximum decay heat, and minimum cooling time for the fuel assembly are 60 GWd/MTU, 0.80 kW/assembly, and 12 years respectively.

The maximum total heat generation rate of the stored fuel is limited to 32 kW in order to keep the maximum fuel cladding temperature below the limit necessary to ensure cladding integrity for 25 years storage. The fuel cladding integrity is assured by the cask and basket design which limits fuel cladding temperature and maintains a non-oxidizing environment in the cask cavity.

The containment vessel (body and lid) is designed and fabricated to the maximum practicable extent as a Class I component in accordance with the rules of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB, Article NB-3200. The alternatives to ASME code requirements are documented in Section A3.5. The cask design, fabrication and testing are covered by a Quality Assurance Program which conforms to the criteria in Subpart G of 10 CFR72.

The cask is designed to maintain a subcritical configuration during loading, handling, storage and accident conditions. Poison materials in the fuel basket are employed to maintain the upper subcritical limit of 0.95 minus benchmarking bias and modeling bias.

The TN-40HT basket is designed and fabricated to the maximum practicable extent in accordance with the rules of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NG, Article NG-3200 (Reference 2). The alternatives to ASME code requirements are documented in Section A3.5.

The TN-40HT cask is designed to withstand the effects of severe environmental conditions and natural phenomena such as earthquakes, tornadoes, lightning and floods. Section A8 describes the cask behavior under these accident conditions.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.4-2 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.5-1 A3.5 ASME CODE ALTERNATIVES The cask containment boundary is designed, fabricated and inspected in accordance with the ASME Code Subsection NB (Reference 2) to the maximum practical extent.

The gamma shielding, which is primarily for shielding, but also provides structural support to the containment boundary, was designed in accordance with Subsection NF of the code. Inspections of the gamma shielding are performed in accordance with ASME code Subsection NF as detailed in the SAR. Other cask components, such as the protective cover, outer shell and neutron shielding are not governed by the ASME Code. The Code Alternatives are described below.

Reference ASME Component Code/Section Code Requirement Alternatives, Justification & Compensatory Measures TN-40HT NB/NF/ Stamping and The TN-40HT cask is not stamped, nor is there a code Cask, NG-1 100 preparation of reports design specification or stress report generated. A design Basket NB/NF/ by the Certificate criteria document is generated in accordance with NG-2130 Holder, Use of ASME Transnuclears (TN) QA Program and the design and NB/NF/ Certificate Holders analysis is performed under TNs QA Program. The cask NG-4121 may also be fabricated by other than N-stamp holders and materials may be supplied by other than ASME Certificate holders.

TN-40HT NCA All Not compliant with NCA. TN Quality Assurance Cask, requirements, which are based on 10 CFR72 Subpart G, Basket are used in lieu of NCA-4000. Fabrication oversight is performed by TN personnel in lieu of an Authorized

____________ _______________ ___________________ Nuclear Inspector.

Pressure NB-6000 Hydrostatic Testing The containment vessel is hydrostatically tested in Test of the accordance with the requirements of the ASME B&PV Containment Code,Section III, Articles NB-6200 with the exception that Boundary some of the containment vessel may be installed in the shield shell during testing. The containment vessel is supported by the shield shell during all design and accident events.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.5-2 Reference ASME Component Code/Section Code Requirement Alternatives, Justification & Compensatory Measures Weld of NB-5231 Full penetration The joint may be welded after the containment shell is bottom inner corner welded joints shrink-fit into the shield shell. The geometry of the joint plate to the require the fusion does not allow for UT inspection. In this case, the joint will containment zone and the parent be examined by RT and either PT or MT methods in shell metal beneath the accordance with ASME subsection NB requirements. If attachment surface to the containment shell is welded complete before shrink be UTd after welding. fitting, UT examination per NB-5231 will be performed.

Containment NB-4213 The rolling process If the plates are made from less than three heats, each Shell Rolling used to form the inner heat will be tested to verify the impact properties.

Qualification vessel should be qualified to determine that the required impact properties of NB-2300 are met after straining by taking test specimens from three different heats.

Welds of the NB-4243 and Category C weld Certain welds are partial penetration welds. As an Bottom NB-5230 joints in vessels and alternative to the NDE requirements of NB-5230, for Shield to similar weld joints in Category C welds, all of these closure welds are multi-Shield Shell other components layer welds that are progressive PT examined.

and Shield shall be full Shell to penetration joints.

Shell Flange These welds shall be examined by UT or RT and either PT or

________ ___________ MT. ___________________________________

Containment NB-7000 Vessels are required No overpressure protection is provided. Function of Vessel to have overpressure containment vessel is to contain radioactive contents protection under normal and accident conditions. The containment vessel is designed to withstand maximum internal pressure considering 100% fuel rod failure and maximum accident temperatures.

Containment NB-8000 Requirements for The TN-40HT cask is to be marked and identified in Vessel, NG-8000 nameplates, accordance with 10 CFR 72 requirements. Code stamping Basket stamping and reports is not required. QA data package to be in accordance with per NCA-8000 TN approved QA program.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.5-3 Reference ASME Component Code/Section Code Requirement Alternatives, Justification & Compensatory Measures Weld of NB-4335 Impact testing of weld The lid shield plate is not in the component support path, Shield Plate NB-4620 and heat affected and has no pressure-retaining function; it is a non-to Lid Outer zone of lid to shield structural attachment, and NB jurisdiction does not apply Plate plate to the plate or weld. The weld must conform to NB-4430.

Post weld heat treatment Gamma NB-1 132 Attachments in the The gamma shield shell and trunnions are not fabricated Shielding NF-1 132 component support completely in accordance with Subsection NF. The shield and load path and not shells primary function is not structural. The weld of the Trunnion performing a bottom shield plate to the shield shell is subject to pressure retaining multilevel PT or MT to prevent complete loss of the bottom function shall conform shielding in an accident. Other shield shell weld (shield to Subsection NF. shell to the shell flange) failures would not lead to loss of shielding.

The trunnions and trunnion welds are designed to load factors much higher than those of subsection NF, the trunnion weld is subject to root and final PT or MT, and the

____________ _______________ ___________________ trunnions are tested to 1 .5 times design load.

Basket NG-2000 Use of ASME The basket neutron poison material is not considered in neutron Materials the structural analysis of the basket. The material provides poison criticality control and adds a heat transfer path. The material poison material is not a code material.

Basket NG-3352 Table NG 3352-1 lists The fusion welds between the stainless steel insert plates the permissible and the stainless fuel compartment tube are not included welded joints. in Table NG-3352-1. The required minimum tested capacity of the welded connection (at each side of the tube) shall be 35 kips (at room temperature). The capacity shall be demonstrated by qualification and production testing.

ASME Code Section IX does not provide tests for qualification of these types of welds. Therefore, these welds are qualified using Section IX to the degree applicable together with the testing described here.

The welds will be visually inspected to confirm that they are located over the insert plates, in lieu of the visual acceptance criteria of NG-5260 which are not appropriate for this type of weld.

A joint efficiency (quality) factor of 1.0 is utilized for the fuel compartment longitudinal seam welds. Table NG-3352-1 permits a joint efficiency (quality) factor of 0.5 to be used for full penetration weld examined by ASME Section V visual examination (VT). For the TN-40HT basket, the compartment seam weld is thin (0.188 thick) and the weld will be made in one pass. Both surfaces of weld (inside and outside) will be fully examined by VT and therefore a factor of 2 x 0.5=1.0 will be used in the analysis. This is justified as both surfaces of the single weld pass/layer will be fully examined, and the stainless steel material that

____________ _______________ ___________________ comprises the fuel compartment tubes is very ductile.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.5-4 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.6-1 A

3.6 REFERENCES

1 0CFR Part 72, Licensing Requirements For The Independent Storage Of Spent Nuclear Fuel, High-Level Radioactive Waste, And Reactor-Related Greater Than Class C Waste.

2. American Society of Mechanical Engineers, ASME Boiler And Pressure Vessel Code,Section III, Division 1 - Subsections NB, NF and NG, 2004 edition including 2006 addenda.

3. NUREG-0612 Control of Heavy Loads at Nuclear Power Plants, July 1980.

4. Standard Review Plan, NRC NUREG-0800, Rev. 2, July, 1981.

5. T. W. Singell, Wind Forces on Structures: Forces on Enclosed Structures, ASCE Structural Journal, July 1958.

6. Baumeister and Marks, Standard Handbook for Mechanical Engineers, Seventh Edition, McGraw-Hill Book Co.

7. 1 0CFR 100, Reactor Site Criteria.

8. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Revision 1, February, 1976.

9. American National Standards Institute, "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials," ANSI N14.6, New York, 1986.

10. Letter, R.O.Anderson (NSP) to U S Nuclear Regulatory Commission, dated March 27, 1995, "Response to Request for Additional Information Regarding NUREG-0612, 'Control of Heavy Loads'".

11. Letter, B.A.Wetzel (NRC) to R.O.Anderson (NSP), transmittal date June 12, 1995, "Safety Evaluation and Safety Assessment Related to Dry Cask Storage At Prairie Island".

12. Not Used.

13. Safety Evaluation 72-448, TN-40 Cask Weight/Storage Slab Design, May 14, 1996.

14. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers,"

NUREG/CR-0200, Rev. 6, September, 1998.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.6-2

15. Prairie Island Nuclear Generating Plant Updated Safety Analysis Report, Units 1 and 2, License Number DPR-42 and DPR-60.

16. Not used.

17. Warren C. Young, Richard G. Budynas, Roark Formulas for Stress and Strain, Seventh Edition, 2002, McGraw Hill.`

18. U.S. Nuclear Regulatory Commission, "Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages," NUREG/CR-6361 April, 1997, ORNL/TM-11936.

19. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.60, Rev. 1, Design Response Spectra for Seismic Design of Nuclear Plants, December, 1973.

20. U.S. Nuclear Regulatory Commission, Recommendations for Preparing the Criticality Safety Evaluation of Transportation Packages, NUREG/CR-5661, Published April 1997, ORNL/TM-1 1936.

21. Not used.

22. Not used.

23. Not used.

24. Formulas for Stress and Strain, Roark, 4th Edition.

25. USNRC, SFPO, Interim Staff Guidance - 11, Rev. 3, Cladding Considerations for the Transportation and Storage of Spent Fuel.

26. Department of Energy, Report No. DOE / RW0472, Rev. 2, Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages, 1998.

27. Rohsenow, W. M., Hartnett, J. P., Cho, Y. I., Handbook of Heat Transfer 3rd Fundamentals, Edition, 1998.

28. ASME Boiler and Pressure Vessel Code,Section II, Part D, Material Properties, 2004.

5th

29. Perry, R. H., Chilton, C. H., Chemical Engineers Handbook, Edition, 1973.

30. Zoldners, N. G., Thermal Properties of Concrete under Sustained Elevated Temperatures, ACI Publications, Paper SP 25-1, American Concrete Institute, Detroit, MI, 1970.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.6-3

31. Bentz, D. P., A Computer Model to Predict the Surface Temperature and Time-of-Wetness of Concrete Pavements and Bridge Decks, Report # NISTIR 6551, National Institute of Standards and Technology, 2000.

4th

32. Siegel, Robert, Howell, R. H., Thermal Radiation Heat Transfer, Edition, 2002.

33. AAR Brooks & Perkins, Advanced Structures Division, Standard Specification for Boral TM Composite Sheets.

34. Not used.

35. USNRC, SFPO, NUREG-1 536, Standard Review Plan for Dry Cask Storage Systems - Final Report, 1997.

36. ANSYS Computer Code and Users Manuals, Version 8.0.

37. TN Letter to NRC, Revision 1 to Transnuclear, Inc. (TN) Application for Amendment 1 to the NUHOMS HD System, Response to Request for Additional Information (Docket No. 72-1030; TAC No. L241 53), Enclosure 2, Response to RAI 4.1, TN Document No. E-27377, December 15, 2008.

38. USNRC, Code of Federal Regulations, Part 71, Packaging and Transportation of Radioactive Material, 2004.

39. Gregory, J. J., Mata, R., Keltner, N. R., Thermal Measurements in a Series of Long Pool Fires, SANDIA Report, SAND 85-0196, TTC-0659, 1987.

40. IAEA Safety Standards, Regulations for the Safe Transport of Radioactive Material, 1985.

41. USNRC, SFPO, NUREG/CR-0497, A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, MATPRO - Version II (Revison 2), EG&G Idaho, Inc., TREE-1280, 1981.

42 Letter, Ashok Thadani (NRC) to S. R. Tritch (Westinghouse), Acceptance for Referencing of Topical Report WCAP-12610 Vantage+ Fuel Assembly Reference Core Report (TAC No. 77258) July 1, 1991.

43. SANDIA Report, SAND90-2406, A Method for Determining the Spent Fuel Contribution to Transport Cask Containment Requirements, 1992.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3.6-4 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.1-1 PRAIRIE ISLAND FUEL ASSEMBLY DESIGN CHARACTERISTICS Exxon/AN F Exxon/ANF (ANP) Exxon/ANF Westinghouse (ANP) High (ANP) Westinghouse OFA (Including (14x14) Burnup TOPROD Standard Vantage+)

Fuel Designations Standard (14x14) (14x14) (14x14) (14x14)

Max Length (in) 161.3 161.3 161.3 161.3 161.3 Max Width (in) 7.763 7.763 7.763 7.763 7.763 Maximum No. of Fuel Rods 179 179 179 179 179 Nominal Fuel Rod OD (in) 0.4240 0.4260 0.4170 0.4220 0.4000 Clad Material Zr-4 Zr-4 Zr-4 Zr-4 Zr-4/ZIRLO Guide Tube # 16 16 16 16 16 Instrument Tube # 1 1 1 1 1 Maximum (1)

MTU/assembly 0.370 0.370 0.370 0.410 0.360 Note:

(1) The maximum MTU/assembly is calculated based on the theoretical density.

The calculated value is higher than the actual value.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.1-2 THERMAL, GAMMA AND NEUTRON SOURCES FOR THE DESIGN BASIS 14X14 WESTINGHOUSE STANDARD FUEL ASSEMBLY U 235 Bundle Average Initial Enrichment (% wt) 3.4 Burnup (MWD/MTU) 60,000 Cooling Time (years) 18 Decay Heat (kW/assembly) 0.845 (2)

(1)

Gamma Source (photon/sec/assembly) 3.37E+ 15 Neutron Source (neutrons/sec/assembly) 7.59E+08 Note:

(1) Includes fuel insert (BPRA/TPA) activation gamma source (2) 0.845 kW is used for source term calculation.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.1-3 FUEL QUALIFICATION CRITERIA Low Range 2.0 ^ Enrichment < 3.4 wt% U-235

^ 44 GWD/MTU Minimum cooling time = 12 years (verify decay heat ^ 800 watts with equation below)

High Range 3.4 ^ Enrichment ^ 5.0 wt% U-235

^ 60 GWD/MTU Minimum cooling time = 12 years (verify decay heat ^ 800 watts with equation below)

The Decay Heat (DH) in watts is expressed as:

F1 = A + B*X1 + C*X2 + D*X1 2 + E*X1*X2 + F*X2 2 H I DH = F 1 *Exp({G*[1 -(1 2/X3)]}*[(X3/X1 ) ])*[(X2/X1 ) ])

where, F1 Intermediate Function, the thermal source at 12 year cooling X1 Assembly Burn-up in GWD/MTU X2 Initial Enrichment in wt. % U-235 (max 5% wt)

X3 Cooling Time in Years (min 12 yrs)

A=18.76 B=11.27 C= 6.506 D= 0.163 E= -1.826 F= 6.617 G = -0.309 H= 0.431 I= -0.374

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-1 TN-40HT CASK COMPONENT WEIGHTS AND CENTER OF GRAVITY LOCATIONS CG Location Component Weight (lbs)

(in.)

Cask Shell 105,849 93 Lid 13,314 177.05 Bottom Assembly 15,579 4.375 Neutron Shield 22,062 89.25 Top Neutron Shield 1,638 184.0 Overpressure Tank 131 190.56 Protective cover 1,362 186.6 Upper Trunnion 667 155.25 Lower Trunnion 253 7.5 Basket 28,488 88.75 40 Fuel Assemblies 52,000 89.4 (2)

Misc 1,000 91.58 (1)

Total 242,343 91.58 (1) Actual weight(s) will vary. Fully loaded cask weight is 244,000 lbs max (Reference 13).

(2) Additional weight to account for small items not explicitly included in the weight calculation of the other components, e.g. head of bolts, drain tube, etc.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-2

SUMMARY

OF INTERNAL AND EXTERNAL PRESSURES ACTING ON THE TN-40HT CASK Individual Load Conditions Maximum Pressure, psig Internal Pressure: ______________________________

(a) Initial Cavity Pressurization 13.0 (b) With 10% Fuel & BPRA Failure 17.5 (c) With 100% Fuel & BPRA Failure See condition (d)

(d) In a Minor Fire (assumed 100% fuel &

BPRA failure) 74.3 (e) Beginning of Life Unloading 57.3 (f) Cask Burial (assumes 100% fuel &

BPRA failure) 94.9 3*

(g) Tornado (h) Selected Bounding Pressure 100 External Pressure _____________________________

(a) Flood 25 (b) Snow and Ice Loading 0.35 (c) Explosion 2.25 (d) Selected bounding Pressure 25

• This is due to a reduced external pressure.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-3

SUMMARY

OF LIFTING LOADS USED FOR UPPER TRUNNIONS (1)

Load at Cask CG Handling Condition Vertical Lifting- Cask Vertical ___________________________________

Yield Evaluation 1 .500x 106 lbs.

Ultimate Evaluation 2.500x 106lbs.

____________________________________ Load at each Trunnion (2)

Yield Evaluation 0.750 x 106 lbs.

Ultimate Evaluation 1.250x 106 lbs.

Notes:

(1) Based on a cask weight of 250,000 lbs.

(2) Load evenly divided between one pair of upper trunnions.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-4

SUMMARY

OF LOADS ACTING ON THE TN-40HT CASK DUE TO ENVIRONMENTAL AND NATURAL PHENOMENA Distributed Loads Lateral Loading:

(a) Wind (external force on cask body) 46,480 lb (2)

(b) Seismic (inertial force throughout system) 0.18W 45,000 lb.

(2)

Selected Bounding Load W x 1 G 250,000 lb.

(1)

Vertical Loading  :

(2)

(a) Seismic (inertial force throughout system) 0.12W 30,000 lb.

(2)

Selected Bounding Load W x 1 G 250,000 lb.

Notes:

(1) Does not include dead weight or lifting loads (2) A low weight is used for stability analysis. A high weight of the cask is used for stress analysis.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-5 TN-40HT CASK LOADING CONDITIONS Normal Assembly Loads (bolt preload and seal compression)

Pressure (internal and external)

Weight Lifting Loads Handling Wind Thermal variations (e.g., insolation, decay heat, ambient)

Man-Made (Accident)

Fuel cladding failure (due to loading or unloading error)

Minor Fire Explosion Natural Phenomena (Accident)

Earthquakes Tornadoes Cask Burial Flood Lightning

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-6 TN-40HT CASK DESIGN LOADS (NORMAL CONDITIONS)

App lied Load Loading Condition Internal Pressure Note (1)

External Pressure Note (2)

Distributed Loads Weight (Cask Body, Contents)

Snow Ice Wind (Tornado)

Lifting Attachment Loads Lifting Bolt Loads Preload for 100 psig and metal seal compression Notes:

(1) Cask design pressure is 22 psig but evaluated for 100 psig internal pressure which envelopes all internal pressure levels including fission gas release .

(2) Cask designed for 25 psig external pressure which envelopes all external pressure effects.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-7 LEVEL A SERVICE LOADS (NORMAL CONDITIONS)

App lied Load Loading Condition Internal Pressure Note (1)

External Pressure Note (2)

Distributed Loads Weight (Cask Body, Contents)

Snow Ice Wind (Tornado)

Lifting Attachment Loads Lifting Bolt Loads Preload for 100 psig and metal seal compression Thermal Effects Decay Heat Insolation Notes:

(1) Cask design pressure is 22 psig but evaluated for 100 psig internal pressure which envelopes all internal pressure levels including fission gas release.

(2) Cask designed for 25 psig external pressure which envelopes all external pressure effects.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-8 LEVEL D SERVICE LOADS (ACCIDENT CONDITIONS)

Load Cause Internal Pressure Note (1)

External Pressure Note (2)

Distributed Loads Weight (Cask Body, Contents)

Tornado Wind Flood Water Seismic Bolt Loads Preload for 100 psig, metal seal compression and drop impact 50 g Bottom Impact 18" Vertical Drop (Handling Accident)

Notes:

(1) Cask design pressure is 22 psig but evaluated for 100 psig internal pressure which envelopes all internal pressure levels including fission gas release.

(2) Cask designed for 25 psig external pressure which envelopes all external pressure effects including flood water level, cask burial and explosion.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-9 NORMAL CONDITION LOAD COMBINATIONS Individual Load Internal External Trunnion Combined Bolt 1g Pressure Pressure Thermal Thermal 3g Local Load Preload Fabrication Down 100 psig 25 psig (Hot) (Cold) Lifting Stress N1 X X X X ______ _____ ____ ____ _____

N2 X X X X _____ _____ ____ ______

N3 X X X X X N4 X X X X X ____ ______

(1)

N5 X X X X X (1)

N6 X X X X X (2)

N7 X X X (2)

N8 X X X Notes:

(1) Calculated cask global stresses due to 3g lifting load (2) Calculated cask local stresses at trunnion/shield shell due to 3g lifting load

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.2-10 ACCIDENT CONDITION LOAD COMBINATIONS Seismic, Internal External Tornado, or Individual Load Bolt 18" Bottom Pressure Pressure Flood 1 g-Lateral Combined Load Preload Fabrication End Drop 50 g 100 psig 25 psig + 2 g-Down A1 X X X _______ _______ ___________

A2 X X X X _______ ___________

A3 X X X X ___________

A4 X X X X A5 X X X X

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-1 AXIAL DECAY HEAT PROFILE

% of Core Height Corresponding Length Peaking Factor for Active Fuel of 144

__________________ (in) ________________

0.00 0.00 0 2.78 4.00 0.652 8.33 12.00 0.967 13.89 20.00 1.074 19.44 27.99 1.103 25.00 36.00 1.108 30.56 44.01 1.106 36.11 52.00 1.102 41.67 60.00 1.097 47.22 68.00 1.094 52.78 76.00 1.094 58.33 84.00 1.095 63.89 92.00 1.096 69.44 99.99 1.095 75.00 108.00 1.086 80.56 116.01 1.059 86.11 124.00 0.971 91.67 132.00 0.738 97.22 140.00 0.462 100.00 144.00 0

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-2 PEAKING FACTORS APPLIED IN THE FINITE ELEMENT MODEL Fuel z i (in) Local Average Average Area Region Peaking Height Peaking underneath No. Factor (in) Factor Profile

_________ ________ PF z,i PF m,i (A i )

0.000 0 1 1.320 0.215 0.660 0.107 0.1419 2 7.070 0.773 4.195 0.617 3.5501 3 14.500 1.000 10.785 0.940 6.9833 4 22.070 1.082 18.285 1.072 8.1167 5 29.500 1.104 25.785 1.093 8.1189 6 37.070 1.108 33.285 1.106 8.3712 7 52.070 1.102 44.570 1.105 16.573 8 67.070 1.094 59.570 1.098 16.472 9 82.070 1.095 74.570 1.095 16.418 10 97.070 1.095 89.570 1.095 16.426 11 102.750 1.092 99.910 1.094 6.2118 12 111.940 1.073 107.345 1.082 9.9464 13 117.690 1.040 114.815 1.067 6.1361 14 126.940 0.885 122.315 1.001 9.2634 15 134.500 0.652 130.720 0.799 6.0431 16 141.940 0.238 138.220 0.534 3.971 17 144.000 0.000 142.970 0.119 0.2449 SUM 143.0 Average 0.993

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-3 MAXIMUM TEMPERATURES FOR HOT NORMAL/OFF-NORMAL CONDITIONS Normal/Off-Normal Storage, 100 °F Ambient Component Maximum Temperature Temperature Limit

________________ (°F) (°F)

Fuel Cladding 680 752 Fuel Compartment 630 ---

Basket Rails 459 ---

Cask Inner Shell

• 305 ---

Shield Shell* 298 ---

Radial Resin 285 300 Cask Outer Shell 260 ---

Top Shield Plate 194 ---

Cask Lid 191 ---

Top Resin 191 300 Protective Cover 159 ---

Inner Bottom Plate 364 ---

Bottom Shield 264 ---

Lid Seal 184 536 Vent & Port Seal 185 536

• This value is the average temperature at the hottest cross section.

This value is the volumetric average temperature at the hottest cross section plus 18 °F to bound the effects of storage pad on the cask view factor. See Section A3.3.2.2.4 for discussion.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-4 AVERAGE TEMPERATURES AT THE HOTTEST CROSS SECTION FOR 100 °F STORAGE CONDITIONS Component Average Temperature

___________________________ (°F)

Basket 479 Aluminum Shims 358 Cask Inner Shell 305 Basket Support Bar 90-270 535 Basket Support Bar 0-180 560 Basket Aluminum Plate 90-270 536 Basket Aluminum Plate 0-180 572 Shield Shell 298

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-5 MAXIMUM TEMPERATURES FOR COLD NORMAL/OFF-NORMAL CONDITIONS Normal/Off-Normal Storage, -40 °F Ambient Component Maximum Temperature

_________________ (°F)

Fuel Cladding 582 Fuel Compartment 528 Basket Rails 352 Cask Inner Shell

• 188 Shield Shell* 181 Radial Resin 167 Cask Outer Shell 141 Top Shield Plate 62 Cask Lid 58 Top Resin 58 Protective Cover 21 Inner Bottom Plate 247 Bottom Shield 136 Lid Seal 52 Vent & Port Seal 52

• This value is the average temperature at the hottest cross section.

This value is the volumetric average temperature at the hottest cross section plus 18 °F to bound the effects of storage pad on the cask view factor. See Section A3.3.2.2.4 for discussion.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-6 MAXIMUM COMPONENT TEMPERATURE FOR FIRE ACCIDENT Component Temperature Time Limit

________________________ (°F) (hr) (°F)

Fuel Cladding 772 ¥ 1058 Fuel Compartment 725 ¥ ---

Basket Rails 542 ¥ ---

Cask Inner Shell 370 ¥ ---

Shield Shell 361 ¥ ---

Radial Resin N/A --- ---

Cask Outer Shell 913 End of fire ---

Top Shield Plate 253 1.2 hr after fire ---

Cask Lid 367 End of fire ---

Top Resin N/A --- ---

Protective Cover 1238 End of fire ---

Lid Seal 374 End of fire 536 Vent & Port Seal 265 1.2 hr after fire 536 This value is the average temperature at the hottest cross section.

This value is the volumetric average temperature at the hottest cross section plus 18 °F to bound the view factor effects on the initial temperatures. See Section A3.3.2.2.4 for discussion.

Neutron shield resin is assumed to be burnt and decomposed during the fire

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-7 MAXIMUM COMPONENT TEMPERATURES FOR BURIED CASK Component Temperature Time Limit

________________________ (°F) (hr) (°F)

Fuel Cladding 1058 95.75 1058 Fuel Compartment 1024 95.75 ---

Basket Rails 892 95.75 ---

Cask Inner Shell 804 95.75 ---

Shield Shell 799 95.75 ---

Radial Resin 300 1.85 300 Cask Outer Shell 791 95.75 ---

Top Shield Plate 319 95.75 ---

Cask Lid 316 95.75 ---

Top Resin 300 85.75 300 Protective Cover 316 95.75 ---

Lid Seal 317 95.75 536 Vent & Port Seal 316 95.75 536 This value is the volumetric average temperature at the hottest cross section plus 18 °F to bound the view factor effects on the initial temperatures. See Section A3.3.2.2.4 for discussion.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-8 MATERIAL THERMAL PROPERTIES (PAGE 1 OF 7)

Homogenized PWR Fuel Temperature ktrans in Helium Temperature ktrans in Vacuum (Air)

(°F) (Btu/hr-in-°F) (°F) (Btu/hr-in-°F) 136 0.0161 195 0.007345 233 0.0177 276 0.009133 330 0.0193 361 0.01138 428 0.0210 449 0.01417 526 0.0228 540 0.01735 624 0.0246 633 0.02136 722 0.0263 727 0.02571 821 0.0281 823 0.03085 920 0.0298 919 0.03653 1019 0.0317 1016 0.04338 Temperature Temperature u/hr-in-° (°F) u/lbm-° 212 0.0558 80 0.059 392 0.0587 260 0.065 572 0.0623 692 0.073 752 0.0673 1502 0.078 932 0.0738 1112 0.0824 r eff = 0.135 lb/in 3

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-8 MATERIAL THERMAL PROPERTIES (PAGE 2 OF 7)

Stainless Steel, Type 304 (Fuel Compartments, Basket Plate Support Bars, and Rail Plates)

(The Material group, conductivity, and diffusivity values are from Reference 28. The density value is from Reference 29.)

Mat. group J Thermal conductivity Thermal Diffusivity Specific heat capacity *

(ft /hr) 2 Temperature (°F) (Btu/hr-ft-°F) (Btu/hr-in-°F) (Btu/lbm-° F) 70 8.6 0.717 0.151 0.116 100 8.7 0.725 0.152 0.117 150 9.0 0.750 0.154 0.119 200 9.3 0.775 0.156 0.121 250 9.6 0.800 0.158 0.124 300 9.8 0.817 0.160 0.125 350 10.1 0.842 0.162 0.127 400 10.4 0.867 0.165 0.128 450 10.6 0.883 0.167 0.129 500 10.9 0.908 0.169 0.131 550 11.1 0.925 0.172 0.132 600 11.3 0.942 0.174 0.132 650 11.6 0.967 0.177 0.134 700 11.8 0.983 0.179 0.134 750 12.0 1.000 0.182 0.134 800 12.3 1.025 0.184 0.136 850 12.5 1.042 0.186 0.137 900 12.7 1.058 0.189 0.137 950 12.9 1.075 0.191 0.138 1000 13.1 1.092 0.194 0.138 1050 13.4 1.117 0.196 0.139 1100 13.6 1.133 0.198 0.140 1150 13.8 1.150 0.201 0.140 1200 14.0 1.167 0.203 0.141 3

r = 0.29 lbm/in 3

Thermal diffusivity, a = k , is used to calculate the specific heat with a density of 0.284 lbm/in for r cP steel alloys.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-8 MATERIAL THERMAL PROPERTIES (PAGE 3 OF 7)

Low Alloy Steel, SA 203, Gr. E or SA350, LF3 (Cask Inner Shell, Cask Lid Outer Plate, Shell Flange, and Bottom Inner Plate)

(The Material group, conductivity, and diffusivity values are from Reference 28. The density value is from Reference 29.)

Mat. group C Thermal conductivity Thermal Diffusivity Specific heat capacity 2

Temperature (°F) (Btu/hr-ft-°F) (Btu/hr-in-°F) (ft /hr) (Btu/lbm-° F) 70 23.7 1.975 0.459 0.105 100 23.6 1.967 0.451 0.107 200 23.5 1.958 0.424 0.113 300 23.4 1.950 0.401 0.119 400 23.1 1.925 0.379 0.124 500 22.7 1.892 0.357 0.130 600 22.2 1.850 0.336 0.135 700 21.6 1.800 0.314 0.140 800 21.0 1.750 0.292 0.147 900 20.3 1.692 0.269 0.154 1000 19.7 1.642 0.247 0.163 1100 19.1 1.592 0.223 0.175 1200 18.3 1.525 0.200 0.186 1300 16.6 1.383 0.164 0.206 1400 15.3 1.275 0.077 0.405 r = 0.284 lbm/in 3

3 Thermal diffusivity, a = k , is used to calculate the specific heat with a density of 0.284 lbm/in for r cP steel alloys.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-8 MATERIAL THERMAL PROPERTIES (PAGE 4 OF 7)

Carbon Steel, SA 266, CL. 2 or SA516, Gr. 70 or SA 105 (Gamma Shield, Cask Outer Shell, Shield Plate, Bottom Shield, & Protective Cover)

(The Material group, conductivity, and diffusivity values are from Reference 28. The density value is from Reference 29.)

Mat. group B Thermal conductivity Thermal Diffusivity Specific heat capacity (ft /hr) 2 Temperature (°F) (Btu/hr-ft-°F) (Btu/hr-in-°F) (Btu/lbm-°F) 70 27.3 2.275 0.530 0.105 100 27.6 2.300 0.520 0.108 200 27.8 2.317 0.487 0.116 300 27.3 2.275 0.455 0.122 400 26.5 2.208 0.426 0.127 500 25.7 2.142 0.399 0.131 600 24.9 2.075 0.373 0.136 700 24.1 2.008 0.346 0.142 800 23.2 1.933 0.319 0.148 900 22.3 1.858 0.291 0.156 1000 21.1 1.758 0.263 0.163 1100 19.8 1.650 0.234 0.172 1200 18.3 1.525 0.204 0.183 1300 16.9 1.408 0.157 0.219 1400 15.7 1.308 0.078 0.410 r = 0.280 lbm/in 3

3 Thermal diffusivity, a = k , is used to calculate the specific heat with a density of 0.284 lbm/in for r cP steel alloys.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-8 MATERIAL THERMAL PROPERTIES (PAGE 5 OF 7)

Aluminum Alloy 1100 (Basket Aluminum Plates)

(The conductivity, diffusivity and density values are from Reference 28.)

Al91 100 Thermal conductivity Thermal Diffusivity Specific heat capacity Temperature (°F) (Btu/hr-ft-°F) (Btu/hr-in-°F) (Btu/lbm-° F) 2 (ft /hr) 70 133.1 11.092 3.671 0.214 100 131.8 10.983 3.606 0.216 150 130.0 10.833 3.505 0.219 200 128.5 10.708 3.418 0.222 250 127.3 10.608 3.347 0.225 300 126.2 10.517 3.285 0.227 350 125.3 10.442 3.227 0.229 400 124.5 10.375 3.170 0.232 3

r = 0.098 lbm/in Aluminum Alloy 6061 (Basket Rail Shims)

(The conductivity, diffusivity and density values are from Reference 28.)

Al96061 Thermal conductivity Thermal Diffusivity Specific heat capacity*

Temperature (°F) (Btu/hr-ft-°F) (Btu/hr-in-°F) (ft /hr) 2 (Btu/lbm-° F) 70 96.1 8.008 2.661 0.213 100 96.9 8.075 2.660 0.215 150 98.0 8.167 2.650 0.218 200 99.0 8.250 2.649 0.221 250 99.8 8.317 2.641 0.223 300 100.6 8.383 2.629 0.226 350 101.3 8.442 2.620 0.228 400 101.9 8.492 2.620 0.230 3

r = 0.098 lbm/in 3

Thermal diffusivity, a = k , is used to calculate the specific heat with a density of 0.098 lbm/in for r cP aluminum alloys.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-8 MATERIAL THERMAL PROPERTIES (PAGE 6 OF 7)

Aluminum Alloy 6063 (Radial Neutron Shield Boxes)

(The conductivity, diffusivity and density values are from Reference 28.)

Al96063 Thermal conductivity Thermal Diffusivity Specific heat capacity

• Temperature (°F) (ft /hr) 2 (Btu/hr-ft-°F) (Btu/hr-in-°F) (Btu/lbm-° F) 70 120.8 10.067 3.340 0.214 100 120.3 10.025 3.299 0.215 150 119.7 9.975 3.232 0.219 200 119.0 9.917 3.177 0.221 250 118.5 9.875 3.133 0.223 300 118.1 9.842 3.088 0.226 350 118.0 9.833 3.040 0.229 400 117.6 9.800 3.000 0.231 3

r = 0.098 lbm/in Neutron Absorber (Poison) Plates Specific heat of Poison Plate BoralTM Thermal conductivity of Poison Plate (Cp )

t Temperature (°F) (Btu/hr-in-°F) (Btu/lbm-° F) 100 4.14 0.22 500 3.70 0.33 Solid Neutron Shield Resin (Borated Polyester - also used for Polypropylene) 3 Thermal diffusivity, a = k , is used to calculate the specific heat with a density of 0.098 lbm/in for r cP aluminum alloys.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-8 MATERIAL THERMAL PROPERTIES (PAGE 7 OF 7)

Helium Temperature Thermal conductivity Temperature Thermal

__________________ ___________________ ________________ conductivity (K) (W/m-K) (°F) (Btu/hr-in-°F) 300 0.1499 80 0.0072 400 0.1795 260 0.0086 500 0.2115 440 0.0102 600 0.2466 620 0.0119 800 0.3073 980 0.0148 1000 0.3622 1340 0.0174 1050 0.3757 1430 0.0181 Air Temperature Thermal conductivity Temperature Thermal

__________________ ___________________ ________________ conductivity (K) (W/m-K) (°F) (Btu/hr-in-°F) 250 0.02228 -10 0.0011 300 0.02607 80 0.0013 400 0.03304 260 0.0016 500 0.03948 440 0.0019 600 0.04557 620 0.0022 800 0.05698 980 0.0027 1000 0.06721 1340 0.0032 Concrete and Soil (Storage Pad)

(The conductivity values for concrete are from Reference 30 and the values for soil are from Reference 31.)

Concrete Soil Temperature Conductivity Conductivity

(°F) (Btu/hr-in-°F) (Btu/hr-in-°F) 70 0.0958 0.0144 1382 0.0479 _______________

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-9 TRANSVERSE EFFECTIVE FUEL CONDUCTIVITY IN HELIUM T OT CT avg keff

(°F) (°F) (°F) (Btu/hr-in-°F) 100 171 136 1.96E-02 200 262 231 2.24E-02 300 353 327 2.62E-02 400 446 423 3.02E-02 500 540 520 3.47E-02 600 635 618 3.97E-02 700 731 716 4.48E-02 800 827 814 5.14E-02 900 924 912 5.78E-02 1000 1021 1011 6.61E-02

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-10 COMPARISON OF MAXIMUM AND AVERAGE TEMPERATURES Storage Array Model TN-40HT Detailed Model Maximum Average Maximum Average Temp (°F) Temp (°F) Temp (°F) Temp (°F)

Cask Inner Shell 331 306 310 305 Cask Outer Shell 276 260 244 242 Radial Resin 283 267

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-1 1 COMPARISON OF VIEW FACTORS View Factors Detailed TN-40HT Model Storage Array Model Cask to Casks 0.1862 0.156 Cask to Pad 0 0.305 Cask to Ambient 0.8138 0.538

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-12 EFFECTIVE CONDUCTIVITIES FOR CASK SECTIONS Section 1, Mat # 1 Section 2, Mat # 2 T avg

(°F) u/hr-in-° F) (Btu/hr-in-° u/hr-in-° F) (Btu/hr-in-°

-50 0.98 2.21 2.23 1.99 50 1.02 2.21 2.23 1.99 150 1.07 2.24 2.25 2.01 250 1.10 2.23 2.24 2.00 350 1.11 2.18 2.19 1.96 450 1.12 2.12 2.13 1.92 550 1.13 2.06 2.07 1.87 650 1.12 1.99 2.01 1.82 750 1.12 1.92 1.94 1.76 Section Mat # 3 Section Mat # 4 T avg

(°F) /hr-in-°F) (Btu/hr-in-°F) (B /hr-in-° Btu/hr-in-°

-50 1.73 0.50 - 1.14 1.69 50 1.73 0.52 - 1.15 1.69 150 1.74 0.56 1.19 1.69 250 1.73 0.58 - 1.22 1.69 350 1.70 0.60 - 1.25 1.68 450 1.66 0.61 - 1.27 1.66 550 1.62 0.62 - 1.28 1.63 650 1.59 0.63 1.29 1.59 750 1.54 0.64 - 1.28 1.55 Section 5, Mat # 5 T avg keffaxl T avg I

(°F) /hr-in-°F u/hr-in-°

-50 0.83 70 1.975 50 0.87 100 1.967 150 0.93 200 1.958 250 0.98 300 1.950 350 1.01 400 1.925 450 1.04 500 1.892 550 1.05 600 1.850 650 1.07 700 1.800 750 1.07 800 1.750

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-13 AVERAGE TEMPERATURES AT THE HOTTEST CROSS SECTION FOR VACUUM DRYING CONDITIONS WITH ZERO GAP BETWEEN BASKET AND CASK INNER SHELL Component Average Temperature

___________________________ (°F)

Basket 573 Aluminum Shims 367 Cask Inner Shell 364 Basket Support Bar 90-270 601 Basket Support Bar 0-180 628 Basket Aluminum Plate 90-270 600 Basket Aluminum Plate 0-180 637 Shield Shell 356

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-14 MAXIMUM COMPONENT TEMPERATURE AT THE END OF VACUUM DRYING OPERATIONS Time Temperature Limit Component

_______________________ (hr) (°F) (°F)

Fuel Cladding 34 725 752 Fuel Compartment 34 685 ---

Basket Rails 34 537 ---

Cask Inner Shell 34 237 ---

Shield Shell 34 234 ---

Radial Resin 34 221 300 Top Resin 34 <237 300 Lid Seal 34 <237 536 Vent and Port Seal 34 <237 536 Temperature is bounded by maximum temperature of the cask inner shell.

The average temperature of the cask inner shell at the hottest cross section is 234°F.

The average temperature of the cask shield shell at the hottest cross section is 229°F.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-15 CASK INTERNAL PRESSURE (FUEL ASSEMBLY WITHOUT BPRAS)

Operating n back fB n free n total P cavity Condition Tcavity

___________ (lb-mole) (---) (lb-mole) (lb-mole) (°F) (psia) (psig)

Normal 0.580 0.01 0.889 0.589 456 27.6 12.9 Off-Normal 0.580 0.10 0.889 0.669 456 31.4 16.7 Fire 0.580 1.00 0.889 1.469 592 79.1 64.4 Accident Buried Cask Accident @ 0.580 1.00 0.889 1.469 929 104.4 89.7 95.75 hr

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-16 CASK INTERNAL PRESSURE (FUEL ASSEMBLY WITH BPRAS)

Operating n back fB n free n total P cavity Condition Tcavity

___________ (lb-mole) (---) (lb-mole) (lb-mole) (°F) (psia) (psig)

Normal 0.549 0.01 1.017 0.559 456 27.7 13.0 Off-Normal 0.549 0.10 1.017 0.651 456 32.2 17.5 Fire 0.549 1.00 1.017 1.566 592 89.0 74.3 Accident Buried Cask Accident @ 0.549 1.00 1.017 1.566 835 109.6 94.9 75 hr

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-17 MAXIMUM ENRICHMENT AND MINIMUM B10 CONTENT Maximum Assembly B10 Content Average Initial Minimum B10 Minimum B10 Used in Criticality (1)

Enrichment Content for BoralContent for B-Al Evaluation (wt. % U-235) (mg/cm2) (mg/cm2) (mg/cm2) 5.00 45.0 37.5 33.7 Notes:

(1) B-Al = Metal Matrix Composites and Borated Aluminum Alloys.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-18 AUTHORIZED CONTENTS FOR TN-40HT SYSTEM Assembly Assembly (1)

Assembly Type Array Class ID Exxon/ANF (ANP) 14x14 WE 14x14 WE 14x14 EXSTD Exxon/ANF (ANP) 14x14 High Burnup 14x14 WE 14x14 EXHBU Exxon/ANF (ANP) 14x14 Toprod 14x14 WE 14x14 EXTOP Westinghouse 14x14 Standard 14x14 WE 14x14 WESTD Westinghouse 14x14 OFA 14x14 WE 14x14 WEOFA Notes:

(1) Assembly ID is a simplification for the purpose of this evaluation to identify these fuel assembly Types.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-19 PARAMETERS FOR WE 14X14 CLASS FUEL ASSEMBLIES Assembly ID (1)(2)

Parameters EXSTD EXHBU EXTOP WESTD WOFA Active fuel length (in) 144 144 144 144 144 Number of fuel rods per assembly 179 179 179 179 179 Pitch (in) 0.556 0.556 0.556 0.556 0.556 Fuel pellet OD (in) 0.3565 0.3565 0.3505 0.3659 0.3444 Clad thickness (in) 0.0300 0.0310 0.0295 0.0243 0.0243 Clad OD (in) 0.424 0.426 0.417 0.422 0.400 16@0.541 16@0.541 16@0.541 16@0.539 16@0.528 Guide/instrument tube OD (in) 1@0.424 1@0.424 1@0.424 1 @0.422 1@0.4015 16@0.507 16@0.507 16@0.507 16@0.505 16@0.490 Guide/Instrument tube ID (in) 1@0.374 1@0.374 1@0.374 1 @0.3734 1@0.3499 Notes:

(1) Reload fuel from other manufacturers with these parameters are also acceptable.

(2) All dimensions shown are nominal

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-20 TN-40HT BASKET DIMENSIONS Actual Dimension Dimension Used in Model Basket Component Description inches inches (cm)

Compartment inside (maximum) 8.10 8.10 (20.574)

Compartment inside (nominal) 8.05 8.05 (20.447)

Compartment inside (minimum) 8.00 8.00 (20.320)

Compartment wall (maximum) 0.2325 0.2325 (0.59055)

Compartment wall (nominal) 0.1875 0.1870 (0.47490)

Compartment wall (minimum) 0.1775 0.1775 (0.45085)

Stainless steel insert bar height 1.75 1.75 (4.445)

SS insert bar thickness (maximum) 0.4925 0.4925 (1.2509)

SS insert bar thickness (nominal) 0.4375 0.4375 (1.1111)

SS insert bar thickness (minimum) 0.4275 0.4275 (1.0858)

Poison/Al plate height 13.18 12.67 (32.1772)

(1)

Poison plate thickness 0.075 0.075 (0.19049)

(1)

Al plate thickness 0.3625 0.3625 (0.92075)

Modeled as aluminum Horizontal gap 0.07 0.07 (0.1778) 1.00 / No slot Vertical slot width/height 5.75 (Replaced with basket)

Section height 15.00 14.49 (36.80)

Cask inside radius 36.00 36.00 (91.44)

Thickness of the inner shell 1.50 1.50 (3.81)

Thickness of the gamma shield shell 7.25 7.65 (19.42)

Thickness of the neutron shield shell 5.25 5.00 (12.70)

Not modeled (included in Thickness of the skin shell 0.50 gamma shield shell)

Notes:

(1) Dimensions given are based on 0.075 inch thick poison plate. Since the SS insert bar thickness is varied from 0.4925 inches to 0.4275 inches, the aluminum plate thickness is also varied from 0.4175 to 0.3525 inches and the poison plate thickness is kept constant at 0.075 inches.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-21 DESCRIPTION OF THE BASIC KENO MODEL UNITS Geometry Units Description 1 Fuel Pin Cell 2 Guide Tube 3 Instrument Tube 21 - 23 Basket Cells with Poison along the West Face of F/A 31 - 33 Basket Cells without Poison along North Face of F/A 41 - 43 Basket Cells without Poison along the East Face of F/A 51 - 53 Basket Cells with Poison along the South Face of F/A Arrays that define the West, North, East and South Faces of the 25, 35, 45, 55 Basket Cell without fuel 61 - 63 Basket Cells without Poison along the West Face of F/A 71 - 73 Basket Cells without Poison along North Face of F/A 81 - 83 Basket Cells without Poison along the East Face of F/A 91 - 93 Basket Cells without Poison along the South Face of F/A Arrays that define the West, North, East and South Faces of the 65, 75, 85, 95 Basket Cell without fuel and poison Basket Cell with Fuel Assembly Positions 201, 202, 205, 206 201 representing the South West Interior Positions 101, 102, 103 Units that model the vertical and horizontal rail plates.

206 Unit describing fuel assembly positions 201, 202, 205 and 206 207 Unit describing fuel assembly positions 203, 204, 207 and 208 210 Unit describing fuel assembly positions 209, 210 and 223 211 Unit describing fuel assembly positions 211, 212 and 228 214 Unit describing fuel assembly positions 213, 214, 217 and 218 215 Unit describing fuel assembly positions 215, 216, 219 and 220 221 Unit describing fuel assembly positions 221 and 222 225 Unit describing fuel assembly positions 224 and 225 226 Unit describing fuel assembly positions 226 and 227 230 Unit describing fuel assembly positions 229 and 230 231 - 240 Units describing fuel assembly positions 231 - 240 241 - 245 6x5 Array of Basket Cells defining the inner 30 locations Global Unit includes the fuel compartments, rails and the cask radial 10 shells

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-22 MATERIAL PROPERTY DATA Density Atom Density Material ID g/cm 3 Element Weight % (atoms/b-cm)

U-235 4.407 1.1 9435E-03 UO 2 1 10.576 U-238 83.743 2.24060E-02 (Enrichment - 5.0 wt. %)

___________________ _____ ______ O 11.850 4.72006E-02 Zr 98.23 4.2541E-02 Sn 1.45 4.8254E-04 Zircaloy-4 2 6.56 Fe 0.21 1.4856E-04 Cr 0.10 7.5978E-05

_____________________ _____ _______ Hf 0.01 2.2133E-06 H 11.1 6.6769E-02 Water (Pellet Clad Gap) 3 0.998

_______________________ ______ ________ O 88.9 3.3385E-02 C 0.080 3.1877E-04 Si 1.000 1.7025E-03 P 0.045 6.9468E-05 Stainless Steel (SS304) 4 7.94 Cr 19.000 1.7473E-02 Mn 2.000 1.7407E-03 Fe 68.375 5.8545E-02

______________________ ______ _______ Ni 9.500 7.7402E-03 H 11.165 6.67692E-02 Borated Water O 88.590 3.33846E-02 5 1.00 (2450 ppm Boron) B10 0.045 2.71583E-05 B11 0.200 1.09316E-04 B11 78.586 1.0988E-01 11 B 4 C in CC 7 2.556

____________________ _____ ______ C 21 .414 2.7470E-02 Aluminum 8 2.702 Al 100.0 6.0307E-02 B10 5.973 8.8013E-03 BORAL Poison B11 26.434 3.5426E-02 0.05" core thickness 9 2.450 2 C 8.993 1.1057E-02 (18.7 mg B-1 0/cm )

______________________ ______ _______ Al 58.600 3.2044E-02 Aluminum - Boron Poison B10 3.636 5.88913E-03 0.075" thickness 9 2.693 B11 0.404 5.95126E-04 2

(18.7 mg B-1 0/cm ) Al 95.960 5.76775E-02 Aluminum - Boron Poison B10 3.93 6.37017E-03 0.125" thickness 9 2.693 B11 0.44 6.43737E-04 2

(33.7 mg B-1 0/cm ) Al 95.63 5.74791E-02 H 11.1 6.6769E-02 Water 10 0.998

_______________________ ______ ________ O 88.9 3.3385E-02 C 1.00 3.92503E-03 Carbonsteel 11 7.821

______________________ ______ _______ Fe 99.0 8.34982E-02

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-23 MOST REACTIVE FUEL TYPE EVALUATION RESULTS (PAGE 1 OF 2)

Description k KENO 1s k eff Exxon 14x14 High Burnup ________

Centered, 80% IMD 0.9052 0.0009 0.9070 Centered, 90% IMD 0.9043 0.0007 0.9057 Centered, 100% IMD 0.9014 0.0008 0.9030 Inward, 80% IMD 0.9079 0.0007 0.9093 Inward, 90% IMD 0.9097 0.0007 0.9111 Inward, 100% IMD 0.9052 0.0007 0.9066 Exxon 14x14 Standard ________

Centered, 80% IMD 0.9054 0.0007 0.9068 Centered, 90% IMD 0.9059 0.0007 0.9073 Centered, 100% IMD 0.9023 0.0008 0.9039 Inward, 80% IMD 0.9099 0.0007 0.9113 Inward, 90% IMD 0.9083 0.0008 0.9099 Inward, 100% IMD 0.9064 0.0008 0.9080 Exxon 14x14 Top Rod _______

Centered, 80% IMD 0.8995 0.0008 0.9011 Centered, 90% IMD 0.8987 0.0008 0.9003 Centered, 100% IMD 0.8930 0.0007 0.8944 Inward, 80% IMD 0.9022 0.0007 0.9036 Inward, 90% IMD 0.9016 0.0008 0.9032 Inward, 100% IMD 0.8974 0.0008 0.8990

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-23 MOST REACTIVE FUEL TYPE EVALUATION RESULTS (PAGE 2 OF 2)

Description k KENO 1s keff WE 14x14 OFA ________

Centered, 80% IMD 0.8921 0.0007 0.8935 Centered, 90% IMD 0.8884 0.0007 0.8898 Centered, 100% IMD 0.8840 0.0007 0.8854 Inward, 80% IMD 0.8954 0.0007 0.8968 Inward, 90% IMD 0.8928 0.0008 0.8944 Inward, 100% IMD 0.8867 0.0008 0.8883 WE 14x14 Standard _________

Centered, 80% IMD 0.9175 0.0008 0.9191 Centered, 90% IMD 0.9189 0.0007 0.9203 Centered, 100% IMD 0.9147 0.0007 0.9161 Inward, 80% IMD 0.9218 0.0009 0.9236 Inward, 90% IMD 0.9215 0.0008 0.9231 Inward, 100% IMD 0.9182 0.0007 0.9196

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-24 FUEL COMPARTMENT TUBE INTERNAL DIMENSION EVALUATION RESULTS Description k KENO 1s k eff Nominal ID = 8.05" from previous evaluation 80% IMD 0.9218 0.0009 0.9236 90% IMD 0.9215 0.0008 0.9231 100% IMD 0.9182 0.0007 0.9196

_________ Maximum ID = 8.10" ________

80% IMD 0.9187 0.0008 0.9203 90% IMD 0.9201 0.0008 0.9217 100% IMD 0.9163 0.0007 0.9177

__________ Minimum ID = 8.00" _________

80% IMD 0.9251 0.0007 0.9265 90% IMD 0.9238 0.0006 0.9250 100% IMD 0.9210 0.0007 0.9224

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-25 COMPARTMENT TUBE AND STEEL BAR THICKNESS EVALUATION RESULTS Description k KENO 1s keff Nominal Tube Thickness = 0.1875", Nominal Steel Bar Thickness = 0.4375"

_____________________________ (from previous evaluation) _______________________

80% IMD 0.9251 0.0007 0.9265 90% IMD 0.9238 0.0006 0.9250 100% IMD 0.9210 0.0007 0.9224 Maximum Tube Thickness = 0.2325", Maximum Steel Bar Thickness = 0.4925" 80% IMD 0.9241 0.0008 0.9257 90% IMD 0.9229 0.0007 0.9243 100% IMD 0.9212 0.0007 0.9226 Maximum Tube Thickness = 0.2325", Minimum Steel Bar Thickness = 0.4275" 80% IMD 0.9226 0.0007 0.9240 90% IMD 0.9252 0.0007 0.9266 100% IMD 0.9213 0.0007 0.9227 Maximum Tube Thickness = 0.2325", Nominal Steel Bar Thickness = 0.4375" 80% IMD 0.9237 0.0008 0.9253 90% IMD 0.9237 0.0007 0.9251 100% IMD 0.9204 0.0007 0.9218 Minimum Tube Thickness = 0.1775", Maximum Steel Bar Thickness = 0.4925" 80% IMD 0.9227 0.0009 0.9245 90% IMD 0.9237 0.0007 0.9251 100% IMD 0.9215 0.0007 0.9229 Minimum Tube Thickness = 0.1775", Minimum Steel Bar Thickness = 0.4275" 80% IMD 0.9226 0.0008 0.9242 90% IMD 0.9248 0.0008 0.9264 100% IMD 0.9229 0.0008 0.9245 Minimum Tube Thickness = 0.1775", Nominal Steel Bar Thickness = 0.4375" 80% IMD 0.9234 0.0007 0.9248 90% IMD 0.9249 0.0008 0.9265 100% IMD 0.9219 0.0007 0.9233

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-26 POISON BAR THICKNESS AND MATERIAL EVALUATION RESULTS Description k KENO 1s keff Nominal Tube Thickness = 0.1875", Nominal Steel Bar Thickness = 0.4375", BORAL 80% IMD 0.9244 0.0008 0.9260 90% IMD 0.9238 0.0008 0.9254 100% IMD 0.9206 0.0008 0.9222 Maximum Tube Thickness = 0.2325", Minimum Steel Bar Thickness = 0.4275", BORAL 80% IMD 0.9241 0.0007 0.9255 90% IMD 0.9241 0.0008 0.9257 100% IMD 0.9208 0.0008 0.9224 Minimum Tube Thickness = 0.1775", Minimum Steel Bar Thickness = 0.4275", BORAL 80% IMD 0.9237 0.0007 0.9251 90% IMD 0.9242 0.0007 0.9256 100% IMD 0.9216 0.0008 0.9232

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-27 RAIL MODELING EVALUATION RESULTS Description k KENO 1s keff Nominal Tube Thickness = 0.1875", Nominal Steel Bar Thickness = 0.4375", Rail1 80% IMD 0.9235 0.0008 0.9251 90% IMD 0.9240 0.0007 0.9254 100% IMD 0.9204 0.0007 0.9218 Nominal Tube Thickness = 0.1875", Nominal Steel Bar Thickness = 0.4375", Rail2 80% IMD 0.9242 0.0008 0.9258 90% IMD 0.9233 0.0008 0.9249 100% IMD 0.9213 0.0008 0.9229 Nominal Tube Thickness = 0.1875", Nominal Steel Bar Thickness = 0.4375", Rail3 80% IMD 0.9271 0.0008 0.9287 90% IMD 0.9280 0.0007 0.9294 100% IMD 0.9245 0.0009 0.9263 Maximum Tube Thickness = 0.2325", Minimum Steel Bar Thickness = 0.4275", Rail3 80% IMD 0.9273 0.0008 0.9289 90% IMD 0.9297 0.0008 0.9313 100% IMD 0.9212 0.0007 0.9226 Minimum Tube Thickness = 0.1775", Minimum Steel Bar Thickness = 0.4275", Rail3 80% IMD 0.9264 0.0007 0.9278 90% IMD 0.9267 0.0007 0.9281 100% IMD 0.9250 0.0007 0.9264 Minimum Tube Thickness = 0.1775", Nominal Steel Bar Thickness = 0.4375", Rail3 80% IMD 0.9260 0.0007 0.9274 90% IMD 0.9270 0.0007 0.9284 100% IMD 0.9228 0.0006 0.9240

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-28 WE 14X14 CLASS ASSEMBLY WITHOUT CCS FINAL RESULTS Model Description k KENO 1s keff 2

Enrichment = 5.00 wt. % U-235, Soluble Boron = 2450 ppm, 33.7 mg B-1 0/cm Internal Moderator Density = 01% 0.5705 0.0004 0.5713 Internal Moderator Density = 10% 0.6496 0.0005 0.6506 Internal Moderator Density = 20% 0.7286 0.0007 0.7300 Internal Moderator Density = 30% 0.7926 0.0007 0.7940 Internal Moderator Density = 40% 0.8444 0.0007 0.8458 Internal Moderator Density = 50% 0.8805 0.0008 0.882 1 Internal Moderator Density = 60% 0.9037 0.0008 0.9053 Internal Moderator Density = 70% 0.9200 0.0007 0.9214 Internal Moderator Density = 80% 0.9293 0.0008 0.9309 Internal Moderator Density = 90% 0.9313 0.0008 0.9329 Internal Moderator Density = 100% 0.9310 0.0007 0.9324

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-29 WE 14X14 CLASS ASSEMBLY WITH CCS FINAL RESULTS Model Description k KENO 1s k eff 2

Enrichment = 5.00 wt. % U-235, Soluble Boron = 2450 ppm, 33.7 mg B-1 0/cm Internal Moderator Density = 01% 0.5787 0.0004 0.5795 Internal Moderator Density = 10% 0.6480 0.0005 0.6490 Internal Moderator Density = 20% 0.7206 0.0006 0.7218 Internal Moderator Density = 30% 0.7808 0.0007 0.7822 Internal Moderator Density = 40% 0.8305 0.0007 0.8319 Internal Moderator Density = 50% 0.8673 0.0007 0.8687 Internal Moderator Density = 60% 0.8944 0.0008 0.8960 Internal Moderator Density = 70% 0.9132 0.0007 0.9146 Internal Moderator Density = 80% 0.9266 0.0007 0.9280 Internal Moderator Density = 90% 0.9333 0.0007 0.9347 Internal Moderator Density = 100% 0.9357 0.0008 0.9373

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-30 BENCHMARKING RESULTS (PAGE 1 OF 4)

Separation U of Enrich. Pitch H 2 O/fuel assemblies Run ID Wt% (cm) volume (cm) AEG k eff 1s B1645SO1 2.46 1.41 1.015 1.78 32.8118 0.9965 0.0008 B1645SO2 2.46 1.41 1.015 1.78 32.7528 1.0006 0.0008 BW1231B1 4.02 1.511 1.139 31.1429 0.9966 0.0009 BW1231B2 4.02 1.511 1.139 29.8872 0.9990 0.0007 BW1273M 2.46 1.511 1.376 32.2213 0.9961 0.0007 BW1484A1 2.46 1.636 1.841 1.636 34.5373 0.9975 0.0008 BW1484A2 2.46 1.636 1.841 4.908 35.1630 0.9934 0.0008 BW1484B1 2.46 1.636 1.841 33.9415 0.9984 0.0008 BW1484B2 2.46 1.636 1.841 1.636 34.5780 0.9961 0.0009 BW1484B3 2.46 1.636 1.841 4.908 35.2638 0.9978 0.0008 BW1484C1 2.46 1.636 1.841 1.636 34.6547 0.9936 0.0009 BW1484C2 2.46 1.636 1.841 1.636 35.2469 0.9944 0.0010 BW1484S1 2.46 1.636 1.841 1.636 34.5159 1.0002 0.0008 BW1484S2 2.46 1.636 1.841 1.636 34.5530 0.9990 0.0008 BW1484SL 2.46 1.636 1.841 6.544 35.4203 0.9944 0.0009 BW1645S1 2.46 1.209 0.383 1.778 30.1060 0.9987 0.0008 BW1645S2 2.46 1.209 0.383 1.778 29.9920 1.0049 0.0008 BW1810A 2.46 1.636 1.841 33.9524 0.9987 0.0006 BW1810B 2.46 1.636 1.841 33.9711 0.9995 0.0006 BW1810CR 2.46 1.636 1.841 33.1556 0.9995 0.0008 BW1810D 2.46 1.636 1.841 33.0876 0.9981 0.0010 BW1810E 2.46 1.636 1.841 33.1520 0.9991 0.0007 BW1810F 2.46 1.636 1.841 33.9581 1.0029 0.0007 BW1810GR 2.46 1.636 1.841 32.9478 0.9986 0.0007 BW1810H 2.46 1.636 1.841 32.9370 0.9981 0.0008 BW1810I 2.46 1.636 1.841 33.9613 1.0028 0.0007 BW1810J 2.46 1.636 1.841 33.1379 0.9995 0.0008 EPRU65 2.35 1.562 1.196 33.9138 0.9959 0.0008 EPRU65B 2.35 1.562 1.196 33.4073 1.0000 0.0009 EPRU75 2.35 1.905 2.408 35.8676 0.9968 0.0009 EPRU75B 2.35 1.905 2.408 35.3074 1.0002 0.0008 EPRU87 2.35 2.21 3.687 36.6120 1.0011 0.0009 EPRU87B 2.35 2.21 3.687 36.3460 1.0003 0.0008 NSE71SQ 4.74 1.26 1.823 33.7627 0.9978 0.0009 NSE71W1 4.74 1.26 1.823 34.0088 0.9981 0.0010 NSE71W2 4.74 1.26 1.823 34.3856 0.9995 0.0010

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-30 BENCHMARKING RESULTS (PAGE 2 OF 4)

Separation U of Enrich. Pitch H 2 O/fuel assemblies Run ID Wt% (cm) volume (cm) AEG k eff 1 s P2438BA 2.35 2.032 2.918 5.05 36.2244 0.9973 0.0009 P2438SLG 2.35 2.032 2.918 8.39 36.2906 0.9985 0.0009 P2438SS 2.35 2.032 2.918 6.88 36.2690 0.9979 0.0009 P2438ZR 2.35 2.032 2.918 8.79 36.2891 0.9976 0.0009 P2615BA 4.31 2.54 3.883 6.72 35.7276 1.0005 0.0011 P2615SS 4.31 2.54 3.883 8.58 35.7456 0.9959 0.0011 P2615ZR 4.31 2.54 3.883 10.92 35.7709 0.9980 0.0010 P2827L1 2.35 2.032 2.918 13.72 36.2491 1.0051 0.0008 P2827L2 2.35 2.032 2.918 11.25 36.2939 1.0005 0.0010 P2827L3 4.31 2.54 3.883 20.78 35.6740 1.0095 0.0009 P2827L4 4.31 2.54 3.883 19.04 35.7173 1.0066 0.0010 P2827SLG 2.35 2.032 2.918 8.31 36.3010 0.9957 0.0008 P3314BA 4.31 1.892 1.6 2.83 33.1874 1.0000 0.0009 P3314BC 4.31 1.892 1.6 2.83 33.2334 0.9992 0.0009 P3314BF1 4.31 1.892 1.6 2.83 33.2422 1.0024 0.0009 P3314BF2 4.31 1.892 1.6 2.83 33.2121 1.0001 0.0010 P3314BS1 2.35 1.684 1.6 3.86 34.8545 0.9957 0.0010 P3314BS2 2.35 1.684 1.6 3.46 34.8324 0.9940 0.0008 P3314BS3 4.31 1.892 1.6 7.23 33.4328 0.9996 0.0009 P3314BS4 4.31 1.892 1.6 6.63 33.4152 1.0000 0.0008 P3314SLG 4.31 1.892 1.6 2.83 34.0109 0.9971 0.0010 P3314SS1 4.31 1.892 1.6 2.83 33.9613 0.9984 0.0010 P3314SS2 4.31 1.892 1.6 2.83 33.7719 1.0014 0.0009 P3314SS3 4.31 1.892 1.6 2.83 33.8956 0.9995 0.0010 P3314SS4 4.31 1.892 1.6 2.83 33.7604 0.9962 0.0009 P3314SS5 2.35 1.684 1.6 7.8 34.9476 0.9947 0.0010 P3314SS6 4.31 1.892 1.6 10.52 33.5406 1.0010 0.0008 P3314W1 4.31 1.892 1.6 34.3962 1.0009 0.0010 P3314W2 2.35 1.684 1.6 35.2153 0.9972 0.0008 P3314ZR 4.31 1.892 1.6 2.83 33.9897 0.9977 0.0010 P3602BB 4.31 1.892 1.6 8.3 33.3198 1.0031 0.0010 P3602BS1 2.35 1.684 1.6 4.8 34.7746 1.0034 0.0009 P3602BS2 4.31 1.892 1.6 9.83 33.3649 1.0047 0.0010 P3602N11 2.35 1.684 1.6 8.98 34.7410 1.0025 0.0008 P3602N12 2.35 1.684 1.6 9.58 34.8378 1.0048 0.0009

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-30 BENCHMARKING RESULTS (PAGE 3 OF 4)

Separation U of Enrich. Pitch H 2 O/fuel assemblies Run ID Wt% (cm) volume (cm) AEG k eff 1 s P3602N13 2.35 1.684 1.6 9.66 34.9334 1.0006 0.0009 P3602N14 2.35 1.684 1.6 8.54 35.0287 0.9969 0.0010 P3602N21 2.35 2.032 2.918 10.36 36.2787 0.9999 0.0009 P3602N22 2.35 2.032 2.918 11.2 36.1963 1.0014 0.0008 P3602N31 4.31 1.892 1.6 14.87 33.2015 1.0063 0.0010 P3602N32 4.31 1.892 1.6 15.74 33.3085 1.0072 0.0010 P3602N33 4.31 1.892 1.6 15.87 33.4168 1.0084 0.0010 P3602N34 4.31 1.892 1.6 15.84 33.4653 1.0028 0.0010 P3602N35 4.31 1.892 1.6 15.45 33.5169 1.0030 0.0009 P3602N36 4.31 1.892 1.6 13.82 33.5832 1.0003 0.0010 P3602N41 4.31 2.54 3.883 12.89 35.5269 1.0127 0.0010 P3602N42 4.31 2.54 3.883 14.12 35.6711 1.0068 0.0009 P3602N43 4.31 2.54 3.883 12.44 35.7505 1.0049 0.0009 P3602SS1 2.35 1.684 1.6 8.28 34.8708 1.0007 0.0009 P3602SS2 4.31 1.892 1.6 13.75 33.4133 1.0026 0.0010 P3926L1 2.35 1.684 1.6 10.06 34.8569 1.0003 0.0009 P3926L2 2.35 1.684 1.6 10.11 34.9374 1.0020 0.0008 P3926L3 2.35 1.684 1.6 8.5 35.0657 0.9967 0.0010 P3926L4 4.31 1.892 1.6 17.74 33.3262 1.0066 0.0009 P3926L5 4.31 1.892 1.6 18.18 33.4035 1.0054 0.0010 P3926L6 4.31 1.892 1.6 17.43 33.5141 1.0038 0.0009 P3926SL1 2.35 1.684 1.6 6.59 35.0674 0.9950 0.0009 P3926SL2 4.31 1.892 1.6 12.79 33.5810 0.9998 0.0009 P4267B1 4.31 1.89 1.59 31.7989 0.9992 0.0008 P4267B2 4.31 1.89 1.59 31.5288 1.0027 0.0007 P4267B3 4.31 1.715 1.09 30.9907 1.0057 0.0009 P4267B4 4.31 1.715 1.09 30.5098 0.9993 0.0008 P4267B5 4.31 1.715 1.09 30.1008 1.0009 0.0008 P4267SL1 4.31 1.89 1.59 33.4692 0.9987 0.0011 P4267SL2 4.31 1.715 1.09 31.9346 0.9995 0.0011 P62FT231 4.31 1.891 1.6 5.67 32.9228 1.0020 0.0009 P71F14F3 4.31 1.891 1.6 5.19 32.8227 1.0009 0.0010 P71F14V3 4.31 1.891 1.6 5.19 32.8587 0.9977 0.0010 P71F14V5 4.31 1.891 1.6 5.19 32.8662 0.9980 0.0010 P71F214R 4.31 1.891 1.6 5.19 32.8669 0.9976 0.0009

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-30 BENCHMARKING RESULTS (PAGE 4 OF 4)

Separation of U Enrich. Pitch H 2 O/fuel assemblies Run ID Wt% (cm) volume (cm) AEG keff 1s PAT80L1 4.74 1.6 3.807 2 35.0276 1.0014 0.0009 PAT80L2 4.74 1.6 3.807 2 35.1079 0.9986 0.0011 PAT80SS1 4.74 1.6 3.807 2 35.0125 0.9998 0.0009 PAT80SS2 4.74 1.6 3.807 2 35.1128 0.9967 0.0010 W3269A 5.7 1.422 1.93 33.1383 0.9976 0.0009 W3269B1 3.7 1.105 1.432 32.4010 0.9962 0.0008 W3269B2 3.7 1.105 1.432 32.3940 0.9965 0.0008 W3269B3 3.7 1.105 1.432 32.2464 0.9945 0.0008 W3269C 2.72 1.524 1.494 33.7731 0.9979 0.0009 W3269SL1 2.72 1.524 1.494 33.3854 0.9973 0.0010 W3269SL2 5.7 1.422 1.93 33.1006 1.0024 0.0010 W3269W1 2.72 1.524 1.494 33.5160 0.9972 0.0012 W3269W2 5.7 1.422 1.93 33.1786 1.0015 0.0010 W3385SL1 5.74 1.422 1.932 33.2320 1.0004 0.0009 W3385SL2 5.74 2.012 5.067 35.8876 1.0014 0.0010 Correlation 0.32 0.41 0.19 0.66 -0.04 N/A N/A

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-31 USL-1 RESULTS Range of Formula for USL-1 Parameter applicability (0.05 D keff Margin) 0.9404 + ( 1 .0545E-03)*X (X < 3.5906 )

U Enrichment (wt% U-235) 2.35 - 5.74 0.9442 (X >= 3.5906 )

0.9357 + ( 4.7885E-03)*X (X < 1.7997 )

Fuel Rod Pitch (cm) 1.105 - 2.540 0.9443 (X >= 1.7997 )

0.9421 + ( 7.5929E-04)*X (X < 2.1362 )

Water/Fuel Volume Ratio 0.383 - 5.067 0.9438 (X >= 2.1362 )

0.9409 + ( 5.0511E-04)*X (X < 7.1170 )

Assembly Separation (cm) 1.636 - 20.78 0.9442 (X >= 7.1170 )

Average Energy Group 0.9438 (X < 32.694 )

Causing Fission (AEG) 29.9 - 36.6 0.9467 + (-8.8965E-05)*X (X >= 32.694 )

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.3-32 USL DETERMINATION FOR CRITICALITY ANALYSIS Value from Limiting Parameter Analysis Bounding USL-1 U Enrichment (wt. % U-235) 3.6 (1)0.9442 Fuel Rod Pitch (cm) 1.412 0.9425 Water/Fuel Ratio 1.610 (2)0.9433 (3)

Assembly Separation (cm) 2.063 0.9419 Average Energy Group Causing (4)

Fission (AEG) 30 0.9438 (1) A bounding lower enrichment value is utilized and is conservative, since the minimum enrichment credited herein is 5.00 wt. % U-235.

(2) The water to fuel volume ratio is calculated using 179 rods.

(3) Separation Distance = 2*0 . 1875 11 + 0 . 4375 11 = 0 . 8125 11 = 2.06375 cm, calculated with nominal dimensions for the stainless steel in the fuel compartment, nominal stainless steel tie bar width and inward fuel assembly positioning.

(4) Examination of the results shows that the value is between 30 and 34 and hence a conservative value that produces the minimum USL was chosen.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A3.4-1 DESIGN CRITERIA FOR TN-40HT CASKS Maximum Gross Weight on Crane (with lift beams, without water) 125 tons Maximum Cask Height with Lid Removed 16 ft. 1 in.

Minimum Design Life 25 years Upper Subcritical Limit < 0.95- minus biases Payload Capacity ^40 intact 14x14 PWR assemblies (Including inserts)

Spent Fuel Characteristics a) Design Basis Initial Enrichment(max) 5.0 wt % U-235 b) Burnup (max) 60 GWD/MTU, c) Cooling time (min) 12 years d) Decay Heat 32 kW (total)

Max Clad Temperature 400 °C (752 °F) - Normal 570 °C (1058 °F) -Accident Cask Cavity Atmosphere helium gas Maximum Internal Pressure for Stress Evaluation 100 psig Minimum/Maximum Ambient Temperature -40 to 120ºF Daily Averaged Ambient Temperature -40 to 100 °F Over 24 hr. period (min-max)

Maximum Solar Heat Load 61.488 Btu/Hr-ft2 (curved surfaces) 2 (Averaged over 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 122.83 Btu/H r-ft (horizontal surfaces)

Tornado Wind 360 mph (rotational plus translational)

Tornado Missiles 4000 lb. auto at 50 mph 4" x 12" x 12' wood plank at 300 mph Cask Drop 18" Drop onto concrete pad or equivalent end drop resulting in 50 g Seismic Design Earthquake 0.12 g horizontal 0.08 g vertical Snow and Ice 50 psf load

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Wind or Water (a)

TI 177.25 199.63 1

Reaction force Seismic (b) 177.25 199.63 ii Reaction force I U I .0 1< >

FIGURE A3.2-1 EARTHQUAKE, WIND, AND WATER LOADS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Missile Missile A 13 177.25 199.63

'l' B 91.58 1

Pivot point I Missile B 0.8V o Vo Missil cop Missile A (p--

13 d CG a

FIGURE A3.2-2 TORNADO MISSILE IMPACT LOADS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 6W-6X 250,000- [50X O61ELO EVALUATiON) 2.50 X O6WUTMATE EVALUATION) 3W (YELD) 3W (YIELD) 5W (ULTIMA 5W CUL11IATE)

ND OW LOADS APPLIED NE PAiR TRUNONS.

PER TRUNNON S OF TOTAL.

FIGURE A3.2-3 LIFTING LOADS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 DRAIN PORT (VENT PORT SiMILAR)

OVERPRESSURE TANK DRFLLED PASSAGE FRC LID SEAL TO DRAIN POE 1-PROTECTIVE COVER AND VF-IJT PORT SEALS SINGLE METALLiC SEAL DRILLED PASSAGE TO LID SEAL E ELASTOMER SEAL N1ETALLIC SEAL CONTROL BOX FIGURE A3.3-1 TN-40HT CASK SEAL PRESSURE MONITORING SYSTEM

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Protective Cover Top Ne Shield Cask Li Basket Concrete Pad Radial N Shield Soil Shield S Cask In Shell Cask Bottom Basket Cross Section Plate Aluminum Boxes Neutron Shield Resin Baske Steel Bar Basket Aluminum Plate FIGURE A3.3-2 FULL-LENGTH CASK MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 oss Section Mesh in Axial Direction Mesh in Homogenized Fuel Assembl y Region Mesh in Neutron Shield Region FIGURE A3.3-3 MESH DENSITY OF FINITE ELEMENT MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 0.020 air g

0. 010 air 0.125 helium gap between basket and cask inner bottom plate 0.010 air gap bet shield shell and b shield 0.125 air gap between cask inner bottom plate and bottom shield 0.06 helium gap between basket rail segments FIGURE A3.3-4 GAPS IN AXIAL DIRECTION

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Detail A Detail B 0.020 gap 0.010 gap 0.020 gap 0.100 gap 0.130 gap FIGURE A3.3-5 GAPS IN RADIAL DIRECTION (PAGE 1 OF 2)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Detail C Detail D 0.020 gap 0.010 gap 0.100 gap 0.010 gap 0.020 gap Detail E Detail F 0.010 gap 0.130 gap 0.010 Shrink fit 0.010 gap gap with effective conductivity FIGURE A3.3-5 GAPS IN RADIAL DIRECTION (PAGE 2 OF 2)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Overpressure Tank Protective Cover Top Neutron Shield Box Polypropylene resin Cask Lid Top Shield Plate Cask Flange Top of Shield Plate Helium at the Top of Basket Mesh Density Model without Gas Elements FIGURE A3.3-6 TOP CASK SUB-MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Convection & Radiation to Ambient and Fixed Tem peratures at Soil ANSYS :3 0 ELEMENTS PowerLraphics C ONV-HC cE

_1II4

-1I...

-icii Solar Heat Flux ANE.YS :3 - ci EL EMENT S Pot,erC.raphirs E FACE T = 1 MAT NTJH HF LU 1 ..30 ii S

- LI :: b LI b

369801

4. 0398 S3 0996 61159:3 692191

- 853386 FIGURE A3.3-7 BOUNDARY CONDITIONS FOR FULL-LENGTH CASK MODEL (PAGE 1 OF 2)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Heat Generatin g Rates in Fuel Assemblies A1TSYS EL U EL EHENT S HC.EN PATES QHIN = i:

':= :32S32:3

- i:

ri ii 144 RI 2:

1 B 1016

- 217219 FIGURE A3.3-7 BOUNDARY CONDITIONS FOR FULL-LENGTH CASK MODEL (PAGE 2 OF 2)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Heat Flux to Cask Top Convection + Radiation to Ambient ANSYS 8Cl CUNV-HCO E

- -10 Solar Heat Flux AM.YS 8.0 HFLU 1_F ff 1/ fi - . .

T i

.'* / J *1 1cF4c4 I I -

J \I i ii

- i;i.

FIGURE A3.3-8 BOUNDARY CONDITIONS FOR CASK TOP SUB-MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Cross-Section Sub-Model Lid-Seal Re g ion Model FIGURE A3.3-9 FINITE ELEMENT MODELS FOR ACCIDENT CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 1.2 1.0 I-0 0.8 0.6 0.4 0.2 0.0 0 25 50 75 100 125 150 Height (in)

FIGURE A3.3-10 COMPARISON OF THE AXIAL HEAT PROFILES IN THE MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 8.1:1 PLd::YI N':::'. 8 NO'DL SOL)_ITION SnlEP=i SUB =1 T]JE=i TEL.,.IED Pc:'e:r'raphicE-EFcET=i 3 *.1 =11:3.

=

LIII .-..

FIGURE A3.3-11 TEMPERATURE DISTRIBUTIONS FOR FULL-LENGTH MODEL NORMAL/OFF-NORMAL STORAGE CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ANSYS 8.0 Homo g enized Fuel Assemblies ixr NC. Fuel Compartments NCFAL SCIUrTCN STP=.

SUB =. [LC 1D. 2 TJE=1 OTI1Tfl TEJMF (A) 1E1 sYs=0 UR I PowerGraphic TFM (M bYCI AESt SM71 =258.L8 6Fcr=1 SMX =680.335 258.18

___ 305.086 =&30. 3 351. 992

___ 398.898

___ 443.804 ___ 1l1G

___ 492.7. ___ 3E2.2O5

___ 539. b.7 4'- M 23

___ 586.523 633.429 517. C1 680. 335 rD. q 2.NSY 8.0 Basket Rails NcSYS 8.0 Radial Resin PTXP NO. 6 P1XT NO. 3 TOI.?L SOLUTION NOE2L SOI.UTIOT STEP=1 STP=1 SUB =1 SUB =1 TIME=1 TI=1 TEMF (A)

TFJF (A)

RYS=0 PowerGraphios EowerGraphics EFlT=1 EFET=1 AVES4t AVESXt SMN =207.185 s4q =245.99 SX =282.897 snx =458.922 2O7.185 245.99 ___ 215.597

___ 269.649 ___ 224.01

___ 293.308 ___ 232.422

___ 36. 967 ___ 240.835

___ 340.626 ___ 249.247

___ 364.285 ___ 257.6E

___ 387.945 ___ 266.O72

___ 411.604 274.485 435.263 282.897 458.922 FIGURE A3.3-12 NORMAL / OFF-NORMAL TEMPERATURE DISTRIBUTION FOR COMPONENTS OF FULL-LENGTH MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 113.943 178.224 242.5C:'6 3C:'6.787 371.1:1:168 14E II 4 _1I I F F 4F 411 Cask lid and To p Shield Plate Top Resin 1 1 J 1 1 1 14 1 - 1 liii 1 1 1 4 4 1 J 1 1 4 1414 11 1 1 1. 1 Iii Protective Cover i_I I I 114 1 4 1 14 FIGURE A3.3-13 NORMAL / OFF-NORMAL TEMPERATURE DISTRIBUTION FOR CASK TOP SUB-MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FuelCcrnp p.J-1 Iii' a'

40 U IrgiShli 1n

12.  :? 1.Lx.F Tinie :-r:

10 4U 6U 6 4U U

6U ft U

40 U Out ShI 1 2 4U 4 1L 10 14 1 Time :hr:

FIGURE A3.3-14 COMPONENT TIME-TEMPERATURE HISTORY FOR FIRE ACCIDENT (PAGE 1 OF 2)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 4" II a.

il il L.1.1 il 1751.

iF. n 10 20 40 U

Tirn I:hr:

.ulo 1=.

il t1 1:1 10 Tii I:hr:I FIGURE A3.3-14 COMPONENT TIME-TEMPERATURE HISTORY FOR FIRE ACCIDENT (PAGE 2 OF 2)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Entire Model Fuel Assemblies 7NSYS 8.0 PLJJ NO. 6 NSYS 8.0 PLYT NO. 1 NQEL SOLUTION NCEL SOluTION STEP=1 SIEF=1 SUB =1.

SUB =1 TIMF=1 T1IE=1 TEM (Am) TEF (Am)

ESYS=0 RYS=0 EowerGraphic PowerGraphicri EP.CT=1 EFZLET=1 AVPESat AVRESt S1N =287.16 S1JN =422.614 SMX =771.538 SMX =771.538

- 287.16 - 422.614

___ 340.98 - 461.383

___ 394.8 ___ 500.153

___ 448.619 ___ 538.922

___ 502.439 ___ 577. 61

___ 555.259 ___ 616.46

___ 61.0.079 ___ 655.23

___ 663.898 ___ 693.999

- 717.718 - 732.768 771.538 771.538 MSYS 8.0 Basket Rails ZNSYS 8.0 Fuel Compartments PDT NO. 2 PLDJ7 NO. 3 NOOZL SOUJTION NOEL SOLLTION STP=1 STEP=1 SUB =1 SUB =1 TIE=1 TFE=1 TEMP (A) TE1F (A)

SYs=0 PowsrGraphics FowerGraphics EFKT=1 EFACET=1 AVRESt ZVRESt MI =422.318 S1N =400.874 S< =725.06 SJYC =541.671

- 422.318 - 400.874

___ 455.956 ___ 416.518

___ 489.594 ___ 432.162

___ 523.232 ___ 447.806

___ 556.87 ___ 463.451

___ 590.508 ___ 479.095

___ 624.146 ___ 44. 739

___ 657.784 ___ 510.383

- 691.422 - 526.027 725.06 541.671 FIGURE A3.3-15 TEMPERATURE DISTRIBUTION FOR FIRE ACCIDENT STEADY STATE, CROSS-SECTION MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Initial Temperatures ANSYS 80 MODAL SOLUTION STEP1 SUB =1 TIME bOB-OS TEMP AVG ils ys = U PowerLraphics EEACET1 AVIIESMSt SMN 119.627 SMX 180.987 119. 627

- 126.445

- 133.263 rRoz,2 l'1 flU 596 140.08 146.898 4O92 fl1_24 153.716 301556 160. S 34 331 BBS 167.351 362212 174.169 180.987 Tem peratures at the End of Fire ANSYS 8.0 NODAL SOLUTION ST EP = a ANSYS 8.0 STJB 4 NODAL SOLUTION TIflE TEIIP AVc)

TIME.25 rYS=O TEMP AVG Porrphics FLSYS=O EFACETI P owe r Cr apin Cs AVPES}Ia EFACETI SHN 151C15 S< 1Z38

- 161016 280672

- 400328 400328 519 83 SJ9 983 539639 639.639 7S929S 759.295 878951 878.951 998607 998.607 Tem peratures at 119.75 hrs after End of Fire A1SYS 8.0 MODAL SOLUTION ANSYS 8.0 STEP3 NODAL SOLUTION SUB 70 STEP3 TIM12O SUE =70 TEMP AVG)

TIMEJ20 TEMP Avc It S Y S = U psYs=O Por3erGraphThs Pphi EFACETI EFACETU AflESNat SNN =139 679 SNX =204 202

- 139679 146848

- 1S4 017 ZOO 07 161185 C

0Z5S C 168 35 Z90. 656 C 172S 320.851 182694 3SJ.O6 189863 381. Z2 197 032 411.437 204 FIGURE A3.3-16 TEMPERATURE DISTRIBUTION FOR FIRE ACCIDENT TRANSIENT RUN FOR LID-SEAL REGION MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A3.3-17 FINITE ELEMENT MODEL OF WE14X14 FUEL ASSEMBLY

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ANSYS 8.0 OCT 26 2006 10: 14:5w WEl4he NODAL SOLUTION C) c: C) 00 C) 0 STE P=5 (Th SUB =1 TIME=5 TEMP (AVG) fl fl fl fl

\,___ '--- \,___ '-- \__ RSYS=0

'___/

owe rGraphics EFACET=1 (Th ( (Th ('\ ( ( (Th AVRE S Mat

\ r\ )\ )\\ %___ )\ --- I SNN =500 fl

--- --- --- SNX =540.29 500 504.477 0 0 0 508. 53 513.43 517.907 522.383 fl 526.86 0 U 0' 0 531.337 535. 13 540.29 000 \____ '- n \____ 0\____

(\ (\ (\ (\ (\ (\ (

NX )

)

_1

)

_) ' I FIGURE A3.3-18 TEMPERATURE PLOT FOR WE14X14 FA IN HELIUM

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A3.3-19 COMPARISON OF TRANSVERSE EFFECTIVE CONDUCTIVITIES

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 I:

18 D - D 7 \

H I 124 8.4OD I I I nI 1

) L ) )

5 0

4 0

3 0

2 1

1

. Emitting Cask 0

1. Receiving Cask Covered Cask 2 1 3

0 18 4

5 0

18 FIGURE A3.3-20 STORAGE CONFIGURATION FOR TN-40HT CASKS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

__i__F.1_4L:'Il__i__ 'A!:4:

a ci 3D View Front View Side View FIGURE A3.3-21 STORAGE ARRAY MODEL FOR TN-40HT CASKS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Section 4, Mat # 4 Section 2, Mat # 2 Section 1, Mat # 1 Section 2, Mat # 2 FIGURE A3.3-22 CASK SECTION FOR EFFECTIVE CONDUCTIVITIES

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A3.3-23 TEMPERATURE DISTRIBUTION ON CASK OUTER SURFACES ON STORAGE PADS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Suriace Temperature = 1 92F Surface Temperature = 1 6OP I,p Storage Pad 70 111.091 152.181 193.272 234.363 O.545 131.G3G 172.727 213.917 254.O9 FIGURE A3.3-24 TEMPERATURE DISTRIBUTION ON STORAGE PAD SURFACE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 AN -

FE 2??

AN - .657 74

.61 61 272.rr7 7R4 2G 31:17.667 31:17.657 2: 1

- 33.

- :331 FIGURE A3.3-25 TEMPERATURE DISTRIBUTION ON HALVES OF CASK INNER SHELL (MIDDLE CASK)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 800 700 600 500 400 300 200 I I I I I I 0 5 10 15 20 25 30 35 40 Time, hrs FIGURE A3.3-26 MAXIMUM FUEL CLADDING TEMPERATURE DURING VACUUM DRYING OPERATIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ANSS 8.

NOV 8 18:16:35 NODAL S OLUT I ON STE P = 17 SUB =1 T IME = 3 4 TEMP (AVG:

RElY S =

Powe.rGraphics EFACET1 AVRESMat SMN = 16. .51:1.5 F

4R:LR47 4sn:I.c158

.54a.76 6I:I7.47 666.

FIGURE A3.3-27 TEMPERATURE DISTRIBUTION AT THE END OF VACUUM DRYING OPERATION

Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

/

1.75" (4.445 cm)

'I, Periodic Boundary at the Bottom of Model FIGURE A3.3-30 BASKET MODEL COMPARTMENT WALL (VIEW G)

(NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Periodic Boundary at the Top of Model P

FIGURE A3.3-31 BASKET MODEL COMPARTMENT WALL (VIEW F)

(NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 11111

, Borated Water Aluminum Plate Poison Plate UO Fuel Surrounded by Clad 11111 2

Stainless Steel Tube FIGURE A3.3-32 BASKET MODEL COMPARTMENT WALL WITH FUEL ASSEMBLY (VIEW G)

(NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Stainless Steel Tie Plate FIGURE A3.3-33 BASKET MODEL COMPARTMENT WALL WITH FUEL ASSEMBLY (VIEW F)

(NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A3.3-34 BASKET COMPARTMENT WITH WE 14X14 FUEL ASSEMBLY (SECTION A)

(NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Aluminum + Poison Stainless Steel Plates Total Width, Compartment Width, Nominal = 0.4375" Nominal = 0.1875" FIGURE A3.3-35 BASKET COMPARTMENT WITH WE 14X14 FUEL ASSEMBLY (SECTION B)

(NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 237 238 239 240 225 217 218 219 220 230 224 213 214 215 216 229 235 223 209 210 211 212 228 236 222 205 206 207 208 227 221 I 201 I 202 I 203 I 204 I 226 231 232 233 234 Poison and Aluminum Plate 1 I Aluminum Plate Only FIGURE A3.3-36 FUEL POSITION AND POISON PLATE LOCATION IN THE TN-40HT BASKET (NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Structural Inner Shell (1.50")

Stainless Steel llllllllllllH Cask Cavity IR = 36.0 w This contains the basket with fuel assemblies and the rails Gamma Shield and Skin Shells ffifflffim Onnnnnnn Combined (7.65")

Onnnnnnnn Carbonsteel Water Neutron Shield (5.00")

External Water Reflector and Specular Boundary Conditions on all Four Sides FIGURE A3.3-37 TN-40HT CASK DESCRIPTION IN THE KENO MODEL (NOT TO SCALE)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A3.3-38 EXXON 14X14 TOP ROD FUEL ASSEMBLY (CENTERED) KENO MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 I'

MII FIGURE A3.3-39 WE 14X14 STANDARD FUEL ASSEMBLY (INWARD) KENO MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

'I i

ri_

1.1

!1_

j: 1 1!:! P. .!.G!IiJ -

FIGURE A3.3-40 TN-40HT DESIGN BASIS KENO MODEL WITHOUT CCS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 N

• P:I:H1iP:IU

- - 44 FIGURE A3.3-41 TN-40HT DESIGN BASIS KENO MODEL WITH CCS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A-1 Appendix A3A TN-40HT CRITICALITY EVALUATION COMPUTER INPUT

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A-2 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-1 A3A.1 Most Reactive Fuel Type Example Input File Input file: tn40hb-box-min-090.inp.

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-2 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-3 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-4 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-5 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-6 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-7 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-8 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-9 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.1-10 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A3A.2-1 A3A.2 Input Listing for Case with Maximum Calculated k eff Case with: Enrichment = 5.00 wt. % U-235, Soluble Boron = 2450 ppm, With CCs.

Input File: tn40hb-e500bp-1 00.inp PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.1-1 SECTION A4 STORAGE SYSTEM A4.1 LOCATION AND LAYOUT-The TN-40HT casks will be stored on the same site and use the same transport route as the TN-40 casks. Therefore the location and layout information provided in Section 4.1 is also applicable to the use of the TN-40HT casks.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-1 A4.2 STORAGE SITE The TN-40HT casks are designed is designed in accordance with the General Design Criteria set forth in 1 0CFR72(F) as discussed in Section 4.2. Additional details are provided below.

A4.2.1 STRUCTURES The discussion of the ISFSI structures in Section 4.2.1 is independent of cask design other than that the location of the cask drop accident analyses of the TN-40HT cask is in Section A8.2.8.

A4.2.1 .1 STATIC ANALYSIS The ISFSI reinforced concrete cask pad analysis described in Section 4.2.1.1 was reviewed for the incorporation of the TN-40HT cask parameters and for increasing the allowed cask weight to a maximum of 250,000 lbs. The increase in cask weight input was reviewed in combination with a 1.5 inch decrease in the cask base diameter for the TN-40HT cask. The existing analysis utilizing the TN-40 input parameters contains considerable design margin. For example, as described in Section 4.2.1.1, the cask pad flexure strength margin indicates a safety factor of 2.14.

The review concluded that the increase to the maximum cask weight of 3.9% ((250 kip -

240.7 kip) / 240.7 kip) greater than used in the analysis and the 1.6% ((91 - 89.5) /

91) decrease in the cask base diameter will have a negligible effect on the adequacy determination of the cask pad analysis and design, and a more rigorous reanalysis is not warranted given the large margins that exist.

A4.2.1 .2 DYNAMIC ANALYSIS The ISFSI concrete cask pad dynamic analysis described in Section 4.2.1.2 was reviewed for the incorporation of the TN-40HT cask parameters and for increasing the allowed cask weight to a maximum of 250,000 lbs. The increase in cask weight input was reviewed in combination with a 1.5 in decrease in the cask base diameter for the TN-40HT cask.

As noted in Section 4.2.1.2, ultimate static stresses govern the design, and dynamic seismic load combinations need not be further considered for the reinforced concrete and acceptable bearing stress design.

The review of the TN-40HT cask input section properties affecting dynamic analysis relative to the TN-40 properties indicated that the seismic accelerations at the center of gravity of the casks would not change significantly, and given the conservatism in the dynamic analysis, a reanalysis is not warranted. Therefore, the dynamic stability of the TN-40HT cask was determined from the governing seismic accelerations established by the existing dynamic analysis utilizing five modes for the controlling single cask configuration.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-2 The cask overturning moment resulting from the governing accelerations was compared to the restoring moment resulting from the TN-40HT parameters. The factor of safety is the ratio of the restoring moment to the overturning moment. The minimum factor of safety for overturning was calculated to be 1.35. The minimum factor of safety for sliding was calculated to be 1.14.

A4.2.2 STORAGE SITE LAYOUT The storage site layout described in Section 4.2.2 remains applicable with the use of the TN-40HT casks.

A4.2.3 STORAGE CASK DESCRIPTION This section summarizes the structural analysis of the TN-40HT cask. For purposes of structural analysis, the cask has been divided into four components: the cask body (consisting of containment vessel and gamma shield shell), the basket, the trunnions and the neutron shield outer shell.

A4.2.3.1 DESIGN BASIS A4.2.3.1.1 CASK BODY The cask body is described in Section A1. Drawings are included in Section A1 .5. The containment boundary inner shell and bottom inner plate material is SA-203 Gr. E, and shell flange and lid outer plate is SA-350 Gr. LF3 or SA-203 Gr. E. The shield shell and bottom shield material is SA-266 CL 2 or SA-516 Gr. 70. The lid shield plate material is SA-105 or SA-516 Gr. 70.

The TN-40HT containment vessel is designed, fabricated, examined and tested in accordance with the requirements of Section III, Subsection NB of the ASME Code (Reference 1) to the maximum practical extent. Alternatives taken to the ASME code are specified in Section A3.5. The containment boundary consists of the inner shell and bottom inner plate, shell flange, lid outer plate, lid bolts, vent and drain cover plates and bolts, bolts and the inner portions of the lid seal and the two lid penetrations (vent and drain).

The containment boundary welds are full penetration welds examined volumetrically by radiograph. These welds are also magnetic particle or liquid penetrant examined.

Acceptance standards are in accordance with Section III, Article NB-5000.

Other structural and structural attachment welds are examined by the liquid penetrant or the magnetic particle method in accordance with Section V of the ASME Code (Reference 1). The magnetic particle and liquid penetrant examination acceptance standards are in accordance with Section III, Subsection NF, Acticle NF-5000 (Reference 1).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-3 Seal welds are examined visually or by liquid penetrant or magnetic particle methods in accordance with Section V of the ASME Code (Reference 1). Electrodes, wire, and fluxes used for fabrication comply with the applicable requirements of the ASME Code,Section II, Part C (Reference 1).

The welding procedures, welders and weld operators are qualified in accordance with Section IX (and NB-4300 where required) of the ASME Code (Reference 1).

A4.2.3.1.2 BASKET The basket is a welded assembly of stainless steel boxes and is designed to accommodate 40, 14x14 PWR fuel assemblies. The stainless steel fuel compartments are fusion welded to Type 304 stainless steel structural plates, sandwiched between the box sections. The fusion welds are spaced intermittently along the box sections.

Neutron poison plates and aluminum plates are sandwiched between the sections of the stainless steel walls of the adjacent box and the adjacent stainless steel plates. The Type 304 stainless steel members are the primary structural components. The neutron poison plates provide criticality control and the aluminum plates provide a heat conduction path from the fuel assemblies to the cask shell.

Stainless steel transition rails and aluminum shim are oriented parallel to the axis of the basket and attached to the periphery of the basket to establish and maintain basket orientation and to support the basket.

Tangential alignment between the basket and cask cavity is maintained by a key at the perimeter of the basket. This key is designed to prevent the basket from rotating in the cask cavity wall under normal lateral inertial loadings.

A4.2.3.1 .3 TRUNNIONS The TN-40HT cask has two top and two bottom trunnions each of which are one piece forgings composed of SA-1 05 or SA-266 CL 2 or CL 4 carbon steel.

The two upper trunnions are designed to lift the loaded TN-40HT cask vertically, while the lower trunnions provide capability to rotate the cask prior to loading of spent fuel.

The upper trunnions are designed to meet the requirements of NUREG-0612 (Reference 21) for non-redundant lifting fixture. This is accomplished by evaluating the trunnions to the stress design factors required by ANSI N 14.6 (Reference 2), i.e.

capable of lifting 6 times and 10 times the cask weight without exceeding the yield and ultimate strengths of the material, respectively.

The trunnions are shown in Section A1 .5 drawings.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-4 A4.2.3.1 .4 OUTER SHELL The outer shell of the neutron shield consists of a cylindrical shell section with segmented closure plates at each end. Each segmented closure plate is welded together. The top and bottom closure plates are welded to the outer surface of the cask body shield shell. The outer shell provides an enclosure for the resin-filled aluminum containers and maintains the resin in the proper location with respect to the active length of the fuel assemblies in the cask cavity. The outer shell has no other structural function. The shell is painted carbon steel.

A4.2.3.1.5 TOP NEUTRON SHIELD The top neutron shield consists of a disk of commercial grade polypropylene enclosed in stainless steel. The top neutron shield is attached to and rests on the cask lid. It is protected from the environment by the protective cover.

A4.2.3.2 INDIVIDUAL LOAD CASES This section describes the analyses performed for the TN-40HT cask under the various loading conditions identified in Section A3.2. These loadings include all of the normal events that are expected to occur regularly. In addition, they include severe natural phenomena and man-induced low probability events postulated because of their potential impact on the immediate environs.

The TN-40HT cask loadings are summarized in Table A3.2-5 and described in Section A3. The loads selected for analysis of the cask are discussed in Section A3.2.5.3. Numerical values of these loads are listed in Tables A3.2-2 through A3.2-4.

The TN-40HT cask components have been evaluated using numerical analyses. Finite element models of the cask body and basket have been developed, and detailed computer analyses have been performed using the ANSYS computer program (Reference 3). The stress analysis of the lid bolts is performed based on the methodology of NUREG/CR-6007 (Reference 4). Other components such as trunnions are analyzed using conventional textbook methods. Table A4.2-1 lists the specific individual load cases analyzed for each major TN-40HT cask component. The sections describing the analyses and the tables listing the stress results, where applicable, are also indicated. TN-40HT cask components are not subjected to any significant cyclic loads such as pressure or temperature fluctuations resulting in an appreciable fatigue usage factor. Also, in the operating temperature range, the materials selected are not subject to significant creep.

A4.2.3.3 STRUCTURAL DESIGN CRITERIA This section describes the structural design criteria for the major components of the TN-40HT storage cask. The cask consists of four major types of components:

Containment Boundary Non-containment Structure

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-5

• Basket

• Trunnions The structural design criteria for these components are described below:

A4.2.3.3.1 CONTAINMENT BOUNDARY The containment boundary consists of the inner shell (both cylinder and bottom inner plate) and closure flange out to the seal seating surface and the lid assembly outer plate. The lid bolts and seals and vent and drain plate bolts and seals are also part of the containment boundary. The containment boundary is designed to the maximum practical extent as an ASME Class I component in accordance with the rules of the ASME Code,Section III, Subsection NB. Alternatives to the ASME Code are discussed in Section A3.5 The stresses due to each load are categorized as to the type of stress induced, e.g.,

membrane, bending, etc., and the classification of stress, e.g., primary, secondary, etc.

Stress limits for containment vessel components, other than bolts, for Normal (Design and Level A) and Hypothetical Accident (Level D) Loading Conditions are given in Table A4.2-2. The stress limits used for Level D conditions, determined on an elastic basis, are based on the entire structure (containment shell and shield shell material) resisting the accident load. Local yielding is permitted at the point of contact where the load is applied. If the elastic stress limits cannot be met, the plastic system analysis approach and acceptance criteria of Appendix F of Section III may be used.

The allowable stress limits for all bolts are based on NUREG/CR-6007 (Reference 4) and are listed in Table A4A4-3, Table A4A.4-4, Table A4A.5-1, and Table A4A.5-2.

The allowable stress intensity value, Sm, as defined by the Code, is taken at the maximum temperature calculated for each service load condition.

A4.2.3.3.2 NON-CONTAINMENT STRUCTURES Certain components such as the shield shell, the neutron shield outer shell and the trunnions are not part of the cask containment boundary but do have structural functions. These components, referred to as non-containment structures, are required to react to the containment or environmental loads and in some cases share loadings with the containment structure. The stress limits for the non-containment structures (excluding the basket and trunnions) are given in Table A4.2-3.

The top neutron shield and the radial neutron shielding including the carbon steel enclosure have not been designed to withstand all of the hypothetical accident loads.

The shielding may degrade during the fire or due to cask burial. Also there may be local damage due to tornado impacts. Therefore a bounding analysis assuming that the exterior neutron shielding is completely removed has been performed. This analysis shows that the site boundary accident dose rates are not exceeded. These accidents are described in Section A8.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-6 A4.2.3.3.3 BASKET The basket is designed, fabricated and inspected in accordance with the ASME Code Subsection NG (Reference 1) to the maximum practical extent. Alternatives to the ASME Code are discussed in Section A3.5.

The stress limits for the basket are summarized in Table A4.2-4. The basket fuel compartment wall thickness is established to meet heat transfer, nuclear criticality, and structural requirements. The basket structure must provide sufficient rigidity to maintain a subcritical configuration under the applied loads. The 304 stainless steel members in the TN-40HT basket are the primary structural components. The aluminum plates are the primary heat conductors and neutron poison plates provide the necessary criticality control.

The fusion welds between the stainless steel support bars and the stainless steel fuel compartments shall be qualified by testing. The required minimum tested capacity of the weld connection shall be based on a margin of safety (test to design) of 1.43 (see Appendix F, Section F-1 342 (c) of Reference 1), corrected for temperature difference between testing and basket operating conditions and the maximum weld load at any weld location in the basket.

A4.2.3.4 EVALUATION The stress calculations performed on the cask and basket are presented in Appendices A4A and A4B respectively. The off-normal loads are bound by normal loads and compared with normal load allowables. Finite element models of the cask body and basket have been developed, and detailed computer analyses have been performed using the ANSYS computer program (Reference 3). The stress analysis of the lid bolts is based on the methodology of NUREG/CR-6007 (Reference 4). Other components such as trunnions are analyzed using conventional textbook methods. Table A4.2-1 lists the specific individual load cases analyzed for each major cask component. The SAR sections where these analyses are described and the tables listing the stress results, where applicable, are also indicated.

Section A3.2 categorizes the loads for the cask body as indicated in Tables A3.2-5 through A3.2-8 into Normal (Level A) and Hypothetical Accident (Level D) Service Loadings. Table A3.2-9 and Table A3.2-1 0 lists the load combinations to be evaluated.

Table A4.2-5 and Table A4.2-6 summarize the combination of the cask body individuals loads evaluated for normal conditions and hypothetical accident conditions respectively.

Table A4.2-7 summarizes the basket load combinations. Each combination is a set of loads that are assumed to occur simultaneously.

Key dimensions for the TN-40HT cask body are shown in Figure A4.2-1.

A4.2.3.4.1 CONTAINMENT VESSEL The evaluation of the containment vessel stresses are summarized along with the evaluation of the gamma shield stress discussed in Section A4.2.3.4.2 below.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-7 A4.2.3.4.2 GAMMA SHIELDING The maximum nodal stress intensities in cask body components are given in Table A4.2-8 and Table A4.2-9 for normal and accident condition load combinations, respectively. Since trunnions were not modeled, pressure and thermal load nodal stress intensities are reported in outer shield cylinder at nodes close to the trunnion locations.

The following process is used for the evaluation of the load combination stress results presented in Table A4.2-8 and Table A4.2-9.

1. Compare the maximum nodal primary stress intensities (due to mechanical loads) with the membrane stress allowable and compute factors of safety. If no factor of safety is less than 1.5 (a conservatively selected number), then both the Pm & PL + Pb criteria are met, and the structure is deemed adequate for that loading condition.

2. Compare the maximum nodal primary plus secondary stress intensity (due to mechanical + thermal loads) with 3 S m allowable and compute factors of safety. If no factor of safety is less than 1.5 (a conservatively selected number), then PL +

Pb + Q criteria is met, and the structure is deemed adequate for that loading condition.

3. However, if a nodal stress intensity results in factor of safety less than 1.5, then the nodal stresses are linearized for comparison with the allowable stresses.

Using the above conservative process, normal and accident condition nodal stresses are as presented in Table A4.2-8 and Table A4.2-9, respectively. Only the nodal stresses, where factor of safety is less than 1.5, are linearized and evaluated in Table A4.2-1 0 for normal conditions. For accident condition, margins of safety in Table A4.2-9 are all higher than 1.5 for all load combinations.

Weld stresses for all normal and accident condition load combinations are evaluated in Table 4.2-11. Individual load nodal global forces FX, FY and FZ at weld location nodes are summed up by ANSYS Postprocessor. Weld stresses are calculated for the 2 1/2 summed tensile force (FY) and shear force [FX2 + FZ ] . Weld 1, 2 and 3 locations are defined in Figure A4.2-2, Figure A4.2-3, and Figure A4.2-4 respectively.

• Weld-1 (Between Bottom Plate and Gamma Cylinder), 1.5 inch groove weld

• Weld-2 (Between Flange and Gamma Cylinder), 1 inch groove weld

• Weld-3 (Between Lid and Outer Shield Cylinder), 3/4 inch groove and 1/4 inch fillet weld The results of the analysis show that stresses in TN-40HT cask for normal and accident condition loads are below allowable limits with adequate factors of safety. This ensures the structural integrity of TN-40HT cask under normal and accident storage loads.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-8 A4.2.3.4.3 LID BOLTS The stress intensities in the lid bolts as calculated in Appendix A4A are summarized in Table A4.2-12. The stresses are well below the allowables.

A4.2.3.4.4 BASKET Analyses for individual basket loads for both normal and accident conditions are provided in Appendix A4B. These individual loads are combined as shown in Table A4.2-7. A summary of the basket stresses for the bounding normal and accident conditions is given in Table A4.2-13 and Table A4.2-14, respectively.

During a hypothetical accident end drop, the fuel assemblies and fuel compartments are forced against the bottom of the cask. It is important to note that, for any vertical loading, the fuel assemblies react directly against the bottom or top end of the cask and not through the basket structure as in lateral loading. It is the dead weight of the basket only that causes axial compressive stress during an end drop. Axial compressive stresses were conservatively computed in Appendix A4B by assuming that all load acts on the compartment tubes during an end drop. During the end drop, the rail supports its own weight by contact with the bottom or top of the cask. Therefore there will not be any stress in rail studs during an end drop.

A4.2.3.4.5 TRUNNIONS The two upper trunnions are used for lifting the cask and are designed to meet the requirements of NUREG-0612 (Reference 21) for non-redundant lifting fixture. This is accomplished by evaluating the trunnions to the stress design factors required by ANSI N14.6 (Reference 2), i.e. capable of lifting 6 times and 10 times the cask weight without exceeding the yield and ultimate strengths of the material, respectively. The upper trunnion stresses are summarized in Tables A4.2-15 and Table A4.2-16.

The local stresses in the cask body at the trunnion locations due to the loadings applied through the trunnions are not included in the result tables reported in Section A4.2.3.4.2. These local stresses are superimposed on the ANSYS stress results for the cases where the inertial lifting loads are reacted at the trunnions. The local stresses are calculated in accordance with the methodology of WRC Bulletin 107 (Reference 5) which is based on the Bijlaard analysis for local stresses in cylindrical shells due to external loadings. The stress intensity due to the local trunnion loading are combined with the finite element results at the trunnion attachment locations and presented in Table A4A.6-3.

The lower trunnions are not relied upon to support the cask when loaded with fuel, thus their structural analysis is not presented.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-9 A4.2.3.4.6 OUTER SHELL The neutron shield outer shell stress analyses are summarized in Table A4.2-17. The shell stresses are the highest when the cask is vertical and subjected to 3g inertia load and 25 psig internal pressure. Stresses in the shell will be much lower during normal storage of the TN-40HT cask on the ISFSI pad. The outer shell is not analyzed under tornado missile loading, but it could be damaged by either Missile A or Missile B, as defined in Section A3.2.1 .2. The effect of any damage to the outer shell is bounded by the cask body structural evaluation where the outer shell is assumed to be completely removed.

A4.2.3.5 MATERIAL DURABILITY Materials must maintain the ability to perform their safety functions over at least the casks 25 year lifetime under the casks thermal, radiological, corrosion, and stress environment.

Metallic components Gamma radiation has no significant effect on metals. The effect of fast neutron irradiation of metals is a function of the integrated fast neutron flux, which is on the order of 1014 n/cm2 inside the TN-40HT cask after 25 years. Studies on fast neutron damage in aluminum, stainless steel, and low alloy steels rarely evaluate damage below 10 17 n/cm2 because it is not significant (Reference 14). Extrapolation of the data available down to the 1014 range confirms that there will be virtually no neutron damage to any of the TN-40HT cask metallic components.

The effect of the TN-40HT cask temperature environment on the required structural properties is accounted for in the structural evaluation. There is no long term degradation of metals in the TN-40HT cask temperature environment. The effect of creep at temperature is the basis for establishing the seal temperature limits.

The cask exterior carbon steel components are protected from corrosion by the paint (epoxy, acrylic urethane, or equivalent enamel coating). A thermal spray coating may be applied before painting. The interior is protected by a thermal spray coating during loading, and by the helium environment inside the cask during storage. The aluminum, carbon steel, neutron absorber, and stainless steel components are not subject to significant corrosion as discussed in Section A4.2.3.6.

The neutron absorbers (Boral, borated aluminum, or metal matrix composites) consist of aluminum with boron added in the inert form of boron carbide, aluminum diboride, or titanium diboride. The durability of these materials in the dry storage thermal and radiation environment is similar to that of aluminum.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-10 Non-metallic components:

The radial neutron shield resin is inert with respect to water, and the fire retardant mineral fill makes it self-extinguishing. Furthermore, the resin is contained in aluminum tubes inside a steel shell, so that the material is retained in place, and isolated from both water and from sources of ignition.

Elastomer o-rings or gaskets in the weather cover, quick disconnects, drain tube, and pressure relief valve are not important to safety. The quick disconnects are not part of the containment boundary.

Radiation levels and temperature on the cask exterior are not high enough to damage the paint. This is confirmed by dry cask experience. Paint is subject to routine maintenance and touch-up.

The polypropylene in the top neutron shield is slow burning to non-burning according to Table 24, Section 1 of the Handbook of Plastics and Elastomers (Reference 11).

Polypropylene is inert with respect to water. Furthermore, the weather protective cover and stainless steel enclosure isolates the top neutron shield material from sources of ignition and from water.

A4.2.3.6 CHEMICAL AND GALVANIC REACTIONS This section discusses the chemical, galvanic or other reactions of the TN-40HT cask materials that might occur during any phase of loading, unloading, handling or storage.

The TN-40HT cask components are exposed to the following environments:

• During loading and unloading, the casks are submerged in borated pool water.

This affects the interior and exterior surfaces of the cask body, lid and the basket.

The protective cover, the top neutron shield, and the overpressure system are not submerged in the spent fuel pool. The casks are only kept in the spent fuel pool for a short period of time. Upon removal, the casks are drained, dried, and then backfilled with helium.

• During handling and storage, the exterior of the cask is exposed to normal environmental conditions of temperature, rain, snow, etc. The exterior surfaces with the exception of stainless steel components and trunnion bearing surfaces are protected from environmental exposure by an epoxy, acrylic urethane, or equivalent paint. Thermal spray of the exterior cask surface prior to painting is optional.

• During storage, the interior of the cask is exposed to an inert helium environment. The helium environment does not support the occurrence of chemical or galvanic reactions because moisture or oxygen must be present for

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-1 1 corrosion to occur. The cask is thoroughly dried before storage by a vacuum drying process. It is then sealed and backfilled with helium.

• The radial neutron shielding materials and the aluminum resin boxes are sealed during all normal operations. The free volume in the sealed region is very small.

The resin material is inert after it has cured and does not affect the aluminum boxes or the carbon steel housing.

A4.2.3.6.1 CASK INTERIOR The TN-40HT cask materials are shown in the Parts List on the Drawings included in Section A1 .5. The containment vessel is made from SA-203 Grade E and SA-350 LF3.

This low-alloy carbon steel is grit blasted and coated with a Zn/Al metallic spray for corrosion protection.

The aluminum metallic spray coating is subject to the following service environments:

• After fabrication, closed, and shipped under air.

• At fuel loading, borated spent fuel pool water for a short duration.

• Vacuum-dried and helium backfilled for storage lifetime of 25 years or more.

• At fuel removal, it may again be exposed to borated spent fuel pool water for a short duration.

The coating is not subject to abrasion except for the one time insertion of the basket.

All sealing surfaces are stainless steel clad by weld overlay. The metallic seals have a stainless steel liner and an aluminum jacket.

The basket is assembled from SA-240 Type 304 stainless steel fuel compartments which are joined to SA-240 Type 304 stainless steel strips by a fusion welding process.

Aluminum and neutron absorber plates fit between the fuel compartments. They are not welded or bolted to the stainless steel, but are held in place by the geometry of the compartments and strips. Transition rails made from SA-240 Type 304 stainless steel and aluminum are bolted to the basket.

A4.2.3.6.2 CASK EXTERIOR The exterior of the cask is carbon steel. The exterior of the cask, with the exception of the trunnion bearing surfaces, is painted using an epoxy, acrylic urethane, or equivalent enamel coating. The paint is selected to be compatible with the pool water and easy to decontaminate. Thermal spray coating prior to painting is optional.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-12 A4.2.3.6.3 LUBRICANTS AND CLEANING AGENTS Neolube, Loctite N-5000, or equivalent may be used to coat the threads and bolt shoulders of the TN-40HT cask closure bolts. Loctite N-5000 or equivalent may be used to coat the contact areas of the top and bottom trunnions prior to lifting operations.

The lubricant shall be removed prior to immersing the cask into the spent fuel pool unless the lubricant has been approved for compatibility with the spent fuel pool water.

The cask and basket are cleaned at the fabricator in accordance with approved procedures. The cleaning agents and lubricants have no significant effect on the cask materials and their important to safety functions.

A4.2.3.6.4 HYDROGEN GENERATION Prairie Islands report to the NRC (Reference 7) in response to NRC Bulletin 96-04 demonstrates that galvanic reactions in hydrogen generation are insignificant for the TN-40 cask. This report is also applicable for the TN-40HT cask.

A4.2.3.6.5 EFFECT OF GALVANIC REACTIONS ON THE PERFORMANCE OF THE CASK There are no significant reactions that could reduce the overall integrity of the cask or its contents during storage. The period of immersion in pool water is too short, and any oxidizing gases remaining after vacuum drying and helium backfill is too small to cause corrosion that could have significant effect on the fuel cladding, neutron absorber integrity, or the basket and cask structural performance.

There are no reactions that would cause binding of the mechanical surfaces or the fuel to basket compartment boxes due to galvanic or chemical reactions.

The stainless steel, aluminum, neutron absorber and thermal spray are negligibly affected by the short term exposure to borated water during loading. The three acceptable neutron absorber materials, Boral, borated aluminum, and metal matrix composites, are all aluminum-based, with the addition of boron in the inert form of boron carbide, aluminum diboride, or titanium diboride. The corrosion behavior of these materials is bounded by Boral because of its porous core.

While formation of blisters in Boral during vacuum drying and heating has been reported, this has not been associated with displacement of the Boral core material containing the boron carbide and therefore has no effect on the Boral criticality safety design function (Reference 8). Furthermore, in the TN-40HT cask, the Boral is captured between the structural basket components, including 3/16 inch thick walls of the fuel compartments, to provide it with added mechanical support and durability.

The outer aluminum lid seals may experience some combination of crevice and galvanic corrosion if they are exposed to water for an extended period. However, this would affect only the outer (non-containment) seal, and the overpressure monitoring system would detect any significant leakage.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-13 There is no significant degradation of any important to safety components caused directly by the effects of the reactions or by the effects of the reactions combined with the effects of long term exposure of the materials to neutron or gamma radiation, high temperatures, or other possible ambient or operating conditions.

A4.2.3.7 DIFFERENTIAL THERMAL EXPANSION This section determines the thermal growths among components of fuel cladding, basket, and cask inner shell in TN-40HT cask. This thermal expansion evaluation covers events of vacuum drying, normal/off-normal and accident storage conditions.

A thermal evaluation of the cask was performed in Section A3.3.2.2 to determine the maximum temperature of the cask components under normal conditions. The analysis considers maximum decay heat and maximum solar heat loading.

The temperature dependent thermal expansion coefficients for structural materials used in this calculation are taken from Table A4.2-1 8 for the cask and Table A4.2-1 9 for the basket. The reference temperature for these materials is 70 °F. The thermal expansion

-6 -6 coefficient of zircaloy (including ZIRLO) is 6.72x1 0 (m/m-K) (3.73x1 0 in/in-°F) for temperatures between 300 and 1,073 K (80 to 1,472 °F) as shown in (Reference 9),

Figure B-4.3. The reference temperature for zircaloy is 300 K (80 °F) (Reference 9).

Thermal Expansion between the Length of Fuel Assembl y and Cask Cavity The bounding temperatures for the fuel cladding and cask shell under different cases were obtained from Section A3.3.2.2 and are listed in Table A4.2-20.

The hot lengths of fuel assembly and cask cavity are calculated as follows.

The length of the spent fuel assembly when hot is:

L F,hot = L F + [L Z x a Z ( T max, F - T ref,Z) + L SS x a SS ( T max, F - T ref,SS)] + L irr Where:

L F,hot = Hot length of PWR fuel assembly L F = Total length of fuel assembly at room temperature = 161.3 in.

L Z = Length of Zircaloy guide tube =144 in. (conservatively use active fuel length).

aZ = Zircaloy axial coefficient of thermal expansion = 3.73E-6 in/in-°F (Reference 9)

L SS = Length of stainless steel per fuel assembly = L F - L z = 17.3 in.

a SS = Stainless steel coefficient of thermal expansion (in/in-°F; interpolated using data in Table A4.2-19)

T max,F = Maximum fuel cladding temperature (°F)

T ref,z = Reference temperature for Zircaloy = 80 °F (Reference 9)

T ref,SS = Reference temperature for stainless steel = 70 °F L irr = Irradiation growth of the spent fuel assembly = 1.25 in.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-14 The hot length of the cask cavity is:

L C,hot = L C [1 + a LS ( T avg,C - T ref,C)]

Where:

L C,hot = Hot length of cask cavity L C = Cold length of cask cavity = 163 in.

a LS = Coefficient of thermal expansion for low alloy steel (in/in-°F; interpolated using data in Table A4.2-1 8)

T avg.,C = Average temperature of cask inner shell and gamma shield (°F)

T ref,C = Reference temperature for low alloy steel = 70 °F The hot gap between the fuel assembly and cask cavity is:

G hot = L C,hot - L F,hot The calculated hot lengths and gaps for differential length growth between the fuel cladding and the cask cavity are listed in Table A4.2-20.

As seen in Table A4.2-20, the hot gaps are larger than zero. Therefore, adequate clearance has been provided between the spent fuel assemblies and the cask cavity length to permit free thermal expansion.

Thermal Expansion between the OD of Basket and ID of Cask The average temperatures for the basket, aluminum shim, and cask inner shell at the hottest cross section are retrieved from the result files of thermal analysis performed in Section A3.3.2.2 for storage conditions. Only the stainless steel components of the basket and rails are considered to calculate the average temperature of the basket.

These temperatures are used to calculate the differential diametrical growth between the basket and the cask inner shell for storage conditions. These temperatures are listed in Table A4.2-21.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-15 The nominal diametrical cold gap between the basket rails and the cask inner shell is 0.30 in. The size of this gap decreases at elevated temperatures during vacuum drying.

Additionally, the maximum temperature of the basket decreases as the gap shrinks. In the worst case of expansion, the gap between the basket and cask inner shell shrinks to zero. Expansion stresses result if the temperature causes additional basket growth. To investigate this worst case condition, the thermal model for vacuum drying described in Section A3.3.2.2 is re-run assuming a zero gap between the basket and the cask inner shell. Reduction of the gap size to zero is implemented in the model by giving a very high conductivity (1000 Btu/hr-in-°F) to the elements in the gap region. To bound the problem and determine the maximum achievable temperatures for basket growth, steady state conditions are considered for the vacuum drying model. An ANSYS macro is used to retrieve the average temperatures for the steady state, vacuum drying model with a zero gap. These temperatures are listed in Table A4.2-21.

The hot diameters of the basket and cask inner shell are calculated as follows.

The largest outer diameter of the hot basket is:

OD B,hot = OD B + [ L SS,B x a SS ( Tavg , B - T ref) + L Sh x a Al ( T avg,Sh - T ref)]

Where:

OD B,hot = Hot outer diameter of the basket OD B = Maximum cold outer diameter of the basket = 72 - 0.30 = 71.7 in.

L SS,B = Length of basket at 0-180 direction = OD B - 2x0.46 = 70.83 in.

L Sh = Length of aluminum shim = 2x0.46 = 0.92 in.

a SS = Stainless steel axial coefficient of thermal expansion (in/in-°F; interpolated using data in Table A4.2-1 9) a Al = Aluminum coefficient of thermal expansion (in/in-°F; interpolated using data in Table A4.2-19)

T avg,B = Average basket temperature at the hottest cross section (°F)

T avg,Sh = Average shim temperature at the hottest cross section (°F)

T ref = Reference temperature for stainless steel and aluminum alloys = 70 °F The inner diameter of the hot cask is:

ID c,hot = ID c [1 + a LS ( T avg,C - T ref)]

Where:

ID C,hot = Hot inner diameter of cask cavity ID C = Cold inner diameter cask cavity = 72 in.

a LS = Coefficient of thermal expansion for low alloy steel (in/in-°F; interpolated using data in Table A4.2-18)

T avg,C = Average cask inner shell and Gamma Shield temperature at hottest cross section (°F)

T ref = Reference temperature for low alloy steel = 70 °F

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-16 The hot gap between the basket outer diameter and cask inner diameter is:

G hot = ID C,hot - OD B,hot The calculated hot diameters and gaps for differential diametrical growth between the basket outer diameter and cask inner diameter are listed in Table A4.2-21.

As seen in Table A4.2-21, the hot gaps are larger than zero. Therefore, adequate clearance has been provided between the basket and the cask inner shell to permit free thermal expansion of basket diameter.

Thermal Expansion between Basket Stainless Steel Su pport Bars and Aluminum Plates along Plate Height A cold gap of 0.07 in. is considered between the stainless steel support plates and aluminum plates in the basket along the plate height. The average temperatures for the fuel compartments and aluminum plates at the hottest cross section are retrieved from the result files of thermal analyses performed in Section A3.3.2.2 for storage and vacuum drying conditions. The average temperatures for vacuum drying conditions are retrieved at 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> after the start of drainage. These temperatures are used to calculate the differential growth between the aluminum plates and the space between two stainless steel support bars. These temperatures are listed in Table A4.2-22.

The space between two stainless steel support bars when hot is:

H C,hot = H C [1 + a SS ( T avg,C - T ref)]

Where:

H C,hot = Hot space between two stainless steel support bars H C = Cold space between two SS support bars = 13.25 in.

a SS = Coefficient of thermal expansion for stainless steel (in/in-°F; interpolated using data in Table A4.2-19.

T avg,C = Average fuel compartment temperature at hottest cross section (°F)

T ref = Reference temperature for stainless steel = 70 °F The hot height of aluminum plate is:

H Al,hot = H Al [1 + a Al ( T avg,Al - T ref)]

Where:

H Al,hot = Hot height of aluminum plate H Al = Cold height of aluminum plate = 13.18 in.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-17 a Al = Coefficient of thermal expansion for aluminum alloys (in/in-°F; interpolated using data in Table A4.2-1 9)

T avg,Al = Average aluminum plate temperature at hottest cross section (°F)

T ref = Reference temperature for stainless steel = 70 °F The hot gap between the basket aluminum plate and two stainless steel bars is:

G hot = H C,hot - H Al,hot The calculated hot dimensions and gaps for differential growth between the aluminum plates and the space between two stainless steel support bars are listed in Table A4.2-22 for both 0-180 and 90-270 directions.

As seen in Table A4.2-22, the hot gaps are larger than zero. Therefore, adequate clearance has been provided between the aluminum plates and stainless steel support bars to permit free thermal expansion along the height of aluminum plates.

Thermal Expansion between Basket Stainless Steel Su pport Bars and Aluminum Plates along Plate Length In order to maintain free thermal expansion for basket plates, the edge of an aluminum plate notch should not hit the stainless steel bar located in that notch. Since the aluminum expansion coefficient is larger than stainless steel expansion coefficient, the smallest gap between the notch edge and the stainless steel bar after thermal expansion is located at the furthest notch from the centerline of the aluminum plate.

The average temperatures for the basket aluminum plates and stainless steel support bars at the hottest cross section are retrieved from the result files of thermal analyses performed in Section A3.3.2.2 for storage and vacuum drying conditions. The average temperatures for vacuum drying conditions are retrieved at 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> after the start of drainage. These temperatures are used to calculate the differential growth between the aluminum plates and the stainless steel support bars along the plate length. An ANSYS macro is used to retrieve the average temperatures. These temperatures are listed in Table A4.2-23.

The cold distance between the centerline of the aluminum plate and the edge of the furthest notch is:

L Al,270 = 3 x 8.86 - 0.75/2 = 26.205 in. for 90-270 direction L Al,180 = 2.5 x 8.86 - 0.75/2 = 21.775 in. for 0-180 direction The cold distance between the centerline of the aluminum plate and the edge of the stainless steel plate in the furthest notch is:

L SS,270 = L Al,270 + (0.75-7/16)/2 = 26.361 in. for 90-270 direction L SS,180 = L Al,180 + (0.75-7/16)/2 = 21.931 in. for 0-180 direction

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-18 The hot dimensions for the above distances are:

L Al,180,hot = L Al,180 (1 + a Al ( T avg,Al, 180 - T ref)]

L Al,270,hot = L Al,270 (1 + a Al ( T avg,Al,270 - T ref)]

L SS,180,hot = L SS,180 (1 + a SS ( T avg,SS,180 - T ref)]

L SS,270,hot = L SS,270 (1 + a SS ( T avg,SS,270 - T ref)]

Where:

a Al = Coefficient of thermal expansion for aluminum alloys (in/in-°F; interpolated using data in Table A4.2-1 9) a SS = Coefficient of thermal expansion for stainless steel (in/in-°F; interpolated using data in Table A4.2-1 9)

T avg,xx,yyy = Average temperature of xx plate in yyy direction at hottest cross section (°F)

T ref = Reference temperature for stainless steel = 70 °F The hot gap between the edge of the notch and the stainless steel plate is:

G hot = L Al,180,hot - L SS,180,hot or G hot = L Al,270,hot - L SS,270,hot The calculated hot distances and gaps between the notch edges of the aluminum plates and the stainless steel bars are listed in Table A4.2-23 for both 0-180 and 90-270 directions.

As seen in Table A4.2-23, the hot gaps are larger than zero. Therefore, adequate clearance has been provided between the aluminum plate notch and the stainless steel support bars to permit free thermal expansion along the plates.

Summary of Differential Thermal Expansion Analysis The hot gaps calculated between the TN-40HT cask components after thermal expansion are summarized in Table A4.2-24.

As seen in Table A4.2-24, the hot gaps are larger than zero. Therefore, adequate clearance has been provided between the TN-40HT cask components to permit free thermal expansion.

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-20 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-21 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 A4.2.3.8.3 MATERIAL PROPERTIES OF FUEL CLADDING The fuel cladding is evaluated based on the mechanical properties obtained from Reference 17 which provides expressions to calculate the modulus of elasticity and yield strength for both Zircaloy-2 (BWR cladding) and Zircaloy-4 (PWR cladding).

These expressions were derived from correlations of experimental results of several different investigations. Assumptions used include the following:

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 Temperature is a significant factor in derivation of Zircaloy properties. These properties are calculated over a range of temperatures for Zircaloy-4. An example calculation is carried out below for Zircaloy-4 (PWR cladding) at 750° F.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-22 Modulus of Elasticity (Pag e 4 of Reference 17) 7 10880 -5.475*10 T+K 1 +K 2 E=

K3 Where:

E = elastic modulus, Pa T = temperature, K

= 672.052K (750° F) 11 8 K

1

=(6.61x 10 + 5.912x 10 T)D 9

= 1.270 x 10 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 Where:

CW = cold work, unitless ratio of areas (valid between 0 and 0.75)

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 25 K 3 =0.88+0.12exp(-/10 )

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 c1 = fast neutron fluence, n/m 2 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 Substituting these values into the expression for E above:

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 Yield Stress (Eq uation 3, Page 3 of Reference 17)

K s 3 yLEnL10 JL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-23 Strength coefficient K = K(T) ( 1 +K(CW)+K(F)) / K(Zry)

Where:

9 5 3 2 K(T)= 1 .17628x 10 + 4.54859x 10 T -3.28185* 10 T + 1.72752 T3 8

= 5.24x 10 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 K(F)= 0.731995 for cl > 7.5 X 1025 n/m 2

K ( Zry) =1.0 for Zircaloy-4 Substituting these values into the expression for K above:

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 Strain Hardening Exponent n = n ( T ) n ( F) / n ( Zry)

Where:

n = strain hardening exponent n ( T) = -9.490

• 1 0 T -1.992

• 1 0 T T 10 3 2

+1.165 x 1 0 6 2 + 9.588 x 1 0 419.4< T <1099.0772 K n ()=1.608953 F >7.5*10 25 n/m 2 n ( Zry) =1.0 for Zircaloy-4 Substituting these values into the expression for K above:

n= 0.07938 1.608953/1.0=0.1277

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-24 Strain Rate PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 Strain Rate Exponent m =0.015 T <750 K m = strain rate exponent The values calculated above can then be inserted into the following expression for yield stress, Uy:

1 m

s= I K e PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390 Calculated values for Modulus of Elasticity and Yield Stress The expressions above are used to calculate the modulus of elasticity E and the yield stress yover a range of temperatures. The result for Zircaloy-4 (PWR) is presented in Table A4.2-25, Figure A4.2-6 and Figure A4.2-7. Note that the figures included calculated data points not summarized in the Table.

ZIRLO vs Zircaloy-4 Reference 22 states that ZIRLO and Zircaloy-4 alloys are very similar in terms stress/strain characteristics. Therefore Zircaloy-4 properties above are adequate for modeling ZIRLO cladding.

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-26 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-28 A4.2.3.

8.6 CONCLUSION

From the above results, for an accident condition bottom end drop of a fuel assembly inside a TN-40HT cask, the maximum total strain remains in the elastic range. Since plastic deformation does not occur in this case, it can be concluded that the fuel assembly cladding will not fail in the event of an 18 inch bottom end drop accident.

A4.2.3.9 THERMAL STRESS OF FUEL CLADDING DUE TO UNLOADING OPERATIONS To evaluate the effects of the thermal loads on the fuel cladding during unloading operations, the following assumptions are made:

• A conservative high maximum fuel cladding temperature of 700 °F and quench water temperature of 50 °F are used.

• The Fuel rod is assumed to be simply supported at both ends.

• The outer surface temperatures of the fuel cladding are conservatively assumed as shown in Figure A4.2-13. 50 °F (water), 212 °F (steam), and 700 °F (cladding) temperature occurs at three equal heights.

• The fuel cladding thickness and cladding outside diameter are reduced by 0.00270 inch to account for oxidation.

A4.2.3.9.1 FINITE ELEMENT MODEL The finite element model is shown in Figure A4.2-14. ANSYS (Reference 3) finite element Plane 55 and Plane 42 (Axisymmetric) are used for thermal and structural analysis respectively. The fuel rod with the thinnest cladding (WE14 x 14 STD) is modeled, as this will result in the largest temperature gradient across the cladding (temperatures are kept constant at the inner and outer surfaces). The cladding thickness is 0.0216 inches and the rod outer diameter is 0.4166 inches. A tube length of 2 inches is considered for the analysis such that maximum stresses are not affected by the boundary conditions.

A4.2.3.9.2 MATERIAL PROPERTIES The following material properties are used for the thermal and structural analysis:

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-29 Material Properties for Thermal Analysis Temp Conductivity

°F Btu/hr-in- °F 212 0.655 392 0.689 572 0.732 752 0.790 Material Properties for Structural Analysis Temp E (psi) a in/in- °F S y (psi) at

°F 750 °F 6

300 12.2 x 10 126,102 6

400 11.7 x 10 116,272 6

500 11.2 x 10 108,921

-6 0.404 600 10.7 x 10 6

3.73 x 10 102,512 6

700 10.2 x 10 95,793 6

750 9.93 x 10 92,000 A4.2.3.9.3 THERMAL ANALYSIS Steady state thermal analysis was conducted using the surface nodal temperatures as shown in Figure A4.2-1 3. The inside surface nodal temperatures are all assumed to be 700 °F , and the outside surface temperatures to conservatively represent the quench water temperature. The temperature distribution resulting from this analysis is shown in Figure A4.2-1 5.

A4.2.3.9.4 THERMAL STRESS ANALYSIS AND RESULTS A thermal stress analysis using the same model was conducted using the nodal temperatures obtained from the thermal analysis. The resulting nodal stress intensity distribution is shown in Figure A4.2-16. The maximum nodal stress intensity in the fuel cladding is 24.0 ksi. This stress is less than the yield strength of Zircaloy, which is 92 ksi at 750 °F.

A4.2.4 INSTRUMENTATION SYSTEM DESCRIPTION No important to safety instrumentation is required for the TN-40HT casks due to the passive nature of the ISFSI design.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.2-30 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.3-1 A4.3 TRANSPORT SYSTEM A4.3.1 FUNCTION The information in Section 4.3.1 is independent of cask design.

A4.3.2 COMPONENTS The information in Section 4.3.2 is independent of cask design.

A4.3.3 DESIGN BASIS AND SAFETY ASSURANCE The information in Section 4.3.3 is independent of cask design.

A4.3.4 DETERMINATION OF NATURAL FORCES ON A LOADED TRANSPORT VEHICLE The information in Section 4.3.4 is independent of cask design.

A4.3.4.1 STABILITY OF A LOADED TRANSPORT VEHICLE UNDER TORNADO LOADING The information in Section 4.3.4.1 is independent of cask design.

A4.3.4.1.1 WIND LOADING Comparing the dimensions and weight of a TN-40HT cask to a TN-40 cask shows that

1) the overall out side dimension is the same, 2) the TN-40HT cask is slightly shorter, and 3) the nominal weight of a TN-40HT cask is larger. Thus; 1) the clearance between the side of a TN-40HT cask and the vehicle side members is the same, 2) the center of gravity of the loaded transport vehicle is slightly lower, and 3) the weight of the loaded transport vehicle would be slightly more. Therefore the wind speed required to tip the transport vehicle with a loaded TN-40HT cask is more than that determined in Section 4.3.4.1.1 and the conclusion that a loaded transport vehicle will not tip as a result of the design tornado wind loads is applicable to the TN-40HT casks.

A4.3.4.1 .2 TORNADO MISSILE IMPACT Comparing the dimensions and weight of a TN-40HT cask to a TN-40 cask shows that

1) the TN-40HT cask is slightly shorter, and 2) the nominal weight of a TN-40HT cask is higher. Thus; 1) the center of gravity of the loaded transport vehicle is slightly lower and

2) the weight of the loaded transport vehicle would be slightly more. Therefore the distance the center of gravity would raise due to a tornado missile impact is less than that determined in Section 4.3.4.1.2 and the conclusion that a loaded transport vehicle will not tip as a result of the design tornado missile impact is applicable to the TN-40HT casks.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.3-2 A4.3.4.2 FLOOD LEVEL The information in Section 4.3.4.2 is independent of cask design.

A4.3.4.3 EARTHQUAKE Comparing the dimensions and weight of a TN-40HT cask to a TN-40 cask shows that

1) the TN-40HT cask is slightly shorter, and 2) the nominal weight of a TN-40HT cask is larger. Thus; 1) the center of gravity of the loaded transport vehicle is slightly lower and

2) the weight of the loaded transport vehicle would be slightly more. Therefore the g value necessary to tip the loaded transport vehicle is higher than that determined in Section 4.3.4.3 and the conclusion that a loaded transport vehicle will not tip as a result of the design seismic event is applicable to the TN-40HT casks.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.4-1 A4.4 OPERATING SYSTEMS A4.4.1 LOADING AND UNLOADING SYSTEMS A4.4.1.1 FUNCTION The information in Section 4.4.1.1 is independent of cask design.

A4.4.1.2 MAJOR COMPONENTS AND OPERATING CHARACTERISTICS The information in Section 4.4.1.2 is independent of cask design.

A4.4.1 .3 SAFETY CONSIDERATION AND CONTROLS The information in Section 4.4.1.3 is independent of cask design.

A4.4.2 DECONTAMINATION SYSTEM The information in Section 4.4.2 is independent of cask design.

A4.4.3 STORAGE CASK REPAIR AND MAINTENANCE Maintenance on the TN-40HT casks can be performed as described in Section A5.1 .3.3.

A4.4.4 UTILITY SUPPLIES AND SYSTEMS The TN-40HT storage casks are passive devices. No utility services are needed for operation of the casks.

A4.4.5 OTHER SYSTEMS A4.4.5.1 ELECTRICAL SYSTEMS The information in Section 4.4.5.1 is independent of cask design.

A4.4.5.2 ALARM SYSTEM The information in Section 4.4.5.2 remains applicable with the use of the TN-40HT casks.

A4.4.5.3 FIRE PROTECTION SYSTEM The information in Section 4.4.5.3 is independent of cask design.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.4-2 A4.4.5.4 VACUUM SYSTEMS The information in Section 4.4.5 remains applicable with the use of the TN-40HT casks.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.5-1 A4.5 CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTS The structures, components, and systems of the TN-40HT Dry Storage Casks are classified as important to safety or not important to safety in accordance with the criteria of 10 CFR 72.3. A tabulation of the major systems and components for a TN-40HT cask is shown in Table A4.5-1. Structures, components, and systems are classified as important to safety whose function is:

• to maintain the conditions required to store spent fuel safely,

• to prevent damage to the spent fuel cask during handling and storage, or

• to provide reasonable assurance that spent fuel can be received, handled, packaged, stored, and retrieved without undue risk to the health and safety of the public.

Items that are important to safety are further categorized using a graded quality approach based on the following definitions:

Category A Item Category A items are critical to safe operations. These items include systems and components whose failure could directly result in a condition adversely affecting public health and safety. The failure of a single item could cause loss of primary containment leading to release of radioactive material, loss of shielding, or an unsafe geometry compromising criticality control.

Category B Item Category B items have a major impact on safety. These items include systems and components whose failure or malfunction could indirectly result in a condition adversely affecting public health and safety. The failure of a Category B item, in conjunction with the failure of an additional item, could result in an unsafe condition.

Category C Item Category C items have a minor impact on safety. These items include systems and components whose failure or malfunction would not significantly reduce the packaging effectiveness and would not be likely to create a situation adversely affecting public health and safety.

The drawings contained in Section A1 .5 show the category classification for each part of a TN-40HT cask.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.5-2 A4.5.1 CONTAINMENT VESSEL The containment vessel components are classified as important to safety Category A since they serve as the primary confinement structure for the fuel assemblies. Failure of any of these components could lead to a release of radioactive material.

A4.5.2 PENETRATION GASKETS The metallic seals on the lid, vent port cover, and drain port cover are classified as important to safety Category A since the inner seal is part of the confinement boundary and a failure of a seal could lead to a release of radioactive material.

A4.5.3 SHIELDING The gamma and neutron shielding are classified as important to safety Category B.

During normal operations there is no credible failure mode for the gamma shield that would prevent it from fulfilling its design shielding function. Under accident conditions the analyses show that the gamma shield remains in place and thus still performs its shielding function. While the neutron shielding may be lost during a fire accident, the analyses show that the resultant dose rates still meet the applicable accident dose limits. The failure of an additional item is necessary for there to be an unsafe condition.

Therefore, these components are classified as Category B.

A4.5.4 PROTECTIVE COVER AND OVERPRESSURE SYSTEM The weather cover and overpressure system serve no safety function and are thus classified as not important to safety.

A4.5.5 CONCRETE STORAGE PADS The information in Section 4.5.5 remains applicable with the use of the TN-40HT casks except that the accidents are described and analyzed in Section A8.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.6-1 A4.6 DECOMMISSIONING PLAN The TN-40HT cask design features inherent ease and simplicity of decommissioning.

At the end of its service life, cask decommissioning will be preceded by one of the following options:

Option 1, the TN-40HT cask, including spent fuel in storage, could be shipped to either a monitored retrievable storage system (MRS) or a geological repository for final disposal, or Option 2, the spent fuel could be removed from the TN-40HT cask (either at the utility or at another off site location) and shipped in a DOE approved cask.

The first option does not require any decommissioning of the TN-40HT cask at Prairie Island Nuclear Generating Plant. No residual contamination is expected to be left behind on the concrete base pad. The base pad, fence, and periphery utility structures will require no decontamination or special handling after the last cask is removed. The ISFSI pad could be demolished with normal construction techniques.

The second option would require decontamination of the TN-40HT cask. The sources of contamination in the interior of the cask would primarily be crud left from the spent fuel pool water or crud from the spent fuel pins. These are expected to be low levels of contamination which could simply be removed with high pressure water spray. After decontamination, the TN-40HT cask could either be cut up for scrap or partially scrapped. For surface decontamination of the TN-40HT cask, electropolishing or chemical etching can be used to remove the contaminated surface of the cask if necessary.

Section 4.6 contains the cask activation analyses for the TN-40 cask to quantify the specific activities of the cask materials after 20 years of storage. The results of that analysis are shown in Tables 4.6-1 and 4.6-2.

Since the TN-40 cask and the TN-40HT cask are very similar in design, the TN-40 activation evaluation can be used to estimate the activity of the TN-40HT cask. Factors are determined that can be applied to results of the TN-40 activation analysis.

The cask bodies are very similar; the TN-40HT has a slightly thinner body and a slightly thicker neutron shield. The TN-40HT basket is heaver than the TN-40 basket with more stainless steel in the basket and the basket rails. The mass factors are shown in the table below:

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.6-2 TN-40 TN-40HT TN-40HT Zone Mass (kg) Mass (kg) Factor Basket__________ __________ __________

SS304 2,770 6,250 2.256 Body, Lid, Rails __________ __________ __________

C Steel (SA-105) 56,070 48,563 0.8661 C Steel (SA-203) 12,260 12,555 1.024 SS304 681 3,926 5.765 Neutron Shield __________ __________ __________

Resin 4,858 5,573 1.147 Shell, Prot Cover _________ _________ _________

C Steel (SA-516) 4,131 4,039 0.9778 The neutron source term for the high burnup fuel assembly in the TN-40HT cask is 7.59E+08 n/s, (Table A7.2-8). From Table 3.1-2, the neutron source for the TN-40 cask is 2.1 9E+08 n/s. Therefore, the activation flux in each of the zones (cask centerline, cavity wall, etc.) of the TN-40HT is predicted to be 3.466 times larger (7.59E+08/2. 1 9E+08) than the fluxes determined for the TN-40 (Table 4.6-1). Because the TN-40HT is conservatively analyzed to store fuel for 40 years even though the minimum design life is 25 years, an additional factor of 2 is applied, (TN-40 is 20 years).

Utilizing the nuclide activities reported for the TN-40 in Table 4.6-2 and the mass and activation flux factor shown above, the activities can be estimated for the TN-40HT cask. These values are listed in Table A4.6-1. Note the majority of the activated nuclides come from stainless steel.

To evaluate the TN-40HT cask and basket for disposal, the specific activity of the isotopes listed in Tables 1 and 2 of 10 CFR 61.55 is determined and compared with the limits for Class A waste in those tables.

It is expected that after the application of a surface decontamination method, the radiation levels will be below the acceptable limits of Regulatory Guide 1.86 (Reference 6). The results of the calculation, shown in Table A4.6-2, show that activation of TN-40HT will be far below the specific activity limits for both long and short lived nuclides for Class A waste. A detailed evaluation will be performed at the time of decommissioning to determine the appropriate mode of disposal.

The volume of waste material produced incidental to ISFSI decommissioning is expected to be limited to that resulting from the surface decontamination of the casks if the spent fuel assemblies must be removed.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.6-3 Furthermore, it is estimated that the cask materials will be only slightly activated as a result of their long term exposure to the relatively small neutron flux emanating from the spent fuel, and that the resultant activation level will be well below the allowable limits for general release of the casks as noncontrolled material. Therefore, it is anticipated that the casks, may be decommissioned from nuclear service by surface decontamination alone, which could be performed at the cask decontamination area in the Auxiliary Building.

The costs of decommissioning the ISFSI are expected to represent a small and negligible fraction of the cost of decommissioning the Prairie Island Nuclear Generating Plant.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.6-4 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.7-1 A

4.7 REFERENCES

American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Sections II, III, V and IX, and appendices 2004 through 2006 addenda.

2. American National Standards Institute, ANSI N 14.6, American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials, 1986.

3. ANSYS Engineering Analysis System, Users Manual for ANSYS Release. 8.0 and 8.1, Swanson Analysis Systems, Inc., Houston, PA.

4. NUREG/CR-6007 "Stress Analysis of Closure Bolts for Shipping Casks," By Mok, Fischer, and Hsu, Lawrence Livermore National Laboratory, 1992.

5. WRC Bulletin 107, March 1979 Revision Local Stresses in Spherical and Cylindrical Shells Due to External Loadings.

6. Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors."

7. Northern State Power Company, Prairie Island Nuclear Generating Plant, Response to NRC Bulletin 96-04, Docket No. 72-10, Materials License No. SNM-2506.

8. USNRC, Resolution of Generic Safety Issue 196, ADAMS Accession No.

MC063520459.

9. US NRC, NUREG/CR-0497 - TREE-1280, Rev. 2, MATPRO-Version 11 (Rev.

2) A Handbook of material Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior.

10. Aluminum Association Catalog, Aluminum Standards and Data, 1990.

11. Harper, Charles A., ed., Handbook of Plastics and Elastomers, McGraw-Hill, 1975.

12. H. E. Adkins, Jr., B. J. Koeppel, and D. T. Tang, Spent Nuclear Fuel Structural Response when Subject to an End Impact Accident, PVP-Vol. 483, San Diego, CA, July 25-29 2004.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4.7-2

13. NUREG 1864 A Pilot Probabilistic Risk Assessment of a Dry Cask Storage System at a Nuclear Power Plant.

14. U.S. NRC, Regulatory Guide 1.99, Rev. 2, Radiation Embrittlement of Reactor Vessel Materials, 1988.

15. T Sanders et al., A Method for Determining the Spent-Fuel Contribution to Transport Cask Containment Requirements, Sandia Report SAND90-2406, Sandia National Laboratories, Albuquerque, NM, 1992.

16. UCID - 21246, Dynamic Impact Effects on Spent Fuel Assemblies, Lawrence Livermore National Laboratory, October 20, 1987.

17. K. J. Geelhood and C. E. Beyer, PNNL Stress/Strain Correlation for Zircaloy, March 2005 (summarized in Mechanical Properties of Irradiated Zircaloy, Transactions of the American Nuclear Society, pp 707-708, November 2005).

18. LS-DYNA Keyword Users Manual, Volumes 1 & 2, Version 9.71s, Rev. 7600.398 August 17, 2006, Livermore Software Technology Corporation.

19. L. F. Van Swam, A. A. Strasser, J. D. Cook, and J. M. Burger, Behavior of Zircaloy-4 and Zirconium Linear Zircaloy-4 Cladding at High Burnup, Proceedings of the1 997 International Topical Meeting on LWR Fuel Performance, Portland, Oregon March 2-6, 1997.

20. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Vols. I-III, NUREG/CR-0200, Rev. 5 (ORNL/NUREG/CSD-2/R5), March 1997 (Standard Composition Library, Table M8.2.4).

21. NUREG-61 2 Control of Heavy Loads at Nuclear Power Plants, July 1980

22. Letter, Ashok Thadani (NRC) to S. R. Tritch (Westinghouse), Acceptance for Referencing of Topical Report WCAP-12610 Vantage+ Fuel Assembly Reference Core Report (TAC No. 77258) July 1, 1991.

23. Siegmann, Eric R., J. Kevin McCoy, Robert Howard, Cladding Evaluation in the Yucca Mountain Repository Performance Assessment, Material Research Society Symp. Proc. Vol. 608, 2000.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-1 INDIVIDUAL LOAD CASES ANALYZED Individual Individual Load Description SAR Section Stress Result

______________________________________________ _____________ Tables Cask Body ____________ ______________

Bolt Preload and Lid Seating Pressure A4A.3 (IL-1) A4A.3-2 Fabrication Stress A4A.3 (I L-2) A4A.3-2 1g Down (Cask vertical, supported at bottom) A4A.3 (IL-3) A4A.3-2 Internal Pressure (100 psig) A4A.3 (IL-4) A4A.3-2 External Pressure (25 psig) A4A.3 (IL-5) A4A.3-2 Thermal Stress Due to Hot Environment A4A.3 (IL-6) A4A.3-2 (100 oF ambient)

Thermal Stress Due to Cold Environment A4A.3 (IL-7) A4A.3-2

(-40 oF ambient) 3g on Top Trunnion Lifting Load (Calculating cask global A4A.3 (IL-8) A4A.3-2 stresses, Cask Vertical, 3g Up) _______________ __________________

Bounding Loads for Seismic, Tornado and Flood A4A.3 (IL-9) A4A.3-2 (1g Lateral (resultant) ^ 2g Down) _______________ __________________

Trunnion Local Stress due to 3g Lifting A4A.6 A4A.6-3 (calculating cask local stress at trunnion / cask shell interface) _______________ __________________

Lid _ Bolts _____________ _______________

Bolt Preload A4A.4-5 Gasket Seating Load (1399 lb/in.)

A4A.4 A4A.4-6 Pressure Load (100 psig)

A4A.4-7 Temperature Load (300 °F) _______________ __________________

Basket _____________ _______________

Normal Condition Bounding Loads A4B.15 A4B.1.5 (3g vertical ^ 3g lateral) _______________ A4B.1-6 Accident Condition End Drop (50g) A4B. 1.5 A4.2-1 4 Trunnions _____________ _______________

A4A.6-1 6g/ 10g Vertical Lifting Load A4A. 6

_________________________________ _________ A4A.6-2

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-2 CONTAINMENT VESSEL STRESS LIMITS Classification Stress Intensity Limit (1)

Normal (Level A) Conditions PmS m Pl 1.5 Sm

( Pm or Pl ) + Pb 1.5 Sm Shear Stress 0.6 Sm Bearing Stress Si,

( Pm or Pl ) + Pb + Q 3 Sm

( Pm or Pl ) + Pb + Q + F Sa (3)

Containment Bolt Normal (Level A) Conditions Tensile Stress, Ftb 2/3 Si, Shear Stress, Fvb 0.4 Si, Combined Stress Intensity, S.I. 0.9 Si, Interaction limit 2 s tb 2

t £ yb 1.0

+

2 2

'tb 'yb (2)

Hypothetical Accident (Level D)

Pm Smaller of 2.4 Sm or 0.7 Su Pl Smaller of 3.6 Sm or S u

( Pm or Pl ) + Pb Smaller of 3.6 Sm or Su Shear Stress 0.42 S u (3)

Containment Bolt Hypothetical Accident (Level D)

Tensile Stress, Ftb Minimum (0.7 S u , Si,)

Shear Stress, Fvb Minimum (0.42 S u , 0.6 Si,)

Combined Stress Intensity, S.I. Not Required Interaction Limit 2 s tb 2

t £ yb 2

+

2 1.0

'tb 'yb Notes:

1. Classifications and Stress Intensity Limits are as defined in ASME B&PV Code,Section III, Subsection NB.

2. Stress intensity limits are in accordance with ASME B&PV Code,Section III, Appendix F.

3. Bolt allowables are from Reference 4

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-3 NON-CONTAINMENT STRUCTURES STRESS LIMITS Classification Stress Intensity Limit (1)

Normal (Level A) Conditions PmS m Pl 1.5 Sm

( Pm + Pl ) + Pb 1.5 S m

( Pm + Pl ) + Pb + Q 3 Sm Shear Stress 0.60 Sm Bearing Stress Sy (2)

Hypothetical Accident (Level D)

Pm Smaller of 2.4 Sm or 0.7 S u Pl Smaller of 3.6 Sm or S u

( Pm + Pl ) + Pb Smaller of 3.6 Sm or S u Shear Stress 0.42 S u Notes:

1. Classifications and stress intensity limits are as defined in ASME B&PV Code,Section III, Subsection NB (Reference 1).

2. Stress intensity limits are in accordance with ASME B&PV Code,Section III, Appendix F (Reference 1).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-4 BASKET STRESS LIMITS Classification Stress Intensity Limit (1)

Normal (Level A) Conditions P mS m Pl 1.5 Sm

( Pm + Pl) + Pb 1.5 S m

( Pm + Pl) + Pb + Q 3 Sm

( Pm + Pl) + Pb + Q + F Sa Shear Stress 0.6 Sm (2)

Hypothetical Accident (Level D)

Pm Smaller of 2.4 Sm or 0.7 S u Pl Smaller of 3.6 Sm or S u

( Pm + Pl) + Pb Smaller of 3.6 Sm or S u Shear Stress Smaller of 0.42 S u or 2(0.6 S m )

Notes:

1. Classifications and stress intensity limits are as defined in ASME B&PV Code,Section III, Subsection NG (Reference 1).

2. Limits are in accordance with ASME B&PV Code,Section III, Appendix F (Reference 1).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-5

SUMMARY

OF LOAD COMBINATIONS FOR NORMAL CONDITIONS IL-10 3g Lifting IL-1 IL-2 IL-3 IL-4 IL-5 IL-6 IL-7 IL-8 Trunn.

Load Pre- Fabric- Gravity Internal External Thermal Thermal 3g Local Comb. load ation (1g) Pressure Pressure (Hot) (Cold) Lift. Stress N1 x x x x _______ _______ _______ ____ ______

N2 x x x x _______ _______ _____ ______

N3 x x x x x _______ _____ ______

N4 x x x x x _____ ______

(1 )

N5 x x x x X ______

(1 )

N6 x x x x X ______

(2)

N7 x x X (2)

N8 x x X Notes:

(1) Calculated cask global stresses due to 3g lifting load.

(2) Calculated cask local stresses at Trunnion/Shield shell due to 3g lifting load.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-6

SUMMARY

OF LOAD COMBINATIONS FOR ACCIDENT CONDITIONS IL-9 IL-3 IL-4 IL-5 Seismic, Load IL-1 IL-2 End Drop Internal External Tornado or Comb. Preload Fabrication (50g) Pressure Pressure Flood A1 x x x _________ _________ __________

A2 x x x x _________ __________

A3 x x x x __________

A4 x x x x A5 x x x x Note (1): The stress components for the end drop are generated by multiplying the stresses from the 1g load determined in IL-3 by 50.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-7 BASKET NORMAL AND ACCIDENT CONDITION LOAD COMBINATIONS IL-3 IL-4 IL-5 Normal Condition Load IL-2 3g Vertical, Thermal Thermal Combination 3g Lifting (1)

_____________________ ____________ 3g Lateral (Hot) (Cold)

N1 x ____________ _________ __________

N2 x x ____________

N3 x x N4 x __________ ___________

N5 x x ____________

N6 x x Note:

1. This load case not only bounds the normal and off-normal loads but also bounds the loads due to seismic, tornado, or flood and conservatively compares with normal condition load allowables.

Accident Loading IL-1 Condition 50g Vertical Bottom End Drop A1 x

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-8

SUMMARY

OF LOAD COMBINATION STRESSES FOR NORMAL CONDITIONS

_________ _______ Cask Component Nodal Stress Intensity (ksi) _______ ________

Inner Shell &

Bottom Lid Lid Gamma Bottom Top Bottom Load Stress Inner Shell Outer Shield Shield Shield Trunn. Trunn. Allow. Min.

Type 2

Cmb. P late Flange Plate plate Shell Plate Region Region (ksi) FOS Primary 14.07 (1)11.63 8.68 2.17 8.79 7.96 5.48 8.79 20.5 1.46 N1 Primary+

14.07 11.63 8.68 2.17 8.79 7.96 5.48 8.79 61.5 4.37 Secondary Primary 15.47 (1)12.29 8.22 0.88 6.61 4.37 4.99 6.61 20.5 1.33 N2 Primary+

15.47 12.29 8.22 0.88 6.61 4.37 4.99 6.61 61.5 3.98 Secondary Primary 14.07 (1)11.63 8.68 2.17 8.79 7.96 5.48 8.79 20.5 1.46 N3 Primary+

20.58 12.67 8.55 2.47 10.12 20.81 7.54 10.10 61.5 2.99 Secondary Primary 15.47 (1)12.29 8.22 0.88 6.61 4.37 4.99 6.61 20.5 1.33 N4 Primary+

19.44 13.35 7.95 1.05 10.79 15.24 6.74 10.65 61.5 3.16 Secondary Primary 12.98 11.43 8.69 2.09 9.23 14.52 (1)5.77 9.23 20.5 1.41 N5 Primary+

24.51 12.44 8.45 2.31 10.63 27.61 7.28 10.59 61.5 2.23 Secondary Primary 13.88 (1)12.09 8.22 0.93 7.06 7.89 5.34 7.06 20.5 1.48 N6 Primary+

23.09 13.12 7.95 1.07 10.67 21.94 6.47 10.53 61.5 2.66 Secondary Notes:

(1) These stresses result in factor of safety less than 1.5 and are linearized in Table A4.2-1 0 for revised factors of safety.

(2) Factor of safety for primary loads is calculated by dividing the membrane allowable by nodal stress intensity, hence very conservative.

(3) Primary stress is the nodal stress intensity due to mechanical loads. Primary + Secondary stress is the nodal stress intensity due to mechanical loads + thermal loads.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-9

SUMMARY

OF LOAD COMBINATION STRESSES FOR ACCIDENT CONDITIONS

________ _________ Cask Component Nodal Stress Intensity (ksi) _________ ________

Inner Shell &

Btm. Lid Lid Gamma Btm. Top Btm.

Load Inner Shell Outer Shield Shield Shield Trunn. Trunn. Allow. Min.*

Cmb. Plate Flange Plate Plate Shell Plate Region Region (ksi) FOS A1 19.31 13.33 8.31 3.62 5.62 5.25 4.44 5.41 49.0 2.54 A2 18.21 12.76 8.43 1.75 5.93 8.97 4.74 5.79 49.0 2.69 A3 19.59 13.47 8.28 4.17 5.56 5.05 4.38 5.43 49.0 2.50 A4 14.68 11.69 8.73 2.10 8.96 9.81 5.47 8.76 49.0 3.33 A5 16.07 12.35 8.24 0.99 6.81 4.72 4.98 6.59 49.0 3.05

• Factor of safety is calculated by dividing the membrane allowable by nodal stress intensity, hence very conservative. All factors of safety are higher than 1.5.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-10 LINEARIZED STRESS EVALUATION FOR NORMAL CONDITION LOAD COMBINATIONS Nodal Stress Linearized Stress Intensity Factor Load Intensity Node Magnitude Allow. of Comb. Component (ksi) Nos. Type (ksi) (ksi) Safety Inner Shell & Pm 12.83 22.9 1.78 N1 Bottom Inner 14.07 223-253 Plate P, ^ Pb 12.91 34.35 2.66 Inner Shell & Pm 13.17 22.9 1.74 N2 Bottom Inner 15.47 224-254 Plate P, ^ Pb 13.60 34.35 2.53 Inner Shell & Pm 12.64 22.9 1.81 N3 Bottom Inner 14.07 224-254 Plate P, ^ Pb 13.06 34.35 2.63 Inner Shell & Pm 13.17 22.9 1.74 N4 Bottom Inner 15.47 224-254 Plate P, ^ Pb 13.60 34.35 2.53 m 1.98 20.5 10.35 BottomP N5 14.52 938-1218 ie P, ^ Pb 5.67 30.75 5.42 Inner Shell & Pm 13.12 22.9 1.75 N6 Bottom Inner 13.53 224-254 Plate P, ^ Pb 13.52 34.35 2.54

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-1 1 WELD STRESSES (PAGE 1 OF 3)

Maximum Weld Allow- Factor Load Weld Nodal Area Stress able of Condition Comb. No. Forces (kips) (in2) (ksi) (ksi) Safety Tensile 364 207.35 1.76 21.00 > 10 W-1 Shear 54 207.35 0.26 12.32 > 10 Tensile 88 139.02 0.63 21.00 > 10 Normal N1 W-2 Shear 48 139.02 0.35 12.32 > 10 Tensile 80 104.95 0.76 21.00 > 10 W-3 Shear 96 104.95 0.91 12.32 > 10 Tensile 193 207.35 0.93 21.00 > 10 W-1 Shear 21 207.35 0.10 12.32 > 10 Tensile 66 139.02 0.47 21.00 > 10 Normal N2 W-2 Shear 56 139.02 0.40 12.32 > 10 Tensile 31 104.95 0.30 21.00 > 10 W-3 Shear 70 104.95 0.67 12.32 > 10 Tensile 65 207.35 0.31 21.00 > 10 W-1 Shear 411 207.35 1.98 12.32 6.22 Tensile 149 139.02 1.07 21.00 > 10 Normal N3 W-2 Shear 65 139.02 0.47 12.32 > 10 Tensile 85 104.95 0.81 21.00 > 10 W-3 Shear 128 104.95 1.23 12.32 > 10 Tensile 259 207.35 1.25 21.00 > 10 W-1 Shear 515 207.35 2.48 12.32 4.96 Tensile 126 139.02 0.91 21.00 > 10 Normal N4 W-2 Shear 72 139.02 0.52 12.32 > 10 Tensile 26 104.95 0.25 21.00 > 10 W-3 Shear 47 104.95 0.45 12.32 > 10

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-1 1 WELD STRESSES (PAGE 2 OF 3)

Maximum Weld Allow- Factor Load Weld Nodal Area Stress able of Condition Comb. No. Forces (kips) (in2) (ksi) (ksi) Safety Tensile 164 207.35 0.79 21.00 > 10 W-1 Shear 417 207.35 2.01 12.32 6.13 Tensile 157 139.02 1.13 21.00 > 10 Normal N5 W-2 Shear 103 139.02 0.74 12.32 > 10 Tensile 82 104.95 0.78 21.00 > 10 W-3 Shear 127 104.95 1.21 12.32 > 10 Tensile 358 207.35 1.73 21.00 > 10 W-1 Shear 521 207.35 2.51 12.32 4.91 Tensile 133 139.02 0.96 21.00 > 10 Normal N6 W-2 Shear 109 139.02 0.78 12.32 > 10 Tensile 29 104.95 0.28 21.00 > 10 W-3 Shear 46 104.95 0.44 12.32 > 10 Tensile 1124 207.35 5.42 42.00 7.75 W-1 Shear 1 207.35 0.01 24.64 > 10 Tensile 24 139.02 0.17 42.00 > 10 Accident A1 W-2 Shear 53 139.02 0.38 24.64 > 10 Tensile 159 104.95 1.52 42.00 > 10 W-3 Shear 281 104.95 2.68 24.64 9.23 Tensile 1266 207.35 6.11 42.00 6.89 W-1 Shear 26 207.35 0.13 24.64 > 10 Tensile 45 139.02 0.32 42.00 > 10 Accident A2 W-2 Shear 46 139.02 0.33 24.64 > 10 Tensile 66 104.95 0.63 42.00 > 10 W-3 Shear 156 104.95 1.49 24.64 > 10

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-1 1 WELD STRESSES (PAGE 3 OF 3)

Maximum Weld Allow- Factor Load Weld Nodal Area Stress able of Condition Comb. No. Forces (kips) (in2) (ksi) (ksi) Safety Tensile 1095 207.35 5.28 42.00 7.95 W-1 Shear 8 207.35 0.04 24.64 > 10 Tensile 23 139.02 0.17 42.00 > 10 Accident A3 W-2 Shear 54 139.02 0.39 24.64 > 10 Tensile 173 104.95 1.65 42.00 > 10 W-3 Shear 314 104.95 2.99 24.64 8.24 Tensile 384 207.35 1.85 42.00 > 10 W-1 Shear 100 207.35 0.48 24.64 > 10 Tensile 87 139.02 0.63 42.00 > 10 Accident A4 W-2 Shear 51 139.02 0.37 24.64 > 10 Tensile 77 104.95 0.73 42.00 > 10 W-3 Shear 93 104.95 0.89 24.64 > 10 Tensile 213 207.35 1.03 42.00 > 10 W-1 Shear 85 207.35 0.41 24.64 > 10 Tensile 65 139.02 0.47 42.00 > 10 Accident A5 W-2 Shear 59 139.02 0.42 24.64 > 10 Tensile 32 104.95 0.30 42.00 > 10 W-3 Shear 73 104.95 0.70 24.64 > 10

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-12

SUMMARY

OF LID BOLT STRESSES Normal Condition Accident Condition Stress Type Stress Allowable Stress Allowable Average Tensile (ksi) 50.1 93.5 50.1 115.5 Shear (ksi) 13.5 56.1 13.5 69.3 Combined stress Not Required 57.6 126.3 intensity (ksi) ________ ____________ (Reference 4)

Interaction E.Q.

0.345 1 0.226 1 Rt2 + Rs2 < 1 Not Required Bearing (ksi) 32.2 33.2 (Reference 4)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-13 NORMAL CONDITION LOADS AND BASKET STRESS ANALYSIS RESULTS Loading Component Load Stress Loads Stress Allow.

Comb. Classification (ksi) Stress

_________ ______________ ________ _______________ ___________ ___________ (ksi) 3g Fuel P m ^P ^Q b Primary 18.40 48.6 (1)

Vertical, Compartment N5 plus 3g ______________ Secondary ___________ _________

Lateral ^ Rails P m ^P ^Q b 25.84 34.13 (2)

Thermal Notes:

(1) The component includes fuel compartments and support plates (bars).

(2) The allowable stress in the rail is based on welds with surface PT examination (0.65 factor, per Subsection NG, Table NG-3352-1 (Reference 1).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-14 ACCIDENT CONDITION BASKET STRESS ANALYSIS RESULTS Drop Stress Maximum Allowable Factor Orientation Component Category Stress (ksi) Stress (ksi) of Safety End Drop Fuel Compartments Pm 6.56 38.88 5.92

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-15

SUMMARY

OF UPPER TRUNNION STRESSES (YIELD STRENGTH CASE)

Upper Trunnion 6/3g Load (3g/Trunnion)

(1)

Trunnion Location AA BB Stress Area (in sq) 113.9 79.8 Area Moment of Inertia (in 4 ) 3,082.0 755.6 Shear Force (lb) 750,000 750,000 Applied Moment (in lb) 5,062,500 1,590,000 Shear Stress (psi) 6,586 9,402 Bending Stress (psi) 13,962 11,837 Stress Intensity (psi) 19,195 22,220 Allowable Stress (psi) 31,800 31,800 (1)

The cross section locations AA and BB are as shown on Figure A4A.6-1.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-16

SUMMARY

OF UPPER TRUNNION STRESSES (ULTIMATE STRENGTH CASE)

Upper Trunnion 10g Load (5g/Trunnion)

(1)

Trunnion Location AA BB Stress Area (in sq) 113.9 79.8 Area Moment of Inertia (in 4 ) 3,082.0 755.6 Shear Force (lb) 1,250,000 1,250,000 Applied Moment (in lb) 8,437,500 2,650,000 Shear Stress (psi) 10,976 15,671 Bending Stress (psi) 23,271 19,728 Stress Intensity (psi) 31,991 37,033 Allowable Stress (psi) 70,000 70,000 (1)

The cross section locations AA and BB are as shown on Figure A4A.6-1.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-17

SUMMARY

OF NEUTRON SHIELD OUTER SHELL STRESSES Stress Allowable (1)

Loading Intensities (ksi) Stress (ksi) 25 psig Internal Pressure 4.92 33.6 25 psig + 3g Inertia 7.04 33.6 (Cask in Vertical Orientation) _________________ _____________

Note:

(1) Allowable stress the outer shell material, SA-51 6 GR.70, at 300 °F is 1.5 Sm = 33.6 ksi.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-18 TEMPERATURE DEPENDENT CASK BODY PROPERTIES OF MATERIALS (PAGE 1 OF 2)

Ultimate Yield Youngs Thermal Cask Temp. Allowable Material1) Strength Strength Modulus Expansion Component (°F) S m(ksi)

S (ksi) S(ksi) ___________ E (psi) a (in/in/oF)

-6 70 70 40 23.3 27.8x10 66.4x10 Inner Shell 200 70 36.6 23.3 6

27.1x10 6.7x10 6

SA-203 and Bottom Gr. E -6 Inner Plate 300 70 35.4 23.3 26.7x10 66.9x10

-6 400 70 34.2 22.9 26.2 x10 67.1 x10

-6 70 70 37.5 23.3 27.8x10 66.4x10 SA-350 6 6 Shell LF3, or 200 70 34.3 22.9 27.1x10 6.7x10 Flange SA-203 6 6 Gr. E 300 70 33.2 22.1 26.7x10 6.9x10

-6 400 70 32.0 21.4 26.2 x10 67.1 x10

-6 70 70 37.5 23.3 27.8x10 66.4x10 SA-350 6 6 Lid Outer LF3 or 200 70 34.3 22.0 27.1x10 6.7x10 Plate SA-203 6 6 Gr. E 300 70 33.2 21.2 26.7x10 6.9x10

-6 400 70 32.0 20.5 26.2 x10 67.1 x10

-6 70 70 36 23.3 29.0x10 66.4x10 SA-1 05 6 6 Lid Shield or 200 70 33 22.0 28.5x10 6.7x10 Plate SA-516, 6 6 Gr.70 300 70 31.8 21.2 28.0x10 6.9x10

-6 400 70 30.8 20.5 27.6x10 67.1 x10

-6 70 70 36 23.3 29.0x10 66.4x10 Gamma SA-266 6 6 Shield Shell CL 2 or 200 70 33 22.0 28.5x10 6.7x10 and Bottom SA-516, 6 6 Shield Gr.70 300 70 31.8 21.2 28.0x10 6.9x10

-6 400 70 30.8 20.5 27.6x10 67.1 x10

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-18 TEMPERATURE DEPENDENT CASK BODY PROPERTIES OF MATERIALS (PAGE 2 OF 2)

Ultimate Yield Youngs Thermal Cask Allowable Material1) Temp. Strength Strength Modulus Expansion Component (oF) Sm(ksi)

S (ksi) S (ksi) E (psi) a (in/in/oF)

-6 70 165 150 50.0 27.8x10 66.4x10 SA-540 6 6 Gr. B24 200 165 144 47.8 27.1x10 6.7x10 Lid Bolt or B23, 6 6 Cl.1 300 165 140.3 46.2 26.7x10 6.9x10

-6 400 165 137.9 44.8 26.2x10 67.1 x10

-6 Outer Shell SA-516, 70 70 38 23.3 29.0x10 66.4x10 Gr.70

-6 200 70 34.8 23.2 28.5x10 66.7x10

-6 300 70 33.6 22.4 28.0x10 66.9x10

-6 400 70 32.5 21.6 27.6x10 67.1 x10

-6 Upper and SA-105 70 70 36 23.3 29.0x10 66.4x10 Lower or

-6 Trunnions SA-266 200 70 33 22.0 28.5x10 66.7x10 CL 2 or 6 6 CL 4 300 70 31.8 21.2 28.0x1 0 6.9x1 0

-6 400 70 30.8 20.5 27.6x10 67.1 x10

-6 Protective SA-516, 70 70 36 23.3 29.0x10 66.4x10 Cover Gr.70

-6 200 70 33 22.0 28.5x10 66.7x10 And Flange or -6 300 70 31.8 21.2 28.0x10 66.9x10 SA-105 400 70 30.8 20.5 27.6x10 67.1 x10

-6

-6 Vent /Drain SA-193 70 125 105 35.0 29.0x10 66.4x10 Port Cover 6 6 and Top Gr. B7 200 125 98 32.6 28.5x10 6.7x10 Neutron 6 6 Shield Bolts 300 125 94.1 31.4 28.0x10 6.9x10

-6 400 125 91.5 30.5 27.6x10 67.1 x10 Note:

(1) Lower strength of the different materials is listed in the table.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-19 TEMPERATURE DEPENDENT BASKET MATERIAL PROPERTIES Ultimate Yield Youngs Thermal Temp. Allowable Part Material Strength Strength Modulus Expansion

( û F) Sm (ks)

_________ ________ _______ Su (ksi) Sy (ksi) __________ E (psi) a (in/in/oF) 6 -6 70 75.0 30.0 20.0 28.3x10 8.5 x 10 Stainless 6 -6 300 66.2 22.4 20.0 27.0x10 9.2x 10 SA-240, Steel 6 -6 400 64.0 20.7 18.6 26.4x10 9.5 x 10 Type Plates 500 63.4 19.4 17.5 25.9x10 6

9.7 x 10

-6 304 6 -6 600 63.4 18.4 16.6 25.3x10 9.8 x 10 6 -6 700 63.4 17.6 15.8 24.8x10 10.0x 10

-6 70 12.1x 10

-6 300 13.3x 10 SB-209

-6 Aluminum 400 13.6x 10 Type Plates

-6 500 13.9x 10 6061

-6 600 14.2x 10

_________ ________ 700 70 12.1x 10

-6 13.3x 10

-6 300 SB-209 Aluminum 400 13.6x 10

-6 Type Plates 13.9x 10

-6 500 1100 14.2x 10

-6 600

_________ ________ 700 Note: Material properties are taken from ASME Section II, Part D, 2004 including 2006 Addenda (Reference 1) and Aluminum Standards and Data (Reference 10).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-20 LENGTH GROWTH BETWEEN FUEL AND CASK Hot Tmax,F aZ aSS LF,hot Tmax,C aLS LC,hot Gap Condition (°F) (in/in-°F) (in/in-°F) (in) (°F) (in/in-°F) (in) (in)

Storage 680 3.73E-6 9.96E-6 162.977 302(1) 6.90E-6 163.261 0.284 Vacuum Drying 725 3.73E-6 10.0E-6 163.010 232(1) 6.76E-6 163.179 0.169 Note:

(1) Conservatively use the average temperature of inner shell and gamma shield shell.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-21 DIAMETRICAL GROWTH BETWEEN THE BASKET AND THE CASK Hot Tavg,B Tavg,Sh aSS aAl OD B,hot Tavg,C aLS ID C,hot Gap Conditions (°F) (°F) (in/in-°F) (in/in-°F) (in) (°F) (in/in-°F) (in) (in)

Storage 479 358 9.66E-6 13.43E-6 71.983 302(1) 6.90E-6 72.115 0.132 Vacuum Drying 573 367 9.80E-6 13.47E-6 72.053 360(2) 7.02E-6 72.147 0.094 Notes:

(1) Based on the average temperature of both the inner shell and gamma shield.

(2) Based on the average temperature of both the inner shell and gamma shield calculated from zero gap between the basket and cask inner shell.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-22 AXIAL GROWTH BETWEEN BASKET AND ALUMINUM PLATES AND SS SUPPORT BARS

_________ _____ _______ For 0-180 Direction ________ _____ ____

Hot HC Tavg,c ,180 aSS HC,hot HAl Tavg,Al ,180 aAl HAl, hot Conditions Gap (in) (°F) (in/in-°F) (in) (in) (°F) (in/in-°F) (in)

(in)

Storage 13.25 560 9.80E-06 13.314 13.18 572 1.414E-05 13.274 0.040 Vacuum 13.25 628 9.86E-06 13.323 13.18 637 1.427E-05 13.287 0.036 Drying _____ ________ ________ ______ _____ _______ _________ ______ _____

_________ _____ _______ For 90-270 Direction ________ ______ _____

Hot HC Tavg,c ,70 aSS HC,hot HAl Tavg,A l,270 aAl HAl, hot Conditions Gap (in) (°F) (in/in-°F) (in) (in) (°F) (in/in-°F) (in)

(in)

Storage 13.25 535 9.77E-06 13.310 13.18 536 1.404E-05 13.266 0.044 Vacuum 13.25 601 9.80E-06 13.319 13.18 600 1.420E-05 13.279 0.040 Drying _____ _______ _______ _____ _____ ______ ________ _____ ____

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-23 LENGTH GROWTH BETWEEN ALUMINUM PLATES AND SS SUPPORT BARS

__________ _______ ________ _________ 0-180 Direction ________ _________ ________ _____

L Al,180 Tavg,Al ,180 aAl LAl,180 ,hot LSS, 180 Tavg,SS ,180 aSS L SS, 180,hot Gap Conditions (in) (°F) (in/in-°F) (in) (in) (°F) (in/in-°F) (in) (in)

Storage 21.775 572 14.14E-06 21.930 21.931 560 9.80E-06 22.036 0.106 Vacuum Drying 21.775 637 14.27E-06 21.951 21.931 628 9.86E-06 22.052 0.101

__________ _______ ________ _________ 90-270 Direction _________ _________ _________ _____

L Al,270 Tavg,Al ,270 aAl L Al,270,hot L SS ,270 Tavg,SS ,270 aSS L SS, 270,hot Gap Conditions (in) (°F) (in/in-°F) (in) (in) (°F) (in/in-°F) (in) (in)

Storage 26.205 536 14.04E-06 26.376 26.361 535 9.77E-06 26.481 0.105 Vacuum Drying 26.205 600 14.20E-06 26.402 26.361 601 9.80E-06 26.498 0.096

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.2-24

SUMMARY

OF HOT GAPS FOR THE TN-40HT CASK

__________ ___________ Hot Gap (in) ____________

Al Plate / SS Al Plate / SS Fuel Bar Bar Assembly / Basket OD / Height Length Conditions Cask Cavity Cask ID 0-180 90-270 0-180 90-270 Storage 0.284 0.132 0.040 0.044 0.106 0.105 Vacuum Drying 0.169 0.094 0.036 0.040 0.101 0.096

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.5-1 CLASSIFICATION OF TN-40HT MAJOR COMPONENTS IMPORTANT TO SAFETY NOT IMPORTANT TO SAFETY Containment vessel including lid, Pressure monitoring system, &

flange, inner containment shell & overpressure cover bottom containment plate Protective cover, bolts, & seal Lid bolts Paint on exterior of cask Lid vent and drain covers, & bolts Basket assembly including fuel compartments, poison plates, &

structural plates Trunnions Basket rails Lid, vent & drain seals Radial neutron shield Cask body shield shell Cask body bottom Lid shield plate Top neutron shield including bolts Outer shell

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.6-1 RESULTS OF ACTIVATION ANALYSIS Curies per TN-40HTCask Outer Shell &

Body, Lid Resin and Protective Nuclide Basket and Rails Al Boxes Cover Total 51 Cr 5.959E-02 2.015E-02 ----- ----- 7.973E-02 RA 54 ivin 7.210E-03 5.379E-02 ----- ----- 6.100E-02 I- 55 re 7.914E-02 5.845E-01 ----- 1.064E-04 6.637E-01 I- 59 re .464E-03 1 .088E-02 ----- ----- 1 .234E-02 58 Co

_______ 9.196E-03 6.222E-03 ----- ----- 1.542E-02 60 Co

_______ 1.301E-04 8.692E-05 ----- 2.170E-04 63 Ni

________ 5.693E-03 3.855E-03 ----- 9.548E-03 Zn 65 ----- ----- 9.541E-05 ----- 9.541E-05 Ni 59 7.007E-05 3.915E-05 ----- ----- 1.092E-04 H3 ----- ----- 1 .686E-09 ----- 1 .686E-09 C 14 ----- 1.252E-09 4.071E-09 2.365E-13 5.324E-09 TOTAL 8.422E-01

-5 Note: Only the nuclides with activity greater than 1 0 curies and those listed in 10 CFR 61.55 are reported here.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 TABLE A4.6-2 COMPARISON OF TN-40HT ACTIVITY WITH CLASS A WASTE LIMITS Specific Activity of Long-Lived Isotopes (1 0CFR61 .55 Table 1) 3 3 Nuclide Ci/m Limit (Ci/m ) Volume (m 3 ) Component C 14 ----- 80 1.539 Basket Ni 594.551E-5 220 C 14 1.461E-10 80 8.571 Body Ni 594.567E-6 220 C 14 1.153E-9 80 3.532 Resin C 14 4.587E-13 80 0.516 Shell Specific Activity of Short-Lived Isotopes (1 0CFR61 .55 Table 2)

Nuclide Ci/m 3 "A" Limit (Ci/m 3 ) "B" Volume (m 3 ) Component 60 Co 8.453E-5 700 Ni 633.698E-3 35 1.539 Basket T 1/2 < 5 1.017E-1* 700 60 Co 1.014E-5 700 Ni 634.497E-4 35 8.571 Body T 1/2 < 5 6.755E-1* 700 T 1/2 < 5 2.064E-4* 700 0.516 Shell H 34.773E-10 40 3.532 Resin T 1/2 < 5 2.701E-5* 700 51 54 55 59 58 65

- Sum of isotopes with half-life less than 5 years (Cr ,Mn ,Fe ,Fe ,Co ,Zn )

Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 Weld-1 (3 Nodes

/ between J

P cylinder a plate are coupled)

FIGURE A4.2-2 SHIELD SHELL TO BOTTOM SHIELD PLATE WELD

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 Weld-2 (2 Nodes between Flange and Outer Cylinder are coupled)

FIGURE A4.2-3 SHIELD SHELL TO SHELL FLANGE WELD

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 Weld-3 (3 Nodes between Lid and Shield Plate are coupled)

FIGURE A4.2-4 LID OUTER PLATE TO LID SHIELD PLATE WELD

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 I

I I

I I

I I

700 °F I

I I

I I___

I I

I I 2 in 212°F I 700°F I___

I I I

II I

I I

I I

I 50°F y I x

FIGURE A4.2-13 TN40HT FUEL CLADDING OUTER SURFACE TEMPERATURES

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 FIGURE A4.2-14 TN40HT FUEL CLADDING FINITE ELEMENT MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13 jSYS lU 'JAl

-LY ii 14:45:44 PLOT NO. 2 FEHJ:

TF-fPFF:TGPES Ti-LX=7 c:"::

=1

['I:T=I .1 F =1

-EJFFFP R

- j'*I**

FIGURE A4.2-15 TN40HT FUEL CLADDING TEMPERATURE DISTRIBUTION

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT REVISION: 13

?1JZ:I i':I:'

i: ii 14:45:45 FucT HO. J Tc[L QLL1TICIT

TEF1 i:[1E =1 TIFE=i

IJT ':.'

DI=.

=.

[r =4'::'5 Ir47 1 III-1: I-i

- 4':Ill55 FIGURE A4.2-16 TN40HT FUEL CLADDING STRESS INTENSITY

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.1-1 APPENDIXA4A STRUCTURAL ANALYSIS OF THE TN-40HT CASK BODY A4A.1 INTRODUCTION This appendix presents the structural analysis of the TN-40HT storage cask body which consists of the cask body, the trunnions and the outer shell. Analyses are performed to evaluate the various cask components under the loadings described in Section A3.2.

Additional analyses are provided in Section A8 to evaluate the cask body under hypothetical accident loadings.

The detailed calculations for the cask body are presented in Section A4A.3 and the lid bolt analysis is reported in a separate Section A4A.4. The calculations for the trunnions and outer shell are reported in Sections A4A.6 and A4A.7, respectively. Section A4A.8 contains the calculations for the top neutron shield bolts and Section A4A.9 the facture toughness evaluation of the TN-40HT cask.

The design criteria used in the analyses of the cask components are discussed in Section A4.2.3. Key dimension of the storage cask are shown in Figure A4.2-1.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.1-2 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.2-1 A4A.2 MATERIALS PROPERTIES DATA This section provides the mechanical properties of materials used in the structural evaluation of the TN-40HT storage cask. Table A4A.2-1 lists the materials selected, the applicable components, and the minimum yield, ultimate, and design stress values specified by the ASME Code. All values reported in Table A4A.2-1 are for metal temperature up to 100°F. For higher temperatures, the temperature dependency of the material properties is reported in Table A4.2-1 8.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.2-2 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-1 A4A.3 CASK BODY STRUCTURAL ANALYSIS A4A.

3.1 DESCRIPTION

The cask body consists of an inner shell and an outer shield shell (and a bottom shield).

The inner shell (and bottom inner plate and shell flange) is the primary confinement barrier of the cask. Key dimensions of the cask body are shown in Figure A4.2-1. The inner shell is 1.5 in. thick cylinder welded to the shell flange and bottom inner plate. The inner shell is shrunk fit into the shield shell. The lid outer plate is bolted to the shell flange by 48, 1.375 in. diameter, high strength bolts and sealed with two metallic seals.

The lid outer plate, inner shell, bottom inner plate, shell flange, shield shell and bottom shield components are made of low alloy steel forgings and plates. Two sets of trunnions are welded to the side of the shield shell upper and lower ends for handling and supporting the cask during lifting and handling operations. A basket assembly inside the cask cavity is used to position and support the fuel assemblies. A detailed physical description of the containment components is provided in Section A1.

Section A1 .5 contains reference drawings of the TN-40HT cask on which the analysis models are based.

A4A.3.2 ANSYS CASK MODEL The outer shield shell, bottom shield plate, inner shell, bottom inner plate, shell flange, lid outer plate and lid shield plate are modeled utilizing ANSYS eight-node brick elements (SOLID45) as shown in Figure A4A.3-1. The individual cask components used for reporting stresses are shown in Figure A4A.3-2 through Figure A4A.3-6. Due to the cyclic symmetry of the TN-40HT cask body, some nodes in the FEM are rotated into a local cylindrical coordinate system (csys, 11) for easy application of node couplings and boundary conditions. This local coordinate system is located at the model axis of rotation (i.e. global Cartesian X =Y = Z = 0.0). The lid bolts are modeled using BEAM4 elements. The bolt preload is simulated using pre-strain in beam element real constant.

The nodes at common surfaces between flange and lid are coupled in the axial direction. The contact surface nodes between outer shield shell inner surface and containment cylinder external surface are radially coupled. Also, the common nodes of the bottom shield plate and bottom inner plate are axially coupled. A total of 20,112 elements and 25,920 nodes comprise the TN-40HT cask finite element model. At the cask cut surface, symmetry boundary conditions are applied. Node couplings between the inner shell and outer shield shell are shown in Figure A4A.3-7. The loadings and displacement boundary conditions for individual load runs are shown in Figure A4A.3-8 through Figure A4A.3-15. To avoid congestion, symmetry boundary conditions (UZ) are not shown in these figures.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-2 For all loading conditions, fuel and basket loads are applied as pressures based on specific component weights, accelerations and applied area. For the seismic loading, a radial cosine pressure distribution, derived in Section A4A.3.6, accounting for the fuel and basket inertial load is used. The radial pressures are applied using ANSYS macros.

However, one radial pressure is hand-calculated for load case IL-9 (see below) to check that macro-computed pressures are in order of magnitude.

Since outer shell, radial neutron shield, aluminum boxes, trunnions, protective cover and top neutron shield are not modeled, their weights are accounted for by increasing the shield shell and lid plate density as follows.

i) Shield Shell Density 2 2 Weight Shield Cylinder = p (44.75 - 37.5 )(172.65-6.25) x 0.283 = 88,219 lb.

Weights of components not in model:

Radial neutron shield = 12,285 lb Outer shell = 7,543 lb Trunnions = 920 lb (955 lb is used)

Aluminum Boxes = 2,233 lb Protective Cover = 1,362 lb (1,357 lb is used)

Total weight not in the model = 24,343 lb (24,373 lb is used)

Total Weight associated with shield Cylinder = 88,219 + 24,373 = 112,592 lb Equivalent Density of Shield Cylinder

= (112,592/88,219) x 0.283 = 0.361 lb/in 3 (0.363 is conservatively used in the ANSYS runs) ii) Lid Density Weight of lid outer plate and Shield plate = 13,314 lb.

Weight of Top Neutron shield = 1,638 lb. (1,635 lb is used)

Total Weight = 13,314 + 1,635 = 14,949 lb Equivalent Density of Lid & Shield

= (14,949/1 3,314) x 0.283 = 0.318 lb/in 3

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-3 A4A.3.3 INDIVIDUAL LOAD CASES The loading conditions analyzed simulate or represent various effects due to the normal and hypothetical accident conditions of storage as specified in 1 0CFR72. A total of 10 individual loading conditions listed in Table A4A.3-1 are executed for normal and accident load combinations for storage. Load case numbers IL-1 to IL-9 are analyzed elastically in this attachment, using an ANSYS (Reference 1) 3-D Finite Element Model (FEM) and are described in detail in Section A4A3.2 above. Load case IL-10 requires calculation of local stresses at the top and bottom trunnions / shield shell interface and are executed by hand calculations. Load case IL-1 0 is described in Section A4A.6.

The following describes the individual load cases analyzed using the ANSYS model described above:

1. Bolt Preload and Lid Seating Pressure (IL-1)

A bolt axial pre-stress of 50 ksi (see Section A4A.4) at the bolt shank (1.375 inch diameter) is simulated by specifying an initial strain in BEAM4 elements representing the bolts. The required initial strain value of 0.002128 in/in (in the bolts) was determined by first calculating the initial strain required to produce an 3 6 axial stress of 50 ksi (i.e., e = s / E = 50x10 / 26.45x10 = 0.00189 in/in). Then, an initial analysis with the calculated strain (0.00189 in/in) was conducted and the resulting bolt pre-stress was backed out. Since, a portion of this strain becomes elastic preload strain in the bolts, and a portion becomes strain in the clamped parts, the backed out pre-stress from the initial analysis will not produce the desired 50 ksi. The FEM was then updated by multiplying the 0.00189 in/in strain by the ratio of (desired pre-stress 50 ksi / initial analysis pre-stress), and a second preload analysis was conducted, which resulted in a 50 ksi bolt pre-stress.

Lid seating pressure due to seating of metallic seals is computed due to 660,129 lb seal reaction (see Section A4A.4). Based on the closest nodes in model to the 2

seal location, the Lid Seating Pressure, p = 660,129 / it (38.42 - 37.35 ) = 2,642 psi.

This pressure is applied to the lid and flange seal areas. For the bolt preload and lid seating pressure ANSYS run, the cask is supported as shown in Figure A4A.3-8.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-4

2. Fabrication Stress (IL-2)

The fabrication stresses in the cask are computed due to a 0.040 inch nominal diametric interference between inner shell and shield shell. The shrink fit stresses are based on radial interference of 0.020 inches.

Inner Shell Radius, a Shield Shell Radius, b Inner Shell c

Shield Shell 2 2 2 Interface pressure, p = (Ed / b) [( b 2 - a ) ( c - b2 ) / (2 b 2 ( c - a 2 )) ]

(Reference 2)

Where, 8 = 0.02 in.

E = 29.0x1 0 6 psi a = 36.0 in.

b = 37.5 in.

c = 44.75 in.

Note that using the higher value of E for the two materials at room temperature, is conservative, because it results in a higher interface pressure.

p = (29.0x106 x 0.02 / 37.5) 2 2 2 2 2 2 x [ ( 37.5 - 36.0 2 ) ( 44.75 - 37.5 ) / ( 2 x 37.5 ( 44.75 - 36.0 ) ) ]

= 511.7 psi ... (515 psi is used)

The radial couplings between the inner shell and shield shell are deleted and the interface 515 psi pressure is applied to the two cylinders. During the run, the cask is supported as shown in Figure A4A.3-9.

3. 1g Down - Cask Support at Bottom (IL-3)

The following inertial load (pressure) magnitudes are applied to the FEM due to 1g vertical acceleration.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-5 An axial pressure due to internals (P,) is applied on the inside bottom surface of the cask cavity.

Weight of Fuel and Basket >> 80,100 lb.

P, = 1.0 x 80,100/ ( (it) x 36 . 02 ) = 19.673 psi The bottom nodes of cask are supported in the vertical direction. Loading and displacement boundary conditions are shown in Figure A4A.3-10.

4. Internal Pressure Loading (IL-4)

An internal pressure of 100 psig is applied to the cavity surface as shown in Figure A4A.3-1 1. The pressure is applied up to the metallic seal inner radius.

The lid bolt preload and seal seating loads are removed in this run. During the run, the cask is supported as shown in Figure A4A.3-1 1.

5. External Pressure Loading (IL-5)

An external pressure of 25 psig is applied to the outer surface of the cask body.

The pressure is applied up to the seal outer radius. During the run, the cask is supported as shown in Figure A4A.3-11.

6. Thermal Stress for 100 o F Hot Environment Ambient Temperature (IL-6)

The thermal analysis of the cask body is described in Section A3. The thermal model is used to obtain the steady state metal temperatures in the cask body for the normal condition which includes 100° F daily averaged ambient air temperature, maximum decay heat and maximum solar heat loading. The cask nodal temperatures from thermal results file are used to interpolate nodal temperatures in structural model. An ANSYS macro is used to get the nodal temperatures in the structural model. The resulting temperature distribution in cask is shown in Figure A4A.3-12. These temperatures are then used as ANSYS input for the thermal stress analysis. Temperature dependent material properties are used in this run. During the run, the cask is supported as shown in Figure A4A.3-12. It is noted that the maximum temperature in a small region of bottom inner plate exceeds assumed temperature of 350 °F. All other load cases (except thermal stress) are analyzed using the material properties at 350 °F. The effect of this temperature change on material modulus of elasticity will be negligible. The thermal stresses, however, are calculated using exact temperatures and material properties.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-6

7. Thermal Stresses for -40 o F Cold Environment Ambient Temperature (IL-7)

The thermal analysis of the cask body is described in Section A3. The thermal model is used to obtain the steady state metal temperatures in the cask body for the normal condition which includes -40 o F ambient air temperature, minimum decay heat and minimum solar heat loading. The cask nodal temperatures from thermal results file are used to interpolate nodal temperatures in structural model.

An ANSYS macro is used to get the nodal temperatures in the structural model.

The resulting temperature distribution in cask is shown in Figure A4A.3-1 3.

These temperatures are then used as ANSYS input for the thermal stress analysis. Temperature dependent material properties are used in this run. During the run, the cask is supported as shown in Figure A4A.3-13.

8. 3g Lifting on Front Trunnion (IL-8)

The cask is oriented vertically in space and held by the 2 top trunnions. The inertial loading is simulated by applying a 3g vertical acceleration to the finite element model.

Since the internals are not included in the model, their loading effects are simulated by distributed pressure (P,) acting on the inside bottom surface of the cask cavity.

Weight of Fuel and Basket >> 80,100 lb.

P, = 3.0 x 80,100/ ( t x 36 . 02 ) = 59.02 psi.

Loading and displacement boundary conditions are shown in Figure A4A.3-14.

9. Bounding Loads for Seismic, Tornado and Flood (IL-9)

A bounding load of 1 g lateral and 2g down for seismic, tornado and flood is applied to the cask in vertical orientation. The cask is assumed to be restrained at the bottom nodes. For the inertial loading, a vertical acceleration of 2g is applied in the global Y direction and 1g lateral acceleration in X direction. Since the internals are not included in the model, their effects are calculated as follows:

i) Vertical loading effect of internals is simulated by a distributed pressure (P,)

acting on the inside bottom surface of the cask cavity.

P, = 2.0 x 80,100/ ( t x 36 . 02 ) = 39.347 psi

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-7 ii) The lateral loading effect of internals is simulated by applying a radial cosine varying pressure (internal contact angle 0° to 150° range) to inner shell cavity.

The formula below produces the cosine distribution of internal pressure and is derived in Section A4A.3.6.

Where:

q = 1/2 Angle of contact = 75°

_______ qi = Circumferential angle pressure W Jr0i P=g cos - is applied LR Jr 20 W = Weight of internals ~ 80,100 lb n 1-0 ')

2 L = Length pressure is applied =163 in 2+

Jr + Jr R = Radius pressure is applied = 36 in 1 -1 g = Acceleration = gs 20 20 For example, a pressure applied at an angle of 7.5° would be calculated as follows.

80,100 1 180x7.5 Pr cos

= 1.0 163.0x 36 sin(90+ 75) sin(90 - 75) 2x 75 180 + 180

+1 -1 2x75 2x75 Pr = 13.65(0.7083) (0.9877) = 9.55 psi Loading and displacement boundary conditions are shown in Figure A4A.3-

15. The radial pressures are applied using an ANSYS macro.

A4A.3.4 TRUNNION LOCAL STRESSES The analyses of the local stresses in the TN-40HT cask body due to the trunnion reactions (while lifting the cask) are evaluated in Section A4A.6 below.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-8 A4A.3.5 EVALUATION (LOAD COMBINATIONS VS. ALLOWABLES)

ANSYS linear elastic analyses are performed for the above individual load cases. A summary of maximum nodal stress intensities in each major cask component under each individual load is presented in Table A4A.3-2. These individual loads are to be combined and evaluated for normal operating and accident conditions in Section A4.2.3.4.

A4A.3.6 PRESSURE DISTRIBUTION OVER CONTACT AREA OF CASK FOR IMPACT IN TRANSVERSE DIRECTION Inertial loading from the internal basket acting in the transverse direction is applied as a load over the contact area on the inside surface of the cask. The pressure distribution is assumed to vary with a cosine distribution around the circumference of the cask. The circumferential cosine pressure distribution over a half angle 0 is calculated as follows.

Pi = Pmax cos(r0i / 20)

Where, Pi = Pressure load at angle 0i.

Pmax = Peak pressure load, at point of impact, and 0i = Angle corresponding to point of interest.

The circumferential pressure distribution is illustrated in the following sketch.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-9 Circumferential Pressure Load Distribution The peak pressure load P max is determined by setting the integral of the vertical pressure components Q, equal to the total transverse impact load Ft as follows.

q q q Q , LRdq, = , cos(q, )LRdq, = Pmax co ' 1cos(q, )LRd q,

= si -

-q -q -q 2q q

Pmax LR cos _, + q, + cos , -q, J dq, 2q 2q 2 -[

p p s,n + q s,n - -q 2

DLR2 max f

+ f (p (p

+1 -1 2q 2q Rearranging terms gives the peak pressure P max.

1

• p p -*

sin - + q sin - -q Ft 2 2 Dmax = +

LR p p

+1 -1 2q 2q Therefore, the pressure at any circumferential location is given by,

-1

.p . p sin +q sin - -q I 2 J 2

=- + cosL_,

2q I+1 1 2q t:IJ-where, Ft =gxW W = Weight of internals or impact limiter g = Acceleration in the transverse direction

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.3-10 Therefore, 1

• p p -*

sin +q sin - -q 2 2 pqi WPi=g-- + cos LR p p 2q

+1 - 1 2q 2q

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-1 A4A.4 LID BOLT ANALYSES A4A.

4.1 INTRODUCTION

This section evaluates the ability of the cask lid bolts to maintain the seal under all applicable loads. Also evaluated in this section are the lid bolt thread and internal thread stresses. The stress analysis is performed in accordance with NUREG/CR-6007 (Reference 3).

The 4.5 inch thick lid outer plate with a 5.5 inch shield plate is bolted directly to the shell flange by 48 high strength alloy steel 1.375 in. diameter bolts (with 1 1/2 -8UN threaded portion). Close fitting alignment pins ensure that the lid is centered in the cask. The bolt material is SA-540 Gr. B24 or B23 CL 1 which has a minimum yield strength of 150 ksi at room temperature.

The bolt size, material and design loads for all TN-40HT closures including the lid bolts are provided in Table A4A.4-1.

The following ways to minimize bolt forces and bolt failures for cask are taken directly from Reference 3, page xiii. All of the following design methods are employed in the TN-40HT lid closure system.

• Use materials with similar thermal properties for the closure bolts, the lid, shell flange, and the cask wall to minimize the bolt forces generated by fire accident

• Apply sufficiently large bolt preload to minimize fatigue and loosening of the bolts by vibration.

• Lubricate bolt threads to reduce required preload torque and to increase the predictability of the achieved preload.

• Use closure lid design which minimizes the prying actions of applied loads.

• When choosing a bolt preload, pay special attention to the interactions between the preload and thermal load and between the preload and the prying action.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-2 The following evaluations are presented in this section for the bolts where applicable:

• Bolt preload

• Gasket seating load

• Internal pressure load

• Temperature load

• Thread engagement length evaluation

• Bearing stress

• Load combinations for normal and accident conditions

• Bolt stresses and allowables In this analysis the applicable load combinations from Reference 3 are analyzed using a bounding approach. The loads that are applicable to the lid bolts are preload, gasket seating load, internal pressure, and thermal. Furthermore, load combinations are split into two subsections for loads that are due to preload and loads that act against preload (internal pressure and gasket sealing forces). The specific design loads corresponding to the loading conditions mentioned above are presented in Table A4A.4-1.

It should be noted that tip over is not a credible event for the TN-40HT cask. The only accident load is 18 bottom end drop. The cask bottom end drop does not induce additional stress in the bolt. Since there are no credible accident loads, all normal condition loads are compared to both normal and accident condition allowable stresses.

The design parameters for the lid closure are summarized in Table A4A.4-2. The lid bolt data and material allowables are presented in Table A4A.4-3 and Table A4A.4-4.

The maximum temperature at the lid seal region is less than 200 °F (Table A3.3-3).

Thus, 300 °F is conservatively used for bolt stress evaluations.

A4A.4.2 LID BOLT LOAD CALCULATIONS Symbols and terminology for this analysis are taken from Reference 3 and are reproduced in Table A4A.4-2.

Lid Bolt Preload:

The desired maximum preload stress in lid bolts is 50,000 psi.

2 For a 1.375 inch bolt shank diameter, the Tensile Stress Area is ( p/4)*1.375 = 1.485 in 2 . Therefore, Fa = 50,000 x Stress Area = 50,000 x 1.485 = 74,250 lb.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-3 The torque required to achieve this preload is (Reference 3, Section 4.0)

Q = K D b Fa = 0.135 (1.375) (74,250) = 13,783 in. lb. = 1,149 ft. lb .

A bolt torque range of 1,100 to 1,150 ft. lb . has been selected. Bolt preload for the maximum torque is, Fa = Q/ KD b = 1,150 x 12 / (0.135 x 1.375) = 74,343 lbs, and Preload stress = 74,343 / 1.485 = 50,063 psi.

Bolt Preload for the minimum torque is, Fa = Q/ KDb = 1,100 x 12 / (0.135 x 1.375) = 71,111 lb.

Residual torsional moment for maximum torque of 1,150 ft. lb . is, Mtr = 0.5Q = 0.5(1,150 x 12) = 6,900 in. lb.

Residual torsional moment for minimum torque of 1,100 ft. lb . is, Mtr = 0.5 Q = 0.5(1,100 x 12) = 6,600 in. lb.

Residual tensile bolt force, Far = Fa = 74,343 lbs Lid Closure Gasket Seating Load (Seal - Helicoflex HND 229 seals (Reference 6)):

The diameter of the inner seal, Dis , is 74.315 in., and the diameter of the outer seal, D lg, is 75.882 in. The force to seat the seals is 1399 lbs./in (245 N/mm) (Reference 6, Helium sealing, Aluminum Jacket, 0.260 inch cross section) times the circumference of the seal. Therefore the force required to seat the seals is:

Inner: rt (74.31 5) (1399) = 326,621 lb.

Outer: rt (75.882) (1399) = 333,508 lb.

Total, Fa = 660,129 lb.

Therefore, the gasket seating load is, Fa /48 = 660,129 / 48 = 13,753 lb./bolt.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-4 Lid Bolt Internal Pressure Loads (Reference 3, Table 4.3)

Axial force per bolt due to internal pressure is, pDl2(P;, -)

g Fa=

4 Nb D lg diameter of the outer seal = 75.882 in. Then, 2

p ( 75 . 882 ) ( 100 -0) = 9 Fa = , 422 lb./bolt 4 ( 48)

The fixed edge closure lid force is,

- D ( P - P;o) - 79 . 31 ( 100) = 1

, 983 lb. in.

b  ;, 1 F

f_ -

4 4 The fixed edge closure lid moment is, (P - Po ) D? = 100(79.3 12) = 19,656 1 Mf = in. lb. in.

32 32 The shear bolt force per bolt is, p(26.7* 10 6 )(4.5)(100)(79.3 1) 2 pEl t l (P - P )Dl -

F = lb./bolt 2 N E b c t

c(

1- N ul) -

2(48)(26.7x10 6 )(8.8)(0.7) =15,037 The gap between the bolt shank and the lid (1.625 in. - 1.375 in. = 0.25 in.) is bigger than the gap between the lid and cask (72.87 - 72.67 = 0.20 in.) which ensures that the lid will contact the cask before the bolts get loaded in shear, therefore Fs = 0.

A4A.4.3 LID BOLT TEMPERATURE LOADS The lid bolt material is SA 540 Gr B24 or B23 CL 1. The lid and flange are both made of SA-350 Gr. LF3 or SA-203 GR. E. The bolts, lid and flange have the same coefficient of thermal expansion (6.9 x 10-6 in/in-°F at 300°F). Therefore, heating to the maximum isothermal temperature will generate no bolt loads.

A4A.4.4 LID BOLT LOAD COMBINATIONS (REFERENCE 3, TABLE 4.9)

A summary of lid bolt individual load and load combinations are presented in the Table A4A.4-5 and Table A4A.4-6 respectively.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-5 A4A.4.5 LID BOLT ADDITIONAL PRYING BOLT FORCE (REFERENCE 3, TABLE 2.1)

Since all applied loads are outward acting, and the additional prying bolt force specified in Reference 3 can only be generated by inward acting loads, no additional prying bolt forces act on the closure lid bolts.

A4A.4.6 LID BOLT BENDING MOMENT BOLT (REFERENCE 3, TABLE 2.2)

The maximum bending bolt moment, Mbb , generated by the applied load is evaluated as follows:

p D lbT Kb M f N b IK b +K l bbH The Kb and Kl are based on geometry and material properties and are defined in Reference 3, Table 2.2. By substituting the values given above, 6 4 IEb D 48 26.7x 10 1 375 5 Kb = (N b b

= 2.006 x 10 , and Lb D

lb J64J 4.5J 79.31 64 J 3

E1 t 26.7x 10 6 (10.0)

Kl =

2 2 (79.31 2

3[(1-0.3 )+(1-0.3) I 79.31 2

3[(1- N )+(1- N )

2 D lb l

82.75 U

7

= 8.2507 x 10 Therefore, p79.31 2.006 x 10 5 M bb= =0.01259Mf.

48 I 2.006x10 +8.2507x10 5 7 For load case 2, Mf = 19,656 in. lb. Substituting this value into the equation above gives, Mbb = 247.5 in. lb. / bolt.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-6 A4A.4.7 LID BOLT STRESS CALCULATIONS (REFERENCE 3, TABLE 5.1)

Average Tensile Stress:

The bolt preload is calculated to withstand the worst case load combination and to maintain a clamping (compressive) force on the closure joint, under both normal and accident conditions. Based upon the load combination results (see Table A4A.4-6), it is shown that a positive (compressive) load is maintained on the clamped joint for all loads. Since there are no credible accident loads applicable for the lid bolts, all normal condition loads are compared to both normal and accident condition allowable stresses.

The maximum non-prying tensile force for normal conditions is 74,343 lb, from load case 1 .A. This load is used to compute bolt stresses below.

Normal Condition:

Sba = 1 . 2732 = 1 . 2732 = 50 , 065 psi. = 50.1 ksi.

1 . 375 2

Dba Bending Stress:

Normal Condition:

247.5

=10.186--=10.186 MS =970 psi. = 0.97 ksi.

1.375 bb Dba Shear Stress:

The average shear stress caused by shear bolt force Fs is, Sbs = 0.

For normal conditions the maximum shear stress caused by the torsional moment Mt is, 6,900 bt = 5 . 093--

MS = 5 . 093_ = 13 , 518 psi. = 13.5 ksi.

Dba 1 . 375 Maximum Combined Stress Intensity:

The maximum combined stress intensity is calculated in the following way (Reference 3, Table 5.1).

= [ ( Sba + ) 2 + 4(Sbs +

2 0.5 Sbi Sbb Sbt ) ]

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-7 For normal conditions combine tension, shear, bending, and residual torsion.

2 0.5 Sbi = [(50,065 + 970) 2 + 4 (0 + 13,518) ] = 57,754 psi. = 57.8 ksi.

Stress Ratios:

In order to meet the stress ratio requirement, the following relationship must hold for both normal and accident conditions.

R t2 + R82 <1 Where Rt is the ratio of average tensile stress to allowable average tensile stress, and R8 is the ratio of average shear stress to allowable average shear stress.

For normal conditions:

Rt = 50,065/93,500 = 0.5355, R8 = 13,518/56,100 = 0.241, R t2 + R82 = (0 . 5355)2 + (0 . 241) 2 = 0.345 < 1.

For accident conditions (using maximum normal condition loads):

Rt = 50,065/115,500 = 0.433, Rs = 13,518/69,300 = 0.195, 2 2 Rt + R s = (0 . 433)2 + (0 . 195)2 = 0.226 < 1.

A4A.4.8 LID BOLT BEARING STRESS (UNDER BOLT HEAD)

The maximum axial force is 74,343 lb for normal conditions. A bolt hole of 1.625 inch diameter is used for the shank.

H = 2.25/2 in. Diam = 1.625 B = 1.125 tan(30°) = 0.650 in.

Total Area of one triangle

= (1.125)(0.650) = .731 in 2 .

B

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-8 Total area under Bolt head - Bolt Hole area 2

= 6(0.731) - ( p/4)(1 .625 ) = 2.31 in2.

The total bearing area is 2.31 in. 2 The bearing stress for normal conditions is, Bearing Stress = 74,343/2.31 = 32,179 psi. = 32.2 ksi.

The allowable normal condition bearing stress on the lid is taken to be the yield stress of the lid material at 300 °F. The lid is manufactured out of SA-350 GR. LF3 or SA-203 GR. E, which has a yield strength of 33.2 ksi at 300 °F.

A4A.4.9

SUMMARY

OF LID BOLT STRESSES A summary of the stresses calculated above is listed in the Table A4A.4-7.

A4A.4.10 MINIMUM ENGAGEMENT LENGTH FOR LID BOLT AND FLANGE For a 1 1/2 - 8UN - 2A bolt, the material is SA - 540 Gr. B24 or B23 CL 1, with Su = 165 ksi., and Sy = 150 ksi (at room temperature)

The flange material is SA - 350 Gr LF3 or SA-203 Gr. E, with Su = 70 ksi., and Sy = 37.5 ksi (at room temperature)

The minimum engagement length Le for the bolt and flange is (Reference 7, page 1490),

2A t Le=

3.1416K[1+.57735n(E i

-K 2

Where, At = tensile stress area = 1.4718 in. 2 (Reference 7, page 1490), for Su>100,000 psi n = number of threads per inch = 8, Kn max = maximum minor diameter of internal threads = 1.390 in., (Reference 7, page 1728)

Es min = minimum pitch diameter of external threads = 1.4093 in., (Reference 7, page 1728)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-9 Ds m,n = minimum major diameter of external threads = 1.4828 in. (Reference 7, page 1728)

Substituting the values given above, 2 ( 1 . 4718)

Le = = 1.1442 in.

1

( 3 . 1416 ) 1 . 390[ + . 57735 ( 8 )(1 . 4093-1 . 390 2

A s x S ue

= (Reference 7)

An x S u, Where, S ue is the tensile strength of external thread material, and Su, is the tensile strength of internal thread material.

As = shear area of external threads

= 3.1416 n L e Kn max [1/(2 n ) + 0.57735 ( Es m,n - Kn max)]

An = shear area of internal threads

= 3.1416 n Le D s m,n [1/(2 n ) + 0.57735( D s m,n - En max)]

For the bolt / flange insert connection:

En max = maximum pitch diameter of internal threads = 1.4283 in. (Reference 7 , page 1728).

Therefore, As = 3.1416 (8) (1.1442) (1.390) [1 / (2 x 8) + 0.57735 (1.4093 - 1.390)]

= 2.944 in. 2 An = 3.1416 (8) (1.1442) (1.4828) [1 / (2 x 8) + 0.57735 (1.4828 - 1.4283)]

4.007 in. 2 So, 2.944(165.0)

=1.73 4.007(70.0)

Therefore, the minimum required engagement length, Q = J L e = 1.73 x 1.1442 = 1.98 in. The actual minimum engagement length = (7.00 bolt length - 4.50 lid thickness) =

2.50 in. > 1.98 in.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.4-10 The above calculation bounds the minimum required engagement length if inserts are used because S, of inserts (Reference 8) is higher than the S, for the lid thus lowering the J value.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.5-1 A4A.5 VENT AND DRAIN COVER BOLT ANALYSES A4A.

5.1 INTRODUCTION

This section evaluates the ability of the cask vent/drain cover bolts to maintain the seal under all applicable loads. The stress analysis is performed in accordance with NUREG/CR-6007 (Reference 3).

The bolt size, material and design loads for all TN-40HT vent/drain cover bolts are provided in Table A4A.4-1.

The following ways to minimize bolt forces and bolt failures for cask are taken directly from Reference 3, page xiii. All of the following design methods are employed in the TN-40HT lid closure system.

• Use materials with similar thermal properties for the closure bolts and the lid.

• Apply sufficiently large bolt preload to minimize fatigue and loosening of the bolts by vibration.

• Lubricate bolt threads to reduce required preload torque and to increase the predictability of the achieved preload.

• Use closure lid design which minimizes the prying actions of applied loads.

• When choosing a bolt preload, pay special attention to the interactions between the preload and thermal load and between the preload and the prying action.

The following evaluations are presented in this section for the bolts where applicable:

• Bolt preload

• Gasket seating load

• Internal pressure load

• Temperature load

• Bearing stress

• Load combinations for normal and accident conditions

• Bolt stresses and allowables In this analysis the applicable load combinations from Reference 3 are analyzed using a bounding approach. The loads that are applicable to the vent/drain cover bolts are preload, gasket seating load, internal pressure, and thermal. Furthermore, load combinations are split into two subsections for loads that are due to preload

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.5-2 and loads that act against preload (internal pressure and gasket sealing forces).

The specific design loads corresponding to the loading conditions mentioned above are presented in Table A4A.4-1.

It should be noted that tip over is not a credible event for the TN-40HT cask. The only accident load is 18 bottom end drop. The cask bottom end drop does not induce additional stress in the bolt. Since there are no credible accident loads, all normal condition loads are compared to both normal and accident condition allowable stresses.

The vent/drain cover bolt data and material allowables are presented in Table A4A.5-1 and Table A4A.5-2.

The maximum temperature at the lid seal region is less than 200 °F (Table A3.3-3).

Thus, 300 °F is conservatively used for bolt stress evaluations.

A4A.5.2 VENT AND DRAIN BOLT LOAD CALCULATIONS Vent and Drain Bolt Torque:

A bolt torque range of 60 to 65 ft. lb . has been selected for the vent and drain cover bolts.

Bolt preload for the maximum torque is, Fa = Q/ KD b = 65 x 12 / (0.135 x 0.75) = 7,704 lbs, and Preload stress = 7,704 / 0.334 = 23,065 psi.

Bolt Preload for the minimum torque is, Fa = Q/ KD b = 60 x 12 / (0.135 x 0.75) = 7,111 lb.

Residual torsional moment for maximum torque of 65 ft. lb . is, Mtr = 0.5 Q = 0.5(65 x 12) = 390 in. lb.

Residual torsional moment for minimum torque of 60 ft. lb . is, Mtr = 0.5 Q = 0.5(60 x 12) = 360 in. lb.

Residual tensile bolt force, Far = Fa = 23,065 lbs

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.5-3 Vent and Drain Cover Gasket Seating Load (Seal - Helicoflex HND 229 seals (Reference 6)):

The diameter of the inner seal, D,8 , is 4.881 in., and the diameter of the outer seal, D o8, is 5.689 in. The force to seat the seals is 1,142 lbs./in (245 N/mm) (Reference 6, Helium sealing, Aluminum Jacket, 0.161 inch cross section) times the circumference of the seal. Therefore the force required to seat the seals is:

Inner: rt (4.881) (1,142) = 17,512 lb.

Outer: rt (5.689) (1,142) = 20,410 lb.

Total, Fa = 37,922 lb.

Therefore, the gasket seating load is, Fa/8 = 37,922 / 8 = 4,740 lb./bolt.

A4A.5.3 VENT AND DRAIN BOLT INTERNAL PRESSURE LOADS (REFERENCE 3, TABLE 4.3):

Axial force per bolt due to internal pressure is, pD l 2(P,,

g

-)

Fa =

4 b D ,g diameter of outer seal = 5.689 in. Then, p ( 5 . 6892) ( 100-0) = 318 Fa = lb./bolt 4 ( 8)

The fixed edge cover lid force is, D,b ( P,, - P,o )7.38(100)185 1 Ff- = lb. in.

4 4 The fixed edge cover lid moment is,

( P,, - P,o ) D2= 100 ( 7 . 382) 170 1 Mf = in. lb. in.

32 32

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.5-4 The shear bolt force per bolt is, 2 2 pE t P - P D p(27 . 0x10 6 )(0 . 25)(100)(7 . 38)

F =

, ,( ,i ,o) ,b

=138 lb./bolt 2 NbEctc( 1 - Nu,) = 2(826 . 7x 10 6 )(2 . 8)(0 . 7)

A4A.5.4 VENT AND DRAIN BOLT TEMPERATURE LOADS:

From Reference 5, the vent and drain bolt material is SA-1 93 Gr B7, with a = 6.9x1 0 -6 in/in °F. The vent and drain covers are made from SA-240, type 304 with a = 9.2x10

-6 in/in °F, and the cask lid outer plate is made from SA-350 Gr. LF3 or SA-203 Gr. E, with a = 6.9x1 0 -6 in/in °F. Assuming a maximum temperature difference of 230 °F (300° F -

70° F), then from Reference 3, the thermal loads are computed as follows.

2 Fa = 0.25 p Db Eb (a, T, - ab Tb) 2 6 -6 -6 Fa = 0.25(p)(0.75 )(28.0x10 )[(9.2x10 )(230) - (6.9x10 )(230)] = 6,544 lb.

Even though the vent and drain covers and the cask lid outer plate are constructed from different materials, the shear force per bolt, F8 , due to a temperature change of 230 °F is, 0 psi, since the clearance holes in the lid are oversized (0.781 inch diameter) allowing the covers to grow in the radial direction.

F8 = 0.

The temperature difference between the inside and outside of the covers will always be less than one degree because the vent and drain cover is only 1 inch thick.

Consequently, the resulting bending moment is negligible.

Mf = 0.

A4A.5.5 VENT AND DRAIN COVER BOLT LOAD COMBINATIONS (REFERENCE 3, TABLE 4.9)

A summary of individual load and load combinations are presented in Table A4A.5-3 and Table A4A.5-4, respectively.

A4A.5.6 VENT AND DRAIN COVER BOLT ADDITIONAL PRYING BOLT FORCE (REFERENCE 3, TABLE 2.1)

Since all applied loads are outward acting, and the additional prying bolt force specified in Reference 3 can only be generated by inward acting loads, no additional prying bolt forces act on the cover lid bolts.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.5-5 A4A.5.7 VENT AND DRAIN COVER BOLT BENDING MOMENT BOLT (REFERENCE 3, TABLE 2.2)

The maximum bending bolt moment, Mbb , generated by the applied load is evaluated as follows:

pD,b Kb M

bb = I Mf N b IK b +K, The Kb and K, are based on geometry and material properties and are defined in Reference 3, Table 2.2. By substituting the values given above, 6 4 N b E b D b 8 28 . 0x 10 075 5 Kb = = 6.002 x 10 , and L D

, 1 64 J b

L05 7 . 38 64 J 3 6 3 E, t 27 . 0x10 ( 1 . 00 )

=

K, 2 2 7.38 2

2 D ,b 2 3[(1 - N (1 -

2 D 3[(1 - 03 )+ (1 - 03) 7 . 38

)+ N u,) I II ,b D,o Ij 5

= 9.839 x 10 Therefore, p 6 . 002x10 5

M bb = L 5 5 M f

= 1.098 M f.

8 16 . 002x 10 + 9 . 839x10 For load case 2, Mf = 170 in. lb. Substituting this value into the equation above gives, Mbb = 187.0 in. lb. / bolt.

A4A.5.8 VENT AND DRAIN BOLT STRESS CALCULATIONS (REFERENCE 3, TABLE 5.1)

Average Tensile Stress:

The bolt preload is calculated to withstand the worst case load combination and to maintain a clamping (compressive) force on the closure joint, under both normal and accident conditions. Based upon the load combination results (see Table A4A.5-4), it is shown that a positive (compressive) load is maintained on the clamped joint for all loads. Since there are no credible accident loads applicable for the vent and drain cover bolts, all normal condition loads are compared to both normal and accident condition allowable stresses. The maximum non-prying tensile force for normal conditions is 14,248 lb, from load case 1 .A. This load is used to compute bolt stresses below.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.5-6 The bolt diameter used for stress calculation is computed from the bolt stress area provided in Table A4A.4-1:

1/2 Stress Area = 0.334 = p/4 D ba 2 D ba = [ 0.334

• 4/p ] = 0.652 in.

Normal Condition:

ba = 1 . 2732-- FS = 1 . 2732 14 2482 = 42 , 672 psi. = 42.7 ksi.

D ba 0 . 652 Bending Stress:

Normal Condition:

=10 . 186--=10 MS . 186_187 =6 , 872 psi.=6.9ksi.

0 . 652 bb D ba Shear Stress:

The average shear stress caused by shear bolt force Fs is, 138

=1.2732--=1.2732_

FS =413 psi. = 0.4 ksi.

0.652 2 bs DL For normal conditions the maximum shear stress caused by the torsional moment Mt is, 390 bt = 5 . 093-- MS = 5 . 093_ = 7 , 166 psi. = 7.2 ksi.

D ba 0 . 652 Maximum Combined Stress Intensity:

The maximum combined stress intensity is calculated in the following way (Reference 3, Table 5.1).

= [ ( Sba + + 4( Sbs +

2 0.5 S bi Sbb ) 2 Sbt ) ]

For normal conditions combine tension, shear, bending, and residual torsion.

2 0.5 S bi = [(42,672 + 6 , 872)2 + 4 (413 + 7,166) ] = 51,811 psi. = 51.8 ksi.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.5-7 Stress Ratios:

In order to meet the stress ratio requirement, the following relationship must hold for both normal and accident conditions.

2 +R Rt 82 < 1 Where Rt is the ratio of average tensile stress to allowable average tensile stress, and R8 is the ratio of average shear stress to allowable average shear stress.

For normal conditions:

Rt = 42,672/62,700 = 0.681 R8 = 7,579/37,600 = 0.202 2

Rt + R82 = (0 . 681)2 + (0 . 202) 2 = 0.504 < 1.

For accident conditions (using maximum normal condition loads):

Rt = 42,672/87,500 = 0.488 R8 = 7,579/52,500 = 0.144 2

Rt + R82 = (0 . 488)2 + (0 . 144)2 = 0.259 < 1.

A4A.5.9

SUMMARY

OF VENT AND DRAIN BOLT STRESSES A summary of the stresses calculated above is listed in the Table A4A.5-5.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.6-1 A4A.6 TRUNNION ANALYSIS A4A.

6.1 INTRODUCTION

This section presents the evaluation of the TN-40HT cask trunnion stresses due to all applied loads as well as resulting local stresses generated in the cask shield shell.

The TN-40HT cask has two upper and two lower trunnions each of which are one piece forgings composed of SA-1 05 or SA-266 CL 2 or CL 4 carbon steel.

The upper trunnions are designed to meet the requirements of NUREG-0612 (Reference 17) for non-redundant lifting fixture. This is accomplished by evaluating the trunnions to the stress design factors required by ANSI N 14.6 (Reference 9), i.e.

capable of lifting 6 times and 10 times the cask weight without exceeding the yield and ultimate strengths of the material, respectively. The upper trunnions are also used to support the loaded cask in the vertical position during transfer to the ISFSI by a transport vehicle.

The lower trunnions are used to support the cask when the cask is in the horizontal position prior to fuel loading and also during rotation of the cask from a horizontal orientation to the vertical orientation. Since the lower trunnions are not relied upon to support the cask when loaded with fuel, their structural analysis is not presented.

The trunnion stress analysis is performed using standard bending and shear stress equations based on the trunnion loading configuration. Nominal dimensions of the trunnions and cask shell are used in the analysis.

Weight of TN-40HT cask used in this analysis is conservatively taken to be 250,000 lb when loaded with fuel.

For the trunnion stress analysis, the following load cases are evaluated:

• Lifting Loads (cask lifted from the pool to the decontamination area and then to the ISFSI). The upper trunnions are analyzed for 6g and 1 0g vertical loads.

The local shield shell stress calculation is based on the method described in Reference 10:

• 3g longitudinal lifting load applied to the upper trunnions (cask in the vertical orientation)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.6-2 A spreadsheet was created to aid in the computation. Table A4A.6-4 is a hardcopy of this spreadsheet. The coefficients used in Table A4A.6-4 through are taken from Tables 1A through 4C of Reference 10.

A4A.6.2 MATERIAL PROPERTIES The trunnions are one piece forgings composed of SA-1 05 or SA-266 CL 2 or CL 4 carbon steel with material properties shown in Table A4.2-1 8. These material properties are taken from ASME Code (Reference 5).

The shield shell material properties (SA-266 CL 2 or SA-51 6-70) used for the local stress analysis are also listed in Table A4.2-18.

The material properties used for the trunnion and shield shell stress analyses are taken o

at 300 F, which is conservative based on the thermal evaluation described in Section A3.

A4A.6.3 STRESS CRITERIA Stress in the Trunnion The upper trunnions are designed to meet the requirements of NUREG-0612 (Reference 17) for non-redundant lifting fixture. This is accomplished by evaluating the trunnions to the stress design factors required by ANSI N 14.6 (Reference 9), i.e.

capable of lifting 6 times and 10 times the cask weight without exceeding the yield and ultimate strengths of the material, respectively.

Stress in the Shield Shell The local stresses in the shield shell due to the trunnion loads are compared with the allowable stresses following requirements of the ASME Code,Section III, Subsection NB (Reference 11).

A4A.6.4 TRUNNION STRESS ANALYSIS Analysis for the upper trunnions is based on a cask weight of 250 kips and corresponding 6/10g lifting loads. The trunnion cross sections are shown in Figure A4A.6-1.

Upper Trunnions:

Based on the nominal dimensions given in Figure A4A.6-1 the stress areas and moments of inertia (MOI) for the two different trunnion cross sections are calculated. For cross sections AA and BB the stress areas, MOI, and moment arms are computed as:

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.6-3 2'

,r ( 17 2 _12 2 Stress Area AA = __________ = 113.9 in 4

4

,r ( 17 _ 12) 4 MOIAA= = 3082.0 in 64 Moment Arm AA = 8.94 - 0.63 - 0.31- (3-0.5)/2 = 6.75 in

,r ( 11 . 252_52) 2 Stress Area BB = ___________ = 79.8 in 4

4

,r ( 11 . 25 _ 5 4 MOIBB= 4)= 755.6 in 64 Moment Arm BB = 4 - 0.63 - (3-0.5)/2 = 2.12 in Table A4A.6-1 and Table A4A.6-2 summarize the stress evaluations. The applied shear load per trunnion is computed by multiplying the cask weight by the appropriate g-load per trunnion. The moment per trunnion is computed by multiplying the shear load by the moment arm length computed in this section.

A4A.6.5 TRUNNION LOCAL STRESS ANALYSIS Case 1: 3g Lon g itudinal Liftin g Load (U pper Trunnions)

For the 3g lifting load case, the cask is lifted vertically and supported by the upper trunnions. The g-load applied to the cask is 3g in the longitudinal direction.

At the upper trunnion the g load per trunnion is, 3g (longitudinal) / 2 sides / 1 trunnion = 1 .5g longitudinal per trunnion The following analytical method is taken from Reference 10. Figure A4A.6-2 illustrates the loads and moments as defined in this analysis as well as their orientation to the cask/trunnion system as a whole. Note that the all of the u locations are on the shield shell outer surface whereas l locations are on the inside shield shell surface.

Geometry:

Shield shell thickness at trunnion locations, T = 5.81 in.

Upper Trunnion radius, r0 = 8.5 in.

Mean radius, RM = (86.62 + 75.00)/4 = 40.405 in.

Upper trunnion moment arm = 9.66 in.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.6-4 Shell Parameter (Reference 10) g = RM/ T = 40.405/5.81 = 6.954 Attachment Parameter (Reference 10) b = 0.875

• r0 / RM = 0.875 x 8.5/40.405 =

0.184 Trunnion loads according to Reference 10 sign convention:

P = 0 lb.

ML = 1.5 x (-250,000) x 9.66 = -3,622,500 in. lb.

MC = 0 in. lb.

MT = 0 in. lb.

VL = 1.5 x (-250,000) = -375,000 lb.

VC = 0 lb.

Stress intensity calculation Stress intensities are calculated in the following way.

+4t 2 ]

2

[%[(s X

+)+/-V(s X

-s)

SI.. = max 2

4 2

(s -

s f ) + t X

From Table A4A.6-4, the maximum stress intensity for case 1 is 8,072 psi, on the outside of the shield shell at the front and back of the trunnion (Bu/Au, compression/tension). Where, sf = 6,059 psi, sx = 8,072 psi, and t = 0 psi Membrane Stress Calculation In order to calculate the membrane and bending stresses, only those components associated with membrane or bending stress respectively are summed to calculate sf, sx and t. For example, to calculate the maximum membrane stress on the outside of the vessel for the 3g lifting load (Table A4A.6-4, locations Du, Cu):

sfm = 0 psi.

sxm = 0 psi.

tm = -2,417 psi.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.6-5 Using the above formula for stress intensity, the corresponding membrane stress intensity is 1 2 0)- I(0 - + 4 ( -2 , 417 ) = 4 , 834 psi.

2 S.I. = max ________________

2 I(0 - + 4 ( -2 , 417 ) = 4 , 834 psi.

= 4,834 psi A4A.

6.6 CONCLUSION

S From Table A4A.6-1 and Table A4A.6-2, it is shown that the trunnions are qualified for lifting/handling based on stress intensity according to the loading requirements set forth in Reference 9.

The results for the local shield shell stress analysis are presented in Table A4A.6-3. To evaluate load combinations N7 and N8 (see Table A4.2-5) the local stresses calculated are combined with stresses due to internal pressure and thermal load taken from Table A4A.3-2. A bounding approach was taken using the internal pressure load and the cold thermal load of the shield shell in order to envelope both load combinations N7 and N8. The combined results presented in these tables demonstrate that the cask shield shell is adequate to handle the required lifting and handling loads.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.7-1 A4A.7 OUTER SHELL A4A.

7.1 INTRODUCTION

This attachment presents the structural analysis of the outer shell of the TN-40HT storage cask. The outer shell consists of a cylindrical shell section and outer shell rings (top and bottom closure plates) at each end which connect the shell to the cask body.

The normal loads acting on the outer shell are due to internal pressure and normal handling loads.

A4A.

7.2 DESCRIPTION

The outer shell is constructed from low-alloy carbon steel and is welded to the outer surface of the shield shell. The cylindrical shell section is 0.5 in. thick and the outer shell rings (top and bottom closure plates) are 0.75 inches thick. Pertinent dimensions are shown in Figure A4A.7-1. Section A1 .5 contains reference drawings of TN-40HT storage cask on which the analysis are based.

A4A.7.3 MATERIALS INPUT DATA The outer shell cylindrical section and outer shell rings are SA-516 Gr 70. The material properties are taken from the ASME Code (Reference 5) and are shown on Table A4.2-18.

A4A.7.4 APPLIED LOADS While the cask is transported to the site in the horizontal position, it is always in the vertical position when loaded with fuel. Thus the structural analysis present here assumes the cask is in the vertical position with the following loading conditions:

• Cask in the Vertical Orientation

- Stress due to 25 psig internal pressure

- Stress due to 25 psig internal pressure and 3g inertia load (lifting)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.7-2 A4A.7.5 METHOD OF ANALYSIS ANSYS Model A finite element model is built for the structural analysis of the outer shell and outer shell rings (top and bottom closure plates). The outer shell and outer shell rings (top and bottom closure plates) are modeled with ANSYS Solid 45 elements (Reference 1). The basic geometries of the outer shell, outer shell rings, and weld sizes used for analysis are shown in Figure A4A.7-1. The bounding weld sizes between the outer shell ring and shield shell, the outer shell ring and outer shell, weld between the outer shell ring pieces (ring may be fabricated from multiple pieces, groove weld size shall not be less than 0.55 in.) are used for the analysis. The finite element model is shown in Figure A4A.7-2.

Cask in the Vertical Orientation Stress due to 25 psig internal pressure An internal pressure of 25 psig is used as the maximum pressure acting on the inner surface of the outer shell. The maximum shell stress intensity for this load case was determined to be 4,919 psi.

Stress due to 25 psig internal pressure and 3g inertia load (lifting the cask in the vertical orientation)

The weight of the radial neutron shield and aluminum containers is modeled as an additional pressure on the inner surface of the bottom outer shell ring (closure plate).

The added pressure load is 9.29 psi per g. The effect of the outer shell dead weight is accounted for by using a 3g gravitational load in the longitudinal direction. The maximum stress intensity for this load case was determined to be 7,044 psi.

A4A.7.6 ANALYSIS RESULTS A. Maximum Stress Intensities for Each Loading Condition A summary of the results obtained for maximum stress intensities for each loading condition is listed in Table A4A.7-1. These stresses are within the allowable stress of o

1.5 S m (at 300 F, S m = 22,400 psi, 1.5 S m = 33,600 psi). The outer shell can be fabricated from two half cylindrical shell with 0.38 partial penetration weld. The maximum stress intensity and factors of safety at the weld location is also computed as follows:

Allowable stress at the weld = 0.3S u = 0.3 x 70,000 = 21,000 psi (Reference 12)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.7-3 In the vertical orientation loading condition, the outer shell is mainly in membrane tension/compression loads, the membrane stress is inverse proportion to the element thickness. The maximum weld stress is computed as follows:

25 psig internal pressure:

Max. stress intensity = 4,919 x 0.5 / 0.38 = 6,472 psi Factor of Safety = 21,000 / 6,472 = 3.24 25 psig internal pressure + 3g inertia load:

Max. stress intensity = 7,044 x 0.5 / 0.38 = 9,268 psi Factor of Safety = 21,000 / 9,268 = 2.27 B. Maximum Stress at Weld Locations (see Figure A4A.7-1)

Weld stress intensities are also calculated at the locations noted in Figure A4A.7-1.

These values are listed in Table A4A.7-2.

The weld stress intensities are less than the allowable stress of 21.0 ksi (Reference 12, 0.3Su, SA-516 GR.70, at 300 °F).

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.8-1 A4A.8 TOP NEUTRON SHIELD BOLTS A4A.

8.1 INTRODUCTION

This section evaluates the top neutron shield bolts. The stress analysis is performed in accordance with NUREG/CR-6007 (Reference 3).

The bolt size, material and design loads for all TN-40HT top neutron shield bolts are provided in Table A4A.4-1.

The following evaluations are presented in this section for the bolts where applicable:

• Bolt preload

• Temperature load

• Bearing stress

• Load combinations for normal and accident conditions

• Bolt stresses and allowables In this analysis the applicable load combinations from Reference 3 are analyzed using a bounding approach. The loads that are applicable to the top neutron shield bolts, are only preload and temperature loads. In addition, a hypothetical 3g load is applied for the purpose of conservatively bounding any unforeseen loads in these areas. The specific design loads corresponding to the loading conditions mentioned above are presented in Table A4A.4-1.

It should be noted that tip over is not a credible event for the TN-40HT cask. The only accident load is 18 bottom end drop. The cask bottom end drop does not induce additional stress in the bolt. Since there are no credible accident loads, all normal condition loads are compared to both normal and accident condition allowable stresses.

The top neutron shield lid bolt data and material allowables are presented in Table A4A.5-1 and Table A4A.5-2.

The maximum temperature at the lid seal region is less than 200 °F (Table A3.3-3).

Thus, 300 °F is conservatively used for bolt stress evaluations.

A4A.8.2 TOP NEUTRON SHIELD COVER BOLT LOAD CALCULATIONS Top neutron shield Cover Bolt Torque:

A bolt torque range of 90 to 100 ft. lb . has been selected for the top neutron shield cover bolts.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.8-2 Bolt preload for the maximum torque is, Fa = Q/ KD b = 100 x 12 / (0.135 x 1.25) = 7,111 lbs, and Preload stress = 6,400 / 0.969 = 7,339 psi.

Bolt Preload for the minimum torque is, Fa = Q/ KD b = 90 x 12 / (0.135 x 1.25) = 6,400 lb.

Residual torsional moment for maximum torque of 100 ft. lb . is, M tr = 0.5 Q = 0.5(1 00 x 12) = 600 in. lb.

Residual torsional moment for minimum torque of 90 ft. lb . is, Mtr = 0.5 Q = 0.5(90 x 12) = 540 in. lb.

Residual tensile bolt force, Far = Fa = 7,339 lbs A4A.8.3 TOP NEUTRON SHIELD COVER BOLT TEMPERATURE LOADS The top neutron shield cover bolt material is SA-193 Gr B7, with a = 6.9x10 -6 in/in °F.

-6 The top neutron shield cover is made from SA-516, Gr. 70 with a = 6.9x10 in/in °F.

Since the coefficient of thermal expansion for all materials is the same, heating to the maximum isothermal temperature will generate no bolt loads.

A4A.8.4 TOP NEUTRON SHIELD COVER BOLT 3G ACCELERATION LOAD (REFERENCE 3, TABLE 4.6)

The loads generated in the top neutron shield cover bolts by the 3g vertical load are computed below. The weight of the top neutron shield cover is 1,638 lb. However, 1,800 lb. is used in this analysis for conservatism. Also, a dynamic load factor of 1.1 is conservatively applied to the analysis.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.8-3 The neutron shield bolts have 1 .25-7 UNC threads with tensile area of 0.969 in 2 .

Under design conditions the assembled and loaded (with fuel) TN-40HT never experiences a net upward acceleration or a side load exceeding the 1 .0g bounding load.

Nevertheless, a 3.0g upward or lateral load (not simultaneous) is assumed to conservatively evaluate these shield attachment bolts. The load per bolt is then 1800(1.1)(3)/4 = 1485 lb/bolt. The bolt stress under the 3.0g loading is equal to 3.0g x attached weight divided by the total bolt area. The stress is 3 x 1,800 lbs/(4 x 0.969 in 2 )(1.1) = 1,533 psi. This is a tensile stress in the upward load case and a shear stress in the side load case.

A4A.8.5 TOP NEUTRON SHIELD COVER BOLT LOAD COMBINATIONS (REFERENCE 3, TABLE 4.9)

A summary of top neutron shield bolt individual load and load combinations are presented in and Table A4A.8-1 and Table A4A.8-2, respectively.

A4A.8.6 TOP NEUTRON SHIELD COVER BOLT STRESS CALCULATIONS (REFERENCE 3, TABLE 5.1)

Average Tensile Stress:

The bolt diameter used for stress calculation is computed from the bolt stress area provided in Table A4A.4-1.

Stress Area = 0.969 = rt/4 D ba 2 D ba = [ 0.969

• 4/p ]

1/2

= 1.111 in.

Normal Condition:

, 111 = 7 ba

=1 . 2732-FS - =1 . 2732 7 2

, 335 psi. = 7.3 ksi.

D ba 1 , 111 For normal conditions the maximum shear stress caused by the torsional moment Mt is, 600 bt

= 5 . 093---

MS =._ = 2 , 228 psi. = 2.2 ksi.

D ba 1 . 111 Maximum Combined Stress Intensity:

The maximum combined stress intensity is calculated in the following way (Reference 3, Table 5.1).

= [ ( Sba + + 4( Sbs +

2 0.5 S bi Sbb ) 2 Sbt ) ]

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.8-4 For normal conditions combine tension, shear, bending, and residual torsion.

2 0.5 Sbi = [(7,335 + 0) 2 + 4 (0 + 2,228) ] = 8,582 psi. = 8.6 ksi.

Stress Ratios:

In order to meet the stress ratio requirement, the following relationship must hold for both normal and accident conditions.

2 +R Rt 82 < 1 Where Rt is the ratio of average tensile stress to allowable average tensile stress, and R8 is the ratio of average shear stress to allowable average shear stress.

For normal conditions:

Rt = 7,335/62,700 = 0.117 R8 = 2,228/37,600 = 0.059 2

Rt + R82 = (0 . 117)2 + (0 . 059)2 = 0.017 < 1.

For accident conditions (using maximum normal condition loads):

Rt = 7,335/87,500 = 0.084 R8 = 2,228/52,500 = 0.042 2

Rt + R82 = (0 . 084)2 + (0 . 042)2 = 0.009 < 1.

A4A.8.7

SUMMARY

OF TOP NEUTRON SHIELD COVER BOLT STRESSES A summary of the stresses calculated above is listed in Table A4A.8-3.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.9-1 A4A.9 FRACTURE TOUGHNESS A4A.9.1 FRACTURE TOUGHNESS REQUIREMENTS OF THE CASK CONTAINMENT BOUNDARY The TN-40HT cask material is a ferritic steel (penetration covers are stainless steel) and is therefore subject to fracture toughness requirements in order to assure ductile o

behavior at the lowest service temperature (LST) of -20 F.

The design temperature of the TN-40HT cask is specified as -40°F. However, the brittle fracture evaluations were done for a -20°F operating temperature. The cask is shown to be able to withstand a -40°F environment but during operation, that is, any movement of the cask, the combination of loaded cask heat generation and actual ambient temperature is expected to maintain a temperature greater than -20°F for all cask containment material. This corresponds to the 1 0CFR71 practice of a -20°F minimum transportation temperature combined with a -40°F minimum material temperature.

The inner shell and bottom inner plate are fabricated from SA-203 Gr. E plate material, 1.5 inches thick. The shell flange is 4.6 inches thick, fabricated from SA-350 Gr. LF3 or SA-203 Gr. E material and the lid outer plate is 4.5 inches thick, fabricated from either SA-350 Gr. LF3 or SA-203 Gr. E material.

By interpolating between values provided in NUREG/CR-3826 (Reference 13) and NUREG/CR-1815 (Reference 14), the nil ductility transition temperatures ( T NDT) of the containment boundary materials are:

• Inner shell and bottom inner plates (1.5 in.): -80 °F

• Shell flange (4.6 in.): -137 °F

• Lid outer plate (4.5 in.): -125 °F The 1.5 inch lid closure bolts are fabricated from SA-540 Grade B24 or B23, Class 1 material. All the lid closure bolt materials meet the ASME Code,Section III, Subarticle NB-2300 criteria.

All the material which forms the containment boundary meets the meets the fracture arrest criteria of Reference 13 or 14.

A4A.9.2 FRACTURE TOUGHNESS EVALUATION OF CASK COMPONENTS AND WELDS The following sections document the fracture toughness evaluations of the non-containment boundary components as well as the containment boundary components.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.9-2 A4A.9.3 METHODOLOGY The allowable flaw sizes were determined using linear elastic fracture mechanics (LEFM) methodology from Section XI of the ASME Code (Reference 15). Flaws in the welds, if they occur, are welding defects, rather than initiated cracks. There is no active mechanism for crack initiation and growth at any of the weld locations since all the containment welds are volumetrically examined by RT and/or UT examination to assure no weld defects are present.

A4A.9.4 LOADINGS Figure A4A.9-1 shows the selected locations on the cask numbered 1 through 10 for fracture toughness analysis. Stresses are linearized at these critical locations to determine maximum tensile membrane and bending stresses.

Table A4A.9-1 and Table A4A.9-2 list the maximum membrane and bending stresses at these selected locations under Normal and Accident Conditions of Storage.

A4A.9.5 MATERIAL FRACTURE TOUGHNESS The shell flange is basically a forged cylinder, nominally 4.6 inches thick by 9 inches long, made from SA-350 Gr. LF-3 or SA-203 Gr. E material. The welding of the flange to the shell is performed using processes that meet ASME requirements so that fracture toughness properties are preserved. The lid outer plate is either a forged disc or a plate, nominally 4.5 inches thick with an 82.75 inch diameter, made from either SA-350 Gr. LF3 or SA-203 Gr. E material.

The electrodes used in the shell flange and lid outer plate weldments have high nickel content. The high alloy content of the electrodes and their typical usage in applications where good toughness is required indicate that the expected fracture toughness values for the weld filler material is as good as or better than that of the base material.

The shield shell is a plate or forged cylinder, nominally 7.25 inches thick by 166.4 inches long. The bottom shield is nominally 7.25 inches thick and 89.5 inches in diameter. These components are made from SA-266 Class 2 or SA-516 Gr. 70 material. Similarly, the 5.5 in. thick lid shield plate is made from SA-516 Gr. 70 or SA-105 material. The welding at the top flange and bottom plate is performed using processes that meet ASME requirements so that fracture toughness properties are preserved.

The fracture toughness values of these materials are enveloped by the low fracture toughness of the SA-266 forging material. Therefore, it is conservative to use fracture toughness of the SA-266 forging for the fracture toughness evaluations of the TN40-HT components.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.9-3 Reference 16 is a very thorough review of the correlation between a range of ferritic steel material strength levels and Charpy impact energies.

Figure A4A.9-2 (reproduced from Figure 4-5 of Reference 16) corresponds to Charpy V-notch impact test results for a normalized SA-266 forging. The actual data points are shown along with a smoothed line that connects the average value at each test temperature. This data demonstrates that a lower bound Charpy impact value of 18 ft-o lbs is appropriate for an exposure temperature of -20 F. The various correlations between Kic and Kid given in Table 4-2 of Reference 16 are compared at the 18 ft-lb level. Using the equation for yield strength of 36 to 50 ksi in transition in Table 4-2 of Reference 16, the Charpy impact measurement may be transformed into a fracture toughness value:

1 '2 Kid = [5E(C v )] = 50,289 psi-(in) 1'2 = 50 ksi-(in.) 1'2 Where Kid = Dynamic Fracture Toughness (based on crack arrest), ksi -(in) 1'2 E = Modulus of Elasticity, 28.1 x 10 6 psi (conservatively use 300 F) o C v = Charpy Impact Measurement, 18 ft-lbs For conservatism, the above calculated Kid was reduced to 47 ksi-(in) 1'2 for fracture toughness evaluations of the TN-40HT cask components (containment and non-containment boundary) and welds.

A4A.9.6 FRACTURE TOUGHNESS CRITERIA Using the rule of Section XI, IWB-3613 (Reference 15), the limiting fracture toughness values are reduced by a factor of 1 0 for the normal and 2 for the accident conditions, to define the limiting allowable Kallowable. That is, Kallowable £ K ia '( 10) = 47 '( 10) = 14.86 ksi- in for normal conditions Kallowable £ K ic '( 2) = 47'( 2) = 33.23 ksi- in for accident conditions Where:

Kia = the available fracture toughness based on crack arrest for the corresponding crack tip temperature Kic = the available fracture toughness based on crack initiation for the corresponding crack tip temperature

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.9-4 However, because of the dynamic nature of loadings, it is appropriate to use K id in place of K ia and K ic.

A4A.9.7 STRESS INTENSITY FACTOR CALCULATIONS The total applied stress intensity K I (applied) is determined from the membrane and bending stresses. For purposes of analysis, the postulated surface flaws are oriented in both the axial and circumferential directions. The surface crack depth is assumed as 15% of component thickness. However, the maximum crack depth is limited to 1/2 inch.

The crack length is assumed to be 10 times the crack depth. The assumed crack sizes are such that they can be readily spotted by visual examination. Compared to surface cracks, the same size subsurface cracks are less critical. The results of the applied stress intensity K I calculations for normal and accident conditions are shown in Table A4A.9-3 and Table A4A.9-4, respectively.

A4A.

9.8 CONCLUSION

S Based on the results of fracture analysis of TN-40HT cask components and welds with the postulated surface crack sizes, it is concluded that there is no potential of fracture failure due to normal or accident storage loadings. The postulated surface flaw sizes are such that they can be readily detected by a visual inspection.

Note that the shield shell is not part of the containment boundary. Cracks postulated in the shield shell will not propagate into the containment boundary due to the geometry of the cask. If the shield shell were to fracture along the length or around the circumference or along the weld between the shield shell and shell flange, there is no credible mechanism that would result in the shield shell separating from the inner shell.

The lid shield plate is welded to the underside of the lid outer plate and is captured by the inner shell. If the weld were to completely fail, the shield plate would still remain inside the containment boundary and would not lose its shielding capability. Therefore, even if a fracture were to occur in the shield shell or the weld between the shield shell and shell flange or weld between lid shield plate and lid outer plate, there would be no safety significance, since confinement would be maintained, and shielding would remain in place. The one exception is in the region of the weld joining the shield shell to the bottom shield. In this region, if the weld were to completely fail, the bottom shield could become detached and reduce the shielding capability of the cask. However, the bottom trunnions are independently welded to the shield shell and the bottom shield and these additional attachment points (welds) would resist detachment of the bottom shield from the shield shell.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.10-1 A4A.10 TN-40HT STORAGE CASK END DROP ANALYSIS The purpose of this section is to determine the rigid body accelerations for the TN-40HT Cask during a vertical drop height of 18 inches on concrete.

The rigid body transfer cask accelerations were predicted numerically by the LS-DYNA 3D explicit nonlinear dynamic analysis finite element solver, Version 9.71s (Reference 18). The methodology used in performing this analysis is based on work conducted at the Lawrence Livermore National Laboratory (LLNL), where an analysis methodology was developed and validated through comparisons with test data (Reference 19 and Reference 20). The analysis methodology was benchmarked in Reference 25.

The results of these analyses are used as input to the detailed analyses for the cask body, internal basket and fuel assemblies.

A4A.10.1 FINITE ELEMENT MODEL DESCRIPTION The ANSYS finite element model of the TN-40HT Cask developed for the cask stress analysis (Appendix A4A.3) was simplified for use in the dynamic impact analysis. The TN-40HT Cask model consists of the cask body, simplified basket structure, concrete pad and soil. Each of these components was modeled using 3D 8-node brick elements.

Fully integrated selectively-reduced solid elements were used for all elements to reduce the risk of hourglassing problems.

The finite element model was developed with ANSYS and transferred to LS-DYNA.

Modifications were made to the LS-DYNA input file to add the material definitions, non-reflecting boundaries and equation of state into LS-DYNA. Features of the cask, such as the trunnions and neutron shield were neglected in terms of stiffness but their weight was lumped into the density of the cask.

The fuel and basket were modeled as a solid cylinder inside the cask walls with elastic material properties approximately equivalent to that of the structure as a whole.

The geometry of the cask finite element model including the cask internals, concrete and base soil is shown in Figure A4A.10-1 and Figure A4A.10-2.

Only 1/2 of the cask, internals, concrete and soil were modeled, because the entire arrangement is symmetric about the x-y plane. The concrete modeled was 16-8 long, 6-8 wide, and 3 thick, and the soil modeled was 66-8 long, 1 8-9 wide, and 39-2 deep.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.10-2 A4A.10.2 MATERIAL PROPERTIES The material properties required to perform the analysis include modulus of elasticity, E, Poisons Ratio, v, and material density ( p ) for the cask body, basket, concrete, and soil.

The concrete pad requires a more detailed material model since all of the significant nonlinear deformations occur in the concrete. Material properties used for the concrete and soil were based on those developed at Lawrence Livermore National Labs (Reference 19 and Reference 20).

All material properties were taken at room temperature. This is considered conservative because the cask loaded with spent fuel will typically reach temperatures higher than room temperature, and the lower modulus of elasticity at higher temperatures tends to soften the impact and consequently lower the computed g-loads.

TN-40HT Cask Material The cask material properties were the same at those used in Appendix A4A.3. All cask materials were modeled as elastic.

Elastic Modulus Density (lb- Poissons Cask Component 2 4

________________ (psi) sec /in ) Ratio

-4 Lid Outer Plate 27.8X10 68.230x10 0.3

-4 Shield Plate 29.0X10 68.230x10 0.3

-4 Shell Flange 27.8X10 67.324x10 0.3

-4 Shell 29.0X10 69.394x10 0.3

-4 Bottom Plate 29.0X10 67.324x10 0.3 6 -4 Inner Liner 27.8X10 7.324X10 0.3 Fuel and Basket Material The basket structure material properties were the same as those used in Reference 20 except for density. The density of the basket was adjusted to calibrate the overall weight of the cask and basket assembly. The basket was modeled as elastic.

E = 2.8X1 0 6 psi n = 0.3 p = 3.215X10 -4 lb sec /in 2 4 Total modeled weight of the cask and basket is 121,174 lbs since it is a half model.

Therefore the total modeled weight is 242,348 lbs. Total actual weight of the cask and basket is 242,400 lbs.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.10-3 Concrete Material The concrete was modeled using material law 16 in LS-DYNA, which was developed specifically for granular type materials. The concrete data used in the analysis was originally designed by LLNL for the Shippingport Station Decommissioning Project in 1988. This model was also used in the LLNL (Reference 19) cask drop analysis.

Material constants were implemented into Material Model 16, Mode II.B in LS-DYNA.

The material represents 4,200 psi compressive strength concrete. A summary of the input used in the analysis is as follows.

p = 2.09675x10 -4 lb sec2 / in 4 n = 0.22 a0 = 1606 a1 = 0.418

-5 -1 a2 = 8.35x10 psi b1 =0 a 0f = 0.0 psi a1 f = 0.385 Effective Plastic Strain versus Scale Factor for Concrete Material Effective Plastic Scale Factor, v Strain __________________

0 0 0.00094 0.289 0.00296 0.465 0.00837 0.629 0.01317 0.774 0.0234 0.893 0.04034 1.0 1.0 1.0 The maximum principal stress tensile failure cutoff was set at 870 psi. Strain rate effects were neglected in the analysis. Dilger (Reference 21) suggests that the major impact of strain rate effects is in the softening part of the stress-strain curve. Since the purpose of these analyses is primarily to predict the peak accelerations, the strain rate effects on the material behavior may be neglected.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.10-4 The pressure-volume behavior of the concrete was modeled with the following tabulated pressure versus volumetric strain rate relationship using the equation of state feature in LS-DYNA.

Tabulated Pressure versus Volumetric Strain Rate for the Concrete Material Volumetric Pressure (psi)

Strain, £ 0 0

-0.006 4,600

-0.075 5,400

-0.01 6,200

-0.012 6,600

-0.02 7,800

-0.038 10,000

-0.06 12,600

-0.0755 15,000

-0.097 18,700 An unloading bulk modulus of 700,000 psi was assumed to be constant at any volumetric strain, as was assumed in Reference 19.

One percent deformation was assumed in the concrete pad to account for the pad reinforcement.

The material properties used for the reinforcing bar are as follows.

E = 30 x 106 psi n = 0.3 Sy = 30,000 psi Tangent Modulus, ET = 30 x 104 psi Soil Material The Lawrence Livermore National Labs report (Reference 20) and Brookhaven National Laboratory report (Reference 23) indicates that the stiffness of the soil has little impact on the peak accelerations predicted in the cask. Thus the same soil model was assumed as that used in the Livermore report. The soil material properties assumed for the analysis are:

E = 6,000 psi

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.10-5 n = 0.45 p = 2.0368 x 10 -4 lb-sec2 / in 4 A4A.10.3 BOUNDARY CONDITIONS Only V2 of the cask was modeled with symmetry boundary conditions used to simulate the full structure. Non-reflecting boundaries were applied to the bottom and sides of the modeled soil not aligned with the plane of symmetry (bottom, left side, right side, and back) to prevent artificial stress waves from reflecting back into the model. Both dilatation and shear waves were damped as described in the LS-DYNA *BOUNDARY command.

An automatic surface to surface (contact_automatic_single_surface) contact definition was applied between all parts except the soil. The contact definition has a 0.5 penalty stiffness scale factor to prevent excessive contact stiffness leading to unrealistic part accelerations. A surface to surface (contact_surface_to_surface) contact definition was applied between the concrete and the soil with soft contact option 2. Soft contact option 2 was necessary between the soil and concrete as the materials have very different material stiffness. A conservatively low coefficient of friction (static and kinetic) of 0.25 was applied between all contact surfaces. It is conservative to use a low value for the coefficient of friction because less energy is absorbed due to friction resulting in greater impact acceleration forces.

A4A.10.4 INITIAL CONDITIONS AND LOADING The analysis begins with a 1 gap between the cask and concrete to allow for at least 5 ms of zero acceleration other than gravity. An initial velocity was applied to all parts of the cask model. The initial velocity was computed by equating potential and kinetic energies. Due to the initial 1 gap and gravitational acceleration, initial velocities were computed 1 shorter than the drop heights.

V = potential energy = mgh T = kinetic energy = V2 mv2 For an 18 Drop:

mgh = V2 mv2 ________

v = 2 gh = 2(386.4)(18 -1) = 114.62 in./sec.

A gravitational acceleration of 386.4 in/sec2 was applied to the cask and basket model.

A4A.10.5 RESULTS OF LS-DYNA ANALYSES The resulting rigid body acceleration time histories were computed by LS-DYNA. The rigid body accelerations were computed for the bottom plate, circumferential shell, and basket representation. The parts can be seen in Figure A4A.10-3.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.10-6 The peak filtered accelerations and corresponding time history plot for different parts of th the TN-40HT cask 18 end drop are listed below. All results were filtered with a 4 order low pass butterworth filter with a 350Hz cutoff frequency.

Results Summary Peak Time History Part Acceleration (g) Figure Number Shell 41.5 A4A.10-4 Bottom Plate 44.1 A4A.10-5 Basket Representation 28.8 A4A.10-6 Based on the Results shown in the above table, the maximum acceleration in the TN-40HT cask during the 18 inch accident condition end drop event is 44.1g and occurs in the bottom plate. Also from this table, the highest acceleration in the basket and fuel is 28.8g. However, since the basket and fuel were not modeled explicitly, the maximum acceleration (28.8g) must be multiplied by the appropriate dynamic load factor (DLF).

The maximum DLF for a triangular load is 1.52 (Reference 24). This results in a maximum loading of 43.8g.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.11-1 4A.11 REFERENCES

1. ANSYS Engineering Analysis System, Users Manual for ANSYS Release 8.0.

2. John Harvey, Theory and Design of Modern Pressure Vessel, Second Edition.

3. Stress Analysis of Closure Bolts for Shipping Casks, NUREG/CR-6007, April 1992.

th

4. Baumeister, T. Marks, L.S., Standard Handbook for Mechanical Engineers, 7 edition, McGraw Hill, 1967.

5. ASME Boiler and Pressure Vessel Code,Section II, Materials Specifications, Part D, 2004, through 2006 addenda.

6. Garlock Helicoflex Catalog, High Performance Seals and Sealing Systems, Catalog Reference HEL 1.

7. Machinery Handbook, 26th Edition, Industrial Press, 2000.

8. Helicoil Catalog, Heli-Coil 8-Pitch Inserts.

9. ANSI N14.6-1 986 Special Lifting Devices for Shipping Containers Weighing 10,000 lb or more.

10. WRC Bulletin 107, March 1979, Local Stresses in Spherical and Cylindrical Shells Due to External Loadings.

11. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsections NB and NCA, 2004, through 2006 addenda.

12. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NF, 2004 through 2006 Addenda.

13. NUREG/CR-3826, Recommendations for Protecting against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Greater than Four Inches Thick, April 1984.

14. NUREG/CR-1815, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers up to Four Inches Thick, August 1991.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4A.11-2

15. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code Section V and Section XI, 2004, through 2006 addenda.

16. SIR-98-1 10, Rev. 0, Allowable Flaw Evaluation of the Transnuclear TN-32 Cask, Structural Integrity Associates, Inc. 1998.

17. NUREG-612 Control of Heavy Loads at Nuclear Power Plants, July 1980.

18. LS-DYNA Keyword Users Manual, Volumes 1 & 2, Version 9.71s, Rev. 7600.398 August 17, 2006, Livermore Software Technology Corporation.

19. Witte, M. et. Al. Evaluation of Low-Velocity Impact Testing of Solid Steel Billet onto Concrete Pads and Application to Generic ISFSI Storage Cask for Tipover and Side Drop. Lawrence Livermore National Laboratory. UCRL-ID-1 26295, Livermore, California. March 1997.

20. NUREG/CR-6608, UCRL-1 D-1 2911," Summary and Evaluation of Low-Velocity Impact Tests of Solid Steel/Billet onto Concrete Pad," LLNL, February, 1998

21. Dilger, etc., Ductility of Plain and Confined Concrete under Different Stain Rates, ACI Journal, January-February, 1984.

22. Not Used.

23. BNL-NUREG-71 1 96-2003-CP, Impact Analysis of Spent Fuel Dry Casks Under Accidental Drop Scenarios, BNL, 2003.

24. Methods for Impact Analysis of Shipping Containers, NUREG/CR-3966, UCID-20639, LLNL, 1987.

25 HUHOMS HD Updated Final Safety Analysis Report, Revision 1

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.2-1 MECHANICAL PROPERTIES OF CASK BODY MATERIALS(1)(3)

Minimum Minimum Design Material Specification Yield Ultimate Application Stress (Nominal Composition) Strength Strength Value, psi (2)

______________________ ____________ Sy, psi Su, psi ___________

Shell flange, ASME SA-350, Grade Lid outer 37,500 70,000 Sm = 23,300 LF3 (3-1/2 Ni)

______________________ plate __________ _________ ___________

Lid outer ASME SA-203, Grade E plate, Inner 40,000 70,000 Sm = 23,300 (3-1/2 Ni) shell, Bottom

_______________________ inner plate ___________ __________ ____________

Shield shell, ASME SA-266, Class 2 Bottom 36,000 70,000 Sm = 23,300 or SA - 516 Gr. 70 (C-Si) shield Outer shell Lid shield ASME SA-516, Gr. 70 or plate, 36,000 70,000 Sm = 23,300 SA-105 (C-Mn-Si)

Protective

______________________ cover __________ _________ ___________

Lower and ASME SA-105 or SA-266 Upper 36,000 70,000 Sm = 23,300 CL 2 or CL 4, (C-Si)

________________________ Trunnions ____________ ___________ ____________

ASME SA-540 Gr. B24 or B23 CL 1 Lid Bolts 150,000 165,000 Sm = 50,000 (2Ni-3/4 Cr-1/3 Mo) ___________ __________ _________ __________

Vent / Drain port cover&

SA-193, Gr. B7 105,000 125,000 Sm = 35,000 Top neutron

_______________________ shield bolts ___________ __________ ____________

Notes:

1. Mechanical properties listed are for metal temperatures up to 100 °F to provide a baseline comparison of all structural materials. Temperature dependent properties required for structural analysis are provided in Table A4.2-18.

2. Values listed are the stress parameters which form the basis for structural analysis acceptance criteria.

3. Data are taken from tables in ASME Section II, Part D, 2004 including 2006 Addenda, unless otherwise noted.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.3-1 NORMAL AND HYPOTHETICAL ACCIDENT CONDITION INDIVIDUAL LOADS Load Case No. Individual Load Description IL - 1 Bolt Preload and Lid Seating Pressure IL - 2 Fabrication Stress IL - 3 1g Down (Cask vertical, supported at bottom)

IL - 4 Internal Pressure (100 psi)

IL - 5 External Pressure (25 psi)

IL - 6 Thermal Stress Due to Hot Environment (100 o F ambient)

IL - 7 Thermal Stress Due to Cold Environment (-40 o F ambient) 3g on Top Trunnion Lifting Load ( Calculating cask global stresses, IL - 8 Cask Vertical, 3g Up)

Bounding Loads for Seismic, Tornado and Flood ( 1g Lateral IL - 9 (resultant) + 2g Down)

IL-10 Trunnion Local Stress due to 3g Lifting

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.3-2

SUMMARY

MAXIMUM NODAL STRESS INTENSITIES IN CASK COMPONENTS

________ (See Figure A4A.3-2 through Figure A4A.3-6 for component definition)

___________ _______ Maximum Nodal Stress Intensity (ksi) __________ _________

Load Inner Shell Lid Lid Top Bottom Number Shell Shield Bottom

& Bottom Outer Shield Trunnion Trunnion Flange Shell Shield

_______ Inner Plate Plate Plate Region Region IL - 1 Bolt Pre- 0.07 4.47 8.03 0.25 0.24 0.01 0.24 0.0 load IL - 2 15.11 12.11 1.43 0.73 7.11 4.20 5.01 7.11 Fab.

IL-3 Gravity 0.08 0.11 0.07 0.07 0.08 0.08 0.03 0.08 1g___________ _______ ______ ________ ______ ________ __________ _________

IL - 4 Internal 2.40 3.69 2.12 2.24 1.97 5.62 0.69 1.97 Pressure IL - 5 External 0.60 0.92 0.53 0.56 0.49 1.41 0.17 0.49 Pressure IL - 6 Thermal 23.55 3.33 1.79 1.13 11.22 13.19 2.30 10.73 100ºF IL - 7 Thermal 25.08 3.33 1.77 1.03 12.59 14.45 2.29 12.11

-40° F IL - 8 3.92 1.39 1.32 0.50 4.39 6.98 1.85 2.23 Lifting 3G IL-9 0.71 0.26 0.22 0.16 1.21 2.04 0.19 1.23 Seismic IL-1 0 Trunnion Trunnion local stresses due to 3g are provided in Section A4A.6 Local Stress

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.4-1 BOLT SPECIFICATIONS AND DESIGN LOADS Bolt Bolt Size Material Design Loads

1. Bolt Preload (1100-1150 ft-lbs) 1.375 inch shank SA540, 2. Gasket seating load 1.5-8UN threads Lid Bolt Gr. B24 or B23, (1399 lb./in, Reference 6)

(tensile area = 1.492 2 Cl. 1 3. Pressure load (100 psig) in (Reference 4))

4. Temperature load (300 °F)

5. Bearing stress

1. Bolt Preload (60-65 ft-lbs) 0.75 inch 10UNC 2. Gasket seating load Vent/Drain (tensile area = 0.334 SA-1 93 B7 (1142 lb./in, Reference 6)

Cover Bolt in (Reference 4))

2 3. Pressure load (100 psig)

4. Temperature load (300 °F) 1.25 inch 7UNC 1. Bolt Preload (90-100 ft-lbs)

Top Neutron (tensile area = 0.969 A-193 B7 2. 3g vertical up Shield Bolt in (Reference 4))

2 3. Temperature load (300 °F)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.4-2 DESIGN PARAMETERS FOR LID CLOSURE BOLTS Db Nominal diameter of closure bolt; 1.375 in.

K Nut factor for empirical relation between the applied torque and achieved preload is

_______ 0.135 for neolube Q Applied torque for the preload (in.-lb.)

D ,b Closure lid diameter at bolt circle, 79.31 in.

D is Closure lid diameter at the seal (inner) = 74.315 in D ,g Closure lid diameter at the seal (outer) = 75.882 in.

6 Ec Youngs modulus of cask wall material (SA-350, LF3, 300 °F), 26.7x 10 psi.

________ (Reference 5) 6 E, Youngs modulus of lid material (SA-350, LF3, 300 °F), 26.7 x 10 psi. (Reference 5)

Nb Total number of closure bolts, 48 Nu, Poissons ratio of closure lid, 0.3, (Reference 4, p. 5-6 use nominal value).

Pei Inside pressure of cask, 100 psig.

D ,o Closure lid diameter at outer edge, 82.75 in.

P,i Pressure inside the closure lid, 100 psig.

tc Thickness of cask wall, 8.8 in.

tl Thickness of lid, 10.0, 4.5 in.

-6 -6

,b Thermal coefficient of expansion, bolt (SA-540, Gr B24), 6.4 x 1 0 at R.T., 6.9 x 1 0 in.

-1 -1

_______ in. °F at 300 °F (Reference 5)

-6 -6

,c Thermal coefficient of expansion, cask (SA-350, LF3) 6.4 x 10 at R.T., 6.9 x 10 in.

-1 -1

_______ in. °F at 300 °F (Reference 5)

Thermal coefficient of expansion, lid (SA-350, LF3) 6.4 x 1 0 in. in.

-6 -1

,l -6 R.T., 6.9 x 1 0

-1

_______ °F at 300 °F (Reference 5) 6 Eb Young's modulus of bolt material (SA-540, B24, 300 °F), 26.7 x 10 psi. (Reference 5)

LF A factor to account for any difference between the rigid body acceleration and the

________ acceleration of the contents and closure lid = 1.1 Sy, Yield strength of closure lid material (SA-350, LF3, 300 °F), 33,200 psi. (Reference 5)

S u, Ultimate strength of closure lid (SA-350, LF3, 300 °F), 70,000 psi. (Reference 5)

Syb Yield strength of lid bolt material (see Table A4A.4-3)

Sub Ultimate strength of lid bolt material (see Table A4A.4-4)

P,o Pressure outside the lid Lb Bolt length between the top and bottom surfaces of closure, 4.5 in.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.4-3 ALLOWABLE STRESSES IN LID CLOSURE BOLTS FOR NORMAL CONDITIONS Normal Condition Allowables Temperature Yield Stress(1) ____________ ____________ _____________

(°F) (ksi) (2, 4) (3, 4)

S.I.

(5)

Ftb Fvb

______________ ______________ (ksi) (ksi) (ksi) 70 150.0 100.0 60.0 135.0 200 144.0 96.0 57.6 129.6 300 140.3 93.5 56.1 126.3 400 137.9 91.9 55.2 124.1 Notes:

1. Yield stress values are from Reference 5

2. Allowable Tensile stress, Ftb = 2/3 S, (Reference 3, Table 6.1)

3. Allowable shear stress, Fvb = 0.4 S, (Reference 3, Table 6.1)

4. Tension and shear stresses must be combined using the following interaction equation:

2 2 sttb yb 2

+£ 2

1.0 (Reference 3)

F tb Fyb

5. Stress intensity from combined tensile, shear and residual torsion loads, S.I. £ 0.9 S, (Reference 3, Table 6.1)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.4-4 ALLOWABLE STRESSES IN LID CLOSURE BOLTS FOR ACCIDENT CONDITIONS Accident Condition Allowables (1)

Temperature Yield Stress o

( F) (ksi) 0.6 Si (3)

Ftb (2, 4) (3, 4)

Fvb

______________ ______________ (ksi) (ksi) (ksi) 100 150.0 90.0 115.5 69.3 200 144.0 86.4 115.5 69.3 300 140.3 84.2 115.5 69.3 400 137.9 82.7 115.5 69.3 Notes:

1. Yield and tensile stress values are from Reference 5, Note that Su is 165 ksi at all temperatures of interest

2. Allowable Tensile stress, Ftb = MINIMUM (0.7 Su, Si), where 0.7 Su = 0.7 (165.0) =

115.5 ksi. (Reference 3, Table 6.3)

3. Allowable shear stress, Fvb = MINIMUM (0.42 Su, 0.6 Si), where 0.42 Su = 0.42 (165.0) = 69.3 ksi. (Reference 3, Table 6.3)

4. Tension and shear stresses must be combined using the following interaction equation:

2 2 st tb

-- + --

£ 1 .0 (Reference 3)

Ftb Fib

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.4-5 LID BOLT INDIVIDUAL LOAD

SUMMARY

Prying Prying Non-Prying Torsional Applied Force, Moment, Load Case Tensile Force, Moment, Load (in. lb.) -1 Ff Mf Fa (lb.) Mt (lb.in . ) (in. lb. in.-1)

Maximum 74,343 6,900 0 0 Torque Preload Residual Minimum 71,111 6,600 0 0 Torque Gasket Seating Load 13,753 0 0 0 Internal 100 psig Internal 9,422 0 1,983 19,656 Pressure___________________________ ______________ ___________ _________ _____________

Thermal 300° F 0 0 0 0

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.4-6 LID BOLT LOAD COMBINATIONS Non-Prying Prying Prying Torsional Load Force, Moment, Combination Description Tensile Moment, 1 Case Ff (lb.in. ) Mf Force, Mt (in. lb.) -1

______ ____________ _________ (in. lb. in. )

Fa (lb.) __________ __________ __________

A.

Preload + Maximum 74,343 6,900 0 0 Temperature Torque _______ __________ __________ __________

(Normal B.

1.

Condition) Minimum 71,111 6,600 0 0

______ ____________ Torque _______ __________ __________ __________

2. Gasket Seating Load +

23,175 0 1,983 19,656 Internal Pressure

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.4-7 LID BOLT STRESS

SUMMARY

Normal Condition Accident Condition Stress Type Stress Allowable Stress Allowable Average Tensile 50.1 93.5 50.1 115.5 (ksi) ________ ____________ ________ ___________

Shear (ksi) 13.5 56.1 13.5 69.3 Combined stress Not Required 57.8 126.3 intensity (ksi) ________ ____________ (Reference 3)

Interaction Eq.

0.345 1 0.226 1 R2 + Rs2 < 1 Not Required Bearing (ksi) 32.2 33.2 (Reference 3)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.5-1 ALLOWABLE STRESSES IN VENT/DRAIN, AND TOP NEUTRON SHIELDBOLTS FOR NORMAL CONDITIONS Normal Condition Allowables Temperature Yield Stress(1) ____________ ____________ _____________

(°F) (ksi) (2, 4) (3, 4)

S.I.

(5)

Ftb Fvb

______________ ______________ (ksi) (ksi) (ksi) 70 105.0 70.0 42.0 94.5 200 98.0 65.3 39.2 88.2 300 94.1 62.7 37.6 84.7 400 91.5 61.0 36.6 82.4 Notes:

1. Yield stress values are from Reference 5

2. Allowable Tensile stress, Ftb = 2/3 S, (Reference 3, Table 6.1)

3. Allowable shear stress, Fvb = 0.4 S, (Reference 3, Table 6.1)

4. Tension and shear stresses must be combined using the following interaction equation:

2 2 st 1.0 (Reference 3) tb yb 2 2 F tb Fyb

5. Stress intensity from combined tensile, shear and residual torsion loads, S.I. ^ 0.9 S, (Reference 3, Table 6.1)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.5-2 ALLOWABLE STRESSES IN VENT/DRAIN, AND TOP NEUTRON SHIELDBOLTS FOR ACCIDENT CONDITIONS Accident Condition Allowables (1)

Temperature Yield Stress o

( F) (ksi) 0.6 Si (3)

Ftb (2, 4) (3, 4)

Fvb

______________ ______________ (ksi) (ksi) (ksi) 100 105.0 63.0 87.5 52.5 200 98.0 58.8 87.5 52.5 300 94.1 56.5 87.5 52.5 400 91.5 54.9 87.5 52.5 Notes:

1. Yield and tensile stress values are from Reference 5, Note that Su is 125 ksi at all temperatures of interest

2. Allowable Tensile stress, Ftb = MINIMUM (0.7 Su, Si), where 0.7 Su = 0.7 (125.0) =

87.5 ksi. (Reference 3, Table 6.3)

3. Allowable shear stress, Fvb = MINIMUM (0.42 Su, 0.6 Si), where 0.42 Su = 0.42 (125.0) = 52.5 ksi. (Reference 3, Table 6.3)

4. Tension and shear stresses must be combined using the following interaction equation:

2 2 st tb yb

+£ 1.0 (Reference 3)

FF;2b

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.5-3 VENT AND DRAIN COVER BOLT INDIVIDUAL LOAD

SUMMARY

Non-Prying Prying Prying Torsional Load Applied Force, Moment, Tensile Moment, Case Load Ff Mf Force, Mt (in. lb.) -1 (lb.in . ) (in.lb.in .-1)

________ _________ __________ Fa (lb.) __________ ________ _________

Maximum 7,704 390 0 0 Torque Preload Residual Minimum 7,111 360 0 0 Torque Gasket Seating Load 4,740 0 0 0 Internal 100 psig Internal 318 0 185 170 Pressure_____________________ _________ ___________ ________ __________

Thermal 300° F 6,544 0 0 0

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.5-4 VENT AND DRAIN COVER BOLT LOAD COMBINATIONS Non Prying Prying Prying Torsional Load Force, Moment, Combination Description Tensile Moment, Case Ff Mf Force, Mt (in. lb.) -1 (lb.in . ) (in.lb.in .-1)

______ _____________ _________ Fa (lb.) __________ ________ __________

A.

Preload + Maximum 14,248 390 0 0 Temperature Torque ______ _________ _______ _________

1 (Normal B.

Condition) Minimum 13,655 360 0 0

______ _____________ Torque _______ __________ ________ __________

2. Gasket Seating Load +

5,058 0 185 170 Internal Pressure

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.5-5 VENT AND DRAIN COVER BOLT STRESS

SUMMARY

Normal Condition Accident Condition Stress Type Stress Allowable Stress Allowable Average Tensile (ksi) 42.7 62.7 42.7 87.5 Shear (ksi) 7.6 37.6 7.6 52.5 Combined Stress Not Required 51.8 84.7 Intensity (ksi) ________ ____________ (Reference 3)

Interaction Eq.

0.504 1 0.259 1 R2 + Rs2 < 1

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.6-1

SUMMARY

OF UPPER TRUNNION STRESSES (YIELD STRENGTH CASE)

Upper Trunnion 6/3g Load (3g/Trunnion)

Trunnion Location AA BB Stress Area (in 2 ) 113.9 79.8 Area Moment of Inertia (in4 ) 3,082.0 755.6 Shear Force (lb) 750,000 750,000 Applied Moment (in lb) 5,062,500 1,590,000 Shear Stress (psi) 6,586 9,402 Bending Stress (psi) 13,962 11,837 Stress Intensity (psi) 19,195 22,220 Allowable Stress (psi) 31,800 31,800

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.6-2

SUMMARY

OF UPPER TRUNNION STRESSES (ULTIMATE STRENGTH CASE)

Upper Trunnion 10g Load (5g/Trunnion)

Trunnion Location AA BB Stress Area (in 2 ) 113.9 79.8 Area Moment of Inertia (in4 ) 3082.0 755.6 Shear Force (lb) 1,250,000 1,250,000 Applied Moment (in lb) 8,437,500 2,650,000 Shear Stress (psi) 10,976 15,671 Bending Stress (psi) 23,271 19,728 Stress Intensity (psi) 31,991 37,033 Allowable Stress(psi) 70,000 70,000

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.6-3 UPPER TRUNNION SHIELD SHELL LOCAL STRESS

SUMMARY

FOR 3G LIFTING LOAD Upper Trunnions, 3g Longitudinal Load (T = 300°F)

Allowable Stress Allowable 1.5 3.0 Sm Location Stress Intensity (ksi)

Intensity Sm (ksi) (ksi)

PL Pb Q (ksi) PL (Reference 11) (Reference

____________ __________________ ___________________ ______________________________ 11)

Outside (1) (1) edge of 4.834 + 1.97 (1) 8.072 + 1.97 + 12.59 31 8 63.6 shield = 6.804 . = 22.632 shell. _____________ ______________ ______________________ __________

Inside (1) 4.834+ 1.97(1)

(1) edge of 7.511 + 1.97 + 12.59 31.8 63.6 shield = 6.804 = 22.071 shell. _____________ ______________ ______________________ __________

1. Note that bounding internal pressure and thermal stress values are taken from Table A4A.3-2.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.6-4 UPPER TRUNNION LOCAL STRESS DUE TO 3G LONGITUDINAL LIFTING LOAD 3gLifting Load _____ ___ ___ ___

I ILII1I1IOI1 LOading GeoiiietiyCask Loadiiip ________ _________ ________

Cask Weight lb 2SUUUU Gamma Shield Thickness (in) 5.81U Longitudinal g 3 _______ _______

Mean Radius (in) 4U.405 Vertical g U _________ ________

Moment Arm (in) 9.F38 Trunnion Outer Radius (in) .5UU Lateral g U _________ _________

Circumfrential Trunnion Moment Mc (in Ib) U Geometry Factor Gamma F3.954 _______________ _________ _________ _________

Longitudinal Trunnion Moment ML n lb -322bUU Geometry Factor Bets 0.184 Torsional Trunnion Moment Mt (in l ____________ U ____________

P (Ib) ____________ 0 ____________

Circumfrential Loading Vc (Ib) ____________ U ____________

Longitudinal Loading VL (Ib) -3750UU Refeieiice Figuie Refeience Ct ive fioiii Fij r1tIIti1)IieI AhsoIut Stiess (psi Au Al Bu BI Cu CI Du DI 3c or 4c ____________ 0 ____________ U.OUU U.0 0.0 U.0 U.0 U.0 U.0 U.0 U.0 ILU ic or2c-i ____________ U ____________ U.OUU U.0 0.0 U.0 U.0 U.0 U.0 U.0 U.0 UJJ 3 ____________ U ____________ U.OUU U.0 _____________ _______ ________ _______ U.0 U.0 U.0 ftU la ____________ 0 0.000 0.0 _____________ _______ ________ _______ 0.0 0.0 0.0 ___L 3b ___________U.S -2U74.771 -1U37.4 iU37.4 iU37.4 -1037.4 -iU37.4 _______ _______ _______

lb or lb-i U.U68 -88572.588 -SU2i.2 6U2i.2 -SU2i.2 -6021.2 6U2i.2 ______ _______ ______

Suiiiiiiatioii of Phi Stie ___________________ Ub8. -3983.8 -0b8E 3983.8 U.0 U.0 U.0 ftU 3c or 4c ____________ U_____________ U.OUU U.0 0.0 U.0 U.0 U.0 U.0 U.0 U.0 UJJ ic-i or2c ____________ U_____________ U.OUU U.0 0.0 U.0 U.0 U.0 U.0 U.0 U.0 UJJ 4a ____________ 0 0.000 0.0 _____________ _______ ________ _______ 0.0 0.0 0.0 ___L 2a ____________ U_____________ U.OUU U.0 _____________ _______ ________ _______ U.0 U.0 U.0 ftU 4b U.i36 -2U74.771 -28U.i 23U.i 28U.1 -28U.i -28U.i _______ ________ _______

2b or2b-i UJJ9 -88572.588 -7791.6 7791.6 -7791.6 -7791.6 7791.6 ______ _______ ______

Suiiiiiiatioii of Clii Sties __________________ 8U7i.8 -7511.4 -8071.8 7511.4 U.0 U.0 U.0 UJJ Torsional Shear Stress 0.0 _____________________ 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Circumfrential Shear Stress U.0 _____________________ 0.0 U.0 U.0 U.0 _______ ________ _______

Loncitudinal Shear Stress -2417.1 __________________ ___________ ______ _______ ______ 2417.1 2417.1 -2417.1 -2417.1 Su iii iii atio ii of Tau Sti __________________ 0.0 U.0 U.0 U.0 2417.1 2417.1 -2417.1 -2417.1 Stiess Intensity Root 1 ___________________ 8U7i .8 3983.8 8U58.8 7511.4 2417.1 2417.1 2417.1 2417.1 Stiess Intensity Root 2 8U58.8 7611.4 8U7i.8 3983.8 2417.1 2417.1 2417.1 2417.1 Stiess Intensity Root 3 r1ax Sti e Iiiteiiity rYleiiil)i ane Sties Inteni1y Root 1 rYleIllI)I ane Sties Intensity Root 2 23U.i 28U.1 iU37.4 iU37.4 2417.1 2417.1 2417.1 2417.1 ryleiiihiaiie Stis Iiiteiiity Root 3 757.3 7b7.3 7b7.3 7b7.3 4834.1 4834.1 4834.1 4834.1 r1ax r1eiiibiaiie Stiess Iiiteiisity 4834.1

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Table A4A.7-1 MAXIMUM STRESS INTENSITIES FOR EACH LOADING CONDITION Max. Stress Allowable Factor of LOADING CONDITION

_________________________ Intensity (ksi) (ksi) Safety 25 psig internal pressure 4.92 33.6 6.83 25 psig internal pressure + 3g 7.04 33.6 4.77 inertia load (Cask in Vertical Orientation) ___________________ _________________ _______________

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Table A4A.7-2 STRESS AT WELD LOCATIONS Weld Location Max. Stress (Figure 3.10.3-1) Intensity (ksi) 25 psig Internal Pressure Location 1 1.70 Location 2 4.92 Location 3 4.69 Location 4 1.65 25 psig Internal Pressure + 3g inertia load Location 1 5.27 Location 2 7.04 Location 3 2.67 Location 4 3.37 25 psig Internal Pressure + 3g Downward Load Location 1 2.86 Location 2 5.78 Location 3 5.50 Location 4 2.79 Note: All the stresses are less than the allowable stress of 21.00 ksi.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.8-1 TOP NEUTRON SHIELD COVER BOLT INDIVIDUAL LOAD

SUMMARY

Non-Prying Prying Prying Torsional Applied Tensile Force, Moment, Load Case Moment, Force, Load Ff Mf Mt (in. lb.)

Fa (lb.) (lb.in .-1) (in.lb.in .-1)

Maximum 7,111 600 0 0 Torque Preload Residual Minimum 6,400 540 0 0 Torque Thermal 300 °F 0 0 0 0 Acceleration 3g Vertical 1,485 0 0 0

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.8-2 TOP NEUTRON SHIELD COVER BOLT LOAD COMBINATIONS Non-Prying Prying Prying Torsional Load Force, Moment, Combination Description Tensile Moment, Case Ff Mf Force, Fa Mt (in. lb.)

(lb.in .-1) (in.lb.in .-1)

(lb.)

A.

Preload + Maximum 7,111 600 0 0 Temperature Torque 1.

(Normal B.

Condition) Minimum 6,400 540 0 0 Torque 2.

Acceleration Loads 1,485 0 0 0

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.8-3 TOP NEUTRON SHIELD COVER BOLT STRESS

SUMMARY

Normal Condition Accident Condition Stress Type

_________________ Stress Allowable Stress Allowable Average Tensile 7.3 62.7 7.3 87.5 (ksi) _________ ______________ I Shear (ksi) 2.2 37.6 2.2 52.5 Combined Stress Not Required 8.6 84.7 Intensity (ksi) _________ ______________ (Reference 3)

Interaction E.Q.

0.017 1 0.009 1

+R <1 2

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.9-1

SUMMARY

LINEARIZED STRESS COMPONENTS - NORMAL CONDITION Cask Location Membrane Stress (ksi) Bending Stress (ksi) from Max. Stress Type (1)

FIGURE A4A 9 .1 I (Rad.) (Tang.) (Axial) (Rad.) (Tang.) (Axial)

S Y (N6) 1.74 3.61 0.38 2.65 1.73 2.40

1. Weld SZ (N6) 1.76 3.26 0.38 2.65 1.75 2.59 S Y (N5) -0.30 1.46 -0.70 0.23 0.14 0.01

2. Weld S Z (N5) -0.20 1.53 -0.54 0.22 0.11 0.04

3. Weld S Y, Z (N3) 1.0 1.06 0.19 0.04 0.09 0.22 S X, (N5) 1.89 3.38 -0.22 4.52 1.15 0.84

4. Bottom Shield S Y (N5) 2.05 3.55 0.27 4.41 1.54 0.57 S Z (N5) 0.37 3.45 0.62 1.51 0.48 1.21 S Y , (N4) -0.30 2.82 0.61 0.27 1.34 2.00

5. Shield Shell-Mid S Z (N4) -0.30 2.88 1.35 0.29 0.82 1.66 S Y (N4) 2.18 4.57 0.87 6.99 2.41 0.03

6. Shield Shell-End

________________ S Z (N4) 2.07 4.06 0.91 7.01 2.35 0.06 S X (N4) 1.66 0.38 -3.18 3.73 3.24 7.80

7. Shell Flange S Y (N4) -0.44 1.85 2.52 0.59 0.85 2.31

_______________ S Z (N4) -0.44 1.69 2.55 0.61 0.79 2.29 S X (N3) -0.07 0.65 0.17 0.12 0.37 0.12

8. Shield Plate S Y (N3) -0.06 0.58 0.13 0.10 0.48 0.12

________________ S Z (N3) -0.06 0.65 0.17 0.12 0.38 0.12 S X (N5) 0.17 0.91 -0.17 1.33 0.41 0.06

9. Lid Outer Plate

_______________ S Y,Z (N5) 0.13 1.01 0.37 1.09 0.41 0.43 S X (N5) -0.04 -3.02 -4.72 1.05 2.40 7.69

10. Inner Shell &

Bottom Inner S Y (N5) -9.82 -3.42 -0.26 2.95 3.46 7.23 Plate

_______________ S Z (N5) -13.04 -4.30 -0.10 2.21 1.51 5.64

1. S - Radial Stress S X Y - Tangential Stress S - Axial Stress Z

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.9-2

SUMMARY

LINEARIZED STRESS COMPONENTS - ACCIDENT CONDITION Cask Location Membrane Stress (ksi) Bending Stress (ksi)

From Max. Stress SX SY S ZS XS YS Figure A4A.9-1 Type 1) (Rad.) (Tang.) (Axial) (Rad.) (Tang.) (Axial)

S Y (A2) 0.09 -0.63 -4.92 0.01 0.16 0.60

1. Weld S Z (A2) 0.09 -0.63 -4.92 0.01 0.16 0.60 S Y (A4) -0.10 0.43 -0.32 0.13 0.03 0.03

2. Weld

________________ S Z (A4) -0.10 -0.36 -0.31 0.13 0.0 0.03 S Y (A3) 0.51 0.41 1.06 0.24 0.43 1.19

3. Weld

_______________ S Z (A3) 0.52 0.42 1.09 0.25 0.46 1.29 S X (A4) 0.45 0.64 0.49 2.80 0.78 0.63

4. Bottom Shield S Y (A4) 0.44 0.53 -0.04 1.82 2.01 0.16

_______________ S Z (A4) -0.14 0.76 1.37 1.49 0.72 1.66 S(A4) -0.25 3.08 0.74 0.28 0.38 0.06

5. Shield Shell-Mid S Z (A4) -0.26 3.04 0.90 0.28 0.25 0.0 S(A4) 1.64 1.12 1.23 2.54 1.76 3.37

6. Shield Shell-End S Z (A4) 1.18 0.56 0.36 1.24 1.51 4.26 S X (A3) 1.48 0.09 -2.95 3.67 3.42 8.34

7. Shell Flange S Y (A3) 1.47 0.90 -2.96 3.67 3.42 8.34

_______________ S Z (A3) -0.22 0.85 2.18 1.11 0.85 2.55 S X ,S Y (A3) 0.79 0.79 -0.02 3.36 3.36 0.10

8. Shield Plate

________________ S Z (A3) 0.16 0.60 0.13 0.34 0.28 0.91 S X (A4) 0.47 0.54 -0.19 1.78 0.35 0.0

9. Lid Outer Plate S Y (A4) 0.72 0.72 -0.02 1.48 1.48 0.10

_______________ S Z (A4) 0.42 0.71 0.38 1.42 0.32 0.43 S X (A3) 8.40 2.36 -2.17 4.10 2.48 3.65

10. Inner Shell &

Bottom Inner S Y (A3) 8.40 2.36 -2.17 4.10 2.48 3.65 Plate S Z (A3) -2.43 -1.99 -4.46 0.67 2.55 7.91

1. S X - Radial Stress S Y - Tangential Stress S Z - Axial Stress

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.9-3

SUMMARY

STRESS INTENSITY FACTORS FOR NORMAL CONDITION Stress Allow.

Crack Crack Intens.

Thick. Max. Stress Factors tdepth, length, Critical Crack Factor Component Stress Intens. of a l Direction 1)

(in) Type" Factor Safety (in) (in) (ksi- in) (ksi- in)

1. Weld 1.50 0.225 2.25 S Axial Y 4.64 14.86 3.20

2. Weld 1.00 0.150 1.50 S Axial Y 1.30 14.86 >10

3. Weld 0.75 0.1125 1.125 S Axial Y 0.80 14.86 >10

4. Bottom 7.25 0.500 5.00 S Tangential 8.61 14.86 1.73 Shield X

5. Shield Shell-7.25 0.500 5.00 S Axial 5.64 14.86 2.64 Mid Y

6. Shield Shell 7.25 0.500 5.00 S Axial 9.31 14.86 1.60 End Y

7. Shell Flange 4.6 0.500 5.00 S Hoop Z 9.32 14.86 1.59

8. Shield Plate 5.5 0.500 5.00 S Axial Y 1.69 14.86 9.27

9. Lid Outer 4.5 0.500 5.00 S Axial 2.0 14.86 7.45 Plate Y

10. Inner Shell &

Bottom Inner 1.5 0.225 2.25 S Hoop Z 5.88 14.86 2.53 Plate

1. S Y - Hoop Stress - Results in Axial crack S Z - Axial Stress - Results in Hoop crack S X - Radial Stress - Results in Tangential crack

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4A.9-4

SUMMARY

STRESS INTENSITY FACTORS FOR ACCIDENT CONDITION Stress Allow.

Crack Crack Intens.

Thick. Max. Critical Stress Factors tdepth, length, Factor Component Stress Crack Intens. of a l 1)

Type Direction Factor 1)

(in) Safety (in) (in) (ksi- in) (ksi- in)

1. Weld 1.50 0.225 2.25 S Hoop Z 0.54 33.23 >10

2. Weld 1.00 0.150 1.50 S Axial Y 0.41 33.23 >10

3. Weld 0.75 0.1125 1.125 S Hoop Z 1.40 33.23 >10

4. Bottom Shield 7.25 0.500 5.00 S Tangential X 4.25 33.23 7.82

5. Shield Shell-7.25 0.500 5.00 S Axial 4.68 33.23 7.10 Mid Y

6. Shield Shell-7.25 0.500 5.00 S Hoop 7.15 33.23 4.65 End Z

7. Shell Flange 4.6 0.500 5.00 S Hoop Z 13.00 33.23 2.56

8. Shield Plate 5.5 0.500 5.00 S Tangential X 5.20 33.23 6.40

9. Lid Outer Plate 4.5 0.500 5.00 S Tangential X 3.22 33.23 >10

10. Inner Shell &

Bottom Inner 1.5 0.225 2.25 S Tangential X 10.89 33.23 3.05 Plate S Y - Hoop Stress - Results in Axial crack S Z - Axial Stress - Results in Hoop crack S X - Radial Stress - Results in Tangential crack

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 AI .JE ..

7 : ii F LITh lIT S Powe.rGraphics EFACET=i TYPE IUI.I 211 =1 L'IST=9L 963 YF =9c:' -

• 37*5 PRECISE HIDL:'EI*.J FIGURE A4A.3-1 TN-40HT STORAGE CASK FEM REPRESENTATION

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 AI .JE ..

3:4 1 F LITh lIT S Powe.rGraphics EFACET=i TYPE IUI.I

=1 yr =1

=1 DIST=4:

YF =176.75 ZF PRECISE HIDDEN FIGURE A4A.3-2 FINITE ELEMENT MODELLID OUTER PLATE AND SHIELD PLATE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 AI .JE ..

3. 4 : .5 C:'

F LITh lIT S Powe.rGraphics EFACET=l TYPE IUI.I

=1 yr =i

=1 DIST=.52 .211 YF =173.125 ZF_.: .

PRECISE HIDDEN FIGURE A4A.3-3 FINITE ELEMENT MODEL - SHELL FLANGE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ANS ..3 -

ELEI*IEITTS Pc:we.rGraphicE EFACFT=i TYPE NUll

=1 Yv =1 zv =1 L'IT=97 YE =:;:;iL...E ZF =jj?*75 PRCI HIr:'r:'EI*

FIGURE A4A.3-4 FINITE ELEMENT MODEL - INNER SHELL AND BOTTOM INNER PLATE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 j .J y ' C' ELIII.ITS Powe.rGraphicl5 EFACFT=i TYPE NtJIi

>iv =1 yr =

2W =1 L'IST=i1I:'4 YF =:BcL4 .5

• 37*5 PRCIS HIDDEIJ FIGURE A4A.3-5 FINITE ELEMENT MODEL - SHIELD SHELL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ANS ..3 -

41 : 57 ELEI*IEITTS Pc:we.rGraphicE EFACFT=i TYPE NUll

=1 L'IT = .54 4 .

YE =3.62.5 ZF =_77E:75 PRCI HIr:'r:'EI*

FIGURE A4A.3-6 FINITE ELEMENT MODEL - BOTTOM SHIELD

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ANSYS 80 OCT 30 2006 08:30:28 ELEMENTS Powe rGraphic s EFACET=1 TYPE NUN I I I I I I I I I Ii I 1111111 II LIII 1111111 zv =1

• III III liii DIST=99 963 YF =90.875 ZF =-22375 PRECISE HIDDEN 11 ooiiiiiirLrrr

• III III liii UDHU"' 1111111111 I 11111111' 1111111 I I I I I I I I I I I

ii.

FIGURE A4A.3-7 COUPLING AND BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 AI .JE ..

0012 3'::'

13 : 3.7 : .59 F LITh lIT S Powe.rGraphics EFACET=i FLAT IVJI.I r7r =1 DIST = 99. 963 YE =90 - 87.5 ZF PRCI HIr:'r:'EI*

E L'GE FIGURE A4A.3-8 BOLT PRELOAD AND LID SEATING PRESSURE - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 pj*..JSyS ccI 13 : 44:

ELEI'IEI.JTS Pore.rGraphics EEACET=i i***r_r i*i._*.*i ZV =1 EIIsT=9c. -

YF =9c:i -

ZE =- ..:.*375 PREc2IS HIDDEIJ E DGE PRE S-NORM 515 TN4c:'H - cask Fdbricdtiorl ci - 0 Radial Interference.

FIGURE A4A.3-9 FABRICATION STRESS - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

()CJ 3i2c5 c:'9:i.E:i.E ELEI'IEI.JTS Pc:re.rGraphics EEACET=i i***r_I i*i._***i ACE L ZV =1 DIST=9c. -

YF =9':' -

ZE =- ...375 PREc2IS HIDDEIJ E DGE PRE SNORM 19 . 673 TIT4111H - Lask 1G 1I'cnrn on Bctc.crrL FIGURE A4A.3-10 1G DOWN - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ccl 14 : 44 : c: 7 ELEMENTS PO.te.rGraphicE EFACET=i I*IPI I*.jI_Jr*4 ZV =1 DIST=99 963 YE =9c:'

ZF =-i.

PRECISE HIDDEN E r:'c4E PRE S-I*.JE:RI.1 1 i::

- cask Internal Pre.,i_ire. , ic:' c:' psiq FIGURE A4A.3-11 INTERNAL PRESSURE - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 15:c:iS:..

ELEMENTS TEMPERATURE S TIIIN=211 - 47.5 TI.A.=363:

=1 L:'IST=ic:I.5 . j .

YF =9c:i -

ZE =- ..:.*375 PREc2IS HIDDEN E DGE 11 475

- 4 24.5 4 211 14.5 31 i:: 1 3 346 947

- 1 TI.14c:'H - cask Thermal Stre.. ,

FIGURE A4A.3-12 THERMAL STRESS 100° F ENVIRONMENT - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 P.J*.JSYS i.5:i:;:l7 ELEMEI.JTS TEMPERATURES TMII..J=:E:9.744 TILZ=247 -

DIST=9:E: -

YF =9c:i -

ZE =- ...375 A: ...A.E4Ei:16 PRECISE HIL'L'EN E r:'ctE

- 1II i.247i:'?

14 .7 i:.

i:

iQ4c7:7

- 747 TN4C:'H - Cask Thermal Stress, 4C:'F FIGURE A4A.3-13 THERMAL STRESS -40° F ENVIRONMENT - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 t'E lx x'::'

i: : 4 : 2.0 P.m. rlrpbi s F1ET1

!T N1!!

71 i:i.;?F prI BIDDl ELI PE -NOw

- Liftiriq !Iz FIGURE A4A.3-14 3G VERTICAL LIFTING - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 pj*..JSyS ccI 1.5:46:20 ELEI'IEI.JTS Pore.rGraphics EEACET=i i***r_r i*i._*.*i ACE L ZV =1 DIST=99 963 YE =9c:' :k75 ZE' PRECISE HIL'L'EIJ E r:'ctE PRE S-I..JE:RN

- 4 11

.iT:i:7 72 TN4c:'H - c_.as k Sc. i srtic Loading FIGURE A4A.3-15 SEISMIC - LOADING AND DISPLACEMENT BOUNDARY CONDITIONS

Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ML y i.0 C

V L k--- Tninnii:iii Ru?1:IresentItiI:Ili A .L '..I 1p,TZ% r:)

1

( isk R.u?1'rI-SI-lltati':'ll Ti:'1:' *:if &_is1.:

Lateral Force I I.ii Ei1 ut I

'Eu Fi.:.iit .:.f 4_:isl.:

?1 ic LoHgitudiital Force -

Fr cunE Trrnuuu:uii

':u ttu:ulLL .:uf si:

FIGURE A4A.6-2 ILLUSTRATION OF DEFINING FORCES, MOMENTS, AND STRESS LOCATIONS

Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

'-1 Internal Pressure Figure A4A.7-2 FINITE ELEMENT MODEL - BOUNDARY CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 9

8 AIJSI .

7:11 E LEIIE IT S c:We.rc4rE1I::hic EFACET=i TYPE NTJIVI

=1 DI ST=9 9.963

= g c:' . 875 ZF =2**...75 PREc_ISE HII:'lI:'EN Note: Numbers indicate selected locations for fracture toughness evaluation.

FIGURE A4A.9-1 CASK BODY CRITICAL LOCATIONS FOR STRESS AND FRACTURE EVALUATION

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 70 60 50 III-cii 43 LU 20 10

-40 -20 -ID 0 10 20 30 Test Temperature, F FIGURE A4A.9-2 CHARPY V-NOTCH TEST RESULTS FOR SA-266 FORGING

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A4A.10-1 OVERVIEW OF FINITE ELEMENT MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A4A.10-2 OVERVIEW OF TN-40HT CASK FINITE ELEMENT MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Bas Sh Bottom P FIGURE A4A.10-3 PARTS ANALYZED FOR ACCELERATION TIME HISTORY

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Shell 20000

('1 15000 0) 10000 4-0 -^- Filtered I- 5000

-5000 Time (sec)

FIGURE A4A.10-4 CASK SHELL ACCELERATION TIME HISTORY (350HZ FILTER)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Bottom Plate 20000 15000 c'l 10000 0,

5000 s Filtered 1 -5000 0

ci:

-10000

-15000 Time (sec)

FIGURE A4A.10-5 CASK BOTTOM PLATE ACCELERATION TIME HISTORY (350HZ FILTER)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Basket 12000 10000 C.) 0 0) 6000

-- Filtered 4000 2 2000 C.)

0 Time (sec)

FIGURE A4A.10-6 CASK BASKET ACCELERATION TIME HISTORY (350HZ FILTER)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-1 APPENDIXA4B STRUCTURAL ANALYSIS OF THE TN-40HT BASKET A4B.1 INTRODUCTION This appendix presents the structural analysis of the TN-40HT basket. The basket is a welded assembly of stainless steel boxes designed to accommodate 40 PWR fuel assemblies. The stainless steel fuel compartments are fusion welded to Type 304 stainless steel structural plates, sandwiched between components. The fusion welds are spaced intermittently along the box sections. Neutron poison plates and aluminum plates are sandwiched between the sections of the stainless steel walls of the adjacent boxes and the adjacent stainless steel plates. The Type 304 stainless steel members are the primary structural components. The neutron poison plates provide criticality control and the aluminum plates provide a heat conduction path from the fuel assemblies to the inner shell.

Stainless steel rails are oriented parallel to the longitudinal axis of the cask and are attached to the periphery of the basket to establish and maintain basket orientation and to support the basket.

The basket structure is open at each end and therefore, longitudinal fuel assembly loads are applied directly to the cask body and not to the fuel basket structure. The fuel assemblies are laterally supported by the stainless steel fuel compartments, and the basket is laterally supported by the cask inner shell.

The deformations and stresses induced in the basket structure due to the applied vertical, lateral, and thermal loads are determined using the ANSYS computer program (Reference 1). The individual loads applicable for the TN-40HT basket are summarized in Table A4B.1-1. These include normal condition handling loads, thermal loads and 50g accident condition bottom end drop load. The inertial loads of the fuel assemblies are applied as equivalent pressures on the stainless steel box interior surfaces. Quasi-static stress analyses are performed with applied loads in equilibrium with the reactions at the periphery of the basket. The calculated stresses in the basket structure are compared in Section A4.2.3.4.4 to the allowable stress limits to demonstrate that the established design criteria are met.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-2 A4B.1.1 TN-40HT FUEL BASKET GEOMETRY The details of the TN-40HT basket are shown on TN Drawings in Section A1 .5. A representative cross section of support bars between fuel compartments is shown in Figure A4B.1-1. The support bars are fusion welded to the fuel compartments. This method of construction forms a very strong honeycomb-like structure. The nominal open dimension of each fuel compartment box is 8.05 in. x 8.05 in. which provides a minimum of 1/8 in. clearance around the fuel assemblies. The pitch of the fuel compartments is approximately 8.86 in. The overall basket length (160 in.) is less than the cask cavity length to allow for loading the fuel assemblies and thermal expansion.

Structural transition rails oriented parallel to the longitudinal axis of the cask are attached to the periphery of the basket to establish and maintain basket orientation, to prevent twisting of the basket assembly, and to support the basket due to lateral initial loads.

A4B.1 .2 FUEL BASKET ANALYSIS OVERVIEW The fuel basket is evaluated for normal and accident condition loads specified in Table A4B.1-1. The basket stress analysis is performed using a finite element method for the normal condition handling and thermal load cases. The end drop load case is calculated by hand.

A4B.1.3 WEIGHT The total weight of the TN-40HT basket is approximately 28,500 lbs., and the total weight of 40 fuel assemblies plus inserts is less than 52,000 lb. A uniform fuel weight of 1,300 lb is assumed to be distributed over the active fuel length of 144 inches. This methodology is conservative considering a maximum combined fuel and insert weight of 1,330 would be distributed over the entire fuel assembly length. The inertia of the basket structure (weight of the basket x g-load) is also included in the analysis.

A4B.1 .4 TEMPERATURE Thermal analyses are performed to obtain the temperature distributions in the basket for various conditions. These analyses are presented in Section A3.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-3 A4B.1 .5 TN-40HT FUEL BASKET STRESS ANALYSIS A4B.1 .5.1 APPROACH Bounding inertial loads of 3g vertical plus 3g lateral for the normal conditions and 50g vertical bottom end drop for accident conditions are applied to the basket. These inertial loads bound all applicable basket loads described in Section A3. 0°, 30°, 45°,

60° and 90° azimuth orientations are analyzed in order to bound all possible lateral loads (Figure A4B. 1-2). Nonlinear (gap element) elastic analyses of the basket structure were performed using ANSYS (Reference 1).

A4B.1 .5.2 BASKET ANALYSIS FOR VERTICAL AND LATERAL INERTIAL LOADS A4B.1.5.2.1 FINITE ELEMENT MODEL DESCRIPTION A three-dimensional finite element model of the basket is constructed using shell elements. The overall finite element model of the fuel basket is shown in Figure A4B.1-3. The fuel compartments and transition rails are included in the model.

For conservatism, the strengths of aluminum plates and poison plates in the basket are neglected by excluding them from the finite element model. However, their weights are accounted for by increasing the structural steel plate material densities to 0.39 lbs/in3.

Because of the large number of plates in the basket and large size of the basket, certain modeling approximations were necessary. In view of continuous support of fuel compartment tubes by the transition rails along the entire basket length during s torage condition lateral loads, only a 15.0 inch long axial section of the basket and transition rail is modeled. At the two cut faces of the model, symmetry boundary conditions are applied.

The fusion welds, connecting the fuel compartments and plates, are modeled with pipe elements connected at each end to adjacent fuel compartment boxes. All other interfaces (i.e., between fuel compartments, between fuel compartments and support plates, between fuel compartments and transition rails, and between transition rails and the cask) are modeled by gap elements. For all interfaces through aluminum and poison plates, the plates are assumed to be in contact to simulate support provided by the aluminum and poison plates. For the transition rails and cask interface the gap is varied in the circumferential direction such that it is zero at the point of contact, which depends on the orientation analyzed, and maximum 180 degrees from the point of contact.

The boundary conditions and interfaces for a typical fuel compartment are shown in Figure A4B.1-21 and Figure A4B.1-22.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-4 A4B.1 .5.2.2 MATERIAL PROPERTIES AND ALLOWABLE STRESSES The stainless steel fuel compartment and transition rails are constructed from SA-240, Type 304 stainless steel. Table A4B.1-2 lists the material properties used in all analyses of the TN-40HT basket. Table A4B.1-3 and Table A4B.1-4 summarize the allowable stress for normal and accident conditions, respectively. Note that the transition rail allowable stresses are based on the allowable stress for surface PT weld (0.65 x S).

A4B.1 .5.2.3 VERTICAL AND LATERAL INERTIAL LOADS The basket structure is analyzed for 0°, 30°, 45°, 60° and 90° azimuth lateral loads.

Due to the basket structure symmetry, these orientations are assumed to envelop all other possible loading orientations.

A uniform fuel weight distribution is assumed over 144 inches, which is the active fuel length. A 15.0 inch section of the basket assembly is modeled. The weight of the aluminum plates and poison plates are accounted for by increasing the density of the steel plates. The aluminum plate stiffness and poison plate stiffness are conservatively neglected in the analysis.

The basket temperature is taken as 650 °F, uniform. The rail temperature is taken as 500 °F, uniform. These temperatures are conservatively taken from the normal condition (100 °F) thermal analysis presented in Section A3.

The load resulting from the fuel assembly weight is applied as pressure on the fuel compartment plates. For the 0° orientation, the pressure acts only on the horizontal plates, and for the 90° orientation, the pressure acts only on the vertical plates. For the 30°, 45° and 60° orientation, the pressure was divided into components that act on both horizontal and vertical plates of the fuel compartments. The pressures for all orientations are calculated below for vertical and horizontal plates due to 1g and 3g lateral acceleration.

0° and 90° Drop Orientations Pressure for 1g p = Fuel assembly weight / (Panel span x Panel length)

= 1300 lb./ (8.2375" x 144") = 1.096 psi Pressure for 3g = 3 x 1.096 = 3.288 psi

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-5 30° Orientation Ph on vertical plates for 1g = p sin 30° = 1.096 x 0.5 = 0.548 psi Pv on horizontal plates for 1 g = p cos 30° = 1.096 x 0.86603 = 0.949 psi Ph on vertical plates for 3g = 3 x 0.548 = 1.644 psi Pv on horizontal plates for 3g = 3 x 0.949 = 2.847 psi 45° Orientation o

Ph on vertical plates for 1g = p sin 45 = 1.096 x 0.7071 = 0.775 psi o

Pv on horizontal plates for 1g = p cos 45 = 1.096 x 0.7071 = 0.775 psi Ph on vertical plates for 3g = 3 x 0.775 = 2.325 psi Pv on horizontal plates for 3g = 3 x 0.775 = 2.325 psi 60° Orientation o

Ph on vertical plates for 1 g = p sin 60 = 1.096 x 0.86603 = 0.949 psi o

Pv on horizontal plates for 1 g = p cos 60 = 1.096 x 0.5 = 0.548 psi Ph on vertical plates for 3g = 3 x 0.949 = 2.847 psi Pv on horizontal plates for 3g = 3 x 0.548 = 1.644 psi An increased axial acceleration is applied to the 15 inch section to simulate the compressive load due to the weight of the complete basket. The acceleration is (including the 3gs),

3.0 x (160/15) = 32 g Note: This simplified approach yields the correct vertical reaction force, but yields a compressive stress in the plates varying from zero at the top-most elements to the full 0.39 ksi at the bottom-most elements. (See the end drop analysis in Section A4B.1 .5.5.

Multiplying the 1g stress of 0.13216 ksi by 3 gives a 3g stress of 0.39 ksi). In reality, at the lower extremities of the basket, the entire 15 modeled portion of the basket would have 0.39 ksi direct compressive stress.

The load distributions for the 0°, 30°, 45°, 60° and 90° analyses for the normal condition loads are shown in Figure A4B.1-8 through Figure A4B.1-12, respectively.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-6 The accelerations applied in each run are as follows.

a z (g)

Orientation Inertial Load (g) ax (g) ay (g)

(simulate 3g axial load) 0° 3g vert. & 3g lat. 0 3 32 30° 3g vert. & 3g lat. -1.5 2.598 32 45° 3g vert. & 3g lat. -2.121 2.121 32 60° 3g vert. & 3g lat. -2.598 1.5 32 90° 3g vert. & 3g lat. -3 0 32 A4B.1 .5.2.4 ANSYS 3G VERTICAL & 3G LATERAL ANALYSES AND RESULTS Nonlinear (gap element) elastic analyses of the basket structure were performed using ANSYS for the 0°, 30°, 45°, 60° and 90° vertical and lateral load orientations.

The nodal stress intensity distribution in the stainless steel f uel compartments and transition rails are computed by ANSYS. The membrane plus bending stress intensity distributions for the 45° loading condition are shown in Figure A4B.1-13 and Figure A4B.1-14 as representative sample of the resulting stresses. The shell middle surface nodal stress intensity is the membrane stress intensity and the top or bottom surface stress intensity is the membrane plus bending stress intensity. The maximum membrane and membrane plus bending stresses for each load orientation are summarized in Table A4B.1-5.

A4B.1.5.3 BASKET ANALYSIS FOR THERMAL LOADS A4B.1.5.3.1 FINITE ELEMENT MODEL DESCRIPTION Thermal Stress Model for Basket Fuel Compartments A three-dimensional finite element model of the basket Fuel Compartments is constructed using shell elements. Due to symmetry, only 1/4 of the model (see Figure A4B.1-15) is used in this analysis.

Thermal Stresses for Transition Rails A three-dimensional finite element model of the transition rails is constructed using shell elements. Due to symmetry, only 1/4 of the model (see Figure A4B.1-16) is used in this analysis.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-7 A4B.1 .5.3.2 THERMAL LOADS The transition rails are attached to the basket with bolts in slotted holes, thus permitting free thermal growth of basket boxes. However, some thermal stresses in basket and rails can develop due to radial gradients (hot at center and cooler at periphery) for normal thermal conditions. Basket and Rail thermal stresses are calculated for the 100

°F (hot normal) and -40 °F (cold normal) ambient.

Elastic material properties described in Section A4B.1 .5.2 are used, and conservative temperature gradients are applied (see paragraphs below).

Nodal Temperature for Basket Fuel Compartments Analyses documented in Section A3 were used to obtain temperatures along a radial line from the basket center to the basket perimeter that give the largest radial thermal gradient. These temperatures were used to develop bounding polynomial curve-fit equations that give temperatures as a function of radial location. The temperature gradients were conservatively modified by increasing the temperature in the middle of the basket by 50 º F and decreasing the temperature at the perimeter of the basket by 50 º F for normal and off-normal thermal storage conditions. Temperatures between the middle of the basket and the perimeter of the basket were modified proportionally to maintain the same basic shape of the temperature distribution. For a given bounding curve, the conservative temperature gradient is mapped onto the basket models in all radial directions to give the largest gradient across the entire diameter of the basket.

The temperature distribution in the fuel compartments for 100 °F and -40 °F are shown in Figure A4B.1-4 and Figure A4B.1-5, respectively.

Nodal Temperature for Transition Rails The same conservative temperature gradient for fuel compartments is mapped onto the rail models in all radial directions to give the largest gradient across the entire diameter of the basket. The temperature distribution in the transition rails for 100 °F and -40 °F ambient conditions are shown in Figure A4B.1-6 and Figure A4B.1-7, respectively.

A4B.1 .5.3.3 ANSYS THERMAL ANALYSIS AND RESULTS Basket Compartment Nodal stress intensity distributions in basket fuel compartments and the support plates are plotted at top or bottom surfaces in Figure A4B.1-17 and Figure A4B.1-18 for 100 °F and -40 °F ambient conditions, respectively.

The maximum stress intensities in the fuel compartments are 8.61 ksi and 8.74 ksi for 100 °F and -40 °F ambient conditions, respectively, and are shown in Table A4B.1-6.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-8 Transition Rail Nodal stress intensity distributions in transition rails are plotted at top or bottom surfaces in Figure A4B.1-19 and Figure A4B.1-20 for 100 °F and -40 °F ambient conditions, respectively.

The maximum stress intensities in the transition rails are 21.73 ksi and 21.60 ksi for 100

°F and -40 °F ambient conditions, respectively, and are shown in Table A4B.1-6.

A4B.1.5.4 SHEAR LOAD IN THE FUSION WELD The results of the static load analyses were post-processed for Axial (FX) and Shear (FY and FZ) forces in the pipe elements representing the Fusion Welds. The maximum forces are listed as follow:

o Maximum Force, FX = 314.92 lb. (at Element No. 32728, 60 lateral load orientation) o Maximum Force, FY = 950.88 lb. (at Element No. 32723, 90 lateral load orientation) o Maximum Force, FZ = 99.27 lb. (at Element No. 32553, 60 lateral load orientation)

The results of the thermal load analyses were also post-processed for Axial (FX) and Shear (FY and FZ) forces in the pipe elements representing the Fusion Welds. The maximum forces are listed as follow:

Maximum Force, FX = 310.69 lb. (at Element No. 32559, 100 °F ambient)

Maximum Force, FY = 608.83 lb. (at Element No. 32548, -40 °F ambient)

Maximum Force, FZ = 99.01 lb. (at Element No. 32717, -40 °F ambient)

The maximum combined shear load in a fusion weld is computed by vectorially adding the maximum FX, FY, and FZ (irrespective of their location).

Maximum Combined Shear Force = [(314.92 + 310 . 69)2 + (950.88 + 608 . 83)2 + (99.27 +

2 1/2 99.01 ) ] = 1,683 lb = 1.68 kips (per fusion weld)

For the fusion weld load capacity test at room temperature, a factor of safety of 1.43 is applied and the material strength is corrected for room temperature testing. Therefore the, the required minimum fusion weld test load per weld is:

= 1,683 * (Factor of Safety) * (S u at room temperature / S u at 650ºF)

=1,683 lb

• 1.43 * (75 ksi/63.4 ksi) = 2,847 lb.

2.9 kips

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-9 A4B.1.5.5 FUEL BASKET VERTICAL END DROP ANALYSIS During vertical end drop load case, the fuel assemblies and fuel compartments are forced against the bottom of the TN-40HT cask. It is important to note that, for any vertical or near vertical loading, the fuel assemblies react directly against the bottom of the cask and not through the basket structure as in lateral loading. It is the weight of the basket that causes axial compressive stress during an end drop. Axial compressive stresses are conservatively computed by assuming that all of the basket weight is reacted by the compartment tubes during an end drop. A basket weight of 28.5 kips is used in the end drop stress calculations.

Stainless Steel Basket Components Total weight of basket assembly = 28.5 kips Compressive stress in fuel compartments:

Section area of compartment tubes = 40 x [(w + 2t) 2 - w2] = 40 x [8 . 4252 - 8.05 2]

= 247.125 in 2 Area of drain slots = 40 x 4 slots/compartment x [1 x .1875] = 30 in 2 Stress due to 1g = P / A = -28.5 / [247.125 - 30] = -0.13126 ksi Stress due to 50g end drop is 50 x 0.13126 = 6.56 ksi This stress is much lower than the accident membrane allowable of 38.88 ksi (2.4 S m at o

650 F)

A4B.1 .5.6 EVALUATION OF BASKET ALUMINUM COMPONENTS FOR LONG TERM STORAGE DEADWEIGHT The long term storage load (deadweight) compressive stresses in the limiting aluminum components were compared to allowable stress values that have been reduced to limit the effects due to material creep.

In the TN-40HT basket, the steel support bars are welded between the fuel compartments and form a cavity for the aluminum plates. Thus, there is no additional compressive stress in the aluminum plates, except the dead weight. The maximum height of the basket aluminum plates is 13.14. Since the dead weight compressive stresses in these plates is so small, these are not the limiting aluminum components.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-10 The aluminum rail inserts are bolted at intervals to the R45 and R90 rails. The maximum compressive stress in the aluminum inserts is conservatively calculated as the entire length of the basket supported by itself at the bottom without taking credit for the bolts. The results are compared to allowable stress values that are reduced to limit the effect due to creep.

The compressive stress in the inserts is conservatively calculated as the entire length of the basket insert.

1 g vertical bearing stress = Load / Area =

VolumeXDensity = Length XDensity= 160 x 0.098 = 15.68 psi Area Where, Length = 160 Density = 0.098 lb/in3, PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-11 Based on the above evaluation, the basket aluminum components for the TN40HT Basket design are structurally adequate for long-term storage loads and creep strain is acceptable.

A4B.1.5.7 BASKET STRESS ANALYSIS CONCLUSIONS Basket Compartment As shown in Table A4B.1-5 and A4B.1-6, the maximum stress intensity (P m + P b ) in the o

fuel compartments and support plates (bars) is 9.66 ksi (at 60 load orientation), this o

stress is less than the membrane allowable (P m ) of 16.2 ksi (S m at 650 F).

The maximum primary plus secondary stress intensity (P m + P b + Q) is 18.40 ksi (maximum primary stress 9.66 ksi + 8.74 ksi from the above thermal analysis shown on Table A4B.1-6). This stress intensity is lower than the allowable stress of 48.6 ksi (3 S m o

at 650 F).

Transition Rail As shown in Table A4B.1-5 and A4B.1-6, the maximum stress intensity (P m + P b) in the o

transition rail is 4.11 ksi (at 90 load orientation), this stress is less than the membrane o

allowable (P m ) of 11.38 ksi (S m at 500 F, 17.5 ksi x 0.65 = 11.38 ksi). This allowable is based on rail welds with surface PT requirements (Reference 2).

The maximum primary plus secondary stress intensity (P m + P b + Q) is 25.84 ksi (maximum primary stress 4.11 ksi + 21.73 ksi from the above thermal analysis as shown on Table A4B.1-6). This stress intensity is lower than the allowable stress of o

34.13 ksi (3 S m at 500 F, 3 x 17.5 ksi x 0.65 = 34.13 ksi). This allowable is based on rail welds with surface PT requirements (Reference 2).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.1-12 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.2-1 A4B.2 REFERENCES ANSYS Engineering Analysis System Computer Code and User's Manual,Section II - Part D, release. 8.0.

2. ASME Boiler and Pressure Vessel Code,Section III, Subsection NG &

Appendices, 2004 through 2006 addenda.

3. TN Technical Report No. E-25768, Rev. 0, Evaluation of Creep of NUHOMS Basket Aluminum Components Under Long Term Storage Conditions, November, 2007.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A4B.2-2 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4B.1-1

SUMMARY

OF INDIVIDUAL LOADS FOR STORAGE CONDITIONS-BASKET Individual Load Loads IL-1 50g Vertical (Bottom End Drop)

IL-2 3g Vertical (Lifting)

IL-3 3g Vertical + 3g Lateral (bounds all normal and off normal

_________________ loads)

IL-4 Thermal Stress due to Hot Environment (100°F ambient)

IL-5 Thermal Stress due to Cold Environment (-40°F ambient)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4B.1-2 MATERIAL PROPERTIES FOR TN-40HT FUEL BASKET 6

Temp. Sm E

am (10 Part Material (ksi) in/in/°F)

_______ _________ (F) (ksi) (ksi) (psix106) 70 30.0 75.0 20.0 28.3 8.5 300 22.4 66.2 20.0 27.0 9.2 Steel Boxes SA-240, 400 20.7 64.0 18.6 26.4 9.5 and Type 304 500 19.4 63.4 17.5 25.9 9.7 Rails 600 18.4 63.4 16.6 25.3 9.8 700 17.6 63.4 15.8 24.8 10.0 Notes:

• Material Properties are obtained from ASME code Section II, Part D (Reference 2).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4B.1-3 BASKET STRUCTURAL ALLOWABLE STRESSES, NORMAL CONDITIONS Basket Compartment PmPm +PP +P +Q Temperature b m b Material (S m ) (1.5 S m ) (3 S m )

( F)

______________ ______________ (ksi) (ksi) (ksi)

SA-240, 650 16.2 24.3 48.6 Type 304 Transition Rail PmPm +P P +P +Q Temperature b m b Material (S m x 0.65) (1.5 S m x 0.65) (3 S m x 0.65)

______________ ____________ (ksi) (ksi) (ksi)

SA-240, 11.38 17.06 34.13 500 Type 304 (17.5 x 0.65) (1.5 x 17.5 x 0.65) (3 x 17.5 x 0.65)

Note: Allowable for transition rail is based on welds with surface PT examination

[Reference 2 ASME Section III, Subsection NG, Table NG-3352-1].

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4B.1-4 BASKET STRUCTURAL ALLOWABLE STRESSES, ACCIDENT CONDITIONS Pm Pm + Pb Temperature (lesser of 2.4 S or (lesser of 3.6 Material m Sm (F) 0.7 S )u or S )

u

_____________ _____________ (ksi) (ksi)

SA-240, 500 42.0 63.0 Type 304 650 38.88 58.32

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4B.1-5

SUMMARY

OF BASKET STRESS ANALYSIS RESULTS Lateral Load Stress Maximum Component Orientation Category Stress (ksi)

Fuel Pm 3.98 Compartments (1)

Pm ^ Pb 0

o 5.89 Pm Rails____________ _____________ 1.35 Pm ^ Pb 3.11 Fuel Pm 4.89 Compartments 1)

Pm ^ Pb 30 o 8.70 Pm Rails____________ _____________ 1.71 Pm ^ Pb 3.49 Pm Fuel Compartments 3.92 Pm ^ Pb 45 o

8.90 Pm Rails____________ _____________ 1.68 Pm ^ Pb 3.51 Fuel Pm 3.88 Compartments 1)

Pm ^ Pb 60 o

9.66 Pm Rails____________ _____________ 1.14 Pm ^ Pb 3.50 Fuel Pm 2.63 Compartments (1) o 90 Pm ^ Pb 6.46 Pm Rails____________ _____________ 1.55 Pm ^ Pb 4.11 Note:

1. The component includes fuel compartments and support plates (bars).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A4B.1-6

SUMMARY

OF STRESSES IN FUEL COMPARTMENTS AND RAILS Loading Component Load Stress Loads Stress (3) Allow.

Condition Classification (ksi) Stress

_________ ______________ ___________ ________________ ___________ _______ (ksi) 3g Lifting Fuel P 3g Axial m 0.39 16.2 2

Compartment IL-2 P m ^ P 0.39 b 24.3 Rail P 0.39 11.38 m

(1)

________ ____________ __________ P + P _________ 0 39(1)17.06 m b Fuel P 3g Axial ^

m 4.89 16.2 Compartment IL3 3g Lateral (2) 3g P ^

m Pb 9.66 24.3 Vertical, Rail P 1.71 m 11.38 3g P + P 4.11 m b 17.06 Lateral 100 F o

Fuel Q 8.61 48.6 (2)

Compartment IL-4 Amb.

Rail Q 21.73 34.13 Thermal Fuel

-40 F o

Q 8.74 48.6 (2)

Compartment IL-5 Amb.

Rail Q 21.60 34.13 (1) Assumed same stresses as for fuel compartment. Thinner walls (3/16) of the fuel compartment carry the weight of the entire basket (fuel compartments, rails, poison plates, aluminum plates, and support bars) while the thicker walls (3/8) of the rails carry only the rail self weigh and the thin aluminum inserts.

(2) The component includes fuel compartments and support bars. A quality factor of 1.0 is applied based on single pass weld and 100% VT inspection on both sides.

(3) The smallest margin between the calculated stress and the allowable is 9.67 ksi (11.38- 171).

This margin is sufficient to address the additional 0.39 ksi compressive stress in the lower extremities of the basket discussed in Section A4B.1.5.2.3.

PROPRIETARY INFORMATION Withhold Under 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 U ii6 5

EL EMNT S

w r'rph ic5 FfET= 1 zv -1 tJTSTfl .17 ZF -i.5

- PTJF!ER 90 Deg 60 Deg I

TNIUH asket Model tte.ss rIa1\ .T sis - 0 L'e.qre.e. - 45 Deg 30 Deg 0 Deg FIGURE A4B.1-2 TN-40HT BASKET LOADING ORIENTATION DEFINITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A4B.1-3 BASKET FINITE ELEMENT MODE L

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 INSYS 8.0 DKC 21 2006 09:48:18 PIIIY]7 NO. 1 NEDAL SOLOTIG STEP=1 SUE =1 TIE=1 BF1EMP (AVG)

MIDDLE DMX =. 191931 SN =313.527 SMX =675.628 XV = .4 YV = .38 ZV = .85 DIST=22 .733 XE =18.499 YF =16.209 ZF =-7.082 Z-EUFFER

- 313.527

___ 353.76

___ 434.227

___ 554.927

___ 595.161

- 635.394 675.628 FIGURE A4B.1-4 FUEL COMPARTMENT FINITE ELEMENT MODEL TEMPERATURE DISTRIBUTION 100 °F AMBIENT

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 NSYS 8.0 DEC 21 2006 09:48:21 PLIY]7 NO. 5 NWL SOLLJTIN SThP=2 S[JB =1 TIM=2 BF1EMP (AVG)

MIDDLE DMX =.151523 SFN =203.292 SMX =572.914 xv = . 4 Yv = .38 zv = . 85 DIST=22 .738 XE =18.492 YF =16.21 ZF =-7.068 Z-EUFFER

- 203.292

___ 244.361

___ 285.43

___ 326.499

___ 367.569

___ 408.638

___ 449.707

___ 490.776 531.845 572. 914 FIGURE A4B.1-5 FUEL COMPARTMENT FINITE ELEMENT MODEL TEMPERATURE DISTRIBUTION

-40 °F AMBIENT

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 NSYS 8.0 DEC 21 2006 09:47 :01 PLIY]7 NO. 1 NWL SOLLJTIN sThp=1 S[JB =1 TIM=1 BF1EMP (AVG)

MIDDLE DMX =. 069734 SFN =313.615 SMX =476.851 xv = . 4 Yv = .38 zv = . 85 DIST=22 . 959 XE =18.218 YF =20.371 ZF =-6.753 Z-EUFFER

- 313.615

___ 331.752

___ 349.889

___ 368.027

___ 386.164

___ 404.302

___ 422.439

___ 440.576 458.714 476.851 FIGURE A4B.1- 6 TRANSITION RAILS FINITE ELEMENT MODEL TEMPERATURE DISTRIBUTION 100 °F AMBIENT

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

/NSYS 8.0 DEC 21 2006 09:47:02 PIXY]7 NO. 5 NcDAL SOLUTIN S1P=2 SIIJB =1 TIM=2 EFEThMP (vG)

MIDDLE EJMK =.053021 SM =203.383 SMK =371.479 xv = . 4 Yv = .38 zv = . 85 EJIST=22.988 KE =18.321 YE =20.467 ZE =-6.789 Z-BUFFFR

- 203.383

___ 222.06

___ 240.737

___ 259.415

___ 278.092

___ 296.77

___ 315.447

___ 334.125 352.802 371. 479 FIGURE A4B.1-7 TRANSITION RAILS FINITE ELEMENT MODEL TEMPERATURE DISTRIBUTION

-40 °F AMBIENT

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

/NSYS S .

PILY]I NO. 2 ELEIENIDS

[YPE NfJM

=1 IDIST=.39.

=-7 .5

-E:UFFFR PRES-NIX-I

_ 1:

LIII FIGURE A4B.1-8 BASKET FINITE ELEMENT MODEL APPLIED PRESSURES - 0° AZIMUTH

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 J'I-sYs 8.

DEC 26 1.3 : 4.3 PIfT J.O. 2 IaEMEJTS

[YPF NUM 2* V =1 L'IST=.39. 6 ZF =-7 .5 Z-BUFFER

*i;1 FIGURE A4B.1-9 BASKET FINITE ELEMENT MODEL APPLIED PRESSURES - 30° AZIMUTH

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 S 8.

13:46 PL:IY]7 NO. 2 TYPE N1IIN 1

DIST=39 . 6

=-7 .5 Z-EUFF'FR Ps-Ncdi FIGURE A4B.1-10 BASKET FINITE ELEMENT MODEL APPLIED PRESSURES - 45° AZIMUTH

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 ANSYS 8 .

E: LT r*c'.

FIt-If*17S TYPE NUM 2V =1

[:'IST=.39 . 6 5 F =-7 .5 5 -EUFFER ppjJflp.

FIGURE A4B.1-11 BASKET FINITE ELEMENT MODEL APPLIED PRESSURES - 60° AZIMUTH

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 1TEYS 8.

1I:'KC .6 1.3:55 PLf NC. 2 ELE-H.

TYPE NU-I 7 1 DIST=39.

ZF =-7.5 PPEs-Nc1 FIGURE A4B.1-12 BASKET FINITE ELEMENT MODEL APPLIED PRESSURES - 90° AZIMUTH

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 NSYS 8.0 DEC 22 2006 03:55:16 PLIY]7 NO. 2 NWL SOLLJTIN SThP=2 S[JB =4 TIM=2 S]INT (AVG) pop DMX =. 139897 SFN =5.494 SMX =8751 SM=12386 xv = . 4 Yv = .38 zv = . 85 DIST=48 . 629 Z-EUFFER

- 5.494

___ 977.174

___ 1949

___ 2921

___ 3892

___ 4864

___ 5836

___ 6807

- 7779 8751 TN4OHT Basket Elastic McKJl Handling Lodds - 45 Dgre - Contact FIGURE A4B.1-13 NORMAL CONDITION 45° AZIMUTH LATERAL LOAD - FUEL COMPARTMENTS -

TOP SURFACE MEMBRANE PLUS BENDING STRESS INTENSITY

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 PNSYS 8.0 DEC 22 2006 03:55:20 PLIY]7 NO. 5 NWPL SOLLJTIN STEP=2 sUE =4 TIE=2 S]INT (AVG)

PDP DMX =.139897 SM =15.079 SMX =3049 SMKB=488 9 xv = .4 Yv = .38 zv = . 85 DIST=48 . 629 Z-EUFF'ER

- 15.079

___ 352.177

___ 689.274

___ 1026

___ 2038

___ 2375

- 2712 3049 TN4OHT basket Kiastic Mcdel Handling Loads - 45 Degree - Contact FIGURE A4B.1-14 NORMAL CONDITION 45° AZIMUTH LATERAL LOAD - PERIPHERAL RAILS TOP SURFACE MEMBRANE PLUS BENDING STRESS INTENSITY

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 8.

lB PI.TY]I NO.

FLEME1.TrEi

[YPE NfJM

= .4 IV=.

= OL LI:'IST=21. 69 17.694 YE =15. 47.B ZE =-7 .

Z-E;UFFFR FIGURE A4B.1-15 THERMAL STRESS ANALYSIS FINITE ELEMENT MODEL - FUEL COMPARTMENT

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 P1JSYS B ii:'FiC II J PL:IY]I D.

LEENTS IYPE NUM xv =.4 IV=.

=

DIST=21. 579 KE =17.869 YE =2c'.249 ZE =-7 .5 5BUFFER FIGURE A4B.1-16 THERMAL STRESS ANALYSIS FINITE ELEMENT MODEL - RAILS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 PNSYS S

[iE _1 _i H ii PL:ff r:I:. 3 1N1;[JF\L SOLUTIcN STEFi STIlE: =1 TIME=1 lOP D4::: 1I SN =32.757 S::.:Ei4Ez57 xv =.4 zv =.85 DIST=22.

-]Q

.499 YE =16.

Z-E;UFFER

- II FIGURE A4B.1-17 THERMAL STRESS ANALYSIS - MAXIMUM FUEL COMPARTMENT STRESS INTENSITY FOR 1000 F AMBIENT TOP SURFACE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 fJSYS 8.0 DEC 21 2006 09:48:23 PIXYTNO. 8 NGDAL SOLUTIcN STEP=2 SUB =1 TI1=2 S]INI (vG)

BTM DMX =. 151523 SN =9.559 SMX =8735 SMXR=1457 0 XV =.4 YV =. 38 zv =. 85 DIST=22 .738 KE =18.492 YF =16.21 ZF =-7.068 Z-EUFFFR

- 9.559

___ 978.998 1948 5826 8735 FIGURE A4B.1-18 THERMAL STRESS ANALYSIS - MAXIMUM FUEL COMPARTMENT STRESS INTENSITY FOR -40° F AMBIENT BOTTOM SURFACE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 zrJsYs 8.0 DEC 21 2006 09:47:02 PID]7 No. 4 NDTL SOLUTIEN srrEp=1 S[IB =1 TIME=1 S]1N17 (TVG)

BCffT EJMX =.069734 SMNT =70.577 SMX =21726 SF=32O81 XV = .4

= .38 ZV = .85 DIST=22 . 959 XE =18.218 YE =20.371 ZE =-6.753 Z-EFJFFFR

- 70.577 19320 21726 FIGURE A4B.1-19 THERMAL STRESS ANALYSIS - MAXIMUM RAILS STRESS INTENSITY FOR 100 °F AMBIENT BOTTOM SURFACE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TNSYS 8.0 DEC 21 2006 09:47:03 PLJ]7 NO. 8 NWN SOLUTION STEP=2 SOB =1 TIM=2 S]iN]7 (TVG)

BJTM DMX =.053021 SN =61.768 SMX =21599 SMKB=32068 xv = .4 Yv = .38 zv = .85 DIST=22.988 KE =18.321 YF =20.467 ZF =-6.789 Z-EUFFFR

- 61.768 9634 21599 FIGURE A4B.1-20 THERMAL STRESS ANALYSIS - MAXIMUM RAILS STRESS INTENSITY FOR -40 °F AMBIENT BOTTOM SURFACE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Fusion Welds - Modeled with Pipe Elements 15 in Sector / /

(Symmetry Boundary Condhons /

/

/

/

All other nterlaces modeled with gap elemenis - gapO Vertical displacement constraints at this face are removed when vertical load is appUed FIGURE A4B.1-21 INTERFACES FOR TYPICAL FUEL COMPARTMENT TUBE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 II.

Contact between the transition rails and cask (enlarged for clarity)

FIGURE A4B.1-22 BOUNDARY CONDITIONS FOR THE BASKET ANALYSIS (SYMMETRY BOUNDARY CONDITIONS ARE NOT SHOWN FOR CLARITY)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A5.1-1 SECTION A5 STORAGE SYSTEM OPERATIONS A5.1 OPERATION DESCRIPTION The TN-40HT cask was designed to use the same equipment as the TN-40 cask. As a consequence of this design objective, the operation of the TN-40 cask, as described in Section 5, is applicable to the operation of the TN-40HT cask. Therefore; rather than duplicating discussions and information already presented, the discussion of the TN-40HT storage system operation will be accomplished by referring to the appropriate section in Section 5 noting any necessary exceptions.

A5.1.1 NARRATIVE DESCRIPTION The narrative of the storage system operation in Section 5.1.1 is applicable to the operation of the TN-40HT casks.

A5.1.2 FLOW SHEETS The information outlined in Section 5.1.2 is applicable to the TN-40HT casks except for the location of the radiation exposure determination for the TN-40HT cask which is located in Section A7.

Note that the helium leak test of the lid seals includes all seals, e.g., the main lid seal, vent cover seal, and the drain cover seal.

A 5.1.3 IDENTIFICATION OF SUBJECTS FOR SAFETY AND RELIABILITY ANALYSIS A5.1 .3.1 CRITICALITY PREVENTION The information in Section 5.1.3.1 is applicable to the TN-40HT casks except for the location of the criticality discussion which is located in Section A3.3.4.

A5.1 .3.2 INSTRUMENTATION The information in Section 5.1.3.2 is applicable to the TN-40HT casks except for the location of the description of the transmitters which is located in Section A3.3.3.

A5.1.3.3 MAINTENANCE TECHNIQUES The information in Section 5.1.3.3 is applicable to the TN-40HT casks.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A5.2-1 A5.2 CONTROL ROOM AND CONTROL AREAS The information in Section 5.2 is applicable to the TN-40HT casks.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A5.3-1 A5.3 SPENT FUEL ACCOUNTABILITY PROGRAM The information in Section 5.3 is applicable to the TN-40HT casks.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A5.4-1 A5.4 SPENT FUEL TRANSPORT TO ISFSI The information in Section 5.4 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A5.5-1 A5.5 SPENT FUEL TRANSFER TO TRANSPORT CASK The information in Section 5.5 is applicable to the TN-40HT casks.

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5.6 REFERENCES

None.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A6.1- 1 SECTION A6 WASTE MANAGEMENT A6.1 DESIGN The information outlined in Section 6.1 is applicable to the TN-40HT casks except for the location of the decommissioning plan for the TN-40HT casks which is located in Section A4.6.

A

6.2 REFERENCES

The references listed in Section 6.2 are also applicable to the TN-40HT casks and there are no additional references associated with Section A6.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.1-1 SECTION A7 RADIATION PROTECTION A7.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA)

A7.1 .1 POLICY CONSIDERATIONS AND ORGANIZATION The information in Section 7.1.1 is independent of cask design.

A7.1.2 DESIGN CONSIDERATIONS The information in Section 7.1.2 is also applicable to the TN-40HT casks.

A7.1.3 OPERATIONAL CONSIDERATIONS The information outlined in Section 7.1.3 is applicable to the TN-40HT casks except for the location of the description of the radiation protection design features is located in Section A7.3.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-1 A7.2 RADIATION SOURCES A7.2.1 CHARACTERIZATION OF SOURCES There are five principal sources of radiation associated with cask storage that are of concern for radiation protection:

- Primary gamma radiation from spent fuel;

- Primary neutron radiation from spent fuel (both alpha-n reactions and spontaneous fission);

- Gamma radiation from activated fuel structural materials and fuel inserts;

- Capture gamma radiation produced by attenuation of neutrons by shielding material of the cask; and

- Neutrons produced by sub-critical fission in fuel.

The first three sources of radiation are evaluated using SAS2H/ORIGEN-S modules of the SCALE (Reference 1) code with the 44 group ENDF/B-V library. The capture gamma radiation and sub-critical multiplication are handled as part of the shielding analysis which is performed with MCNP5 (Reference 2).

The fuel assemblies acceptable for storage in the TN-40HT are listed in Table A3.1-1.

The largest uranium loading results in the largest source term at the design basis enrichment and burnup. Table A7.2-1 provides additional fuel assembly design characteristics for the fuel designs.

The 14x14 Westinghouse standard is the design basis fuel for shielding purposes because it has the highest initial heavy metal loading (modeled as 0.410 Mtu), and therefore results in the highest radioactive source terms for a given irradiation history.

Initial enrichment of 3.4 wt% U235, assembly average burnup of 60 GWd/MTU, and cooling time of 18 years complete the specification of the design basis fuel. A conservative two-cycle irradiation at a constant specific power of 20 MW/assy is utilized with a 20 day down time between each cycle to calculate the source terms for the design basis fuel.

The source terms are generated for the active fuel region, the plenum region, and each end region. Irradiation of the fuel assembly structural materials (including the fuel inserts, plenum, and end fittings) are included as an irradiation in the fuel zone with appropriate flux reduction factors as described below. The fuel assembly hardware materials and masses on a per assembly basis are listed in Table A7.2-2. Table A7.2-3 provides the material composition of fuel assembly hardware materials. Cobalt impurities are included in the SAS2H model. In particular, the cobalt impurities in Inconel-718, Zircaloy and Stainless Steel 304 are 0.469%, 0.001% and 0.08%,

respectively.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-2 Table A7.2-4 contains the masses of the various hardware materials present in the four principal regions of the fuel assembly. The masses for the materials in the top end fitting, the plenum, and the bottom fitting regions are multiplied by the activation ratios 0.1, 0.2 and 0.2, respectively (Reference 5), to correct for the spatial and spectral changes of the neutron flux outside of the active fuel zone.

As an example, the effective cobalt mass in the top end region is determined by taking the mass of the material listed in Table A7.2-4 times the cobalt impurity percentage listed above, which are also shown on Table A7.2-3. The resultant product is then adjusted by the above activation ratio:

Co = 6.30 kg (stainless steel, Table A7.2-4)*0.08% (Table A7.2-3) +

0.508 kg (Inconel Table A7.2-4)* 0.469% (Table A7.2-3) = 0.0074 kg effective Co (after applying the 0.1 activation ratio) = 0.00074 kg The material compositions of the fuel assembly hardware are included in the SAS2H/ORIGEN-S model on a per assembly basis. The cobalt content for each fuel assembly region utilized in the source term calculation is obtained from Table A7.2-4 reduced by the activation ratios.

Fuel Insert Thimble Plug Device (TPD)

The TPD materials and masses for each irradiation zone are listed in Table A7.2-5. The TPD is irradiated to an equivalent host assembly life burnup of 125 GWd/MTU. The model assumes that the TPD is irradiated in an assembly with an initial enrichment of 3.85 weight % U-235. The fuel assembly containing the TPD is burned for three cycles with a burnup of 15 GWd/MTU per cycle and a down time of 30 days between cycles.

This is equivalent to an assembly life burnup of 45 GWd/MTU over the three cycles. The results are increased by a factor of 2.7778 to achieve the equivalent 125 GWD/MTU source. The source term for the TPD is taken at 16 years cooling time.

Fuel Insert Burnable Poison Rod Assembl y (BPRA)

The BPRA materials and masses for each irradiation zone are also listed in Table A7.2-5. These materials are irradiated in the appropriate zone for two cycles of operation. The model assumes that the BPRA is irradiated in an assembly with an initial enrichment of 3.85 weight % U-235. The fuel assembly containing the BPRA is burned for two cycles with a burnup of 15 GWd/MTU per cycle and a down time of 30 days between cycles. This is equivalent to an assembly life burnup of 30 GWd/MTU over the three cycles. The source term for the BPRA is taken at 18 years cooling time.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-3 Reactor Coolant S ystem Boron Concentration Soluble boron in the reactor coolant system is the primary reactivity control mechanism in a PWR. It is important to include boron in the depletion model because it is a thermal neutron absorber and its presence hardens the neutron spectrum. The presence of boron reduces thermal absorption in U-235 and as a second-order effect, increases epithermal absorption in U-238, increasing the buildup of actinides such as Cm-244, the major contributor to the neutron source term.

Typical cycle average boron concentration is on the order of 900 ppm. For modeling purposes in the current analysis, 600 ppm was chosen to be the average boron concentration for the first irradiation cycle, with the second having 95% of this value.

Studies were performed showing that the use of a lower boron concentration leads to a tiny underproduction of decay heat, neutron and gamma source strength in the energy groups that contribute the most to casks dose rates. The under predictions are within 1 % of those determined with the lower boron concentration and thus have essentially no effect on dose rates and cooling times.

The results of the fuel qualification sensitivity calculations with soluble boron concentration are shown in Table A7.2-12. Cases C1 through C8 are sensitivity calculations where the soluble boron concentration in Cases B1 through B7 on Table A7.2-1 1 was increased from 600 ppm to 1000 ppm. The results of these evaluations show that this increase in boron concentration results in an increase in the dose rate by approximately 1.5% and an increase in the decay heat by approximately 1%. The fuel qualification, however, ensures that the design basis fuel assembly, i.e. Case A7 on Table A7.2-1 1, utilized in the shielding calculations results in bounding dose rates even though the boron concentration utilized is lower than that of a typical cycle average value.

Reactor Coolant S ystem Temperature o

Moderator temperatures can vary between 500-600 F. The long-term operating temperature affects the end-of-life reactivity (as a second order effect) and the total actinide inventory. Higher average moderator T av reduces the average moderator density, which reduces neutron moderation during operation. This again results in increased epithermal absorption in U-238 which results in an increase in the actinide inventory in the fuel for a given total fuel burnup. A moderator density of 0.733 g/cc (which corresponds to 566 º F at an operating pressure of 2250 psia) is used in the SAS2H calculation.

The results of the fuel qualification sensitivity calculations with moderator temperature are shown in Table A7.2-12. Cases D1 through D8 are sensitivity calculations where the moderator temperature in Cases C1 through C7 on Table A7.2-1 2 was increased from 558 K (545º F, representative of a core average moderator temperature) to 590 K (602 º F, representative of an average hot leg moderator temperature) and the moderator density was correspondingly reduced from 0.733 g/cm 3 to 0.690 g/cm 3 . The results of

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-4 these evaluations show that this increase in moderator temperature and boron concentration results in an increase (when compared to corresponding cases B1 through B7) in the dose rate by approximately 4% and an increase in the decay heat by approximately 2%. The fuel qualification, however, ensures that the design basis fuel assembly, i.e. Case A7 on Table A7.2-1 1, utilized in the shielding calculations results in bounding dose rates. In addition, the use of a moderator density of 0.733 g/cm 3 (which corresponds to a moderator temperature of 566°F) for the design basis is justified because 566°F is representative of a core average moderator temperature.

Fuel and Claddin g Temperature Representative temperatures of 840 K and 620 K were selected for the fuel and clad, respectively.

A7.2.2 AIRBORNE RADIOACTIVE SOURCES In addition to the source terms generated for the shielding analysis, the SAS2H irradiation also provides the bounding radiological source terms for confinement. These results are provided in Table A7.2-6.

A7.2.3 AXIAL SOURCE DISTRIBUTION Reference 4 provides axial PWR burnup profiles as a function of fuel assembly burnup ranges. For the Prairie Island high burnup fuel to be stored in the TN-40HT cask, the burnup profile for > 46 GWD/MTU was selected for use in this analysis. Figure A7.2-1 represents this profile.

The conservative axial profile containing 18 axial zones is utilized in the shielding evaluation. The axial zones are approximately equal and each zone represents between 5% and 6% of the total active fuel zone. The peaking factors range from a slightly more than 0.5 at the bottom and top, to a maximum of 1.11 just below the middle. The gamma source is directly proportional to the burnup and the neutron source is proportional to the fourth power of the burnup. This data is presented in Table A7.2-7 and shown in Figure A7.2-2. For the gamma distribution, the average value is around 1.0. However, for the neutron distribution, the average value of the distribution is greater than 1.0. The average value of the axial neutron distribution may be interpreted as the ratio of the true total neutron source in an assembly to the neutron source calculated by SAS2H/ORIGEN-S for an average assembly burnup. Therefore, to properly correct the magnitude of the neutron source, the neutron source per assembly, reported in Table A7.2-8, is multiplied by the average value of the neutron source distribution as reported in Table A7.2-7.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-5 A7.2.4 GAMMA SOURCE The gamma source terms for the 14x14 design basis fuel assembly are provided in Table A7.2-8. The gamma source spectra are presented in the 18-group structure consistent with the SCALE 27n-1 8g cross section library. The conversion of the source spectra from the default ORIGEN-S energy grouping to the SCALE 27n-1 8 g energy grouping is performed directly through the ORIGEN-S code. The SAS2H/ORIGEN-S input file is provided in Appendix A7B.

The gamma source for the fuel assembly hardware is primarily from the activation of cobalt. This activation contributes primarily to SCALE Energy Groups 36 and 37.

For this analysis, it is assumed that the fuel insert source is a composite source; a BPRA in the core region and a TPD in the plenum and top fitting region. This bounds both types since the TPD does not extend into the core region but has a higher source term in the plenum and top fitting regions than a BPRA. The cumulative burnup of the fuel assembly(s) where the BPRA resided during reactor operation is assumed to be 30 GWd/MTU and the BPRA has cooled for 18 years. The cumulative burnup of the TPD host assemblies is assumed to be 125 GWD/MTU and the TPD has cooled for 16 years.

The fuel insert source is shown in Table A7.2-9.

A7.2.5 NEUTRON SOURCE Table A7.2-8 provides the total neutron source for the 14x14 design basis fuel assembly. The neutron source term consists primarily of spontaneous fission neutrons (largely from Cm-244) with ( a, O-1 8) sources of lesser importance, both causing secondary fission neutrons. The overall spectrum is well represented by the Cm-244 fission spectrum. For the MCNP analyses, the default Cm-244 energy spectrum is utilized.

A7.2.6 FUEL QUALIFICATION This section provides the basis for qualification of the design basis fuel chosen for the shielding analysis of the TN-40HT cask. The objective is to demonstrate that the fuel assemblies with the parameters corresponding to design basis fuel result in the highest calculated dose rate so that bounding shielding analysis can be performed by utilizing these design basis source terms. In order to determine the bounding spent fuel parameters (design basis fuel assembly), the candidate assembly parameters are ranked by their relative radiation source strengths. A simple 2-D shielding calculation based on a representation of the TN-40HT cask is performed and the radiation dose rates at 2m from the side surface are determined. The spent fuel parameters that yield the highest total dose rate (gamma + neutron) are considered the design basis for shielding calculations.

These analyses were carried out using the SAS2H depletion module from the SCALE (Reference 1) computer software and MCNP (Reference 2). For all SAS2H

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-6 calculations the SCALE 44 group ENDF/B-V (44groupndf5) library was used. MCNP calculations used the default cross section libraries.

A simplified MCNP model was utilized to calculate a response function at 2 meters from the side surface of a cask. The response function is simply a source-to-dose conversion function and is representative of the TN-40HT cask shielding configuration.

A representative response function is adequate for ranking the impact of burnup, enrichment and cooling time combination on dose rate. In essence the dose rate at 2 meters is calculated for each gamma energy group of the source as calculated by SAS2H. For neutrons, since a bounding energy spectrum is used, the response function calculated is just a total source to dose factor.

The response function is then utilized with SAS2H models to estimate the 2 meter side dose rate as a function of different burnup, enrichment and cooling time combinations.

Other information, such as the decay heat per fuel assembly and the cooling time are also collected.

The response function is shown in Table A7.2-1 0. As described above, the response function for neutrons and secondary gamma is a total source to dose factor while that for the primary gamma is a function of the energy spectrum. Table A7.2-1 0 also provides the additional dose rate contribution from the active fuel portion of the BPRA.

A comparison of the neutron, gamma and total dose rate results for the design basis fuel based on the response function and the calculational MCNP results (mid-plane average from Table A7A.5-2) indicates that the response function results are adequate (ratio of neutron to gamma) for the purpose of fuel qualification (relative comparison of source terms).

The response function is employed to determine the design basis spent fuel parameters from among seven limiting combinations of burnup and cooling time (BECT). These combinations are selected such that the resulting decay heat is greater than the maximum allowable decay heat of 800 watts per fuel assembly.

Four sets of calculations (A, B, C and D) are performed to determine the design basis spent fuel parameters by a comparison of the resulting response function dose rates for the combinations of spent fuel parameters.

The results of these calculations are shown in Table A7.2-1 1 and Table A7.2-12. Cases A1 through A7 show the results of the response function dose rate calculations for the seven limiting BECT combinations. These calculations show that Case A7 results in the highest dose rate. Cases B1 through B8 show the results of the response function dose rate calculations for eight BECT combinations with a decay heat of approximately 800 watts per fuel assembly. Cases B1 through B8 represent the actual BECT combinations (fuel that would more closely qualify for loading) while A1 through A7 represent conservative combinations. As expected, the dose rates for cases B1 through B7 are lower than those for cases A1 through A7. Based on the results of this evaluation, the design basis source terms for shielding are obtained conservatively from

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-7 the Westinghouse 14x14 standard fuel assembly with an enrichment of 3.40 wt. % U-235, a burnup of 60,000 MWD/MTU and a cooling time of 18 years. Cases B9 and B1 0 represent BECT combinations at enrichments of 2.1 wt. % U-235 and 1.0 wt. % U-235.

Their results are also bounded by A7.

The results of the sensitivity calculations - C and D cases are shown in Table A7.2-

12. Cases C1 through C8 are sensitivity calculations where the soluble boron concentration is increased from 600 ppm to 1000 ppm. A boron concentration of 1000 ppm averaged over the entire depletion is a conservative representation of the boron concentration during actual depletion. The results of these evaluations show that the increase in boron concentration results in an increase in the dose rate by approximately 1.5% and an increase in the decay heat by approximately 1%.

Cases D1 through D8 are sensitivity calculations where the moderator temperature is increased from 558 K (545°F) to 590 K (602°F, representative of an average hot leg moderator temperature) and the moderator density is correspondingly reduced from 0.733 g/cm 3 to 0.690 g/cm 3 . The soluble boron concentration is maintained at 1000 ppm, similar to that of the previous sensitivity evaluation. The results of these evaluations show that the increase in moderator temperature and soluble boron concentration results in an increase in the dose rate by approximately 4% and an increase in the decay heat by approximately 2%.

However, a comparison of the results from the A, B, C and D cases demonstrate that the highest calculated dose rate is obtained from Case A7. Therefore Case A7 represents the design basis case from a fuel qualification standpoint.

The calculated dose rate and decay heat along with the cooling time are then utilized to determine the bounding radiological source term. The final design basis radiological source term is generated by adding the fuel insert source term to the fuel/hardware source term.

The evaluation with the response function showed the design basis fuel assembly with parameters of 3.4 %wt U-235, 60 GWD/MTU burnup and cooled for 18 years to provide the appropriate bounding source terms. It is important to note that the decay heat for this bounding assembly is 845 watts which is in excess of the allowable limit as discussed in Section A3.

A7.2.7 RECONSTITUTED FUEL ASSEMBLIES Reconstituted fuel assemblies are fuel assemblies that have replaced damaged fuel pins with either natural uranium dioxide replacement rods, Zirconium inert rods, or stainless steel rods. These replacement rods have the same dimensions as the damaged fuel pin being replaced. While lower enriched fuel rods will have a higher source term than higher enriched rods with the same burnup, and activated stainless steel rods will initially have a higher source than a fuel pin due to Cobalt-60, the source

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.2-8 term of the design basis fuel described in Section A7.2.1 will bound reconstituted fuel assemblies for the following reasons:

Since the replacement rods will see at least one cycle of exposure less than the damage rods would have, the burnup of the natural uranium dioxide replacement pins will be at most 2/3 of the burnup that the damaged fuel pin would have seen.

This difference is enough to ensure that the source term of the design basis fuel bounds reconstituted fuel assemblies with natural uranium dioxide replacement pin(s).

The source term due to activation of a Zirconium inert rod is much less than the source term would be for the fuel pin being replaced. Thus, the source term of a reconstituted fuel assembly with Zirconium inert rod(s) is bounded by the source term of the design basis fuel.

The source term (primarily Cobalt-60) for the activated steel pin is greater than what the replaced fuel pin would have been at time of discharge. The decay of the steel rod source term is much greater than the replaced rod. After the specified minimum cooling time of 12 years, the Cobalt-60 activity in the steel rod has decayed to less than 1/4 of its original value. This decay is sufficient to ensure that the source term of the design basis fuel bounds reconstituted fuel assemblies with natural uranium replacement pin(s).

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.3-1 A7.3 RADIATION PROTECTION DESIGN FEATURES A7.3.1 STORAGE SYSTEM DESIGN DESCRIPTION The TN-40HT cask has a number of design features which ensure a high degree of integrity for the confinement of radioactive materials and reduction of direct radiation exposures to levels that are as low as practical.

- The casks are loaded, sealed, and decontaminated prior to transfer to the ISFSI.

- The fuel will not be unloaded nor will the casks be opened at the ISFSI.

- The fuel will be stored dry inside the casks, so that no radioactive liquid is available for leakage.

- The casks will be sealed with a helium atmosphere to preclude oxidation of the fuel.

- The casks will be heavily shielded to reduce external dose rates. The shielding design features are discussed below.

- No radioactive material will be discharged during storage.

Shielding for the TN-40HT cask is provided mainly by the thick-walled cask body. For neutron shielding, a borated polyester resin compound surrounds the cask body and a polypropylene disk covers the lid for storage. Additional shielding is provided by the steel shell surrounding the resin layer and by the stainless steel and aluminum/steel structure of the basket.

A7.3.2 SHIELDING The design of the cask shielding, material of construction, and method of calculating shielding parameters are explained in Appendix A7A. No special features or remote handling of the casks in the ISFSI are proposed. Portable shielding may be used during cask maintenance, if appropriate.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.3-2 A7.3.3 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION As indicated in Section A3.3.5.3, there are no credible events associated with use of the TN-40HT casks which could result in releases of radioactive products or unacceptable increases in direct radiation. Area radiation and airborne radioactivity monitors are therefore not needed at the ISFSI; however, TLDs will be used to record dose rates along the outer (nuisance) fence boundary of the ISFSI

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.4-1 A7.4 ESTIMATED ONSITE COLLECTIVE DOSE ASSESSMENT Table A7.4-1 shows the estimated design basis occupational exposures to ISFSI personnel during the loading, transport and emplacement of the storage casks. Dose rates utilized to determine occupational exposures to ISFSI personnel during the loading, transport and emplacement of the storage casks are conservatively estimated from those reported in Appendix A7A.

Table A7.4-2 shows the estimated design basis annual exposure for surveillance and maintenance activities. Visual surveillance was based on a walkdown of each of the two pads at a distance no closer than 2 meters to the casks. The dose rate for visual surveillance activities was increased to account for dose rate due to loaded casks already at the ISFSI and to account for short term activities close to the cask.

To estimate dose rates for operability tests and calibration, the worker was assumed to be at the monitoring panel at the perimeter fence entrance and exposed to a dose rate of 4 mrem/hr. During instrumentation and surface defect repairs the worker is assumed to be exposed to a dose rate of 300 mrem/hr that is representative of high surface dose rates on the cask. For major maintenance work, the worker is assumed to be exposed to the 1 meter side dose rate.

Both Table A7.4-1 and Table A7.4-2 provide for each task the estimated time required for the task, number of personnel required, the design basis dose rates and the exposure. The purpose of these tables is to provide an estimated total dose and not to prescribe limits or restrictions on dose rates, times to complete tasks, or number of persons working on tasks. Actual loading and maintenance activities may deviate from those shown in the tables but will be conducted such that exposure is as low as reasonable achievable.

Localized regions of elevated dose rates should be anticipated and minimized with good ALARA practices. Such regions exist due primarily to radiation streaming, including for example, streaming through the vent and drain ports.

To evaluate the additional dose to station personnel from ISFSI operations, a dose rate analysis has been performed using the dose rate versus distance results given in Appendix A7A.7. The occupational dose calculation considers all workers at the Prairie Island Nuclear Generating Plant to be in buildings (however no credit is taken for shielding of personnel by buildings) or in the plant yard. This population includes a normal work force and contractor personnel as well as the increased staffing required during outages.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.4-2 Table A7.4-3 provides a summary of staffing levels assumed at various site locations along with the distance from the center of the ISFSI. The locations designated on Table A7.4-3 are shown on Figure A7.4-A. This information is considered to be a general estimate of station staffing and does not necessarily reflect actual staffing at any given time. Changes in staffing levels and locations can be expected to occur in the future without affecting the general estimates contained in this section.

Distance dependent dose rates are provided in Table A7A.7-2. The data is based on loading 48 TN-40HT casks over a 22-year period assuming 4 spent fuel casks are loaded every 2 years. The dose rates in Table A7.4-3 were conservatively (using linear interpolation) calculated using the dose rate vs. distance data corresponding to the corner detectors.

Table A7.4-4 summarizes the calculated total doses to full time and outage help at the various locations due to ISFSI operation. The collective onsite dose by location and the total onsite dose estimates are also shown on this table.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.5-1 A7.5 OFFSITE COLLECTIVE DOSE ASSESSMENT Figure 1.2-1 illustrates the Prairie Island Nuclear Generating Plant site boundary, which is also the boundary of the exclusion area as defined in 1 0CFR1 00.3. This exclusion area corresponds to the controlled area for the ISFSI as defined in 1 0CFR72.3.

In calculating the offsite collective dose, the entire permanent population within a 2 mile radius of the plant was conservatively taken to be at the location of the residence subject to the highest exposure, i.e. 0.45 miles NW of the ISFSI. In addition to the permanent population, there is a large transient population of persons employed at or visitors to the Treasure Island Hotel and Casino that is located within a 2 mile radius of the plant. For these calculations it is assumed that this entire transient population is located 0.8 miles from the ISFSI. The estimates of the population (both permanent and transient) within the 2 mile radius were taken from the Prairie Island Nuclear Generating Plants evacuation time study (Reference 6). A description of the off-site locations considered in this evaluation, the relevant population data, distances and occupancy times are shown in Table A7.5-1.

The dose rate resulting from cask storage at the ISFSI as a function of distance is given in Table A7A.7-2. For a distance of 0.45 mile (724 meters) in the corner direction, the total dose rate is 2.20 E-3 rem/year. This is the annual exposure at the nearest resident location.

Table A7.5-2 shows the total dose rates from the ISFSI at each of the assumed off-site locations. In addition, Table A7.5-2 also contains the information summarized in Table A7.5-1. For conservatism, the dose rates calculated at a distance of 1004 m from the north-south face of the ISFSI are utilized to determine the off-site exposures at distances greater than 0.8 miles (1280 m).

Table A7.5-2 summarizes the calculated total doses to the off-site population within a 2-mile radius due to the ISFSI operation. The total collective off-site dose is calculated to be 3.60 person-rem.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.6-1 A7.6 RADIATION PROTECTION PROGRAM The information in Section 7.6 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.7-1 A7.7 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The information in Section 7.7 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7.8-1 A

7.8 REFERENCES

SCALE 4.4, Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers, CCC-545, ORNL.

2. MCNP/MCNPX - Monte Carlo N-Particle Transport Code System Including MCNP5 1.40 and MCNPX 2.5.0 and Data Libraries, CCC-730, Oak Ridge National Laboratory, RSICC Computer Code Collection, January 2006.

3. ORNL/TM-1 1018, Standard- and Extended-Burnup PWR and BWR Reactor Models for the ORIGEN2 Computer Code, Oak Ridge National Laboratory, December 1989.

4. NUREG/CR-680 1 (ORNL/TM-200 1/273), March 2003, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis.

5. Luksic, Spent Fuel Assembly Hardware: Characterization and 10 CFR 61 Classification for Waste Disposal, PNL-6906, UC-85, June 1989.

6. Evacuation Time Estimate Study for the Prairie Island Nuclear Generating Plant Emergency Planning Zone, September 25, 2003

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT 13 TABLE A7.2-1 PRAIRIE ISLAND FUEL ASSEMBLY DESIGN CHARACTERISTICS Fuel Designations Exxon Std Exxon Std Exxon Westinghouse Westinghouse (14x14) High Toprod Standard OFA Burnup (14x14) (14x14) (including (14x14) Vantage+)

_________________ __________ __________ _________ _____________ (14x14)

Max Length (in) 161.3 161.3 161.3 161.3 161.3 Max Width (in) 7.763 7.763 7.763 7.763 7.763 (1)

Fuel Density 93.18 93.20 92.80 93.32 94.80

(% Theoretical) ___________ ___________ __________ ______________ _____________

Rod Pitch (in) 0.556 0.556 0.556 0.556 0.556 No of Fueled Rods 179 179 179 179 179 Maximum Active 144.0 144.0 144.0 144.0 144.0 Fuel Length (in) ____________ ____________ ___________ _______________ _______________

Fuel Rod OD (in) 0.4240 0.4260 0.41 70 0.4220 0.4000 Clad Thickness (in) 0.030 0.031 0.0295 0.0243 0.0243 Fuel Pellet OD (in) 0.3565 0.3565 0.3505 0.3659 0.3444 Clad Material Zr-4 Zr-4 Zr-4 Zr-4 Zr-4/ZIRLO Guide Tube OD (in) 0.541 0.541 0.541 0.539 0.528 Guide Tube Wall 0.017 0.017 0.017 0.017 0.019 Thickness (Zr - 4) (in) ____________ ____________ ___________ _______________ _______________

GuideTube# 16 16 16 16 16 Instrument Tube # 1 1 1 1 1 Instr. Tube OD (in) 0.424 0.424 0.424 0.422 0.4015 Instr. Tube Wall 0.025 0.025 0.025 0.0243 0.0258 Thickness (Zr - 4) (in) ____________ ____________ ___________ _______________ _______________

Maximum (2)

MTU/assembly 0.370 0.370 0.370 0.410 0.360 Free Gas Volume 0.226 0.226 0.226 0.107 0.107 m 3 @STP _________ _________ ________ ___________ ___________

Fill Gas He He He He He Notes:

(1) The fuel density values are as-built, i.e., they incorporate the effect of the pellet dish and chamfer, as well as the theoretical density of the pellet. These densities were calculated from reload data provided by the fuel manufacturer, as follows:

As-built UO density = total amount UO in reload/ total volume of fuel in reload 2 2 Where fuel volume = ( # fuel pins) ( p/4) (pellet OD)2 ( active fuel length)

(2) The maximum MTU/assembly is calculated based on the theoretical density. The calculated value is higher than the actual value.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-2 WESTINGHOUSE 14X14 STANDARD FUEL ASSEMBLY HARDWARE CHARACTERISTICS Average Mass Item Material (kg/assembly)

Fuel Zone Cladding Zircaloy 83.4 Spacers I nconel 5.37 Guide/Instrument Tube* Stainless Steel 7.74 Fuel-Gas Plenum Zone Cladding Zircaloy 4.13 Springs Stainless Steel 5.68 Spacer I nconel 0.68 Guide/Instrument Tube* Stainless Steel 0.38 Top End Fittin g Zone Top Nozzle Stainless Steel 6.30 Hold Down Springs I nconel 0.51 Bottom End Fitting Zone Bottom Nozzle Stainless Steel 7.89 T 122 Conservatively assumed as stainless steel

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-3 MATERIAL COMPOSITIONS FOR FUEL ASSEMBLY HARDWARE MATERIALS Material Composition, grams per kg of

______________ material Atomic Element Number Stainless Zircaloy-4 Inconel-718 Steel 304 H 1 1.30E-02 - -

Li3 - - -

B 5 3.30E-04 - -

C 6 1.20E-01 4.00E-01 8.00E-01 N 7 8.00E-02 1.30E^00 1.30E^00 O 8 9.50E-01 - -

F9 - - -

Na11 - - -

Mg12 - - -

Al 13 2.40E-02 5.99E^00 -

Si 14 - 2.00E^00 1.00E^01 P 15 - - 4.50E-01 S 16 3.50E-02 7.00E-02 3.00E-01 Cl 17 - - -

Ca 20 - - -

Ti 22 2.00E-02 7.99E^00 -

V 23 2.00E-02 - -

Cr 24 1.25E^00 1.90E^02 1.90E^02 Mn 25 2.00E-02 2.00E^00 2.00E^01 Fe 26 2.25E+00 1.80E+02 6.88E+02 Co 27 1.00E-02 4.69E^00 8.00E-01 Ni 28 2.00E-02 5.20E^02 8.92E^01 Cu 29 2.00E-02 9.99E-01 -

Zn 30 - - -

Zr 40 9.79E^02 - -

Nb 41 - 5.55E^01 -

Mo 42 - 3.00E^01 -

Ag47 - - -

Cd 48 2.50E-04 - -

In49 - - -

Sn 50 1 .60E^01 - -

Gd 64 - - -

Hf 72 7.80E-02 - -

W 74 2.00E-02 - -

Pb 82 - - -

U 92 2.00E-04 - -

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-4 MATERIAL COMPOSITIONS FOR THE WESTINGHOUSE STANDARD 14X14 FUEL ASSEMBLY Material Compositions for the Westinghouse Standard 14x14 Fuel Assembly Fuel Height Zircaloy UO 2Inconel - SS-304 Total Volume Zone (cm) (Kg) (Kg) 718 (Kg) (Kg) (Kg) (cm 3 )

Bottom 7.82 7.893 7.893 3.042E+03 Core 365.8 83.36 465.1 5.370 7.736 561 .566 1.422E+05 Plenum 18.14 4.13 0.680 6.067 10.88 7.053E+03 Top 8.89 0.508 6.300 6.808 3.456E+03 Total 87.49 465.1 6.558 27.996 587.144 _________

Note: The PWR fuel assembly is modeled as a homogenized cuboid with a cross section area of 388.80 cm 2 (19.718 cm by 19.718 cm) or about 7.763" by 7.763".

The assembly is conservatively assumed to contain 410 Kg Uranium.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-5 FUEL INSERTS - MATERIALS AND MASSES Component Region Material Mass (kg)

Top Fitting TPD Baseplate, yoke, hold down bar, Type 304 SS 2.5 etc.

spring Inconel 718 0.36 Plenum Thimble plugs Type 304 SS 2.2 Top Fitting BPRA Baseplate, yoke, hold down bar, Type 304 SS 2.5 etc spring Inconel 718 0.36 Plenum Cladding & liner Type 304 SS 0.42 Fuel Zone

___________ Cladding & liner Type 304 SS 7.96

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-6 RADIOACTIVE INVENTORY FOR 14X14 DESIGN BASIS FUEL ASSEMBLY Classification Nuclide 14x14 W std

____________ _____________ (Ci/assembly)

(1) gas H3 1.78E+02 fine Pu238 3.02E+03 fine Pu239 1 .35E+02 fine Pu240 2.67E+02 fine Pu241 3.19E+04 fine Am241 1.50E+03 fine Am243 4.17E+01 fine Cm244 5.28E+03 gas Kr 85 1 .78E+03 volatile Sr 90 3.11E+04 fine Y 90 3.11E+04 gas I129 2.40E-02 volatile Cs134 4.01E+02 volatile Cs137 5.27E+04 fine Ba137m 4.97E+04 fine Pm147 6.53E+02 fine Eu154 1.33E+03 fine Np239 4.17E+01

____________ Total 2.11E+05 SAS2H Results for all (2)

Isotopes Light Elements 1 .44E+03 Actinides 4.22E+04 Fission Products 1.69E+05

____________ Total 2.13E+05 Notes:

1) The total of H3 from Light Elements and Fission Products.

2) A comparison of the total radioactive inventory indicates that the activity of the selected isotopes represents greater than 99.1% of the total.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-7 AXIAL SOURCE TERM PEAKING FACTOR Fractional Core Gamma Neutron Height Profile Profile 0 to 0.0556 0.573 0.108 0.0556 to 0.111 0.917 0.707 0.111 to 0.167 1.066 1.291 0.167 to 0.222 1.106 1496 0.222 to 0.278 1.114 1.540 0.278 to 0.333 1.111 1.524 0.333 to 0.389 1.106 1.496 0.389 to 0.445 1.101 1.469 0.445 to 0.500 1.097 1.448 0.500 to 0.556 1.093 1.427 0.556 to 0.611 1.089 1.406 0.611 to 0.667 1.086 1.391 0.667 to 0.722 1.081 1.366 0.722 to 0.778 1.073 1.326 0.778 to 0.833 1.051 1.220 0.833 to 0.889 0.993 0.972 0.889 to 0.944 0.832 0.479 0.944 to 1.000 0.512 0.069 Average: 1.00 1.15

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-8 SOURCE DISTRIBUTION FOR 14X14 FUEL ASSEMBLY

_________ - _________ Gamma Source(1) (particles/sec/assembly)

E min, MeV to E max, MeV Bottom Nozzle In-core Plenum Top Nozzle 0.00E+00 to 5.00E-02 4.6970E+10 1.1471 E+15 6.41 66E+1 0 2.8144E+1 0 5.00E-02 to 1.00E-01 5.8263E+09 2.2692E+14 7.4846E+09 3.4247E+09 1 .00E-01 to 2.00E-01 1 .4050E+09 1 .5457E+14 2.0286E+09 8.2647E+08 2.00E-01 to 3.00E-01 6.9758E+07 4.6484E+13 1 .0335E+08 4.1 160E+07 3.00E-01 to 4.00E-01 9.1527E+07 3.0384E+13 1.7060E+08 5.3770E+07 4.00E-01 to 6.00E-01 5.7847E+06 3.8369E+13 1.1945E+09 3.3967E+06 6.00E-01 to 8.00E-01 2.0089E+06 1.6010E+15 1.5982E+09 3.6813E+08 8.00E-01 to 1.00E+00 7.9012E+07 2.4369E+13 1.0456E+09 3.9964E+08 (2) (2) 1.6114+12 (2) 1 .00E+00 to 1 .33E+00 1 .7003E+ 12 7.4979E+ 13 3.081 9E+ 12 (2) (2) (2) 1 .33E+00 to 1 .66E+00 4.801 6E+ 11 1 .5249E+ 13 8.7003E+ 11 4.5485+11 1.66E+00 to 2.00E+00 1.6758E-06 7.8481E+10 2.9812E+01 2.4793E-03 2.00E+00 to 2.50E+00 1.1395E+07 4.4802E+09 1.4591E+07 6.6906E+06 2.50E+00 to 3.00E+00 1 .7669E+04 4.4758E+08 2.2624E+04 1 .0374E+04 3.00E+00 to 4.00E+00 1 .5658E-14 7.9748E+07 2.7747E-1 1 1 .0249E-1 1 4.00E+00 to 5.00E+00 0.00 2.6150E+07 0.0 0.0 5.00E+00 to 6.50E+00 0.00 1.0495E+07 0.0 0.0 6.50E+00 to 8.00E+00 0.00 2.0588E+06 0.0 0.0 8.00E+00 to 1.00E+01 0.00 4.3714E+05 0.0 0.0 Total Gamma: 2.2349E+ 12 3.3595E+ 15 4.0297E+ 12 2.0095E+ 12 (1)

________ - _________ Neutron Source (particles/sec/assembly)

E min, MeV to E max, MeV Bottom Nozzle In-core Plenum Top Nozzle 1.00E-01 to 4.00E-01 0.00 2.89E+07 0.00 0.00 4.00E-01 to 9.00E-01 0.00 1.48E+08 0.00 0.00 9.00E-01 to 1.40E+00 0.00 1.35E+08 0.00 0.00 1.40E+00 to 1.85E+00 0.00 9.94E+07 0.00 0.00 1.85E+00 to 3.00E+00 0.00 1.75E+08 0.00 0.00 3.00E+00 to 6.43E+00 0.00 1.59E+08 0.00 0.00 6.43E+00 to 2.00E+01 0.00 1.41E+07 0.00 0.00 (3)

Total Neutrons (n/sec/FA) 7.59E+8 The design basis gamma source term correspond to the Westinghouse 14x14 Standard fuel assembly with 3.40 wt.% U-235 enrichment, 60,000 MWD/MTU burnup, 18 year cooling time with TPD/BPRA insert source term.

Total gamma source from the fuel assembly and the BPRA/ TPD source shown in Table A7.2-9.

The neutron source spectrum shown in the table is from the SAS2H run and thus shows the spectrum due to all spontaneous fissions as well as from the alpha-neutron reactions. For the shielding calculations, the neutron source spectrum is modeled as Cm-244 using the built-in MCNP distribution and the total neutron source.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-9 GAMMA SOURCE TERMS FOR FUEL INSERTS Gamma Source from 125 GWD/MTU, 16 Yr Cooled TPD (gammas/sec/assembly)

In-Core Plenum Top End Fitting Total 1 - 1.33 1.33 - 1 - 1.33 1.33 - 1 - 1.33 1.33 -

MeV 1.66 MeV MeV 1.66 MeV MeV 1.66 MeV

- - 9.05E+11 2.55E+11 6.13E+11 1.73E+11 1.95E+12 Gamma Source from 30 GWD/MTU, 18 Yr Cooled BPRA (gammas/sec/assembly)

In-Core Plenum Top End Fitting Total 1 - 1.33 1.33 - 1 - 1.33 1.33 - 1 - 1.33 1.33 -

MeV 1.66 MeV MeV 1.66 MeV MeV 1.66 MeV 3.10E+12 8.73E+11 - - - - 3.97E+12

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-10 RESPONSE FUNCTION FOR TN-40HT CASK Response Function

((mrem/hour) per

___________________ particle) per cask Neutron 5.38E-09 Secondary Gamma 2.32E-08 Primary Gamma Energy Range Response Function

((mrem/hour) per (MeV) particle) per cask 0.40 to 0.60 8.1 1E-18 0.60 to 0.80 7.86E-16 0.80 to 1.00 5.39E-15 1.00 to 1.33 4.03E-14 1.33 to 1.66 1.84E-13 1.66 to 2.00 5.73E-13 2.00 to 2.50 1.57E-12 2.50 to 3.00 3.43E-12 3.00 to 4.00 7.32E-12 Dose Rate from BPRA 0.29 mrem/hour Calculational Neutron Gamma Total Dose Model (mrem/hour) (mrem/hour) (mrem/hour)

Response

Function 13.52 11.07 24.59

Response

Function (BPRA) 13.52 11.36 24.88 TN-40 HT (1)

Shielding 10.10 9.10 19.20 Ratio 0.75 0.80 0.77 (1) The neutron, gamma and total dose rates are obtained as an average of the dose rates shown in Table A7A.5-2. The dose rates at axial height ranging from -22.9 cm to 27.8 cm are included in the average calculations.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-1 1 FUEL QUALIFICATION CALCULATIONS FOR TN-40HT CASK Cooling Decay Burnup Enrichment Time Heat Dose Rate (mrem/hour)

Case (GWD/MTU) (wt.% U-235) (years) (watts) Neutron Gamma Total

_____ ___________ Design Basis Cases for Fuel Qualification _______ ______

A1 52 3.4 12.2 813 9.82 14.55 24.36 A2 53 3.4 12.8 817 10.31 14.09 24.40 A3 56 3.4 14.9 829 11.77 12.65 24.42 A4 57 3.4 15.6 835 12.25 12.27 24.52 A5 58 3.4 16.4 838 12.68 11.82 24.50 A6 59 3.4 17.2 841 13.10 11.43 24.53 A7 60 3.4 18.0 844 13.52 11.07 24.59

_____ Fuel Qualification for Decay Heat of 800 Watts/Assembly ______

B1 52 3.4 12.7 798 9.63 13.80 23.43 B2 53 3.4 13.5 799 10.05 13.12 23.17 B3 56 3.4 16.1 803 11.25 11.36 22.62 B4 57 3.4 17.0 805 11.63 10.88 22.50 B5 58 3.4 18.1 801 11.90 10.28 22.18 B6 59 3.4 19.1 802 12.21 9.84 22.05 B7 60 3.4 20.2 800 12.45 9.37 21.83 B8 60 4.9 18.0 798 7.46 10.05 17.52 B9 44 2.1 12.0 687 10.03 14.39 24.42 B10 19 1.0 12.0 272 1.25 7.30 8.55

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.2-12 FUEL QUALIFICATION SENSITIVITY CALCULATIONS FOR TN-40HT CASK Cooling Decay Burnup Enrichment Time Heat Dose Rate (mrem/hour)

Case (GWD/MTU) (wt.% U-235) (years) (watts) Neutron Gamma Total

____ Sensitivity - Soluble Boron Concentration of 1000 ppm _____

C1 52 3.4 12.7 806 9.86 13.85 23.72 C2 53 3.4 13.5 806 10.27 13.18 23.45 C3 56 3.4 16.1 811 11.48 11.42 22.90 C4 57 3.4 17.0 812 11.85 10.94 22.79 C5 58 3.4 18.1 811 12.12 10.34 22.46 C6 59 3.4 19.1 811 12.42 9.90 22.32 C7 60 3.4 20.2 809 12.66 9.44 22.10 C8 60 4.9 18.0 806 7.65 10.16 17.81

_____ Sensitivity - Moderator Temperature of 590 K ________ ______

D1 52 3.4 12.7 818 10.26 14.01 24.27 D2 53 3.4 13.5 818 10.67 13.33 24.01 D3 56 3.4 16.1 824 11.87 11.57 23.45 D4 57 3.4 17.0 825 12.24 11.09 23.33 D5 58 3.4 18.1 823 12.50 10.49 22.99 D6 59 3.4 19.1 823 12.81 10.04 22.85 D7 60 3.4 20.2 821 13.04 9.58 22.62 D8 60 4.9 18.0 818 8.01 10.36 18.37

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.4-1 OCCUPATIONAL EXPOSURES FOR CASK LOADING, TRANSPORT, AND EMPLACEMENT Dose Time No. of Rate Required Persons Dose Task Description (mrem/hr) (hrs) (man) (man-rem)

Placement in Pool 2 2 3 0.012 Loading Process 2 5 5 0.050 Removal from Pool 32 5 5 0.800 Transfer to Decontamination 32 1 3 0.096 Area Processing of Cask 43 6.5 2 0.559 Helium Leak Test 43 2 2 0.172 Decontamination 66 2 3 0.396 Install Neutron Shield, 43 3 2 0.258 Pressurize, Test Preparation for Transport 30 1 3 0.090 Transfer of Cask to ISFSI 18 1 3 0.054 Final Cask Emplacement 63 2 5 0.630 TOTAL N/A 30.5 N/A 3.117

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.4-2 DESIGN BASIS ISFSI MAINTENANCE OPERATIONS ANNUAL EXPOSURES Dose Time No. of Rate Required Persons Dose (man-Task Description (mrem/hr) (hrs) (man) rem)

Visual Surveillance of Casks 63 1 2 0.126 Instrumentation Operability Tests 4 1 2 0.008 Instrumentation Calibration 4 2 2 0.016 Instrumentation Repairs 300 1 2 0.600 Surface Defect Repair 300 1 2 0.600 Major Maintenance 32 32.5 3 3.120 TOTAL N/A 38.5 N/A 4.470

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.4-3 STATION PERSONAL LOCATION AND DOSE RATES Location Location Distance Full Outage Dose

1. (Feet From Time Rate, Center mrem/hr

________ ________________________________ ISFSI) ________ ________ ________

1 34 OCA Gatehouse (D-2) 978 2 0 2.36e-2 2 22 Receiving Warehouse (D-5) 2.55e-2 55 13 P1ex Project Office (D-5) 953 30 0

________ A 4 P1ex Project Office (D-5) __________ _______ _______ ________

3 24 NPD Building (E-6) 2.08e-2 33 NPD Annex Building (E-5) 1016 60 1 Quality Assurance Modular Office (E-5) 4 42 Fabrication Shop (C-5) 5.19e-3 25 Steam Generator Mock up Building 1381 2 45

________ (C-5) 48 Fabrication Shop (D-5) ___________ ________ ________ ________

5 30 Warehouse (C-5) 2.07e-3 31 Multiuse Warehouse (C-5) 1650 0 2

________ 21 Warehouse -8 (C-6) __________ _______ _______ ________

6 12 Substation, SBO Structures (B-6) 2236 0 2 3.60e-4 7 23 New Administration Building (D-6) 2.93e-3 1496 287 25

_________ 6 Security Building (D-6) ___________ ________ ________ ________

8 4 Turbine Building (D-7) 2.07e-3 28 Warehouse -2 (E-7) 3 Auxiliary Building (E-7) 13 New Service Building (D-8) 1650 180 192 11 D5/D6 DG Building (D-7) 79 Security DG Building (D-7) 9 Outage Trailers (E-7) 1322 5 40 6.39e-3 10 60 Fabrication Shop (E-7) 2.59e-3 29 Main Plant Warehouse (E-7) 1557 23 45 66 Maintenance Storage Building (E-7) ___________ ________ ________ ________

11 20 Environmental Lab (B-9) 2534 2 0 1 .58e-4 12 Training Center 905 49 0 2.91e-2 13 38 Old Administration Building (D-7) 1 .53e-3 71 Administration Building Addition (D- 1745 68 0

_________ 7) ___________ ________ ________ ________

14 13 New Service Building (D-8) 1905 9 0 9.36e-4

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.4-4 STATION PERSONAL COLLECTIVE DOSE Location Location Distance Full Outage ** Total

1. (Feet From Time

• Exposure, Center person-

______________________________ ISFSI) rem 1 34 OCA Gatehouse (D-2) 978 0.12 0 0.12 22 Receiving Warehouse (D-5) 1.91 0 1.91 2 55 13 P1ex Project Office (D-5) 953

________ A 4 P1ex Project Office (D-5) __________ _______ ________ ________

24 NPD Building (E-6) 3.12 1.12e-2 3.13 33 NPD Annex Building (E-5) 3 1016 Quality Assurance Modular Office (E-

5) ___________ _______ _________ _________

42 Fabrication Shop (C-5) 0.03 0.13 0.15 4 25 Steam Generator Mock up Building 1381

________ (C-5) 48 Fabrication Shop (D-5) ___________ _______ _________ _________

30 Warehouse (C-5) 0 0.00 2.23e-3 5 31 Multiuse Warehouse (C-5) 1650

________ 21 Warehouse -8 (C-6) __________ _______ ________ ________

6 12 Substation, SBO Structures (B-6) 2236 0 3.89e-4 3.89e-4 23 New Administration Building (D-6) 2.10 0.04 2.14 7 1496

________ 6 Security Building (D-6) ___________ _______ _________ _________

4 Turbine Building (D-7) 0.93 0.21 1.14 28 Warehouse -2 (E-7) 3 Auxiliary Building (E-7) 8 13 New Service Building (D-8) 1650 11 D5/D6 DG Building (D-7) 79 Security DG Building (D-7) 9 Outage Trailers (E-7) 1322 0.08 0.14 0.22 60 Fabrication Shop (E-7) 0.15 0.06 0.21 10 29 Main Plant Warehouse (E-7) 1557 66 Maintenance Storage Building (E-7) ___________ _______ _________ _________

11 20 Environmental Lab (B-9) 2534 0.00 0 7.91 e-4 12 Training Center 905 3.57 0 3.57 38 Old Administration Building (D-7) 0.26 0 0.26 13 71 Administration Building Addition (D- 1745

________ 7) ___________ _______ _________ _________

14 13 New Service Building (D-8) 1905 0.02 0 0.02 Total 12.28 0.59 12.88

• For full time employees, assume 2500 hour0.0289 days <br />0.694 hours <br />0.00413 weeks <br />9.5125e-4 months <br />s/year.

• • For outage employees, assume 540 hour0.00625 days <br />0.15 hours <br />8.928571e-4 weeks <br />2.0547e-4 months <br />s/year.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.5-1 OFFISTE POPULATION: LOCATION, DISTANCE, OCCUPANCY TIME Permanent population within 2 mile radius Description Population Occup times Distance 2 mile -radius 452 24hr/da ay/yr >0.45 mile Transient Employee Population ________________

Description Population Occupancy times Distance 2 mile - PI community 65 8hr/ day, 5 day/wk, 50 0.8 mile center, government wk/yr center, and clinic _____________ ___________________

2 mile - Employees at 650 At any give time; i.e. 0.8 mile Treasure Island Casino 24hr/day 365 day/yr Transient Hotels, Motels, and malls Population Description Population Occupancy times Distance 2 mile - Treasure Island 7,000 12hr/ day, 5 day/wk, 52 0.8 mile Casino (day) wk/yr 2 mile - Treasure 9,000 12hr/ day, 5 day/wk, 52 0.8 mile Island Casino (night) wk/yr 2 mile - Treasure Island 15,850 48 hr/weekend, 52 0.8 mile Casino (Week end) weekend/yr Transient recreational Population __________________ ________

Description Population Occupancy times Distance 2 mile - Treasure Island 21,202 3 hr/show, 3 0.8 mile Casino (special events) show/event, 24 events Transient hospitals and clinics population _________

Description Population Occupancy times Distance 2 mile - PI community 140 8hr/ day, 5 day/wk, 0.8 mile center, government 52wk/yr center, and clinic

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7.5-2 OFFSITE POPULATION: DOSE RATES AND COLLECTIVE DOSE Offsite Description Distance Occupancy Population Dose Collective Population times Rate, Exposure, mrem/hr person-

_________________ _________ __________ __________ ________ rem Permanent*

>0.45 population within 2 mile radiu s8760 452 2.42e-4 0.96 miles 2 mile radius ___________________ __________ ___________ ___________ _________ ____________

2 mile -PI community center, government 0.8 mile 2000 65 2.64e-5 3.43e-3 Transient center, and clinic __________ ___________ ___________ _________ ____________

Employee 2 mile -Employees at Population Treasure Island 0.8 mile 8760 650 2.64e-5 0.15

________________ Casino __________ ___________ ___________ _________ ____________

2 mile -Treasure 0.8 mile 3120 7000 2.64e-5 0.58 Island Casino (day) _________ __________ __________ ________ ___________

Transient hotels, 2 mile -Treasure motels, and malls 0.8 mile 3120 9,000 2.64e-5 0.74 Island Casino (night) population 2 mile -Treasure Island Casino 0.8 mile 2496 15,850 2.64e-5 1.04

________________ (Weekend) __________ ___________ ___________ _________ ____________

Transient 2 mile -Treasure recreational Island Casino (special 0.8 mile 216 21,202 2.64e-5 0.12 population events) __________ ___________ ___________ _________ ____________

Transient 2 mile -PI community hospitals and center, government 0.8 mile 2080 140 2.64e-5 7.69e-3 clinics population center, and clinic __________ ___________ ___________ _________ ____________

Total Collective 3.60 Within 2 Mile

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 1.2 1

0.8 C

0.6 /

0.4 0.2 0

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 Fraction of Active Core Height FIGURE A7.2-1 AXIAL BURNUP PROFILE FOR DESIGN BASIS FUEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 1.5 1.25 1

0.75 0.5 0.25 0

-200 -150 -100 -50 0 50 100 150 200 Axial Height (cm)

FIGURE A7.2-2 AXIAL NEUTRON AND GAMMA POWER PROFILES

Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.1-1 Appendix A7A TN-40HT CASK DOSE ANALYSIS A7A.1 SHIELDING DESIGN FEATURES Shielding for the TN-40HT cask is provided mainly by the cask body. For the neutron shielding, a borated polyester resin compound surrounds the cask body and a polypropylene disk covers the lid. Additional shielding is provided by the steel shell surrounding the resin layer and by the steel and aluminum structure of the fuel basket.

Geometric attenuation, enhanced by air and ground attenuation, provides additional dose reduction for distant locations at the restricted area and site boundaries.

Figure A7A.1-1 shows the configuration of shielding in the cask. Table A7A.1-1 lists the compositions of the shielding materials.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.2-1 A7A.2 DISCUSSION AND

SUMMARY

OF RESULTS The shielding evaluation presented in this section is performed utilizing as a design basis the source terms from the14x14 Westinghouse standard fuel assembly with fuel inserts. This fuel assembly is bounding because it contains the highest mass of fuel among the candidate fuel assemblies.

All shielding calculations are performed using the computer code MCNP (Reference 1) which uses a 3-dimensional Monte Carlo method. MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. Point-wise (continuous energy) cross-section data are used. The MCNP code is selected because of its ability to handle thick, multi-layered shields as in the TN-40HT cask body using 3-D geometry.

For neutrons, all reactions given in a particular cross-section evaluation are accounted for in the cross section set. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorption in pair production with local emission of annihilation radiation, and bremsstrahlung. Important standard features that make MCNP very versatile and easy to use include a powerful general source; an extensive collection of cross-section data; and an extensive collection of variance reduction techniques that can be employed to track particles through very complex deep penetration problems. Mesh tallies capabilities were utilized in calculating dose rates distributed over and around the surface of the TN-40HT cask.

Normal and off-normal conditions are modeled with the TN-40HT cask intact. Average dose rates on the side, top and bottom of the TN-40HT cask are calculated. Maximum dose rates are also calculated above and below the radial neutron shield.

Accident conditions assume radially the neutron shield and steel outer shell are removed and axially the polypropylene disk and protective cover are removed. This evaluation bounds the accident conditions in Section A8.

The expected dose rates (for normal, off-normal, and accident conditions) from the TN-40HT cask are provided in Table A7A.2-1. The dose point locations are illustrated in Figure A7A.2-1.

The total (direct + skyshine) dose rates as a function of distance from the cask are presented in Table A7A.2-2.

The total ISFSI dose rate (from 48 casks all containing design basis fuel at the time of initial loading) as a function of distance from each location (N/S side, E/W side, or corner) is summarized in Table A7A.7-2.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.3-1 A7A.3 METHODOLOGY The methodology used in the shielding analysis of the TN-40HT cask is the 3-D MCNP code. MCNP allows for explicit 3-D modeling of any shielding configuration and reduces the number of approximations needed. The methodology used herein is summarized below.

1. Sources are developed for all fuel regions using the source term data from Section A7.2. Source regions include the active fuel region, bottom end fitting (including all materials below the active fuel region), plenum, and top end fitting (including all materials above the active fuel region).

2. Suitable shielding material densities are calculated for all regions modeled.

3. The 3-D Monte Carlo transport code MCNP is used to calculate dose rates on and around the TN-40HT cask.

4. For the cask, importance biasing is utilized for variance reduction for both the near and far field dose rate calculation. F4 type tallies are utilized for determining near field dose rates and far field dose rates.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.4-1 A7A.4 MODEL SPECIFICATION The neutron and gamma dose rates on the surface of the TN-40HT cask (Side, Top and Bottom surfaces) and at 1 m and 2m from the side surface of the cask are evaluated with the 3-D Monte Carlo transport code MCNP (Reference 1). The flux-to-dose conversion factors specified by ANSI/ANS 6.1.1-1 977 (Reference 2) are used and provided in Table A7A.4-1.

The cross-section data used is the continuous energy ENDF/B-VI provided with the MCNP code. The cross-section data allows coupled neutron/gamma-ray dose rate evaluation to be made to account for secondary gamma radiation (n, g), if desired.

All of the near field dose rate calculations account for the dose rate due to secondary gamma radiation. For the far field (long distance) dose rate calculation, the dose rate contribution from the secondary gamma radiation is ignored because it is insignificant.

Figure A7A.4-1 is a plot of the TN-40HT cask MCNP model and shows the axial cross section of the cask. This shielding configuration is utilized to determine the radial (side surface) and axial (top and bottom surface) dose rates around the TN-40HT cask for normal, off-normal and accident conditions. This configuration is also utilized to determine the dose rates at long distances from the cask.

The following are key assumptions used in the development of the MCNP models:

• The condition of the cask during and after an accident assumes the side neutron shield and steel shell, the protective cover and the top neutron shield (polypropylene) are lost.

• The borated neutron absorber sheets in the TN-40HT basket are modeled as aluminum.

• Fuel is homogenized into 4 zones within the fuel assembly perimeter, although the TN-40HT basket is modeled explicitly.

• The basket is modeled as discrete stainless steel boxes surrounded by aluminum plates. The stainless steel support bars are conservatively neglected.

• The spatial distribution of the source is assumed to be uniform within each non-fuel hardware zone and within each axial burnup segment in the active fuel. Isotropic angular distribution is assumed for all sources.

A7A.4.1 MCNP MODEL FOR NORMAL AND OFF-NORMAL CONDITIONS A single shielding configuration is utilized for the TN-40HT design for both normal and off-normal conditions of storage. The cask bottom dose rates are determined as a consequence of off-normal conditions since these become important only during loading and transfer. During normal conditions of storage the cask is upright and is firmly seated on the concrete pad. A three-dimensional MCNP model which includes a discrete fuel assembly model of the TN-40HT cask was developed for this purpose. In

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.4-2 the MCNP model, the TN-40HT cask axis is modeled along the Z-direction. The X and Y axes in the MCNP model represent the cask in the radial direction. The cask is assumed to sit on a concrete pad (Z-direction).

The MCNP model for these shielding configurations is based on a discrete basket with the homogenized fuel assemblies (with an active height of 144 inches) positioned within fuel compartments. The MCNP model developed in this calculation is based on the design details from the TN-40HT cask drawings (within the limitations of the code geometry modeling options), shown in Section A1 .5, except for some conservative representations. Table A7A.1-1 provides the cask material densities and thicknesses as designed and employed in the MCNP models. Figure A7A.1-1 is a sketch of the TN-40HT cask containing the modeled dimensions in the shielding evaluation models.

Cells 2051 through 2133 represent the discrete basket and fuel assembly zones.

Figure A7A.4-1 is a Y-Z plot of the MCNP model of the TN-40HT cask (axial section view). All the major details of the cask model are shown in this figure. The fuel basket extends to 160" in the axial direction from the bottom of the bottom fitting (-190.70 cm) to about 2 above the top of the fuel assembly (209.91 cm). The active fuel zone is 144" long with the center at 0 (+/- 182.88 cm). A simple analog model is used for calculating the neutron dose. For the primary gamma dose rates, a multiple cell sub-layer model is used. The cask trunnions and the resin cutouts (flats) are modeled explicitly. The neutron resin boxes/radial neutron shield, top neutron shield and protective cover are also modeled explicitly.

The basket is modeled discretely using the advanced geometry features of MCNP. The fuel is modeled as a cuboid based on a 7.763" square. The fuel compartment inside dimension is 8.05" and is modeled with stainless steel with a thickness of 0.187" surrounded by a 0.437"-thick basket aluminum plates. The Boral © (or any other poison material) plates were modeled as pure aluminum and the stainless steel strips (tie plates) were not modeled. The stainless steel and aluminum peripheral rails were modeled explicitly. A small air gap of 0.15" (0.38 cm) was assumed between the basket and the cask shell. Figure A7A.4-2 is a radial cross section (X-Y plot) of the MCNP gamma model at Z=1 81 cm that shows the upper trunnions. Figure A7A.4-3 is also a radial cross section plot which shows one quarter of the basket and the rails.

Above the basket, the stainless steel fuel compartments and aluminum plates surrounding the fuel assembly are replaced by air (void).

The spatial distribution of the source is assumed to be uniform within each non-fuel hardware zone and within each axial burnup segment in the active fuel. Isotropic angular distribution is assumed for all sources.

Two MCNP models are developed for determining the normal and off-normal dose rates. The gamma model containing a detailed segmentation of the thicker cask steel body (for variance reduction purposes - implemented employing multiple cell sub-layers)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.4-3 is utilized to calculate the primary gamma dose rates. The neutron model is utilized to calculate the neutron and secondary gamma dose rates.

Tallies are based on the F4 type mesh tally feature of MCNP to provide the average flux in the defined volume. The radial tallies are located just off the cask surface, 1 m from cask outer surface and 2 m from cask outer surface. Similarly, axial tallies are located at the cask top (protective cover) and bottom which determine the average flux at the top and bottom surfaces, and 1m and 2m from the top surface.

A7A.4.2 MCNP MODEL FOR ACCIDENT CONDITIONS The MCNP design basis model for accident conditions is almost identical to that of the normal (and off-normal) conditions except that all neutron shielding and the outer steel shell materials are replaced with void. This scenario is based on a cask drop accident that results in the complete removal of the neutron shielding materials including the polypropylene disk and the protective cover at the top. This is implemented in MCNP by replacing the appropriate materials in the MCNP models with void.

A7A.4.3 MCNP MODEL FOR LONG DISTANCES FROM THE CASK The near field dose rate MCNP model described above is modified to determine the long distance dose rates. The near-field MCNP model is extended further to include additional volumetric detectors beyond the immediate vicinity of the cask and splitting cells are used to transport particles far from the cask (about 2000 m from the cask surface) to obtain dose rates at long distances. A schematic of the far field model is shown in Figure A7A.4-4. This modeling technique splits particles upwards and outwards to better simulate skyshine. At farther regions of the model, cones are used to split particles downward to the detector location.

The MCNP calculated dose rates at far distances consist of contributions from direct, air scatter (skyshine) and limited ground scatter (only in the immediate vicinity of the cask).

Modifications (cell flagging) are also made to separate the direct and skyshine component. Any radiation that crosses an elevation of 20 feet (~1.5 meter above cask) is considered to be skyshine. This is a valid approximation for two main reasons. One, this height corresponds to the earth berm. Second, beyond this elevation it is highly probable the radiation will have to scatter back in order to reach far detectors.

Soil is modeled as the ground surface under and extending away from the cask. It is expected that the maximum contribution to the ground scatter component occurs at the immediate vicinity of the cask. A layer of soil more than 4 feet thick is adequate to account for ground shine. Actual soil depth modeled in MCNP models is about 6 feet.

Axial splitting is not modeled in the soil.

The doses due to capture gamma sources are not calculated since they are insignificant at large distances in comparison to primary gamma and neutron sources.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.4-4 The dose is calculated as F4 tallies (that calculate the volumetric dose rates) in an annular cylindrical detector, 20 cm thick and 3 feet high at about 6 feet off the ground.

The dose rates are calculated at distances ranging from 1 0m to 2000m from the edge of the cask.

A7A.4.4 SHIELD REGIONAL DENSITIES Table A7.2-4 shows the fuel assembly material composition for the four fuel assembly regions. Based on these material compositions and material densities, atom fractions for the fuel assembly regions are determined and provided in Table A7A.4-2. The mass of materials in each fuel assembly region is homogenized over the volume of the region (area = 60.26 in 2 , based on a 7.763" x 7.763" cross-section). Table A7A.4-3 provide the shield regional densities for the TN-40HT cask. The actual fuel layout in the TN-40HT cask is an array of fuel assemblies inside stainless steel compartments surrounded by sheets of aluminum material.

The radial resin and aluminum boxes are discreetly modeled based on the material composition of the polyester resin and the aluminum.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision:13 Page A7A.5-1 A7A.5 NEAR FIELD DOSE RATE CALCULATIONS Two basic three-dimensional MCNP models are employed for shielding analyses of the TN-40HT cask, one model for neutrons and the other for gammas. An additional two models are utilized for the accident configuration. The results of the near field dose rate calculations for normal and off-normal conditions are shown in Table A7A.5-1 and Table A7A.5-2. These results are also shown in Figure A7A.5-1 through Figure A7A.5-5. A sample MCNP input file listing is provided in Section A7B. Additional cask surface dose rate results above and below the neutron shield for normal and off-normal conditions are shown in Table A7A.5-3.

Figure A7A.2-1 shows the locations around the cask where the dose rates are calculated. As expected due to the reduced shielding, there is an increase in the dose rates at the surface above and below the neutron shield as shown in Table A7A.5-3 in comparison to the cask surface side dose rates shown in Table A7A.5-1.

For the TN-40HT cask an accident case is performed assuming the neutron shield and steel neutron shield jacket (outer skin) of each have been torn off and the top neutron shield and protective cover are gone. The results of the near field dose rate calculations for accident conditions are shown in Table A7A.5-4. These results are also shown in Figure A7A.5-6 through Figure A7A.5-9.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.6-1 A7A.6 FAR FIELD DOSE RATE CALCULATIONS MCNP models are developed to determine the far field dose rates as discussed above.

The results of the far field dose rate calculations are shown in Table A7A.2-2 and Table A7A.6-1. These results are also shown in Figure A7A.6-1. The MCNP input file listing for neutron is shown in Section A7B.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.7-1 A7A.7 OFFSITE DOSE RATE CALCULATIONS This Section evaluates off-site dose rates assuming the ISFSI consists of two (2) 2x1 2 array of TN-40HT casks, loaded in sets of four casks every two years, each loaded with design-basis fuel at the time of initial loading (i.e. total of 48 casks and credit for decay while the casks are on the ISFSI pads). This conservatively bounds the currently allowed storage of 48 casks and bounds operation with a mixture of TN-40 and TN-40HT casks. The ISFSI layout used in this evaluation is shown in Figure A7A.7-1.

This evaluation determines the neutron and gamma-ray off-site dose rates including skyshine in the vicinity of the ISFSI. It provides results for distances ranging from 10 to 1,000 meters from the front, side and corner of the arrays. A co-ordinate transformation of these results is performed so that they may be reported as a distance from a common point, i.e., the center of the ISFSI.

The Monte Carlo computer code MCNP (Reference 1) calculates the dose rates at the specified locations around the array of casks. An MCNP Array method is utilized to evaluate the dose rate as a function of distance from the ISFSI.

A7A.7.1 MCNP ARRAY METHOD - METHODOLOGY, MODEL AND ASSUMPTIONS The ISFSI layout as illustrated in Figure A7A.7-1 is explicitly modeled in MCNP using advanced MCNP geometry. The gamma and neutron dose rates are determined as a function of distance from the ISFSI. All the doses are determined using F5 detector tallies.

The ISFSI array MCNP model consists of a two 2x12 arrays of TN40-HT casks. The cask array is modeled using advanced MCNP geometry to represent the ISFSI as shown in Figure A7A.7-1. Source specification is consistent with the loading scheme where four casks are loaded every two years. Due to the symmetry of the layout of the cask array and in the loading schedule, sources are only specified for the north-west row of casks (even-numbered casks 2 through 24 from Figure A7A.7-1). Even though, sources are not specified in the remaining 36 casks, the shielding configurations are identical and ensure that the resulting dose rate distribution around the ISFSI is symmetric. Tallies are constructed from small, spherical, F5 detectors providing for variance reduction capabilities. These detectors are positioned all around the ISFSI (all sides and corner locations) to determine the dose rates as a result of the source symmetry. These detector locations are shown in Table A7A.7-1.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.7-2 Credit is taken for the ISFSI berm. The berm is modeled as an isosceles trapezoid with 14.02 and 2.44 meters bottom and top segments, respectively. The berm is assumed 6.25 meters height and made from soil. Based on the coordinate system depicted on Figure A7A.7-1 (the berm is not shown on the Figure), the planes of symmetry along the west and east sides of the berm are X=-114.30 and X=121.61 meters, respectively.

Similarly, planes of symmetry along the north and south sides of the berm are at Y=-58.22 and Y=58.22 meters, respectively. -

Radiological sources from Bottom Nozzle, In-core, Plenum and Top Nozzle axial regions of the TN40 HT design basis fuel assembly are calculated for cooling times ranging from 18 to 40 years, with 2 years increments and shown in Table A7A.7-4. The SAS2H\ORIGEN-S models utilized in the design basis fuel source term calculations are also employed herein.

In order to simplify the source term specification in the MCNP models, the total gamma and neutron radiation source terms strength can be approximated with an exponential function as a function of decay time. The source strength at any cooling time is expressed as:

• e( - A(t-18))

At = A0 where At is the Source Strength at time t (18 ^ t ^ 40)

A0 is the source strength at 18 years A is a decay constant The decay constants are calculated based on the above equation using the source strengths obtained from the SAS2H calculations. Decay constants for calculating neutron, in-core gamma (active fuel region), and fittings (top and bottom nozzle and plenum regions) gamma radiation source terms strength are calculated to be equal to

-1 -1 0.0358 years (corresponds to ~19.4 years half-life), 0.025 years (corresponds to

-1

~27.7 years half-life) and 0.1315 years (corresponds to ~5.27 years half-life) respectively. These decay constants are calculated in such a way that exponential function produces source terms that match, within 1%, the source terms calculated using SAS2H\ORIGEN-S models.

For ease of input specification in the MCNP models, the total gamma source strength (4.8937e+18 gammas per second including contribution from BPRAs and TPDs.) is calculated utilizing the exponential functions. Spectrum due to the design basis fuel assembly is used to conservatively describe the spectral distribution of gamma radiation source terms in all the casks where source is specified.

For the neutron source specification, the total neutron source strength calculated directly (1 .9253e+12 neutrons per second, including sub-critical multiplication and axial source profile) with the SAS2H\ORIGEN-S models are utilized. Spectral distribution of neutron radiation source terms is due to Cm-244. MCNP provides built-in parameters to describe this distribution in the calculational models.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.7-3 The assumptions for the MCNP methodology are summarized below.

• Due to symmetry of the cask loading configurations, the source specification is simplified and considered only for 12 of the 48 casks while the detectors are positioned to obtain the results as a result of this simplified source specification.

• The "universe" is a sphere surrounding the ISFSI. The radius of this sphere (r=5.0 kilometers) is more than 20 mean free paths for neutrons and more than 20 mean free paths for gammas in energy groups that contribute the most to the gamma radiation dose rates. This ensures that the model is of a sufficient size to include all interactions affecting the dose rate at the detector locations.

• Detectors located at distances less than or equal to 45 meters are within the ISFSI berm perimeter. Dose rates points at larger distances are behind the berm. It is likely that the dose rates beyond the berm perimeter are almost entirely due to skyshine component.

• Dose rates include contribution due to primary gamma and neutron sources as well as due to the secondary gamma radiation from (n,g) interactions.

A7A.7.2 MCNP ARRAY METHOD - DOSE RATE AS A FUNCTION OF DISTANCE The MCNP results provide the dose rate as a function of distance at all the locations (including sides and corner) around the ISFSI. The total and skyshine dose rate results for the N/S sides, E/W sides and the corners are shown in Table A7A.7-2. The skyshine doses are shown in Table A7A.7-3 only for comparison. Due to presence of the ISFSI berm, it is expected that the dose rates at distances greater than 1 00m are dominated by the skyshine component and these results serve to demonstrate this assertion. Co-ordinate transformation is performed such that the results reported in Table A7A.7-2 are based on the distance from the center of the ISFSI.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-1 A7A.8 CONFINEMENT A7A.8.1 CONTAINMENT BOUNDARY The containment boundary consists of the inner shell and bottom plate, shell flange, lid outer plate, lid bolts, penetration cover plate and bolts and the inner portions of the lid seal and the two lid penetrations (vent and drain). The containment vessel prevents leakage of radioactive material from the cask cavity. It also maintains an inert atmosphere (helium) in the cask cavity. Helium assists in heat removal and provides a non-reactive environment to protect fuel assemblies against fuel cladding degradation.

A7A.8.2 SEALS AND WELDS The containment boundary welds consist of the circumferential welds attaching the bottom closure and the top flange to the vessel shell. Also, the longitudinal weld(s) on the rolled plate, closing the cylindrical vessel shell, and the circumferential weld(s) attaching the rolled shells together are containment welds.

Double metal seals are utilized on the lid and the two lid penetrations. Helicoflex HND or equivalent seals may be used. The seals are shown in Figure A7A.8-1. The internal spring and lining maintain the necessary rigidity and sealing force, and provide some elastic recovery capability. The outer aluminum jacket provides a ductile material to seal against the sealing surfaces. The jacket also provides a connecting sheet between the inner outer seals.

Holes in this sheet allow for attachment screws and for communication between the overpressure system and the space between the seals. This sheet, which is about 0.020 in. thick, has insufficient strength to transmit radial forces great enough to overcome the axial compressive forces on the seals, which are over 1000 lb/in. of seal length. Additional information on the seals is provided in Section A3.3.2. The overpressure port seal is a single metal seal of the same design, Helicoflex HN200 or equivalent.

All TN-40HT surfaces which mate with the metal seals are stainless steel.

The use of a double seal system allows the TN-40HT cask to have a pressure monitoring of the interspace between the seals. This combined cover-seal pressure monitoring system always meets or exceeds the requirement of a double barrier closure which guarantees tight, permanent confinement. When the cask is placed in storage a pressure greater than the cavity pressure is set up in the gaps (interspace) between the double metal seals of the lid and the lid penetrations. A decrease in the pressure of the monitoring system would be signaled by a pressure transducer in the overpressure system.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-2 The lid and penetration seals described above are contained in grooves. A high level of sealing over the storage period is assured by utilizing seals in a deformation-controlled design. The deformation of the seals is constant since bolt loads assure that the mating surfaces remain in contact. The seal deformation is set by the original diameter and the depth of the groove.

The nominal cross-section diameter of the lid seal is 0.26 in. and the nominal groove depth is 0.22 in. At 0.04 in. compression, the sealing force is 1,399 lb/inch (Reference 6). The total force of the double seal is 660,129 lb. The total minimum preload of the 48 lid bolts is 71,111 lb/bolt b, which is greater than the combined force of the seals and internal pressure, 23,175 lb/bolt (Appendix A4A.4).

The nominal cross-section diameter of the port seals is 0.161 in. and the nominal groove depth is 0.126 in. At .035 in. compression, the sealing force is 1,142 lb/inch.

The total force of the double seal is 37,922 lb. The total minimum preload of the 8 cover bolts is 7,111 lb/bolt, which is greater than the combined force of the seals and internal pressure, 5,058 lb/bolt (Appendix A4A.5).

The maximum radial force on the seals is from the 5.5 atm abs overpressure system.

This results in a force per unit seal length of about 15 lb./in., which is negligible compared to the compressive (axial) force of over 1000 lb/inch. Because the maximum pressure is between the two seals, the direction of this force is such that the seals are supported by the walls of the seal groove. However, the seals are designed to retain pressure in either direction.

-5 Helicoflex metal seals are all capable of limiting leak rates to much less than 1 x 1 0 ref cm 3/s. After loading, all lid and cover seals are leak tested and the acceptable total

-5 cask leakage (both inner and outer seals combined) is 1 x 1 0 ref cm 3/s with a minimum test sensitivity of 5 x 10-6 atm-cc/sec.

A7A.8.3 CLOSURE The containment vessel contains an integrally-welded bottom closure and a bolted and flanged top closure (lid). The lid plate is attached to the cask body with 48 bolts. The bolt torque required to seal the metal seals located in the lid and maintain confinement under normal and accident conditions is provided in Drawing TN40HT-72-1 in Section A1 .5. The closure bolt analysis is presented in Section A4A.4.

The lid contains two penetrations which are sealed by flanged covers fastened to the lid by 8 bolts each. The bolt torque required to seal the metal seals in the penetration covers and maintain confinement under normal and accident conditions is provided in Drawing TN40HT-72-1 in Section A1 .5. The penetration bolt analysis is presented in Section A4A.5.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-3 A7A.8.4 MONITORING OF SYSTEM CONFINEMENT An overpressure (OP) monitoring system is part of the TN-40HT design. The pressure in the monitoring system is greater than that of the cask cavity and the cask cavity pressure is greater than ambient. In this configuration, neither in-leakage of air nor out-leakage of cavity gas is possible.

If a leak existed in the seals, the design of the TN-40HT overpressure system is such that the leak will either be to the atmosphere or to the cask cavity. Leakage from the cask cavity past the higher pressure of the OP system is physically impossible.

-5 The seals are collectively leak tested to 1 x 1 0 ref cm3/s. Using the methodology of ANSI N 14.5 (Reference 7), an equivalent maximum hole size is estimated based upon test conditions of equivalent air leaking from 1 atm abs to 0.01 atm abs in ambient

-5 temperature conditions (77 °F or 25 °C) and the maximum acceptable leak of 1 x 1 0 ref cm3/s. The leakage hole length is assumed to be the same as the metal seal width, 0.5 cm. The equivalent maximum hole size is calculated below.

L u = (F c + F m )(P u - P d )(P a/P u ) cc/sec at T u , P u Other definitions:

-5 L u = upstream volumetric leakage rate, cc/sec = 1 x 1 0 ref cm 3/s (Test Leak Rate)

F c= coefficient of continuum flow conductance per unit pressure, cc/atm-sec F m= coefficient of free molecular flow conductance per unit pressure, cc/atm-sec P u = fluid upstream pressure, atm abs = 1.0 atm abs P d= fluid downstream pressure, atm abs = 0.01 atm abs D = leakage hole diameter, cm a = leakage hole length, cm = 0.5 cm (assuming leak path length is on the order of the metal seal width) m = fluid viscosity, cP = 0.01 85 cP (from ANSI N14.5, Table B.1)

T = fluid absolute temperature, K = 298 K M = molecular weight, g/mol = 29.0 g/mol (from ANSI N 14.5, Table B.1)

P a= average stream pressure = 1/2 (P u + P d ), atm abs = 0.505 atm abs L u = (F c + F m )(P u - P d )(P a/P u ) cc/sec where:

6 4 F c = (2.49x1 0 x D )/(am) cc/atm-sec 0.5 F m = {3.81x10 3 x D3 x (T/M) } / (aP a ) cc/atm-sec Substituting:

F c = (2.49x106 x D4 )/(0.5 x 0.0185) = 2.69 x 108 D 4 0.5 4 F m = {3.81x10 3 x D3 x (298/29.0) } / {0.5 x 0.505) = 4.84 x 10 D 3 L u = (F c + F m )(P u - P d )(P a/P u ) cc/sec

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-4

-5 1 x 10 = (F c + F m )(1.0 - 0.01 )(0.505 / 1.0)

-5 Fc + F m = 2 x 10

-4 Solving the simultaneous equations, the equivalent hole diameter, D, is 4.825 x 10 cm.

During operations, the OP system is initially back filled with 5.5 atm abs (66.2 psig) of helium at standard temperature.

The temperature of the helium in the OP tank at equilibrium is assumed to be 160 °F (71 °C) 1 . The pressure in the OP system at this temperature will be 6.349 atm abs (79 psig). Assuming the OP system is leaking to the atmosphere, the leak rate is defined using the equations of ANSI N14.5 (Reference 7):

L u,He = (F c + F m )(P u - P d )(P a/P u ) cc/sec 4

F c = (2.49x1 0 6 x D )/(am) cc/atm-sec 0.5 F m = {3.81x10 3 x D3 x (T/M) } / (aP a ) cc/atm-sec Where:

L u,He = helium volumetric leakage rate Pu = 6.349 atm abs Pd = 1.0 atm abs

-4 D = 4.825 x 10 cm a = 0.5 cm m = 0.0219 cP (for helium at 344 K)

T = 344°K M = 4.0 g/mol Pa = 1/2 (P u + P d ) = 3.674 atm abs Substituting:

-4 4 F c = {2.49 x 10 6 x (4.825 x 1 0 ) }/(0.5 x 0.0219) = 1.234 x 1 0

-5 3 -4 3 0.5 F m = {3.81 x 10 x (4.825 x 10 ) x (344/4) } / (0.5 x 3.674) = 2.160 x 10 -6 L u,He = (F c + F m )(P u - P d )(P a/P u )

-5 -6 L u,He = (1.234 x 10 + 2.160 x 10 )(6.349 - 1.0)(3.674/6.349)

-5 L u,He = 4.489 x 1 0 cc/sec of helium 1 The assumed OP gas temperature is the average of the OP Tank and lid seal(s) temperatures.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-5 Over the first year, the maximum volume leaked from the OP system is:

-5 V = 4.489 x 10 cc/sec x (365 days/year x 24 hrs/day x 3600 sec/hr) = 1416 cc at T u ,

Pu The OP system tank consists of a 6 in. diameter schedule 80 pipe (27 in. long) and two 6 in. diameter schedule 80 end caps. The volume of the tank is 835 in3 . The volume of the OP system is increased to 900 in 3 (14,748 cc) to include the OP system tubing and the space between the metallic seals in the lid and penetrations. Correspondingly, the pressure is reduced by the following in the first year:

P OP released = P OP Sys, Initial X {V released / V OP Sys}

P OP released = 6.349 atm (1416cc / 14748cc) = 0.609 atm The overpressure system pressure is also corrected for the corresponding drop in temperature over the first year. At the start of the second year, the overpressure system pressure is:

P OP start of second year = (6.349 atm -0.609 atm )* (339.6°K/344°K) = 5.666 atm abs These calculations are repeated every year for the 25 year life of the cask.

Figure A7A.8-2 illustrates the pressure drop from the OP system to the atmosphere.

Figure A7A.8-2 also illustrates the pressure drop in the cask cavity due to fuel cooling.

If a leak is to the cask cavity rather than the atmosphere, the pressure drop in the OP system is calculated using a downstream pressure the minimum cavity pressure of 1.37 atm abs (5.43 psig). Figure A7A.8-2 also illustrates the results of this analysis. In this scenario, the corresponding increase in the cask cavity pressure is negligible.

As shown above, the monitoring system pressure is greater than the cask cavity or atmospheric pressure assuming a leak based on the conservative initial acceptance test

-5 leak rate of 1 x 1 0 ref cm3/s. Typically, Helicoflex metal seals result in joints with much lower leak rates than the acceptance criteria. Therefore, no leakage will occur from the cask cavity during the storage period.

The pressure in the overpressure system will be monitored over the lifetime of the cask.

To allow time to diagnose and correct any problems, the OP monitoring system is set to alarm if the overpressure system drops below 2.89 atm abs (27.8 psig). This set point is based on the maximum off-normal cask cavity pressure (32.2 psia from Table A3.3-16) plus 0.7 atm for margin. It ensures that pressure decreases in the overpressure monitoring system are identified well before any potential out leakage from the cask cavity occurs.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-6 A7A.8.5 CONFINEMENT REQUIREMENTS FOR NORMAL CONDITIONS OF STORAGE The TN-40HT dry storage cask is designed to provide storage of spent fuel for at least 25 years. The cask cavity pressure is always above ambient during the storage period as a precaution against the in-leakage of air which might be harmful to the fuel. Since the containment vessel consists of a steel cylinder with an integrally-welded bottom closure, the cavity gas can escape only through the lid closure system. In order to ensure no release of radioactivity material, two systems are employed. First, all bolted closures are provided with double seals. Second, the interspace between the seals is pressurized to provide a positive pressure gradient. If the inner seals were to leak, helium would flow into the cask cavity and radioactive material would not be released. If the outer seals were to leak, helium would leak from the overpressure system to the exterior, and no radioactive material would be released.

The cask loadings for normal conditions of storage are given in Section A3.2.5. It is shown that the seals are not disturbed by any of the loadings and thus, the cask confinement is maintained.

A7A.8.6 CONFINEMENT REQUIREMENTS FOR HYPOTHETICAL ACCIDENT CONDITIONS A7A.8.6.1 SOURCE TERMS FOR CONFINEMENT CALCULATIONS Table A7.2-6 lists the activity representing the fission gases, volatiles, and fines contributing more than 0.1% of the activity contained in a design basis fuel, plus Iodine 129.

The releasable source term is first determined. The release fractions (References 8 and 9) applied to the source term are provided below.

Off-Normal Accident Variable Conditions Conditions Fraction of crud that spalls off rods, f C0.15 1.0 Fraction of Rods that develop cladding breaches, fB 0.10 1.0 Fraction of Gases that are released due to a cladding breach, f G 0.3 0.3

-6* -6*

Fraction of Fines that are released due to a cladding breach, fF 3 x 10 3 x 10

-4 -4 Fraction of Volatiles that are released due to a cladding breach, fV2 x 10 2 x 10

-5

• Per NUREG-1617, 3 x 10 of the fines are released during a cladding breach. Per

-5 SAND90-2406, (Reference 10), page IV-7, of the 3 x 10 of the fines released recommends that only 10% of the fuel fines ejected remain airborne.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-7 The releasable source term also depends on the leak rate from the TN-40HT. Under off-normal conditions, it is assumed that the OP system is not functioning properly. In

-5 this case, the cask cavity gas is assumed to leak out at a rate of 1 x 10 std cc /sec.

Assuming the cask cavity gas acts like helium (including the gases, volatiles, fines and crud), the leak rate is adjusted to a helium leak rate at cask cavity conditions using the equations of ANSI N14.5 (Reference 7). This calculation is shown below.

P u = 2.42 atm abs , 20.9 psig (off-normal cask cavity pressure assuming 10% of the fuel rods have failed - Selected to be conservatively higher than value in Table A3.3-16)

P d = 1.0 atm abs

-4 D = 4.825 x 10 cm a = 0.5 cm m = 0.0283 cP (for helium at 500K)

T = fluid absolute temp = average cavity gas temp = 456 °F = 508 K M = 4.0 P a = 1/2 (P u + P d ) = 1.708 atm abs Substituting:

F c = 9.530E-06 F m = 5.645E-06 L u,he = 1.51 8E-05 cc/sec of helium for off-normal conditions Similarly, under hypothetical accident conditions, it is assumed that the OP system has stopped functioning and fire conditions exist.

P u = 6.34 atm abs, 78.5 psig (cask cavity pressure following hypothetical fire and assuming 100% fuel rod failure - Selected to be conservatively higher than value in Table A3.3-16)

P d = 1.0 atm abs

-4 D = 4.825 x 10 cm a = 0.5 cm m = 0.0292 cP (for helium at 575 K)

T = fluid absolute temp = average cavity gas temp following fire = 592 °F = 584 K M = 4.0 P a = 1/2 (P u + P d ) = 3.67 atm abs Substituting in to the equations of ANSI N14.5:

F c= 9.236E-06 F m= 2.816E-06 L u,he = 3.725E-05 cc/sec of helium for hypothetical accident conditions.

The releasable contents from the TN-40HT during off-normal and hypothetical accident conditions are provided in Tables A7A.8-1 and A7A.8-2, respectively.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-8 A7A.8.6.2 RELEASE OF CONTENTS Two scenarios are considered:

Off-Normal Conditions - This condition exists over a 45 day period, seals are

-5 leaking at the test leak rate of 1 x 1 0 ref cm3/s and the fraction of rods that have failed is 10%. Stability category D and a 5 m/s wind speed are used for this analysis. This scenario assumes one cask is in off-normal condition at the ISFSI.

The 45 day exposure duration will serve as the bases for the allowed completion times for the Cask Interseal Pressure Technical Specification.

Hypothetical Accident Conditions - This condition exists over a 30 day period,

-5 seals are leaking at the test leak rate of 1 x 1 0 ref cm 3/sec, the fraction of rods that have failed is 100%, and the temperature inside the cask is comparable to the fire accident conditions. Stability category F and 1 m/s wind speed is used for this analysis. This scenario assumes one cask is in the hypothetical accident condition at the ISFSI.

In the first scenario, the release is assumed to occur for more than a 20 minute period.

The methodology of Reg Guide 1.145 (Reference 11) is applied. The atmospheric diffusion from a ground level point source at 110 meters is based on the following parameters.

Wind speed = 5 meter/second sy = 9 meters (Reference 11, Figure 1) sz = 5 meters (Reference 11, Figure 2)

M = 1.1, (Reference 11, Figure 3]

Sy = M s y = 9.9 meters 2

A = cross sectional area of the TN-40HT = 9.1 m Using the methodology of Reg Guide 1.145, { c/Q} 110 meters during off-normal conditions is 1 .29E-03 sec/m 3 . Similarly, the atmospheric diffusion for 700 meters which corresponds to the distance from the ISFSI to the nearest residence (700 meters is a conservative value, 724 meters is the distance from the center of the ISFSI to the nearest resident),

during off-normal conditions is calculated using the following parameters.

Wind speed = 5 meter/second sy = 50 meters sz = 25 meters M = 1.1 Sy = M s y= 55 meters During off normal conditions { c/Q} 700 meters is 4.63E-05 sec/m 3

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-9 In the second scenario the release is assumed to be a short term ground level release (occurring, however, over a 30 day period) assuming the methodology of Regulatory Guide 1.25 (Reference 12). The atmospheric stability classification of F and a wind speed of 1 m/sec are used. The atmospheric diffusion from a ground level point source at 110 and 724 meters is taken from Table 8.2-1 and Table 2.3-2 to be:

c /Q 110 meters = 6.63E-03 sec/m 3 and 3

c /Q 0.45miles = 2.66E-04 sec/m (nearest resident is 0.45 miles, about 724 meters from center of ISFSI)

A7A.8.6.2.1 DOSE CALCULATIONS Dose components are calculated following the method of Regulatory Guide 1.109 (Reference 13) and utilizing dose conversion factors from EPA Federal Guidance Reports Numbers 11 and 12 (References 14 and 15). (Note: Two sets of dose conversion factors (DCFs) depending upon the chemical state of Sr-90 are reported in Federal Guidance Report Number 11. One set of DCF values is for Sr in the form of SrTiO 3 and the other set is for Sr in all other forms. The Sr-90 fission product should not form SrTiO 3 within the storage cask and therefore the DCF for this compound was not used.)

To determine the committed doses (from air inhalation), the following equation is used:

Dose inhalation = R x c/Q x Q x DCF inhalation x Time Where:

R = Inhalation Rate = 8,000 m 3/year = 2.54E-04 m 3/sec (References 8 and13) c/Q = Short term average centerline value of atmospheric diffusion for a ground level release (sec/m 3 )

Q = amount of material released (m Ci/sec)

DCF inhalation = Exposure Dose Conversion Factor (mrem/ m Ci), from (Reference 14).

Time = Time of Exposure (Seconds)

To determine the deep doses (from air immersion), the following equation is used:

Dose air immersion = {c/Q x Q x DCF air immersion} x Time Where:

c / Q = Short term average centerline value of atmospheric diffusion for a ground level release (sec/m 3 )

Q = amount of material released (m Ci/sec) 3 DCF immersion = Exposure Dose Conversion Factor (mrem/year per m Ci/cm ),

(Reference 15)

Time = Time of Exposure (Seconds)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-10 For off-normal conditions, the estimated 45 day airborne doses (internal and external) at 110 meters from a single TN-40HT cask are provided in Table A7A.8-3 and Table A7A.8-4. The sum of the deep dose (external) and the committed dose (internal) on an organ basis and total effective dose for distances of 110, and 700 meters are summarized below:

Dose at 110 Dose at 700 meters meters Target (mrem) (mrem)

Gonad 4.69E-02 1 .69E-03 Breast 1 .48E-02 5.34E-04 Lung 3.85E-01 1.39E-02 Red Marrow 3.32E-01 1 .20E-02 Bone Surface 2.90E+00 1 .04E-01 Thyroid 1 .43E-02 5.1 5E-04 Remainder 1 .26E-01 4.52E-03 Effective 2.95E-01 1 .06E-02 Skin 1 .73E-03 6.24E-05 The values presented in bold print above demonstrate that the criteria of 10CFR-72.104 are met under off-normal conditions.

For hypothetical accident conditions, the committed doses (internal) and the deep doses (external) at 110 meters from a single TN-40HT cask for a 30 day exposure are provided in Table A7A.8-5 and Table A7A.8-6. The total effective dose equivalent at 110 m and at 724 m from the TN-40HT cask is summarized in the table below.

Dose at 110 Dose at 724 meters meters Target (mrem/yr) (mrem/yr)

Gonad 3.88E+00 1.56E-01 Breast 9.79E-01 3.93E-02 Lung 2.75E+01 1.10E+00 Red Marrow 2.78E+01 1.11E+00 Bone Surface 2.44E+02 9.79E+00 Thyroid 9.66E-01 3.88E-02 Remainder 1.01 E+01 4.04E-01 Effective (TEDE) 2.40E+01 9.63E-01 Skin 1.38E-01 5.54E-03 Lens Dose (Skin + TEDE) 2.41 E+01 9.68E-01

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-1 1 The maximum 30-day TEDE value is 0.024 rem. The corresponding 10CFR 72.106 limit is 5 rem.

The maximum 30-day Lens Dose Equivalent value is 0.0241 rem. The corresponding 10CFR 72.106 limit is 15 rem.

The maximum 30-day dose to any organ/tissue is 0.244 rem and it occurs at the bone surface. The corresponding 10CFR 72.106 limit is 50 rem.

Therefore all the criteria of 72.106 are met at 110 m.

A summary of the doses at 110 m and their corresponding regulatory limits is shown below.

Off-Normal Conditions 10CFR72.104(a) 110 meter Organ Limit (mrem) Dose (mrem)

Whole Body (TEDE) 25 0.30 Thyroid 75 0.01 2.90 Other Critical Organ 25 (Bone Surface)

Accident Conditions ________________

10CFR72.106(b) 110 meter Organ Limit (mrem) Dose (mrem)

Whole Body (TEDE) 5000 24.0 244 Organ (TODE) 50000 (Bone Surface)

Lens of Eye (LDE) 15000 24.1 Skin (SDE) 50000 0.14 A7A.8.6.3 LATENT SEAL FAILURE By design the overpressure monitoring system does not immediately alarm if there is a leak in a seal or the overpressure system. The time period from when a leak begins to occur and when the overpressure system alarm is activated is dependent on the size of the leak. Two conditions which could exist within the TN-40HT confinement system are:

(1) The outer seal (or the overpressure system) is leaking to the atmosphere. In this case the inner seal is intact and there is no release of the contents of the cask cavity to the atmosphere.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-12 (2) The inner seal is leaking (or the overpressure system is leaking into the cask cavity). In this case the outer seal is still intact and there is no release of the cask cavity contents to the atmosphere.

If a latent seal leak has occurred, the tables below provide some examples of the time to alarm based on assumed leakage rates.

Case 1 - Leakage of Overpressure System to the Atmosphere Estimated Time to Alarm Estimated Time to Loss of OP Leak Rate (from Start of Latent Seal System Pressure (from Start of 3

(ref cm /s) Failure) Latent Seal Failure)

-3 1 x 10 19 days 34 days

-4 1 x 10 195 days 358 days

-4 5 x 10 1.2 year 6 years

-5 1x10 (see Figure A7A.8-2) 9.5 years over 40 years Case 2 - Leakage of Overpressure System to Cask Cavity Estimated Time to Equalize OP System Pressure with Cask Estimated Time to Alarm Cavity Pressure Leak Rate (from Start of Latent Seal (from Start of Latent Seal 3

(ref cm /s) Failure) Failure)

-3 1 x 10 19 days 28 days

-4 1 x 10 201 days 288 days

-4 5 x 10 1.4 years 13 years

-5 1 x 10 (see Figure A7A.8-2) 9.5 years over 40 years As shown in the tables above, the alarm is set such that for any credible leak, there is time to evaluate the leaking condition and correct the condition provided that the overpressure system remains pressurized. This period can be extended by re-pressurizing the overpressure tank.

Another condition which has been considered is that a latent seal failure has occurred and the overpressure system is removed due to an accident.

(1) If the outer seal has the latent failure and the OP system is removed then there is no release of cask cavity contents to the atmosphere.

(2) If the inner seal has a latent failure and the OP system is removed then the table below provides the time before 10 CFR 72.106(b) limits will be exceeded.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-13 Standard Leak Rate Time to exceed (ref cm 3/sec) 10 CFR 72.1 06(b) Limits

-3 1 x 10 59 days

-4 1 x 10 602 days

-5 5 x 10 1213 days 5

1 x 10 6145 days The times above demonstrate that a latent failure up to 100 times greater than the test value could occur and still allow recovery.

The time to reach the accident release limits is dependent on the size of the leak. Due to the reliability of the metal seals used in static applications, it is not considered credible that the inner seals could leak at a rate significantly higher than the test leak rate. The probability that a gross leak of an inner seal in combination with a gross leak in an outer seal or the overpressure system, such that the overpressure system could not hold pressure, is not considered a credible event.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.8-14 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.9-1 A7A.9 REFERENCES MCNP/MCNPX - Monte Carlo N-Particle Transport Code System Including MCNP5 1.40 and MCNPX 2.5.0 and Data Libraries, CCC-730, Oak Ridge National Laboratory, RSICC Computer Code Collection, January 2006.

2. ANSI/ANS 6.1-1-1 977, "Neutron and Gamma Ray Flux-to-Dose Rate Factors."

3. Title 10 Code of Federal Regulations Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste.

4. Title 10 Code of Federal Regulations Part 20, Standards for Protection Against Radiation.

5. Title 40 Code of Federal Regulations Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations.

6. Helicoflex Catalogue ET 507 E5930.

7. ANSI N14.5-1 997, Leakage Tests on Packages for Shipment, February 1998.

8. USNRC, Spent Fuel Project Office, Interim Staff Guidance No. 5, Revision 1.

9. NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, Draft Report for Comment, US Nuclear Regulatory Commission, March 1998.

10. SAND90-2406, A Method for Determining the Spent Fuel Contribution to Transport Cask Containment Requirements, Sandia National Laboratories, November 1992.

11. USNRC Regulatory Guide 1 .145,"Atmospheric Dispersion Models for Potential Accident Consequence Assessment at Nuclear Power Plants," Rev 1, 1983.

12. USNRC Regulatory Guide 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling Storage Facility for Boiling and Pressurized Water Reactors.

13. USNRC Regulatory Guide 1.109, "Calculation of Annual Doses to Men from Routing Releases of Reactor Effluent for the Purpose of Evaluating Compliance with 10CFR50, Appendix I," Rev 1, 1977.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7A.9-2

14. USEPA Federal Guidance Report No 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion," EPA-520/1 0202, September 1988.

15. USEPA Federal Guidance Report No 12, "External Exposure to Radionuclides in Air, Water, and Soil," EPA-402-R-93-081, September, 1983.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.1-1 TN-40HT CASK SHIELD MATERIALS Density Thickness Com ponent Material 3 (g/cm ) (inches)

Cask Body Wall Carbon Steel 7.82 8.75 Lid Carbon Steel 7.82 10.00 Bottom Carbon Steel 7.82 8.75 Polyester Resin Styrene (1)

Resin 1.58 5.00 Aluminum Hydrate Zinc Borate Aluminum Box Aluminum 2.7 0.12 Outer Shell Carbon Steel 7.82 0.50 Stainless Steel 7.94 0.19 (Inserts)

Stainless Steel 794 044 Basket (2)

(Strips)

Aluminum Neutron Poison 2.7 0.44 (3)

Material Protective Cover Carbon Steel 7.82 0.38 Polypropylene Disc Polypropylene 0.90 4.00 Notes:

(1) The neutron shielding is borated polyester resin compound with a density of 1.58 g/cc. The four major constituents are listed in the table.

(2) The stainless steel inserts of the basket are 0.19 inches thick. The intermittent stainless steel strips are not modeled.

(3) The neutron poison material is modeled as aluminum for shielding purposes.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.2-1

SUMMARY

OF AVERAGE DOSE RATES Average Dose Rates (mrem/hr) for Normal and Off-Normal Conditions Side Side Side Top Side Bottom 1 Meter 2 Meter Above Along Below Surface3 Surface 7 Surface4 from from Shield 1 Shield 5 Shield 2 (Protective Side Side

_______ ______ _______ _______ Cover) _______ ________ ______ ______

Gamma 61.3 32.7 52.9 18.5 35.5 272 12.8 7.61 Neutron 146 28.2 706 7.6 78.7 1515 18.9 10.6 Total 208 60.9 759 26.1 114 1790 31.7 18.2

______ Average Dose Rates (mrem/hr) for Accident Conditions 6 Top Side Bottom 1 Meter 2 Meter Surface8 Surface 7 Surface from from (Protective Side Side

_______ ______ _______ _______ Cover) _______ ________ ______ ______

Gamma - - - 63.4 116 - 50.6 31.0 Neutron - - - 135 1980 - 785 444 Total - - - 198 2095 - 835 475 Notes:

(1) Maximum surface dose rates above the neutron shield are 81.5 mrem/hr gamma and 189 mrem/hr neutron.

(2) Maximum surface dose rates below the neutron shield are 86.5 mrem/hr gamma and 928 mrem/hr neutron.

(3) Maximum surface dose rates on the protective cover are 45.5 mrem/hr gamma and 12.7 mrem/hr neutron.

(4) Maximum surface dose rates on the cask bottom are 804 mrem/hr gamma and 3980 mrem/hr neutron.

(5) Maximum surface dose rates on the cask side are 39.3 mrem/hr gamma and 34.8 mrem/hr neutron (6) The radial neutron shield and outer shell are removed from the sides. The polypropylene disk and protective cover are removed from the top.

(7) Includes surfaces above and below the radial neutron shield.

(8) Top surface dose rate is on the Top Surface of the Cask Lid Note: Maximum gamma and neutron dose rates may not occur at the same location.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.2-2 DOSE RATES AT A DISTANCE FROM ONE TN-40HT CASK Total Dose Rate Results Distance Gamma Neutron (meter) (mrem/hr) Error (mrem/hr) Error 10 1.02E+00 0.0048 2.01E+00 0.0047 20 2.68E-01 0.0064 5.43E-01 0.0056 25 1.69E-01 0.0050 3.51E-01 0.0058 30 1.16E-01 0.0051 2.44E-01 0.0061 40 6.30E-02 0.0054 1.35E-01 0.0065 50 3.86E-02 0.0056 8.35E-02 0.0069 75 1.53E-02 0.0062 3.35E-02 0.0081 100 7.59E-03 0.0069 1.67E-02 0.0090 150 2.62E-03 0.0076 5.50E-03 0.0112 175 1.68E-03 0.0081 3.48E-03 0.0141 250 5.35E-04 0.0086 1.00E-03 0.0155 300 2.74E-04 0.0097 4.94E-04 0.0161 350 1.52E-04 0.0123 2.52E-04 0.0193 400 8.37E-05 0.0133 1.36E-04 0.0226 500 2.89E-05 0.0147 4.44E-05 0.0247 600 1.07E-05 0.0161 1.52E-05 0.0253 800 1.76E-06 0.0220 2.30E-06 0.0296 1,100 1.52E-07 0.0427 1.77E-07 0.0348 1,500 7.30E-09 0.1118 8.75E-09 0.0267 2,000 2.76E-10 0.2795 3.10E-10 0.0239

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.4-1 ANSI STANDARD-6.1.1-1977 FLUX-TO-DOSE FACTORS Photon energy Response Neutron Response (MeV) Function energy Function 2

(rem/hr)/(y/cm2-s) (MeV) (rem/h r)/(n/cm

-s) 0.01 3.96E-06 2.5E-08 3.67E-06 0.03 5.82E-07 1 .0E-07 3.67E-06 0.05 2.90E-07 1 .0E-06 4.46E-06 0.07 2.58E-07 1 .0E-05 4.54E-06 0.10 2.83E-07 1.0E-04 4.18E-06 0.15 3.79E-07 1.0E-03 3.76E-06 0.20 5.01E-07 1.0E-02 3.56E-06 0.25 6.31E-07 1.0E-01 2.17E-05 0.30 7.59E-07 5.0E-01 9.26E-05 0.35 8.78E-07 1.0 1.32E-04 0.40 9.85E-07 2.5 1.25E-04 0.45 1.08E-06 5.0 1.56E-04 0.50 1.17E-06 7.0 1.47E-04 0.55 1.27E-06 10.0 1.47E-04 0.60 1.36E-06 14.0 2.08E-04 0.65 1.44E-06 20.0 2.27E-04 0.70 1.52E-06 0.80 1.68E-06 1.0 1.98E-06 1.4 2.51E-06 1.8 2.99E-06 2.2 3.42E-06 2.6 3.82E-06 2.8 4.01 E-06 3.25 4.41 E-06 3.75 4.83E-06 4.25 5.23E-06 4.75 5.60E-06 5.0 5.80E-06 5.25 6.01 E-06 5.75 6.37E-06 6.25 6.74E-06 6.75 7.11E-06 7.5 7.66E-06 9.0 8.77E-06 11.0 1.03E-05 13.0 1.18E-05 15 .0 1.33E-05 ______________________________

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.4-2 ATOMIC FRACTIONS FOR THE HOMOGENIZED DESIGN BASIS 14X14 FUEL Atomic Nuclide ID Element Fraction 3

Active Fuel Region, Density = 3.949 g/cm 8016 Oxygen 0.5453 92235 Uranium 0.0094 92238 Uranium 0.2631 40000 Zirconium 0.1420 24000 Chromium 0.0072 26000 Iron 0.0165 28000 Nickel 0.0125 Bottom Nozzle Region, Density = 2.595 g/cm 3 14000 Silicon 0.0194 24000 Chromium 0.1995 25055 Manganese 0.01 99 26000 Iron 0.6684 28000 Nickel 0.0884 Plenum Region, Density = 1.543 g/cm 3 40000 Zirconium 0.2640 14000 Silicon 0.0164 24000 Chromium 0.1436 25055 Manganese 0.0131 26000 Iron 0.4465 28000 Nickel 0.1084 Top Nozzle Region, Density = 1.970 g/cm 3 14000 Silicon 0.0216 24000 Chromium 0.1966 25055 Manganese 0.01 84 26000 Iron 0.6245 28000 Nickel 0.1326

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.4-3 SHIELD REGIONAL DENSITIES FOR TN-40HT MCNP MODEL (PAGE 1 OF 2)

Carbon Stainless (1)

_________________ Steel Steel 304(1)

(2) (1)

Air Aluminum gram density (g/cc) 7.8212 7.94 1.23E-03 2.702 atom density (atom/barn-cm) 0.08742 0.08759 0.00051 0.06031 Element Z M wt. % at. % wt. % at. % wt. % at. % wt. % at. %

H 1 1.0 ____ _____ _____ _____ ____ ____ ____ ____

C 6 12.0 1 4.486 0.08 0.364 0.014 0.017 _____ _____

N 7 14.0 75.52 79.44 _____ _____

O 8 16.0 23.18 21.08 _____ _____

Al 13 27.0 100 100 Si 14 28.1 ____ 1 1.944 _____ _____ ____ ____

P 15 31.0 0.045 0.079 _____ _____ _____ _____

Ar 18 39.9 1.29 0.47 _____ _____

Cr 24 52.0 19 19.949 _____ _____ _____ _____

Ti 22 47.9 _____ ______ ______ ______ _____ _____ _____ _____

Mn 25 54.9 2 1.987 _____ _____ _____ ____

Fe 26 55.8 99 95.514 68.375 66.841 ______ ______ _____ _____

Ni 28 58.7 9.5 8.836 ______ ______ _____ _____

Zr 40 91.2 ____ _____ _____ _____ _____ _____ ____ ____

Sn 50 118.7 ____ _____ _____ _____ _____ _____ ____ ____

Hf 72 178.5 _____ ______ ______ ______ _____ _____ _____ _____

Total 100 100 100 100 100 100 100 100 Notes:

(1) - SCALE Standard Material Composition (2) - ANSI/ANS-6.6. 1, Dry Air

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.4-3 SHIELD REGIONAL DENSITIES FOR TN-40HT MCNP MODEL (Page 2 of 2)

(2)

(1) (3)

Concrete Polyester Resin Polypropylene Soil gram density (g/cc) 2.32 1.58 0.9 1.625 atom density (atom/barn-cm) 0.074447 0.106815 0.115918 0.068588 Element Z M wt. % at. % wt. % at. % wt. % at. % wt. % at. %

H 1 1.0 0.56 10.427 5.05 44.631 14.37 66.667 2.10 29.73 C 6 12.0 35.13 26.054 85.63 33.333 1.60 1.90 Na 11 23.0 1.71 1.396 _____ ______ _____ _______ _____ _____

O 8 16.0 49.83 58.449 41.73 23.234 _____ _______ 57.70 51.46 Mg 12 24.3 0.24 0.185 ______ ______ ______ ________ _____ _____

Al 13 27.0 4.56 3.172 14.93 4.929 ______ ________ 5.00 2.64 Si 14 28.1 31.58 21.102 _____ ______ _____ _______ 27.10 12.06 S 16 32.1 0.12 0.070 _____ ______ _____ _______ _____ _____

K 19 39.1 1.92 0.922 _____ ______ _____ _______ 1.30 0.47 Ca 20 40.1 8.26 3.868 _____ ______ _____ _______ 4.10 1.46 Fe 26 55.8 1.22 0.410 _____ ______ _____ _______ 1.10 0.28 B 5 10.8 1.05 0.865 _____ _______ _____ _____

Zn 30 65.4 2.11 0.288 _____ _______ _____ _____

________ Total 100 100 100 100 100 100 100 100 Notes:

(1) - ANSI/ANS-6.4.3 (2) - Chemestry & Physics Handbook (3) - Jacob, Radiation Protection Dosimetry 14, 299, 1986

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.5-1 TN-40HT NEAR FIELD DOSE RATES - NORMAL AND OFF-NORMAL CONDITIONS Axial Cask Side Surface Location (mrem/hr) _______

(cm) gamma neutron total

-205.3 19.4 484 503

-189.8 86.5 928 1014

-176.0 55.9 116 172

-164.0 72.1 43.2 115

-143.1 29.7 31.5 61.2

-113.3 33.5 34.1 67.6

-83.6 34.6 34.8 69.5

-53.8 35.6 31.0 66.6

-24.0 32.6 28.5 61.1 5.8 31.3 27.2 58.4 35.6 39.3 29.1 68.4 65.4 33.8 25.8 59.6 95.1 31.0 23.6 54.6 124.9 25.4 16.7 42.1 148.8 23.7 13.3 36.9 166.3 17.5 25.0 42.4 183.8 19.9 25.4 45.3 201.3 22.1 17.6 39.7 216.5 81.5 189 271 231.5 41.1 104 145 Average 35.5 78.7 114 Top Neutron Radial Protective Cover Cask Bottom ie l Location (mrem/hr) (mrem/hr)

(mrem/hr) ____ ______ ______

(cm) g n total g n total g n total 5.7 45.3 2.16 47.5 76.8 9.07 85.9 770 3979 4749 17.1 45.5 2.97 48.5 81.5 6.17 87.7 797 3905 4702 28.5 43.3 2.87 46.2 81.7 5.08 86.8 804 3670 4474 39.9 41.0 2.04 43.1 79.0 4.84 83.8 781 3446 4227 51.3 38.0 2.25 40.3 76.5 5.83 82.3 784 3119 3903 62.7 31.5 2.12 33.6 66.2 5.35 71.6 713 2686 3399 74.1 25.2 2.82 28.1 50.2 5.02 55.2 530 2261 2791 85.5 18.8 5.22 24.0 31.2 7.48 38.7 272 1724 1996 96.9 13.4 9.80 23.2 19.0 38.9 57.9 95.8 1104 1200 108.3 9.8 12.7 22.5 11.3 41.9 53.1 30.5 545 575 Average 18.5 7.6 26.1 31.7 24.1 55.7 272 1515 1790

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.5-2 TN-40HT NEAR FIELD DOSE RATES - NORMAL AND OFF-NORMAL CONDITIONS 1 M from cask side surface 2 M from cask side surface Axial _______ (mrem/hr) ______ _______ (mrem/hr) ______

Location (cm) gamma neutron total error* gamma neutron total error*

-200.3 12.1 52.6 64.7 0.012 6.52 18.6 25.1 0.013

-175.0 12.6 46.8 59.4 0.012 7.30 18.8 26.1 0.013

-149.6 13.0 34.5 47.5 0.012 7.64 17.2 24.8 0.013

-124.3 13.7 26.6 40.3 0.014 7.98 15.7 23.6 0.013

-98.9 14.7 22.3 37.0 0.014 8.53 14.2 22.8 0.013

-73.6 15.4 19.1 34.4 0.013 8.87 12.8 21.7 0.013

-48.2 15.7 17.1 32.8 0.014 9.06 11.3 20.4 0.013

-22.9 15.7 16.4 32.2 0.014 9.17 11.0 20.2 0.014 2.5 15.7 14.6 30.3 0.014 9.16 10.0 19.2 0.013 27.8 15.7 13.7 29.4 0.015 9.15 9.2 18.4 0.014 53.2 15.2 12.4 27.6 0.014 8.92 8.8 17.7 0.014 78.5 14.4 11.5 25.9 0.018 8.53 8.3 16.8 0.014 103.9 13.3 11.8 25.1 0.024 7.98 8.1 16.1 0.015 129.2 11.9 10.4 22.3 0.016 7.59 7.7 15.3 0.015 154.6 10.8 9.6 20.4 0.016 7.00 7.4 14.4 0.015 179.9 10.1 10.3 20.4 0.016 6.54 6.9 13.4 0.015 205.3 9.9 11.8 21.8 0.016 6.18 7.0 13.2 0.016 230.6 10.1 12.9 23.0 0.020 5.83 6.8 12.6 0.017 256.0 8.9 12.4 21.3 0.017 5.43 6.6 12.0 0.021 281.3 6.5 12.0 18.5 0.020 4.84 6.4 11.2 0.018 Average 12.8 18.9 31.7 7.61 10.6 18.3 _____

Radial 1 M from Protective Cover 2 M from Protective Cover Location (mrem/hr) (mrem/hr) ______

(cm) gamma neutron total error* gamma neutron total error*

5.7 17.9 1.91 19.8 0.046 8.99 0.52 9.51 0.060 17.1 17.4 1.87 19.2 0.030 8.59 1.40 10.0 0.050 28.5 17.2 1.37 18.6 0.024 8.69 1.19 9.88 0.038 39.9 16.4 1.57 18.0 0.022 8.34 1.28 9.61 0.032 51.3 15.0 1.94 17.0 0.022 7.90 1.34 9.24 0.029 62.7 13.6 1.45 15.0 0.019 7.34 1.59 8.93 0.033 74.1 12.6 2.27 14.9 0.022 7.07 1.20 8.27 0.028 85.5 10.9 2.71 13.6 0.022 6.44 1.60 8.04 0.028 96.9 9.5 3.36 12.9 0.025 6.00 1.78 7.78 0.028 108.3 8.4 3.34 11.7 0.025 5.77 1.98 7.75 0.030 Average 10.7 2.79 13.5 _____ 6.44 1.67 8.11 _____

• - fractional standard deviation

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.5-3 TN-40HT NEAR FIELD DOSE RATES - NORMAL AND OFF-NORMAL CONDITIONS (DOSE RATES ABOVE AND BELOW THE RADIAL NEUTRON SHIELD)

Axial Surface Dose (mrem/hr)

Location (cm) gamma neutron total error* _______________

216.5 81.5 189 271 0.014 above N shield 231 .5 41.1 104 145 0.017 ______________

-205.3 19.4 484 503 0.018 below N shield

-189.8 86.5 928 1014 0.012 _______________

• - fractional standard deviation

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.5-4 TN-40HT NEAR FIELD DOSE RATES - ACCIDENT CONDITIONS (PAGE 1 OF 2)

Axial Cask Surface 1 M from Cask Surface 2 M from Cask Location (mrem/hr) (mrem/hr) Surface (mrem/hr)

(cm) g n total g n total g n total

-200.3 36.4 650 687 33.9 642 676 21.6 384 406

-175.0 125 1396 1522 45.0 743 788 26.5 433 459

-149.6 137 2243 2381 54.7 863 917 30.5 479 509

-124.3 150 2919 3069 60.9 975 1036 34.1 522 556

-98.9 152 3276 3428 64.1 1077 1141 36.5 551 588

-73.6 152 3404 3556 65.6 1137 1202 38.2 573 611

-48.2 150 3347 3497 66.0 1178 1244 39.5 594 633

-22.9 135 3236 3372 65.8 1193 1259 40.4 597 637 2.5 110 3219 3319 66.7 1182 1249 40.4 590 630 27.8 188 3403 3592 67.3 1143 1210 40.2 581 621 53.2 162 3295 3457 67.0 1073 1140 39.0 552 591 78.5 146 2933 3079 64.7 974 1039 37.3 516 553 103.9 141 2417 2558 60.0 854 914 35.5 476 512 129.2 120 1741 1861 53.6 718 771 32.3 428 460 154.6 89.3 1062 1151 47.3 581 628 28.9 378 407 179.9 101 553 655 40.4 450 490 25.2 329 355 205.3 114 254 367 33.4 340 374 21.9 283 305 230.6 48.6 109 158 25.2 248 274 18.0 242 260 256.0 33.5 60.5 93.9 17.9 183 201 14.7 203 217 281.3 18.4 61.7 80.1 12.4 139 152 11.7 167 179 Average 116 1980 2095 50.6 785 835 30.6 444 475

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.5-4 TN-40HT NEAR FIELD DOSE RATES - ACCIDENT CONDITIONS (PAGE 2 OF 2) 1 Radial Lid Surface 1 M from Lid 2 M from Lid Location (mrem/h r) (m rem/h r) (m rem/h r)

(cm) _____

g n total g n total g n total 5.7 179 334 513 89.8 82.2 172 44.0 27.9 71.9 17.1 175 318 493 89.3 82.8 172 42.7 36.1 78.8 28.5 172 305 477 88.0 85.3 173 41.4 33.9 75.3 39.9 173 287 461 79.9 77.0 157 39.8 30.2 70.0 51.3 168 260 428 74.3 79.2 153 37.7 32.9 70.7 62.7 155 228 383 65.1 70.8 136 35.0 30.8 65.8 74.1 112 200 312 55.9 64.4 120 32.3 32.5 64.8 85.5 57.6 155 212 46.6 59.1 106 29.4 31.4 60.8 96.9 18.9 94.4 113 37.5 54.0 91.5 26.3 31.2 57.6 108.3 24.3 59.9 84.7 29.8 51.6 81.5 22.9 28.2 51.2 Average 63.4 135 198 44.9 59.0 104 28.4 30.6 59.1 Notes:

1. - Lid surface dose same as in normal conditions

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.6-1 TN-40HT - SINGLE CASK FARFIELD DOSE RATES

_________ Skyshine Dose Rate Results Direct Dose Rate Results Distance Gamma Neutron Distance Gamma Neutron (meter) (mrem/hr) Error* (mrem/hr) Error* (meter) (mrem/yr) (mrem/yr) 10 1.17E-02 0.0072 1.29E-01 0.0052 10 1.01E+00 1.88E+00 20 9.35E-03 0.0057 9.30E-02 0.0041 20 2.59E-01 4.50E-01 25 8.36E-03 0.0054 7.91E-02 0.0041 25 1.61E-01 2.72E-01 30 7.52E-03 0.0055 6.73E-02 0.0040 30 1.09E-01 1.77E-01 40 6.09E-03 0.0053 4.89E-02 0.0040 40 5.69E-02 8.61 E-02 50 4.99E-03 0.0052 3.68E-02 0.0053 50 3.36E-02 4.68E-02 75 3.19E-03 0.0057 1.93E-02 0.0079 75 1.21E-02 1.42E-02 100 2.10E-03 0.0058 1.10E-02 0.0085 100 5.49E-03 5.72E-03 150 1.02E-03 0.0065 4.19E-03 0.0110 150 1.61E-03 1.32E-03 175 7.19E-04 0.0073 2.79E-03 0.01 50 175 9.56E-04 6.89E-04 250 2.80E-04 0.0081 8.68E-04 0.0154 250 2.55E-04 1.33E-04 300 1.57E-04 0.0102 4.38E-04 0.0148 300 1.17E-04 5.58E-05 350 9.40E-05 0.0137 2.29E-04 0.0181 350 5.84E-05 2.21E-05 400 5.39E-05 0.0147 1.27E-04 0.0217 400 2.98E-05 9.47E-06 500 1.99E-05 0.0117 4.20E-05 0.0220 500 8.98E-06 2.41E-06 600 7.72E-06 0.0129 1.42E-05 0.0230 600 2.97E-06 1.02E-06 800 1.35E-06 0.0130 2.19E-06 0.0261 800 4.07E-07 1.13E-07 1,100 1.20E-07 0.0191 1.73E-07 0.0331 1,100 3.14E-08 3.88E-09 1,500 5.91E-09 0.0408 8.72E-09 0.0266 1,500 1.39E-09 2.21E-11 2,000 1.91E-10 0.1226 3.1E-10 0.0234 = 2,000 8.52E-11 1.93E-12

- fractional standard deviation

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.7-1 TN-40HT ISFSI DETECTOR LOCATIONS FOR SITE DOSE EVALUATION Distance"North Detectors": "South Detectors": "West Detectors": "East Detectors":

(meters) Distance is measured from Distance is measured from Distance is measured from Distance is measured (X,Y)=( -548.64, 676.91) (X,Y)=( -548.64, -128.27) (X,Y)=( -7595.87, 274.32) from (X,Y)=( 6498.59, cm. "X" is constant. Tally cm. "X" is constant. Tally cm. "Y" is constant. Tally 274.32) cm. "Y" is F5. F15. F25. constant. Tally F35.

Point "Y" Point "Y" Point "X" Point "X" Detector # Coordinate, Detector # Coordinate, Detector # Coordinate, Detector Coordinate,

__________________ (meters) (meters) (meters) _________

1. (meters) 0 1 6.77 14 -1.28 27 -75.96 40 64.99 10 2 16.77 15 -11.28 28 -85.96 41 74.99 45 3 51.77 16 -46.28 29 -120.96 42 109.99 100 4 106.77 17 -101.28 30 -175.96 43 164.99 200 5 206.77 18 -201.28 31 -275.96 44 264.99 300 6 306.77 19 -301.28 32 -375.96 45 364.99 400 7 406.77 20 -401.28 33 -475.96 46 464.99 500 8 506.77 21 -501.28 34 -575.96 47 564.99 600 9 606.77 22 -601.28 35 -675.96 48 664.99 700 10 706.77 23 -701.28 36 -775.96 49 764.99 800 11 806.77 24 -801.28 37 -875.96 50 864.99 900 12 906.77 25 -901.28 38 -975.96 51 964.99 1000 13 1006.77 26 -1001.28 39 -1075.96 52 1064.99 Distanc "North-West Detectors": "North-East "South-West Detectors": "North-East e Distance is measured from Detectors": Distance Distance is measured from Detectors":

(meters (X,Y)=(-7558.30, 639.34) cm. is measured from (X,Y)=(-7558.30, -90.70) cm. "X" Distance is

) "X" is constant. Tally F45. (X,Y)=(6461.02, is constant. Tally F65. measured from 639.34) cm. "Y" is as (X,Y)=(6461 .02, -

in Tally F45. Tally 90.70) cm. "Y" is as

________ ________ ________ F55. ________ ________ ________ in Tall y F65. Tally Point "X" "Y" Point "X" Point "X" "Y" Point "X" Detecto Coord, Coord, Detector Coord, Detector Coord, Coord, Detector Coord, r# (meters (meters # (meters # (meters) (meters) # (meters)

) ) )

10 53 -82.65 13.46 65 71.68 77 -82.65 -7.98 89 71.68 45 54 -107.40 38.21 66 96.43 78 -107.40 -32.73 90 96.43 100 55 -146.29 77.10 67 135.32 79 -146.29 -71.62 91 135.32 200 56 -217.00 147.81 68 206.03 80 -217.00 -142.33 92 206.03 300 57 -287.72 218.53 69 276.74 81 -287.72 -213.04 93 276.74 400 58 -358.43 289.24 70 347.45 82 -358.43 -283.75 94 347.45 500 59 -429.14 359.95 71 418.16 83 -429.14 -354.46 95 418.16 600 60 -499.85 430.66 72 488.87 84 -499.85 -425.17 96 488.87 700 61 -570.56 501.37 73 559.59 85 -570.56 -495.88 97 559.59 800 62 -641.27 572.08 74 630.30 86 -641.27 -566.59 98 630.30 900 63 -711.98 642.79 75 701.01 87 -711.98 -637.30 99 701.01 1000 64 -782.69 713.50 76 771.72 88 -782.69 -708.01 100 771.72

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.7-2 DOSE RATES AS A FUNCTION OF DISTANCE FROM THE ISFSI (PAGE 1 OF 2)

From the ISFSI "North\South" Faces Distance Gamma Radiation Neutron Radiation Total, from the Gamma+Neutron, ISFSI Radiation Center Dose Dose Dose Rate Dose Dose Dose (meters) Rate Rate Value, Rate Rate Rate Value, Relative mrem/hr Relative Value, Relative mrem/hr Error Error mrem/hr Error 4.03 5.51 0.02 6.94 0.04 12.45 0.02 14.03 3.57 0.01 4.84 0.03 8.42 0.02 49.03 1.09 0.01 1.56 0.02 2.65 0.02 104.03 0.06 0.01 0.22 0.02 0.28 0.02 204.03 0.02 0.01 0.04 0.03 0.06 0.02 304.03 0.005 0.01 0.010 0.04 0.015 0.02 404.03 0.002 0.02 0.003 0.05 0.005 0.03 504.03 6.29e-4 0.03 9.06e-4 0.05 1.53e-3 0.03 604.03 2.52e-4 0.03 3.24e-4 0.09 5.77e-4 0.05 704.03 1.42e-4 0.18 1.30e-4 0.11 2.72e-4 0.11 804.03 6.13e-5 0.07 6.02e-5 0.24 1.22e-4 0.12 904.03 2.80e-5 0.06 1.95e-5 0.12 4.75e-5 0.06 1004.03 1.72e-5 0.09 9.21e-6 0.15 2.64e-5 0.08 From the ISFSI "East\West" Faces Distance Gamma Radiation Neutron Radiation Total, from the Gamma+Neutron, ISFSI Radiation Center Dose Rate Dose Dose Rate Dose Dose Rate Dose (meters) Value, Rate Value, Rate Value, Rate mrem/hr Relative mrem/hr Relative mrem/hr Relative

________ _________ Error Error Error 70.47 33.29 0.03 28.73 0.04 62.03 0.02 80.47 2.77 0.02 4.43 0.04 7.20 0.03 115.47 0.06 0.01 0.25 0.02 0.30 0.02 170.47 0.03 0.01 0.10 0.02 0.13 0.02 270.47 0.01 0.02 0.02 0.02 0.03 0.02 370.47 0.002 0.02 0.005 0.05 0.008 0.04 470.47 0.001 0.03 0.002 0.06 0.002 0.04 570.47 3.06e-4 0.03 4.77e-4 0.06 7.83e-4 0.04 670.47 1.25e-4 0.03 2.33e-4 0.12 3.58e-4 0.08 770.47 5.77e-5 0.05 8.38e-5 0.14 1.41e-4 0.09 870.47 2.71e-5 0.04 2.43e-5 0.12 5.14e-5 0.06 970.47 1.48e-5 0.06 1.15e-5 0.27 2.62e-5 0.12 1070.47 8.36e-6 0.08 4.44e-6 0.11 1.28e-5 0.06

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.7-2 DOSE RATES AS A FUNCTION OF DISTANCE FROM THE ISFSI (PAGE 2 OF 2)

From ISFSI Corners Distance Gamma Radiation Neutron Radiation Total, from the Gamma+Neutron, ISFSI Radiation Center Dose Rate Dose Dose Rate Dose Dose Rate Dose (meters) Value, Rate Value, Rate Value, Rate mrem/hr Relative mrem/hr Relative mrem/hr Relative Error Error Error 77.91 3.67 0.01 5.13 0.03 8.80 0.02 107.91 0.50 0.01 0.80 0.02 1.30 0.01 159.24 0.04 0.01 0.12 0.01 0.15 0.01 256.49 0.01 0.01 0.02 0.02 0.03 0.01 355.27 0.003 0.02 0.01 0.04 0.01 0.03 454.58 0.001 0.02 0.002 0.03 0.003 0.02 554.14 4.53e-4 0.02 6.67e-4 0.06 1.12e-3 0.04 653.84 1.94e-4 0.03 2.37e-4 0.07 4.31e-4 0.04 753.61 9.07e-5 0.04 8.48e-5 0.09 1.75e-4 0.05 853.44 4.36e-5 0.03 3.94e-5 0.14 8.30e-5 0.07 953.31 3.95e-5 0.38 1.17e-5 0.10 5.12e-5 0.29 1053.20 1.31e-5 0.04 5.59e-6 0.10 1.87e-5 0.04

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.7-3 SKYSHINE DOSE RATES AS A FUNCTION OF DISTANCE FROM THE ISFSI (PAGE 1 OF 2)

From the ISFSI "North\South" Faces Distance Gamma Radiation Neutron Radiation Total Skyshine, from the Gamma+Neutron, ISFSI Radiation Center Dose Dose Dose Dose Dose Dose (meters) Rate Rate Rate Rate Rate Rate Value, Relative Value, Relative Value, Relative mrem/hr Error mrem/hr Error mrem/hr Error 4.03 4.77 0.02 6.94 0.04 11.71 0.02 14.03 3.08 0.02 4.84 0.03 7.93 0.02 49.03 0.94 0.01 1.56 0.02 2.50 0.02 104.03 0.06 0.01 0.22 0.02 0.27 0.02 204.03 0.01 0.01 0.04 0.03 0.06 0.02 304.03 0.004 0.02 0.010 0.04 0.015 0.03 404.03 0.001 0.02 0.003 0.05 0.004 0.03 504.03 5.77e-4 0.04 9.06e-4 0.05 1.48e-3 0.03 604.03 2.27e-4 0.04 3.24e-4 0.09 5.51e-4 0.05 704.03 1.29e-4 0.20 1.30e-4 0.11 2.59e-4 0.11 804.03 5.48e-5 0.08 6.02e-5 0.24 1.15e-4 0.13 904.03 2.46e-5 0.06 1.95e-5 0.12 4.41e-5 0.06 1004.03 1.52e-5 0.11 9.21e-6 0.15 2.44e-5 0.09 From the ISFSI "East\West" Faces Distance Gamma Radiation Neutron Radiation Total Skyshine, from the Gamma+Neutron, ISFSI _______ Radiation Center Dose Dose Rate Dose Dose Dose Dose (meters) Rate Relative Rate Rate Rate Rate Value, Error Value, Relative Value, Relative mrem/hr mrem/hr Error mrem/hr Error 70.47 28.19 0.04 28.71 0.04 56.902 0.03 80.47 2.38 0.03 4.43 0.04 6.812 0.03 115.47 0.05 0.01 0.25 0.02 0.297 0.02 170.47 0.03 0.01 0.10 0.02 0.127 0.02 270.47 0.01 0.02 0.02 0.02 0.027 0.02 370.47 0.002 0.02 0.005 0.05 0.007 0.04 470.47 0.001 0.04 0.002 0.06 0.002 0.04 570.47 2.72e-4 0.03 4.77e-4 0.06 7.49e-4 0.04 670.47 1.10e-4 0.04 2.33e-4 0.12 3.43e-4 0.08 770.47 4.97e-5 0.06 8.38e-5 0.14 1.34e-4 0.09 870.47 2.32e-5 0.04 2.43e-5 0.12 4.75e-5 0.06 970.47 1.23e-5 0.07 1.15e-5 0.27 2.38e-5 0.13 1070.47 6.99e-6 0.09 4.44e-6 0.11 1.14e-5 0.07

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.7-3 SKYSHINE DOSE RATES AS A FUNCTION OF DISTANCE FROM THE ISFSI (PAGE 2 OF 2)

From ISFSI Corners Distance Gamma Radiation Neutron Radiation Total Skyshine, from the Gamma+Neutron, ISFSI Radiation Center Dose Dose Dose Dose Dose Dose (meters) Rate Rate Rate Rate Rate Rate Value, Relative Value, Relative Value, Relative mrem/hr Error mrem/hr Error mrem/hr Error 77.91 3.15 0.02 5.12 0.03 8.28 0.02 107.91 0.43 0.01 0.80 0.02 1.23 0.01 159.24 0.03 0.01 0.12 0.01 0.15 0.01 256.49 0.01 0.01 0.02 0.02 0.03 0.01 355.27 0.003 0.02 0.006 0.04 0.009 0.03 454.58 0.001 0.02 0.002 0.03 0.003 0.02 554.14 4.13e-4 0.02 6.67e-4 0.06 1.08e-3 0.04 653.84 1.76e-4 0.03 2.37e-4 0.07 4.14e-4 0.04 753.61 8.24e-5 0.04 8.48e-5 0.09 1.67e-4 0.05 853.44 3.91e-5 0.03 3.94e-5 0.14 7.85e-5 0.07 953.31 3.69e-5 0.40 1.17e-5 0.10 4.86e-5 0.31 1053.20 1.14e-5 0.05 5.59e-6 0.10 1.70e-5 0.05

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.7-4 SAS2H SOURCE TERMS AS A FUNCTION OF COOLING TIME Decay Time Source Strength (particles/sec) _________

Bottom (years) Nozzle In-Core Plenum Top Nozzle Neutron 18 2.235E+12 3.303E+15 2.870E+12 1.314E+12 7.59E+08 20 1.718E+12 3.142E+15 2.206E+12 1.010E+12 7.05E+08 22 1.321E+12 2.989E+15 1.696E+12 7.763E+11 6.54E+08 24 1.015E+12 2.843E+15 1.304E+12 5.967E+11 6.08E+08 26 7.805E+11 2.704E+15 1.002E+12 4.587E+11 5.65E+08 28 6.000E+1 1 2.573E+15 7.705E+1 1 3.527E+1 1 5.25E+08 30 4.613E+11 2.447E+15 5.923E+11 2.711E+11 4.88E+08 32 3.546E+1 1 2.328E+15 4.553E+1 1 2.084E+1 1 4.54E+08 34 2.726E+11 2.214E+15 3.500E+11 1.602E+11 4.22E+08 36 2.095E+11 2.106E+15 2.691E+11 1.232E+11 3.93E+08 38 1.611E+11 2.004E+15 2.068E+11 9.468E+10 3.65E+08 40 1 .238E+ 11 1 .906E+ 15 1 .590E+ 11 7.278E+ 10 3.40E+08

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.8-1 TN-40HT RELEASABLE SOURCE TERM FOR OFF-NORMAL CONDITIONS -

DESIGN BASIS 14X14 FUEL Concentration in Material Released 2 Activity Release Void Space of TN- Q 3

Isotope (Ci/assembly) Fraction 40HT 1 (Ci/cm ) (jCi/sec)

H3 1 .78E+02 0.30 3.80E-05 5.76E-04 (3)

Co60 6.73E+00 1.50E-01 7.18E-06 1.09E-04 Pu238 3.02E+03 3.00E-06 6.44E-09 9.78E-08 Pu239 1 .35E+02 3.00E-06 2.88E-1 0 4.37E-09 Pu240 2.67E+02 3.00E-06 5.70E-10 8.65E-09 Pu241 3.19E+04 3.00E-06 6.81E-08 1.03E-06 Am241 1 .50E+03 3.00E-06 3.20E-09 4.86E-08 Am243 4.17E+01 3.00E-06 8.90E-11 1.35E-09 Cm244 5.28E+03 3.00E-06 1.13E-08 1.71E-07 Kr 85 1 .78E+03 0.30 3.80E-04 5.77E-03 Sr 90 3.11 E+04 2.00E-04 4.42E-06 6.72E-05 Y 90 3.11E+04 3.00E-06 6.64E-08 1.01E-06 I129 2.40E-02 0.30 5.12E-09 7.77E-08 Cs134 4.01E+02 2.00E-04 5.70E-08 8.66E-07 Cs137 5.27E+04 2.00E-04 7.50E-06 1.14E-04 Ba137m 4.97E+04 3.00E-06 1.06E-07 1.61E-06 Pm147 6.53E+02 3.00E-06 1.39E-09 2.12E-08 Eu154 1.33E+03 3.00E-06 2.84E-09 4.31E-08 Np239 4.17E+01 3.00E-06 8.90E-11 1.35E-09

1. Values are based on 10% failure of the fuel rods and cask free volume of 5.63 m3.

2. Values are based on 1.51 8E-05 cm 3/sec helium leak from containment.

3. The Co-60 source is calculated using the methodology of Reference (Reference 8). It is based on a 14x14 PWR fuel assembly with surface area of 1300 cm2/rod and a crud surface concentration of 140.iCi /cm 2 (per Table 7.1 of Reference 8) at the time of discharge. (The value listed above includes a minimum cooling time of twelve years.)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Table A7A.8-2 TN-40HT RELEASABLE SOURCE TERM FOR ACCIDENT CONDITIONS -

DESIGN BASIS 14X14 FUEL (2)

Concentration in Material Released Activity Release Void Space of TN- Q (1) 3 Isotope (Ci/assem bly) Fraction 40HT (C i/cm ) (m Ci/sec)

H3 1.78E+02 0.30 3.80E-04 1.41E-02 (3)

Co60 6.73+00 1 .00E-00 4.79E-05 1 .78E-03 Pu238 3.02E+03 3.00E-06 6.44E-08 2.40E-06 Pu239 1 .35E+02 3.00E-06 2.88E-09 1 .07E-07 Pu240 2.67E+02 3.00E-06 5.70E-09 2.12E-07 Pu241 3.19E+04 3.00E-06 6.81E-07 2.54E-05 Am241 1.50E+03 3.00E-06 3.20E-08 1.19E-06 Am243 4.17E+01 3.00E-06 8.90E-10 3.31E-08 Cm244 5.28E+03 3.00E-06 1.13E-07 4.20E-06 Kr85 1.78E+03 0.30 3.80E-03 1.41E-01 Sr90 3.11E+04 2.00E-04 4.42E-05 1.65E-03 Y90 3.11 E+04 3.00E-06 6.64E-07 2.47E-05 I129 2.40E-02 0.30 5.12E-08 1.91E-06 Cs134 4.01E+02 2.00E-04 5.70E-07 2.12E-05 Cs137 5.27E+04 2.00E-04 7.50E-05 2.79E-03 Ba137m 4.97E+04 3.00E-06 1.06E-06 3.95E-05 Pm147 6.53E+02 3.00E-06 1.39E-08 5.19E-07 Eu154 1.33E+03 3.00E-06 2.84E-08 1.06E-06 Np239 4.17E+01 3.00E-06 8.90E-10 3.31E-08

1. Values are based on 100% failure of the fuel rods and cask free volume of 5.63 m3.

3

2. Values are based on 3.72E-05 cm /sec helium leak from containment.

3. The Co-60 source is calculated using the methodology of Reference (Reference 8). It is based on a 14x14 PWR fuel assembly with surface area of 1300 cm2/rod and a crud surface concentration of 140 .iCi / cm 2 at the time of discharge. (The value listed above includes a minimum cooling time of twelve years.)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.8-3 OFF-SITE AIRBORNE DOSES FROM OFF-NORMAL CONDITIONS AT 110 M (INTERNAL + EXTERNAL)

Design Basis 14x14 Fuel, Committed Doses (Internal) + Deep Dose (External) mrem for 45 days R. B.

Isotope Gonad Breast Lung Marrow Surface Thyroid Remainder Effective Skin H3 4.68E-05 4.68E-05 4.69E-05 4.68E-05 4.68E-05 4.68E-05 4.68E-05 4.68E-05 0.00E+00 Co60 2.69E-03 9.71 E-03 1 .77E-01 9.06E-03 7.28E-03 8.56E-03 1 .87E-02 3.05E-02 2.93E-04 Pu238 1 .29E-02 4.60E-07 1 .47E-01 6.99E-02 8.73E-01 4.42E-07 3.23E-02 4.87E-02 7.40E-1 1 Pu239 6.53E-04 1.89E-08 6.64E-03 3.47E-03 4.34E-02 1.86E-08 1.55E-03 2.38E-03 1.51E-12 Pu240 1.29E-03 3.87E-08 1.31E-02 6.87E-03 8.58E-02 3.68E-08 3.07E-03 4.71E-03 6.27E-12 Pu241 3.31E-03 1.49E-07 1.54E-02 1.63E-02 2.04E-01 6.02E-08 6.36E-03 1.08E-02 2.24E-12 Am241 7.42E-03 6.11E-07 4.20E-03 3.97E-02 4.95E-01 3.66E-07 1.79E-02 2.74E-02 1.15E-09 Am243 2.07E-04 9.65E-08 1.13E-04 1.10E-03 1.38E-02 5.27E-08 4.91E-04 7.55E-04 6.87E-11 Cm244 1.28E-02 8.36E-07 1.55E-02 7.54E-02 9.40E-01 8.12E-07 3.84E-02 5.38E-02 1.24E-10 Kr85 1.25E-05 1.43E-05 1.22E-05 1.16E-05 2.35E-05 1.26E-05 1.16E-05 1.27E-05 1.41E-03 Sr90 8.33E-04 8.33E-04 1.18E-03 1.06E-01 2.29E-01 8.33E-04 1.81E-03 1.11E-01 1.14E-05 Y90 4.86E-08 4.92E-08 4.41E-05 1.32E-06 1.32E-06 4.86E-08 1.83E-05 1.08E-05 1.16E-06 I129 3.24E-08 7.73E-08 1.15E-07 5.14E-08 5.20E-08 5.70E-04 4.34E-08 1.71E-05 1.58E-09 Cs134 5.41E-05 4.53E-05 4.92E-05 4.92E-05 4.67E-05 4.64E-05 5.77E-05 5.21E-05 1.51E-06 Cs137 4.68E-03 4.19E-03 4.72E-03 4.44E-03 4.25E-03 4.24E-03 4.88E-03 4.62E-03 1.82E-05 Ba137m 8.40E-07 9.59E-07 8.34E-07 8.13E-07 1.38E-06 8.58E-07 7.98E-07 8.58E-07 1.11E-06 Pm147 2.16E-12 3.95E-12 7.69E-06 8.11E-07 1.01E-05 2.23E-12 5.85E-07 1.05E-06 3.17E-10 Eu154 2.42E-06 3.19E-06 1.61E-05 2.15E-05 1.06E-04 1.49E-06 2.29E-05 1.57E-05 6.61E-08 Np239 6.61E-10 3.22E-10 1.52E-08 1.48E-09 1.34E-08 2.36E-10 6.26E-09 4.50E-09 4.00E-10 Total 4.69E-02 1 .48E-02 3.85E-01 3.32E-01 2.90E+00 1 .43E-02 1 .26E-01 2.95E-01 1 .73E-03

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.8-4 OFF-SITE AIRBORNE DOSES FROM OFF-NORMAL CONDITIONS AT 110 M (EXTERNAL)

Design Basis14x14 Fuel, Deep Doses (External) mrem for 45 days R. B.

Isotope Gonad Breast Lung Marrow Surface Thyroid Remainder Effective Skin H3 0.00E+00 0.00E+00 2.93E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.53E-09 0.00E+00 Co60 2.48E-04 2.81 E-04 2.50E-04 2.48E-04 3.59E-04 2.56E-04 2.42E-04 2.54E-04 2.93E-04 Pu238 1.19E-11 2.30E-11 1.92E-12 3.04E-12 1.68E-11 7.26E-12 3.60E-12 8.83E-12 7.40E-11 Pu239 3.92E-13 6.11E-13 2.14E-13 2.16E-13 7.66E-13 3.14E-13 2.31E-13 3.43E-13 1.51E-12 Pu240 1.02E-12 1.97E-12 1.74E-13 2.64E-13 1.48E-12 6.27E-13 3.14E-13 7.60E-13 6.27E-12 Pu241 1.37E-12 1.66E-12 1.24E-12 1.08E-12 4.19E-12 1.33E-12 1.16E-12 1.39E-12 2.24E-12 Am241 7.72E-10 9.62E-10 6.06E-09 4.68E-10 2.58E-09 7.04E-10 5.70E-10 7.36E-10 1.15E-09 Am243 5.47E-11 6.52E-11 4.80E-11 3.87E-11 1.87E-10 5.22E-11 4.47E-11 5.45E-11 6.87E-11 Cm244 2.18E-11 4.21E-11 2.24E-12 4.62E-12 2.79E-11 1.33E-11 5.73E-12 1.55E-11 1.24E-10 Kr85 1.25E-05 1.43E-05 1.22E-05 1.16E-05 2.35E-05 1.26E-05 1.16E-05 1.27E-05 1.41E-03 Sr90 9.67E-09 1.18E-08 8.00E-09 6.76E-09 2.83E-08 9.11E-09 7.59E-09 9.36E-09 1.14E-05 Y90 3.52E-09 4.10E-09 3.30E-09 3.02E-09 8.28E-09 3.49E-09 3.13E-09 3.54E-09 1.16E-06 I129 6.95E-10 9.58E-10 3.08E-10 2.36E-10 1.58E-09 5.55E-10 3.31 E-10 5.47E-10 1.58E-09 Cs134 1.19E-06 1.35E-06 1.18E-06 1.15E-06 1.92E-06 1.21E-06 1.13E-06 1.21E-06 1.51E-06 Cs137 1.68E-08 2.04E-08 1.41E-08 1.20E-08 4.82E-08 1.59E-08 1.34E-08 1.63E-08 1.82E-05 Ba137m 8.40E-07 9.59E-07 8.34E-07 8.13E-07 1.38E-06 8.58E-07 7.98E-07 8.58E-07 1.11E-06 Pm147 2.93E-13 3.74E-13 2.13E-13 1.75E-13 8.53E-13 2.64E-13 2.06E-13 2.71E-13 3.17E-10 Eu154 4.78E-08 5.43E-08 4.78E-08 4.68E-08 7.52E-08 4.90E-08 4.58E-08 4.90E-08 6.61E-08 Np239 1.88E-10 2.18E-10 1.79E-10 1.62E-10 5.00E-10 1.88E-10 1.69E-10 1.92E-10 4.00E-10 Total 2.63E-04 2.97E-04 2.65E-04 2.62E-04 3.86E-04 2.71 E-04 2.56E-04 2.69E-04 1 .73E-03

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.8-5 OFF-SITE AIRBORNE DOSES FROM ACCIDENT CONDITIONS AT 110 M (INTERNAL)

Design Basis 14x14 Fuel, mrem/30 Days, Committed Doses (Internal)

Isotope Gonad Breast Lung R. Marrow B. Surface Thyroid Remainder Effective H3 3.95E-03 3.95E-03 3.95E-03 3.95E-03 3.95E-03 3.95E-03 3.95E-03 3.95E-03 Co60 1 .37E-01 5.30E-01 9.94E+00 4.96E-01 3.89E-01 4.67E-01 1 .04E+00 1 .70E+00 Pu238 1.09E+00 3.88E-05 1.24E+01 5.89E+00 7.37E+01 3.73E-05 2.72E+00 4.11E+00 Pu239 5.51E-02 1.60E-06 5.60E-01 2.93E-01 3.66E+00 1.56E-06 1.31E-01 2.01E-01 Pu240 1.09E-01 3.26E-06 1.11E+00 5.79E-01 7.23E+00 3.10E-06 2.59E-01 3.98E-01 Pu241 2.79E-01 1.25E-05 1.30E+00 1.38E+00 1.72E+01 5.08E-06 5.36E-01 9.13E-01 Am241 6.26E-01 5.14E-05 3.54E-01 3.35E+00 4.18E+01 3.08E-05 1.51 E+00 2.31 E+00 Am243 1.75E-02 8.14E-06 9.53E-03 9.26E-02 1.16E+00 4.44E-06 4.14E-02 6.37E-02 Cm244 1 .08E+00 7.05E-05 1.31 E+00 6.36E+00 7.93E+01 6.85E-05 3.24E+00 4.54E+00 Kr85 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr90 7.03E-02 7.03E-02 9.93E-02 8.94E+00 1 .93E+01 7.03E-02 1 .53E-01 9.34E+00 Y90 3.80E-06 3.80E-06 3.72E-03 1.11E-04 1.11E-04 3.80E-06 1.54E-03 9.10E-04 I129 2.68E-06 6.44E-06 9.67E-06 4.31 E-06 4.25E-06 4.81 E-02 3.64E-06 1.44E-03 Cs134 4.46E-03 3.71 E-03 4.05E-03 4.05E-03 3.77E-03 3.81 E-03 4.77E-03 4.29E-03 Cs137 3.95E-01 3.54E-01 3.98E-01 3.74E-01 3.58E-01 3.58E-01 4.11E-01 3.89E-01 Ba137m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm147 1.58E-10 3.02E-10 6.49E-04 6.84E-05 8.55E-04 1.66E-10 4.94E-05 8.89E-05 Eu154 2.00E-04 2.65E-04 1.35E-03 1.81 E-03 8.93E-03 1.22E-04 1.93E-03 1.32E-03 Np239 3.99E-08 8.73E-09 1.26E-06 1.11E-07 1.09E-06 4.08E-09 5.13E-07 3.63E-07 Total 3.86E+00 9.62E-01 2.75E+01 2.78E+01 2.44E+02 9.51 E-01 1 .00E+01 2.40E+01

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A7A.8-6 OFF-SITE AIRBORNE DOSES FROM ACCIDENT CONDITIONS AT 110 M (EXTERNAL)

Design Basis 14x14 Fuel, mrem/30 Days, Deep Doses (External)

R. B.

________ Gonad Breast Lung Marrow Surface Thyroid Remainder Effective Skin H3 0.00E+00 0.00E+00 2.47E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.98E-07 0.00E+00 Co60 1 .40E-02 1 .58E-02 1.41 E-02 1 .40E-02 2.02E-02 1 .44E-02 1 .36E-02 1 .43E-02 1 .65E-02 Pu238 1.00E-09 1.94E-09 1.62E-10 2.56E-10 1.42E-09 6.12E-10 3.04E-10 7.45E-10 6.24E-09 Pu239 3.30E-11 5.15E-11 1.81E-11 1.82E-11 6.46E-11 2.65E-11 1.95E-11 2.89E-11 1.27E-10 Pu240 8.58E-11 1.66E-10 1.47E-11 2.23E-11 1.25E-10 5.29E-11 2.65E-11 6.41E-11 5.29E-10 Pu241 1.16E-10 1.40E-10 1.05E-10 9.08E-11 3.53E-10 1.13E-10 9.82E-11 1.17E-10 1.89E-10 Am241 6.51E-08 8.11E-08 5.11E-07 3.95E-08 2.18E-07 5.94E-08 4.81E-08 6.20E-08 9.71E-08 Am243 4.62E-09 5.50E-09 4.05E-09 3.27E-09 1 .57E-08 4.41 E-09 3.77E-09 4.60E-09 5.80E-09 Cm244 1.84E-09 3.55E-09 1.89E-10 3.90E-10 2.35E-09 1.12E-09 4.83E-10 1.31E-09 1.04E-08 Kr85 1.05E-03 1.21E-03 1.03E-03 9.81E-04 1.98E-03 1.06E-03 9.81 E-04 1.07E-03 1.19E-01 Sr90 8.15E-07 9.95E-07 6.75E-07 5.70E-07 2.39E-06 7.68E-07 6.40E-07 7.89E-07 9.64E-04 Y90 2.97E-07 3.46E-07 2.78E-07 2.55E-07 6.98E-07 2.94E-07 2.64E-07 2.99E-07 9.81 E-05 I129 5.86E-08 8.08E-08 2.60E-08 1.99E-08 1.33E-07 4.68E-08 2.79E-08 4.61E-08 1.33E-07 Cs134 1.00E-04 1.14E-04 9.96E-05 9.72E-05 1.62E-04 1.02E-04 9.54E-05 1.02E-04 1.28E-04 Cs137 1.41E-06 1.72E-06 1.19E-06 1.01E-06 4.07E-06 1.34E-06 1.13E-06 1.37E-06 1.53E-03 Ba137m 7.09E-05 8.09E-05 7.04E-05 6.86E-05 1.16E-04 7.24E-05 6.73E-05 7.24E-05 9.37E-05 Pm147 2.47E-11 3.16E-11 1.80E-11 1.47E-11 7.20E-11 2.23E-11 1.74E-11 2.29E-11 2.68E-08 Eu154 4.03E-06 4.58E-06 4.03E-06 3.95E-06 6.34E-06 4.14E-06 3.87E-06 4.13E-06 5.57E-06 Np239 1 .59E-08 1 .84E-08 1.51 E-08 1 .37E-08 4.22E-08 1 .59E-08 1 .43E-08 1 .62E-08 3.37E-08 Total 1 .52E-02 1 .72E-02 1 .53E-02 1.51 E-02 2.25E-02 1 .57E-02 1 .48E-02 1 .55E-02 1 .38E-01

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 THIS PAGE IS LEFT INTENTIONALLY BLANK

Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Protective Cover Surface Side surface Cask Lid above N-shield Top Surface

___ Side surface Side surface below N-shield Bottom Surface FIGURE A7A.2-1 TN-40HT NORMAL CONDITIONS DOSE POINT LOCATIONS

Figure Withheld Under 10 CFR 2.390 Withhold Figure Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 FIGURE A7A.4-3 TN-40HT MCNP QUARTER GAMMA MODEL - XY PLOT AT Z=0 CM

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 nical ieJ eJ 15 c

SL)iI layer 1km FIGURE A7A.4-4 TN-40HT FAR FIELD MCNP MODEL

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13

_;IEEEE__ElLEE 225 175 125 75

• gamma

___________

• neutron 25

• total

-25

-75

__EEIhI

-125

-175

-225 10.0 100.0 1000.0 mrem/hr Note: 0.0 is the center line of the active fuel FIGURE A7A.5-1 TN-40HT CASK SIDE SURFACE DOSE RATES NORMAL AND OFF-NORMAL CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 60 50 40 I-I::

10 0

0 20 40 60 80 100 120 radial location (cm)

FIGURE A7A.5-2 TN-40HT CASK PROTECTIVE COVER SURFACE DOSE RATES NORMAL AND OFF-NORMAL CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 10000 1000 I 100 10 0 20 40 60 80 100 120 radial location (cm)

FIGURE A7A.5-3 TN-40HT CASK BOTTOM SURFACE DOSE RATES NORMAL AND OFF-NORMAL CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 300 250 200 150 100 50 0

-50

-100

-150

-200 0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 mrem/hr FIGURE A7A.5-4 TN-40HT CASK DOSE RATES 1 M FROM SIDE SURFACE NORMAL AND OFF-NORMAL CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 300 250 200 150 100 50 0

0

! 0

-50

-100

-150

-200 0.00 10.00 20.00 30.00 mrem/hr FIGURE A7A.5-5 TN-40HT CASK DOSE RATES 2 M FROM SIDE SURFACE NORMAL AND OFF-NORMAL CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 300 H

250 gamma - -

U neutron total 200 ____

• ____

I 150 100 50 0

-50

-100

-150

-200 ____,:H ----- -

10.0 100.0 1000.0 10000.0 mrem/hr FIGURE A7A.5-6 TN-40HT CASK SIDE SURFACE DOSE RATES - ACCIDENT CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 300 250 200 150 100 50 0

-50

-100

-150

-200 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 m rem/hr FIGURE A7A.5-7 TN-40HT CASK DOSE RATES 1 M FROM SIDE SURFACE - ACCIDENT CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 300 250 200 150 100 50 0

-50

-100

-150

-200 0 100 200 300 400 500 600 700 m re m/hr FIGUREA7A.5-8 TN-40HT CASK DOSE RATES 2 M FROM SIDE SURFACE - ACCIDENT CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 600 500 400 300 200 100 0

0 20 40 60 80 100 120 radial location (cm)

FIGURE A7A.5-9 TN-40HT CASK LID SURFACE DOSE RATES - NORMAL AND ACCIDENT CONDITIONS

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Gamma 1.0E+01 1 .0E+00 1 .0E-01 1 .0E-02 1 .0E-03 1.0E-04 1.0E-05 1.0E-06 1 .0E-07 1 .0E-08 1 .0E-09 1 .0E-1 0 1 .0E-1 1 110 100 1000 10000 meter Total -h Skyshine Neutron 1 .0E+01 1 .0E-01 1.0E-03 1.0E-05 E

1 .0E-07 1 .0E-09 1.0E-11 1 10 100 1000 10000 meter

-- Total -h - Skyshine - - - - Direct FIGURE A7A.6-1 TN-40HT CASK - DOSE RATE AS A FUNCTION OF DISTANCE

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Direction towards West

• 45 o

W detectors. Distance is measured West face of the Direction towards North-West 4 1 2 casks array. The reference r5i1 corner detectors. Distance for point is between two rows of5 dose rates is measured from 3 4 2x1 2 array. (X,Y) of the Cask #2 (X,Y)=(-7467.60, reference point is (-7595.87, 5 5 6 548.64 ) cm.

274.32) cm.

7 8 910 Array 1112

1. 1 1314 56 1516 Pitch=18 j 57 1718 1920 Origin of 2122 Coordinate System Z

2324 Y

38 ft - -

2 ft Direction towards North 11'1 X

2526 detectors. Distance is measured from North face of 27 28 the casks array. The reference point is between two 2x12 29 30 arrays. (X,Y) of the reference point is (-548.64, 676.91) cm.

31 32 33 34 35 36 37 38 Array #2 3940 I

41 42 43 44 45 46 47 48 FIGURE A7A.7-1 ARRANGEMENT OF CASKS AND DETECTORS IN MCNP CALCULATION MODEL FOR ISFSI DOSE RATE ESTIMATES

Figure Withheld Under 10 CFR 2.390 PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 7O 65 ------B

---

B 60 1J---------------------------------------------------------

50 ---------1 45 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I E 4O OP System Mowtoirng  :  :

AlaimSetpoint 3.5 -----------

i ------------------------------

30 -- - - -- ----- --------------------

2 5 --- - OP System Piessure Leak toCask Cavity  : =

I I I I I I I I I I 20 OPSystem Pressure Leak to Cask Cavity Piessure  :

05 -------------

00 - -

246 810121416182022242628303234363840 Time since Cask Loading (years)

FIGURE A7A.8-2 OVERPRESSURE MONITORING SYSTEM PRESSURE DROP WITH TIME

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B-1 Appendix A7B TN-40HT SHIELDING EVALUATION COMPUTER INPUT

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B-2 A7B SAMPLE SHIELDING INPUT FILES This Section provides representative input files for the computer codes, SAS2H and MCNP, utilized to calculate the source terms and dose rates for the TN-40HT Cask. A listing of the file names and a brief description are provided below, followed by the files themselves.

Section # File Name Description A7B.1 I_6034.in SAS2H Input File for Design Basis Fuel (In-core zone)

A7B.2 htgnfn MCNP Input for Primary Gamma Dose

_________ ____________ (Near Field, Normal and Off-Normal Storage Conditions)

A7B.3 Htgnfa MCNP Input for Primary Gamma Dose

_________ ____________ (Near Field, Accident Storage Conditions)

A7B.4 gfwd2_ MCNP Input for Gamma Dose

_________ ____________ (Far Field)

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.1-1 A7B.1 SAS2H INPUT FILE FOR DESIGN BASIS FUEL (IN-CORE ZONE)

I_6043.in PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.1-2 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.1-3 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.1-4 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-1 A7B.2 MCNP INPUT FOR PRIMARY GAMMA DOSE (NEAR FIELD, NORMAL AND OFF-NORMAL STORAGE CONDITIONS) htgnfn PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-2 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-3 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-4 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-5 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-6 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-7 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-8 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-9 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-10 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-1 1 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-12 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-13 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-14 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-15 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-16 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-17 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.2-18 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-1 A7B.3 MCNP INPUT FOR PRIMARY GAMMA DOSE (NEAR FIELD, ACCIDENT STORAGE CONDITIONS)

Htgnfa PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-2 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-3 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-4 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-5 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-6 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-7 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-8 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-9 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-10 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-1 1 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-12 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-13 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-14 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-15 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-16 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-17 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.3-18 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.4-1 A7B.4 MCNP INPUT FOR GAMMA DOSE (FAR FIELD) gfwd2_

PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.4-2 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A7B.4-3 PROPRIETARY - TRADE SECRET INFORMATION WITHHELD PURSUANT TO 10 CFR 2.390

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.1-1 SECTION A8 ACCIDENT ANALYSIS A8.1 OFF-NORMAL OPERATIONS Off-normal operations are design events of the second type (Design Event II) as defined in ANSI/ANS 57.9 (Reference 1). Design Event II consists of that set of events that, although not occurring regularly, can be expected to occur with moderate frequency or on the order of once during a calendar year of ISFSI operation.

Of the various types of Design Event II conditions described in Reference 1, only the loss of external power supply for a limited duration is considered to be applicable and credible to ISFSI operations. Loss of electric power as an off-normal condition is analyzed below.

A8.1.1 LOSS OF ELECTRICAL POWER A total loss of ac power is postulated to occur in the feeder cabling which supplies power to the ISFSI. The failure could be either an open or a short to ground circuit, or any other mechanism capable of producing an interruption of power.

A8.1.1.1 POSTULATED CAUSE OF THE EVENT A loss of power to the ISFSI may occur as a result of natural phenomena, such as lightning or extreme wind, or as a result of undefined disturbances in the non-safety-related portion of the electric power system of the Prairie Island Nuclear Generating Plant (PINGP).

If electric power is lost, the following systems could be de-energized and rendered nonfunctional:

Area lighting Area receptacles Cask pressure monitoring instrumentation A8.1.1.2 DETECTION OF EVENTS A loss of power at the PINGP site would be indicated and/or alarmed in the main control room. If the loss of power were localized at the ISFSI site, this would be detected during periodic surveillance by noting that area lighting is not operational.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.1-2 A8.1.1.3 ANALYSIS OF EFFECTS AND CONSEQUENCES This event has no safety or radiological consequences because a loss of power will not affect the integrity of the storage casks, jeopardize the safe storage of the fuel, or result in radiological releases. None of the systems whose failure could be caused by this event are necessary for the accomplishment of the safety function of the ISFSI. The lighting system functions merely for convenience and visual monitoring, and the instrumentation monitors the long-term performance of the storage casks with respect to the cask seals. None of these parameters are expected to change rapidly and their status is not dependent upon electric power.

A8.1.1.4 CORRECTIVE ACTION Following a loss of electric power to the ISFSI, plant maintenance personnel will be informed and will isolate the fault and restore service by conventional means. Such an operation is straightforward and routine for the maintenance personnel of an electric utility.

A8.1.2 RADIOLOGICAL IMPACT FROM OFF-NORMAL OPERATIONS No radiological impact from off-normal operations is postulated.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-1 A8.2 ACCIDENTS Accidents are design events of the third and fourth type (Design Events III and IV) as defined in ANSI/ANS 57.9 (Reference 1). Design Event III consists of that set of infrequent events that could reasonably be expected to occur during the lifetime of the ISFSI.

Design Event IV consists of the events that are postulated because their consequences may result in the maximum potential impact on the immediate environs. Their consideration establishes a conservative design basis for certain systems with important confinement features.

A8.2.1 EARTHQUAKE A8.2.1 .1 CAUSE OF ACCIDENT The design earthquake (DE) is postulated to occur as a design basis extreme natural phenomenon.

A8.2.1 .2 ACCIDENT ANALYSIS Cask response to a seismic event is evaluated in Sections A3.2.3 and A4.2.3. Results of these analyses show that the cask does not tip over or slide and that the containment vessel stresses resulting from the seismic loads are below ASME code allowable stresses for accident conditions. The integrity of the cask is not compromised. No damage to the cask is postulated. The basket stresses are also low and do not result in deformations that would prevent fuel from being unloaded from the cask.

A8.2.1.3 ACCIDENT DOSE CALCULATIONS The DE does not damage the cask. Hence, no radioactivity is released and there is no associated dose increase due to this event.

A8.2.2 EXTREME WIND A8.2.2.1 CAUSE OF ACCIDENT The extreme winds due to passage of the design tornado as defined in Section A3.2.1 are postulated to occur as an extreme natural phenomenon.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-2 A8.2.2.2 ACCIDENT ANALYSIS In Section A3.2.1, it is shown that extreme winds do not result in a cask tip over or sliding of the cask. The pressure due to high winds on the surface of the cask is bounded by the assumed external pressure of 25 psig. The stresses in the cask resulting from this external pressure are presented in Appendix A4A. High winds have no effect on the integrity of the cask, and do not result in damage to the cask. High winds do not affect the basket or the ability to retrieve the spent fuel from the cask. The effect of tornado missiles hitting the cask has been evaluated in Section A3.2.1. These analyses show that the stresses in the cask as a result of missile impact are well below the ASME Code allowable stresses for Accident (Level D) conditions. It is also shown in Section A3.2.1 that the tornado missile impact will not result in a cask tip over. Local damage to the neutron shield may result from the tornado missile impact.

A8.2.2.3 ACCIDENT DOSE CALCULATIONS Extreme winds are not capable of overturning the casks nor of damaging the cask seals. The overpressure system and the neutron shielding may be damaged.

To determine the bounding dose rate, loss of neutron shielding (322 mrem from Section A8.2.5) is combined with the total effective dose equivalent (TEDE) from the loss of one confinement barrier and 100% fuel cladding failure (24 mrem from Section A8.2.9). The resulting site boundary accident dose, 346 mrem, is below the 5 rem TEDE limit as specified in 10 CFR 72.106(b) (Reference 2).

A8.2.3 FLOOD A8.2.3.1 CAUSE OF ACCIDENT The probable maximum flood has been calculated to reach a level of 703.6 ft., with wave action to a maximum level of 706.7 ft.

A8.2.3.2 ACCIDENT ANALYSIS The casks are designed to withstand the forces developed by the probable maximum flood without damage to cask integrity or tipping of the casks. The height of the cask seals will be above the level of the probable maximum flood and associated wave action. Accordingly no fuel damage or criticality is postulated to occur as a result of flooding. Analyses are contained in Section A3.2.2.

A8.2.3.3 ACCIDENT DOSE CALCULATIONS The probable maximum flood is not capable of overturning the casks or of damaging their seals. Therefore, no resultant doses are projected.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-3 A8.2.4 EXPLOSION A8.2.4.1 CAUSE OF ACCIDENT A munition barge explosion has been postulated to occur at a location approximately 2600 feet from the ISFSI. This occurrence is described in detail in Section 2.2. The pressure wave of 2.25 psi is estimated to occur at the ISFSI.

A8.2.4.2 ACCIDENT ANALYSIS The TN-40HT cask is designed for an external pressure of 25 psig as described in Appendix A4A.

A8.2.4.3 ACCIDENT DOSE CALCULATIONS The cask will not tip as a result of the postulated pressure wave. Accordingly, no cask damage or release of radioactivity is postulated. Since no radioactivity is released, no resultant dose increase is associated with this event.

A8.2.5 FIRE A8.2.5.1 CAUSE OF ACCIDENT The only combustible materials in the ISFSI are in the form of insulation on instrumentation wiring and paint on the outside surface of the storage casks. In addition, the tow vehicle will contain a small amount of gasoline or diesel fuel. No other combustible or explosive materials are allowed to be stored on the ISFSI slabs. The ISFSI area is cleared of trees. The entire area surrounding the Equipment Storage Building and concrete pad within the perimeter road is covered with crushed rock. In addition, other equipment in the area is adequately separated from the ISFSI slabs.

Therefore, no fires other than small electrical fires are considered credible at the ISFSI.

However a hypothetical fire accident is evaluated for the TN-40HT cask based on a fuel fire, the source of fuel being that from a ruptured fuel tank of the cask transporter tow vehicle. The bounding capacity of the fuel tank is 200 gallons and the bounding hypothetical fire is an engulfing fire around the cask.

A8.2.5.2 ACCIDENT ANALYSIS The evaluation of the hypothetical fire event is presented in Section A3.3.2.2.2. The fire thermal evaluation is performed primarily to demonstrate the containment integrity of the TN-40HT cask. This is assured as long as the metal lid seals remain below 536 °F and the cavity pressure is less than 100 psig.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-4 Based on the thermal analyses for the fire accident conditions, the TN-40HT cask can withstand the hypothetical fire accident event without compromising its containment integrity. No melting of the metallic cask components occurs. Peak cask component temperatures are summarized in Table A3.3-6. The maximum seal temperature is well below the temperature limit of the metal seals. Table A3.3-15 and Table A3.3-16 show that at the elevated temperatures corresponding to the hypothetical fire, the cask cavity pressure remains below the analyzed pressure of 100 psig.

A8.2.5.3 ACCIDENT DOSE CALCULATIONS Local damage to the neutron shielding may result from the fire. This is bounded by removal of all the neutron shielding which is evaluated in Appendix A7A. Even with this conservative assumption, the site boundary accident dose rates are below 5 rem to the whole body or any organ as specified in 10CFR72.106(b) (Reference 2).

The off-site doses are evaluated for the following accident condition:

1) loss of radial neutron shielding and

2) loss of the protective cover and top neutron shield.

For accident conditions, the following assumptions are made:

a) the nearest postulated site boundary is 100 meters distant from the cask b) the accident involves a single cask c) the accident duration is 30 days d) a person remains at the postulated site boundary 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day for the entire duration The normal condition total dose rates at 100 meters are scaled by the ratio of accident to normal surface dose rates as shown in the following table. All units are mrem/hr.

Normal Dose Accident, Average, Accident, Rate Surface Table Surface Table 100 m 100 m A7A.2-1 A7A.2-1 mrem/hr

_______ Table A7A.2-2 _____________ _____________ _____________

Gamma 7.59E-03 116 35.5 2.48E-02 Neutron 1.67E-02 1980 78.7 4.20E-01

_______ ________________ _____________ Total: 4.45E-01 The total dose over 30 days would be 320 mrem. The background from the rest of the ISFSI would be less than 1/12 of the 25 mrem/year limit (10 CFR 72.104) (Reference 2),

or 2 mrem. The combined total accident dose would be 322 mrem. There is no activity release from the cask containment.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-5 A8.2.6 INADVERTENT LOADING OF A NEWLY DISCHARGED FUEL ASSEMBLY A8.2.6.1 CAUSE OF ACCIDENT The possibility of a spent fuel assembly, with a heat generation rate greater than 0.8 kw, being erroneously selected for storage in a cask has been considered. The cause of this accident is postulated to be an error during the loading operations, e.g., wrong assembly picked by the fuel handling crane, or a failure in the administrative controls governing the fuel handling operations.

A8.2.6.2 ACCIDENT ANALYSIS The fuel assemblies require several years of storage in the spent fuel pool before the heat generation decays to a rate below 0.8 kw. This accident scenario postulates the inadvertent loading of an assembly not intended for storage in the storage canister, with a heat generation rate in excess of the design basis specified in Section A3.1 .1.

In order to preclude this accident from going undetected, and to ensure that appropriate rectification actions can take place prior to the sealing of the casks, a final verification of the assemblies loaded into the casks and a comparison with fuel management records will be performed to ensure that the loaded assemblies do not exceed any of the specified limits.

These administrative controls and the records associated with them will be included in the procedures described in Section A9 and will comply with the applicable requirements of the Quality Assurance Program described in Section A1 1.

Therefore, appropriate and sufficient actions will be taken to ensure that an erroneously loaded fuel assembly does not remain undetected. In particular, the storage of a fuel assembly with a heat generation in excess of 0.8 kw is not considered credible in view of the multiple administrative controls.

A8.2.6.3 ACCIDENT DOSE CALCULATIONS The inadvertent loading of a fuel assembly not intended for storage in a storage cask is not considered to be a credible occurrence. Therefore, no doses are postulated.

A8.2.7 CASK SEAL LEAKAGE The storage casks feature redundant seals in conjunction with an extremely rugged body design. Additional barriers to the release of radioactivity are presented by the sintered fuel pellet matrix and the Zircaloy cladding which surrounds the fuel pellets.

Furthermore, the interseal gaps are pressurized in excess of the cask cavity. As a result, no credible mechanisms that could result in leakage of radioactive products have been identified. Nevertheless, a loss of the storage cask confinement capability is postulated in Section A8.2.9, and the results found to be negligible.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-6 A8.2.8 HYPOTHETICAL CASK DROP ACCIDENT A8.2.8.1 CAUSE OF ACCIDENT The stability of the TN-40HT storage cask in the upright position on the ISFSI concrete storage pad is demonstrated in Section A3.2. The effects of tornado wind and missiles, flood water and earthquakes are described in Sections A3.2.1, A3.2.2 and A3.2.3, respectively. It is shown in those sections that the cask will not tip over under the most severe natural phenomena specified in the Prairie Island Updated Safety Analysis Report.

The cask is lifted at Prairie Island using a single failure proof crane. The upper trunnions are designed to meet the requirements of NUREG-0612 (Reference 5) for non-redundant lifting fixture. This is accomplished by evaluating the trunnions to the stress design factors required by ANSI N 14.6 (Reference 3), i.e. capable of lifting 6 times and 10 times the cask weight without exceeding the yield and ultimate strengths of the material, respectively. The loaded cask will be handled by the transport vehicle in a vertical orientation and not lifted higher than 18 in.

However section of the SAR considers design events of the third and fourth types (includes accidents) as defined in ANSI/ANS 57.9. The third type of events are those that could reasonably be expected to occur over the lifetime of the ISFSI (does not include dropping of the cask). The fourth type of event includes severe natural phenomena (described in Section A8.2.1 through A8.2.5) and man-induced low probability events postulated because their consequences could result in the maximum potential impact on the immediate environs. Therefore the cask is examined for a dropping accident which is an impact event that is extremely unlikely to occur.

A8.2.8.2 ACCIDENT ANALYSIS In this section the cask is evaluated under bottom end impact on the ISFSI storage pad after a drop from a height of 18 in. The storage pad is the hardest concrete surface outside of the containment building. The cask is always oriented vertically and is never lifted higher than 18 in. once it leaves the containment building. Therefore this case is an upper bound drop event since impact onto a softer surface would result in lower cask deceleration and a lower impact force.

A8.2.8.2.1 DYNAMIC IMPACT LOADS The peak decelerations in the cask and basket during the 18 inch end drop were calculated by a dynamic nonlinear analysis described in Section A4A.10. The analysis showed a maximum acceleration in the TN40HT cask body of 44.1g. This occurred in the bottom plate. The highest acceleration in the basket and fuel was 28.8g. However, since the basket and fuel were not modeled explicitly, the maximum acceleration (28.8g) must be multiplied by the dynamic load factor of 1.52 resulting in a maximum loading of 43.8g.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-7 A8.2.8.2.2 CASK BODY ANALYSIS The cask is analyzed conservatively for a 50 g vertical load simulating the end drop.

The evaluation is presented in Section A4.2.3.4. All calculated stresses meet code allowables.

A8.2.8.2.3 LID BOLT ANALYSIS During a bottom end drop, the rim of the lid is forced against the flange of the cask body. The lid is initially seated against the flange by preloading (torquing) the bolts.

The bolt preload will not be affected if compressive yielding of the contact bearing area does not occur.

The evaluation of the cask presented in Section A4.2.3.4 shows that during a drop cask accident the maximum stresses in the lid outer plate and the shell flange are less than the yield stress. Thus the bolt preload will not be affected by the bottom drop.

Therefore, this hypothetical accident case will not affect the bolts.

A8.2.8.2.4 BASKET ANALYSIS The basket is analyzed conservatively for a 50 g vertical load simulating the end drop.

The evaluation is presented in, Section A4.2.3.4. All calculated stresses meet code allowables.

A8.2.8.3 ACCIDENT DOSE CALCULATIONS Cask drop will not breach the cask confinement barrier. No radioactivity will be released and no resultant doses will occur.

However, a bounding dose can be determined. The loss of neutron shielding (322 mrem from Section A8.2.5) is combined with the total effective dose equivalent (TEDE) from the loss of one confinement barrier and 100% fuel cladding failure (24 mrem from Section A8.2.9). The resulting site boundary accident dose, 346 mrem, is below the 5 rem TEDE limit as specified in 10 CFR 72.106(b) (Reference 2).

A8.2.9 LOSS OF CONFINEMENT BARRIER A8.2.9.1 CAUSE OF ACCIDENT A combined event of failure of one of the seals in addition to a failure of the pressure monitoring system is assessed. This could also be a failure of the pressure boundary of the overpressure system.

A8.2.9.2 ACCIDENT ANALYSIS Analysis has been performed in Appendix A4A to show that the bolts will be able to maintain the seal under accident conditions. Thus the leak rate is limited to the test leak rate of 1 x1 0 -5 ref cm 3/s.

A description of the three possible leaks which could occur is presented below:

. In any of the inner containment seals (lid seal, inner vent seal or inner drain seal

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-8 The lid and lid penetration cover bolts and seals are designed to prevent leakage during all postulated accident events. Therefore, this is a very unlikely event.

In this case the overpressure system, which has a higher pressure than the cask cavity, would leak helium into the cask cavity. Since the pressure is higher in the overpressure tank, it would prevent leakage of radioactive materials out of the cask cavity until the pressure between the overpressure tank and the cask cavity equalized. This would take several years, depending on the size of the leak. At the test leak rate, the overpressure system pressure would always exceed the cask cavity pressure, as shown in Appendix A7A. Therefore no leakage of radioactive material can occur, even if the alarm were to fail. Appendix A7A also demonstrates that even if the inner seal has experienced a latent seal failure there is ample time for identifying the leak through routine surveillances.

. In any of the outer seals (lid, overpressure port cover, vent cover or drain cover)

The lid and lid penetration cover bolts and seals are designed to prevent leakage during all postulated accident events. Therefore, this is a very unlikely event.

In this case, leakage out of the interspace to the atmosphere would occur. This would not result in release of radioactive material from the cask cavity since the inner seal is intact. Again, as demonstrated in Appendix A7A, a latent seal failure of the outer seals would not result in a release of any radioactive material to the environment. There is also ample time for identifying the leak through routine surveillances.

. A leak in the overpressure system This is the most likely cause of a leak.

In this case two scenarios could exist:

The overpressure system is not functioning and the inner seal is intact. In this case there is no release of radioactive material to the environment; or The overpressure system is not functioning and the inner seal is leaking at some rate.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.2-9 In this latter case, leakage out of the interspace to the atmosphere and the cask cavity could occur. This would not result in release of radioactive material from the cask cavity until the pressure fell to the cask cavity pressure.

At the test leak rate of 1 x 1 0 -5 ref cm3/s, this would not occur during a 25 year storage period. However, a leak of this magnitude in combination with a loss of the over pressure system has been evaluated in Appendix A7A.

A8.2.9.3 ACCIDENT DOSE CALCULATIONS The results of the calculations in Appendix A7A assuming accident conditions indicated that at the site boundary (11 0m from the cask), for a 30 day release, the total effective dose equivalent is 24 mrem. The total organ dose equivalent to any individual organ (the critical organ in this case is the bone surface) is 244 mrem for a 30 day release.

The lens dose equivalent to the lens of the eye is 24.1 mrem for a 30 day release.

These values are well below the limiting off site doses defined in 10 CFR 72.106(b).

Another accident condition under consideration is that the overpressure system is not functioning and the inner seal has experienced a latent seal failure. This analysis is presented in Appendix A7A. This accident analysis demonstrates that a latent failure up to 100 times greater that the test value could occur and there is ample time for recovery before the limiting off site doses in 10 CFR 72.106(b) are met. The probability that a gross leak of an inner seal in combination with a gross leak in the outer seal is not considered a credible event.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A8.3-1 A8.3 SITE CHARACTERISTICS AFFECTING SAFETY ANALYSIS Site characteristics have been considered in the formation of the bases for these safety analyses. Conservative assumptions concerning meteorology were used in the determination of c /Q. The characteristics of extreme winds and their contribution to maximum flood level were considered. Regional and site seismology and geology were used to help define the design earthquake acceleration value and analyze for liquefaction potential. Population distribution and other demographic data were used to determine radiation doses.

Other site characteristics affecting safety analyses include the proximity to the Mississippi River to the ISFSI assumptions concerning barge traffic was used to develop the bases for the explosion (Section A8.2.4).

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8.4 REFERENCES

1. American Nuclear Society, ANSI/ANS-57.9, Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type), 1992.

2. 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste.

3. American National Standards Institute, ANSI N 14.6, Special Lifting Devices for Shipping Containers Weighing 10,000 pounds or More, 1986.

4. Structural Design of Concrete Storage Pads for Spent Fuel Casks, EPRI NP-7551, August 1991 by Rashid, Nickel and James.

5. NUREG-612 Control of Heavy Loads at Nuclear Power Plants, July 1980

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.1-1 SECTION A9 CONDUCT OF OPERATIONS A9.1 ORGANIZATIONAL STRUCTURE A9.1.1 CORPORATE ORGANIZATION A9.1.1.1 CORPORATE FUNCTIONS, RESPONSIBILITIES AND AUTHORITIES The information in Section 9.1.1.1 is independent of cask design.

A9.1.1.2 ISFSI PROJECT ORGANIZATION The information in Section 9.1.1.2 is independent of cask design.

A9.1.1.3 RELATIONSHIP WITH CONTRACTORS AND SUPPLIERS The information in Section 9.1.1.3 is independent of cask design.

A9.1.1.4 TECHNICAL STAFF The information in Section 9.1.1.4 is independent of cask design.

A9.1.2 OPERATING ORGANIZATION, MANAGEMENT AND ADMINISTRATIVE CONTROL SYSTEM A9.1.2.1 ONSITE ORGANIZATION The information in Section 9.1.2.1 is independent of cask design.

A9.1.2.2 PERSONNEL FUNCTIONS, RESPONSIBILITIES AND AUTHORITIES The information in Section 9.1.2.2 is independent of cask design.

A9.1.3 PERSONNEL QUALIFICATION REQUIREMENTS A9.1.3.1 MINIMUM QUALIFICATION REQUIREMENTS The information in Section 9.1.3.1 is independent of cask design.

A9.1.3.2 QUALIFICATIONS OF PERSONNEL The information in Section 9.1.3.2 is independent of cask design.

A9.1 .4 LIAISON WITH OUTSIDE ORGANIZATIONS The information in Section 9.1.4 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.2-1 A9.2 STARTUP TESTING AND OPERATION A9.2.1 ADMINISTRATIVE PROCEDURES FOR CONDUCTING TEST PROGRAM The information in Section 9.2.1 is independent of cask design.

A9.2.2 TEST PROGRAM DESCRIPTION A9.2.2.1 PHYSICAL FACILITIES The information in Section 9.2.2.1 is independent of cask design.

A9.2.2.2 OPERATIONS The information in Section 9.2.2.2 is independent of cask design.

A9.2.3 TEST DISCUSSION The information in Section 9.2.3 is independent of cask design.

A9.2.4 COMPLETION OF PRE-OPERATIONAL TEST PROGRAM The information in Section 9.2.4 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.3-1 A9.3 TRAINING PROGRAM The information in Section 9.3 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.4-1 A9.4 NORMAL OPERATIONS A9.4.1 PROCEDURES The information in Section 9.4.1 is independent of cask design.

A9.4.1 .1 ADMINISTRATIVE PROCEDURES The information in Section 9.4.1.1 is independent of cask design.

A9.4.1.2 ANNUNCIATOR RESPONSE GUIDES The information in Section 9.4.1.2 is independent of cask design.

A9.4.1.3 RADIATION PROTECTION PROCEDURES The information in Section 9.4.1.3 is independent of cask design.

A9.4.1.4 MAINTENANCE PROCEDURES The information in Section 9.4.1.4 is independent of cask design.

A9.4.1.5 OPERATING PROCEDURES The information in Section 9.4.1.5 is independent of cask design.

A9.4.1.6 TEST PROCEDURES The information in Section 9.4.1.6 is independent of cask design.

A9.4.1.7 PREOPERATIONAL TEST PROCEDURES The information in Section 9.4.1.7 is independent of cask design.

A9.4.1 .8 QUALITY ASSURANCE PROCEDURES The information in Section 9.4.1.8 is independent of cask design.

A9.4.2 RECORDS The information in Section 9.4.2 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.5-1 A9.5 EMERGENCY PLANNING The information in Section 9.5 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.6-1 A9.6 PHYSICAL SECURITY PLAN The information in Section 9.6 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-1 A9.7 ADDITIONAL FABRICATION TESTING AND INSPECTIONS A9.7.1 CHARPY IMPACT TESTING The base metals for the TN-40HT shield shell and bottom shield shall be subject to Charpy impact testing in accordance with ASME Code (Reference 4) NF-2320 at -20oF during cask fabrication. The acceptance standard shall be a minimum energy absorption of 18 ft-lb.

The weld filler material and Heat Affected Zone (HAZ) shall be subject to Charpy impact testing per ASME Code NF-2431.1(a) through (d), except that:

a) In lieu of the base materials specified for weld test assemblies in the governing weld material specification (SFA), the weld test assemblies for Charpy impact testing shall be prepared using the same base metals that are used for the shield shell and bottom shield.

b) Charpy impact testing shall be performed for both the weld filler material and the heat affected zone of each base metal.

c) The acceptance standard shall be a minimum energy absorption of 18 ft-lb.

A9.7.2 WELDING REQUIREMENTS AND INSPECTIONS Qualification of welding procedures and welders shall be determined using Section IX of the ASME Code, Reference 4.

The ASME Code qualified materials (i.e. containment boundary) used in the construction of the TN-40HT shall be examined following the requirements of ASME Code Section II.

Section V of the ASME Code shall be used in producing Non-destructive examination (NDE) specifications and procedures. NDE requirements for welds are specified on the drawings provided in Chapter A1. Acceptance criteria are as specified by the governing code. NDE personnel shall be qualified in accordance with SNT-TC-1A, Reference 5.

The confinement welds on the TN40HT shall be inspected in accordance with ASME Code Subsection NB including alternatives to ASME Code specified in SAR Section A3.5.

Non-confinement welds shall be inspected in accordance with ASME Code Subsection NF including alternatives to the Code as specified in SAR Section A3.5.

Basket welds shall be inspected to the NDE acceptance criteria of ASME Code Subsection NG as described on the drawings in Section A1. Alternatives to the ASME Code are specified in SAR Section A3.5.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-2 A9.7.3 NEUTRON ABSORBER REQUIREMENTS The neutron absorber used for criticality control in the TN-40HT basket may consist any of the following types of material:

(a) Boron-aluminum alloy (borated aluminum)

(b) Boron carbide / aluminum metal matrix composite (MMC)

(c) Boral The TN-40HT safety analyses do not rely upon the tensile strength of these materials.

The radiation and temperature environment in the cask is not sufficiently severe to damage these metallic/ceramic materials.

To assure performance of the neutron absorbers design function only visual inspections, thermal conductivity testing, and the presence / uniformity of B1 0 need to be verified with testing requirements specific to each material.

References to metal matrix composites throughout this chapter are not intended to refer to borated aluminum or Boral .

A9.7.3.1 BORON ALUMINUM ALLOY (BORATED ALUMINUM)

Description The material is produced by direct chill (DC) or permanent mold casting with boron precipitating as a uniform fine dispersion of discrete aluminum diboride (AlB 2 ) or Titanium diboride (TiB 2 ) particles in the matrix of aluminum or aluminum alloy. For extruded products, the TiB 2 form of the alloy shall be used. For rolled products, the AlB 2 , the TiB 2 , or a hybrid may be used.

Boron is added to the aluminum in the quantity necessary to provide the specified minimum B10 areal density in the final product. The boron may have the natural isotopic distribution or may be enriched in B10.

The criticality calculations take credit for 90% of the minimum specified B10 areal density of borated aluminum. The basis for this credit is the B10 areal density acceptance testing, which shall be as specified in Section A9.7.4.3.

Requirements The boron content in the aluminum or aluminum alloy shall not exceed 5% by weight.

The neutron absorbers shall be 100% visually inspected in accordance with the inspection requirements described in Section A9.7.4.1.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-3 The thermal conductivity of the material shall be tested in accordance with the testing requirements in Section A9.7.4.2.

The minimum B10 areal density specified in Table A3.3-17 shall be confirmed via neutron transmission testing as described in Section A9.7.4.3.

A9.7.3.2 BORON CARBIDE / ALUMINUM METAL MATRIX COMPOSITES (MMC)

Description The material is a composite of fine boron carbide particles in an aluminum or aluminum alloy matrix. The material shall be produced by either direct chill casting, permanent mold casting, powder metallurgy, or thermal spray techniques. It is a low-porosity product, with a metallurgically bonded matrix.

The criticality calculations take credit for 90% of the minimum specified B10 areal density of MMCs. The basis for this credit is the B1 0 areal density acceptance testing, which is specified in Section A9.7.4.3.

Requirements For non-clad MMC products, the boron carbide content shall not exceed 40% by volume. The boron carbide content for MMCs with an integral aluminum cladding shall not exceed 50% by volume.

Non-clad MMC products shall have a density greater than 98% of theoretical density, with no more than 0.5 volume % interconnected porosity. For MMC with an integral cladding, the final density of the core shall be greater than 97% of theoretical density, with no more than 0.5 volume % interconnected porosity of the core and cladding as a unit of the final product.

Boron carbide particles for the products considered here typically have an average size in the range 10-40 microns, although the actual specification may be by mesh size, rather than by average particle size. No more than 10% of the particles shall be over 60 microns.

The neutron absorbers shall be 100% visually inspected in accordance with the inspection requirements described in Section A9.7.4.1.

The thermal conductivity of the material shall be tested in accordance with the testing requirements in Section A9.7.4.2.

The minimum B10 areal density specified in Table A3.3-17 shall be confirmed via neutron transmission testing as described in Section A9.7.4.3.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-4 The MMCs material shall be qualified in accordance with the requirements specified in Section A9.7.5, and shall subsequently be subject to the process controls specified in Section A9.7.6.

A9.7.3.3 BORAL Description This material consists of a core of aluminum and boron carbide powders between two outer layers of aluminum, mechanically bonded by hot-rolling an ingot consisting of an aluminum box filled with blended boron carbide and aluminum powders. The core, which is exposed at the edges of the sheet, is slightly porous. The average size of the boron carbide particles in the finished product is approximately 50 microns after rolling.

The criticality calculations take credit for 75% of the minimum specified B1 0 areal density of Boral .

Requirements The nominal boron carbide content shall be limited to 65% (+ 2% tolerance limit) of the core by weight.

The neutron absorbers shall be 100% visually inspected in accordance with the inspection requirements described in Section A9.7.4.1.

The thermal conductivity of the material shall be tested in accordance with the testing requirements in Section A9.7.4.2.

The minimum B10 areal density specified in Table A3.3-17 shall be confirmed via chemical analysis and by certification of the B1 0 isotopic fraction for the boron carbide powder, or by neutron transmission testing described in Section A9.7.4.3. Areal density testing shall be performed on a coupon taken from the sheet produced from each ingot.

If the measured areal density is below that specified, all the material produced from that ingot will be either rejected, or accepted only on the basis of alternate verification of B1 0 areal density for each of the final pieces produced from that ingot.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-5 A9.7.4 NEUTRON ABSORBERS ACCEPTANCE TESTING A9.7.4.1 VISUAL INSPECTIONS OF NEUTRON ABSORBERS For borated aluminum and MMCs, visual inspections shall follow the recommendations in Aluminum Standards and Data (Reference 6), Chapter 4 Quality Control, Visual Inspection of Aluminum Mill Products and Castings. Local or cosmetic conditions such as scratches, nicks, die lines, inclusions, abrasion, isolated pores, or discoloration are acceptable. Widespread blisters, rough surface, or cracking shall be treated as non-conforming. Inspection of MMCs with an integral aluminum cladding shall also include verification that the matrix is not exposed through the faces of the aluminum cladding and that solid aluminum is not present at the edges.

For Boral, visual inspection shall verify that there are no cracks through the cladding, exposed core on the face of the sheet, or solid aluminum at the edge of the sheet.

A9.7.4.2 THERMAL CONDUCTIVITY TESTING OF NEUTRON ABSORBERS Testing shall conform to ASTM E1225 (Reference 7), ASTM E1461 (Reference 8), or equivalent method, performed at room temperature on coupons taken from the rolled or extruded production material. Previous testing of borated aluminum and metal matrix composite, Table A9.7-1, shows that thermal conductivity increases slightly with temperature. Initial sampling shall be one test per lot, defined by the heat or ingot, and may be reduced if the first five tests meet the specified minimum thermal conductivity.

If a thermal conductivity test result is below the specified minimum, additional tests may be performed on the material from that lot. If the mean value of those tests falls below the specified minimum the associated lot shall be rejected.

After twenty five tests of a single type of material, with the same aluminum alloy matrix, the same boron content, and the boron appearing in the same phase, e.g., B4C, TiB2, or AlB2, if the mean value of all the test results less two standard deviations meets the specified thermal conductivity, no further testing of that material is required. This exemption may also be applied to the same type of material if the matrix of the material changes to a more thermally conductive alloy (e.g., from 6000 to 1000 series aluminum), or if the boron content is reduced without changing the boron phase.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-6 The thermal analysis in Chapter A3.3.2.2 considers a dual plate basket construction base model with 0.125 thick neutron absorber with a 0.312 thick aluminum 1100 plate.

This model gives the bounding values for the maximum component temperatures.

Either a dual plate basket construction or an alternate single plate (borated aluminum or MMC) construction basket may be utilized. For the dual plate construction, the specified thickness of the neutron absorber may vary, and the thermal conductivity acceptance criterion for the neutron absorber will be based on the nominal thickness specified. In either construction type, to maintain the thermal performance of the basket, the minimum thermal conductivity shall be such that the total thermal conductance (sum of conductivity

• thickness) of the neutron absorber and the aluminum 1100 plate shall at least equal the conductance assumed in the analysis for the base model. Samples of the acceptance criteria for various neutron absorber thicknesses are highlighted in Table A9.7-2.

The aluminum 1100 plate does not need to be tested for thermal conductivity; the material may be credited with the values published in the ASME Code Section II part D.

The neutron absorber material need not be tested for thermal conductivity if the nominal thickness of the aluminum 1100 plate is 0.359 inch or greater.

A9.7.4.3 NEUTRON TRANSMISSION TESTING OF NEUTRON ABSORBERS Neutron Transmission acceptance testing procedures shall be subject to approval by Transnuclear . Test coupons shall be removed from the rolled or extruded production material at locations that are systematically or probabilistically distributed throughout the lot. Test coupons shall not exhibit physical defects that would not be acceptable in the finished product, or that would preclude an accurate measurement of the coupons physical thickness.

A lot is defined as all the pieces produced from a single ingot or heat or from a group of billets from the same heat. If this definition results in a lot size too small to provide a meaningful statistical analysis of results, an alternate larger lot definition may be used, so long as it results in accumulating material that is uniform for sampling purposes.

The sampling rate for neutron transmission measurements shall be such that there is at least one neutron transmission measurement for each 2000 square inches of final product in each lot.

The B10 areal density is measured using a collimated thermal neutron beam of up to 1 inch diameter.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-7 The neutron transmission through the test coupons is converted to B10 areal density by comparison with transmission through calibrated standards. These standards are composed of a homogeneous boron compound without other significant neutron absorbers. For example, boron carbide, zirconium diboride or titanium diboride sheets are acceptable standards. These standards are paired with aluminum shims sized to match the effect of neutron scattering by aluminum in the test coupons. Uniform but non-homogeneous materials such as metal matrix composites may be used for standards, provided that testing shows them to provide neutron attenuation equivalent to a homogeneous standard. Standards will be calibrated, traceable to nationally recognized standards, or by attenuation of a monoenergetic neutron beam correlated to the known cross section of boron 10 at that energy.

Alternatively, digital image analysis may be used to compare neutron radioscopic images of the test coupon to images of the standards. The area of image analysis shall be up to 0.75 sq. inch.

The minimum areal density specified shall be verified for each lot at the 95% probability, 95% confidence level or better. If a goodness-of-fit test demonstrates that the sample comes from a normal population, the one-sided tolerance limit for a normal distribution may be used for this purpose. Otherwise, a non-parametric (distribution-free) method of determining the one-sided tolerance limit may be used. Demonstration of the one-sided tolerance limit shall be evaluated for acceptance in accordance with Transnuclear s QA procedures.

The following illustrates one acceptable method and is intended to be utilized as an example. The acceptance criterion for individual plates is determined from a statistical analysis of the test results for their lot. The B1 0 areal densities determined by neutron transmission are converted to volume density, i.e., the B10 areal density is divided by the thickness at the location of the neutron transmission measurement or the maximum thickness of the coupon. The lower tolerance limit of B1 0 volume density is then determined as the mean value of B10 volume density for the sample less K times the standard deviation, where K is the one-sided tolerance limit factor with 95% probability and 95% confidence (Reference 9).

Finally, the minimum specified value of B1 0 areal density is divided by the lower tolerance limit of B1 0 volume density to arrive at the minimum plate thickness which provides the specified B10 areal density.

Any plate which is thinner than the statistically derived minimum thickness or the minimum design thickness, whichever is greater, shall be treated as non-conforming, with the following exception. Local depressions are acceptable, so long as they total no more than 0.5% of the area on any given plate, and the thickness at their location is not less than 90% of the minimum design thickness.

Non-conforming material shall be evaluated for acceptance in accordance with Transnuclear s QA procedures.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-8 A9.7.5 QUALIFICATION TESTING OF METAL MATRIX COMPOSITES A9.7.5.1 APPLICABILITY AND SCOPE Prior to initial use in a spent fuel dry storage system, new MMCs shall be subjected to qualification testing that will verify that the product satisfies the design function. Key process controls shall be identified per Section A9.7.6 so that the production material is equivalent to or better than the qualification test material. Changes to key processes shall be subject to qualification before use of such material in a spent fuel dry storage system.

ASTM methods and practices are referenced below for guidance. Alternative methods may be used with the approval of Transnuclear .

A9.7.5.2 DURABILITY There is no need to include accelerated radiation damage testing in the qualification.

Metals and ceramics do not experience measurable changes in mechanical properties due to fast neutron fluences typical over the lifetime of spent fuel storage.

Thermal damage and corrosion (hydrogen generation) testing shall be performed unless such tests on materials of the same chemical composition have already been performed and found acceptable. The following paragraphs illustrate two cases where such testing is not required.

Thermal damage testing is not required for unclad MMCs consisting only of boron carbide in an aluminum 1100 matrix, because there is no reaction between aluminum and boron carbide below 842 °F (Reference 10), well above the basket temperature under normal conditions of storage or transport.

Corrosion testing is not required for MMCs (clad or unclad) consisting only of boron carbide in an aluminum 1100 matrix, because testing on one such material has already been performed by Transnuclear (Reference 11).

A9.7.5.3 DELAMINATION TESTING OF CLAD MMC Clad MMCs shall be subjected to thermal damage testing following water immersion to ensure that delamination does not occur under normal conditions of storage.

A9.7.5.4 REQUIRED TESTS AND EXAMINATIONS TO DEMONSTRATE MECHANICAL INTEGRITY At least three samples, one each from approximately the two ends and middle of the test material production run shall be subjected to:

a) room temperature tensile testing (ASTM- B557 (Reference 12)) demonstrating that the material:

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-9

• has a 0.2% offset yield strength no less than 1.5 ksi;

• has an ultimate strength no less than 5.0 ksi; and

• has minimum elongation in two inches no less than 0.5%.

As an alternative to the elongation requirement, ductility may be demonstrated by bend testing per ASTM E290 (Reference 13). The radius of the pin or mandrel shall be no greater than three times the material thickness, and the material shall be bent at least 90 degrees without complete fracture.

b) testing by ASTM-B31 1 (Reference 14) to verify more than 98% theoretical density for non-clad MMCs and 97% for the matrix of clad MMCs. Testing or examination for interconnected porosity on the faces and edges of unclad MMC, and on the edges of clad MMC shall be performed by a method to be approved by Transnuclear. The maximum interconnect porosity is 0.5 volume %.

And for at least one sample, c) for MMCs with an integral aluminum cladding, thermal durability testing demonstrating that after a minimum 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> soak in either pure or borated water, then insertion into a preheated oven at approximately 825°F for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the specimens are free of blisters and delamination and pass the mechanical testing requirements described in test a of this section.

A9.7.5.5 REQUIRED TESTS AND EXAMINATIONS TO DEMONSTRATE B10 UNIFORMITY Uniformity of the boron distribution shall be verified either by:

(a) Neutron radioscopy or radiography (ASTM E94 (Reference 15), E142 (Reference 16), and E545 (Reference 17)) of material from the ends and middle of the test material production run, verifying no more than 10% difference between the minimum and maximum B10 areal density, or (b) Quantitative testing for the B10 areal density, B1 0 density, or the boron carbide weight fraction, on locations distributed over the test material production run, verifying that one standard deviation in the sample is less than 10% of the sample mean. Testing may be performed by a neutron transmission method similar to that specified in Section A9.7.4.3, or by chemical analysis for boron carbide content in the composite.

A9.7.5.6 APPROVAL OF PROCEDURES Qualification procedures shall be subject to approval by Transnuclear.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-10 A9.7.6 PROCESS CONTROLS FOR METAL MATRIX COMPOSITES This section provides process controls to ensure that the material delivered for use is equivalent to the qualification test material.

A9.7.6.1 APPLICABILITY AND SCOPE Key processing changes shall be subject to qualification prior to use of the material produced by the revised process. Transnuclear shall determine whether a complete or partial re-qualification program per Section A9.7.5 is required, depending on the characteristics of the material that could be affected by the process change.

A9.7.6.2 DEFINITION OF KEY PROCESS CHANGES Key process changes are those which could adversely affect the uniform distribution of the boron carbide in the aluminum, reduced density, reduced corrosion resistance, or reduce the mechanical strength or ductility of the MMC.

A9.7.6.3 IDENTIFICATION AND CONTROL OF KEY PROCESS CHANGES The manufacturer shall provide Transnuclear with a description of materials and process controls used in producing the MMC. Transnuclear and the manufacturer shall identify key process changes as defined in Section A9.7.6.2.

An increase in nominal boron carbide content over that previously qualified shall always be regarded as a key process change.

The following are examples of other changes that are established as key process changes, as determined by Transnuclears review of the specific applications and production processes:

(a) Changes in the boron carbide particle size specification that increase the average particle size by more than 5 microns, or that increase the amount of particles larger than 60 microns from the previously qualified material by more than 5% of the total distribution but less than the 10% limit; (b) Change of the billet production process, e.g., from vacuum hot pressing to cold isostatic pressing followed by vacuum sintering; (c) Change in the nominal matrix alloy; (d) Changes in mechanical processing that could result in reduced density of the final product, e.g., for powder metallurgy or thermal spray MMCs that were qualified with extruded material, or a change to direct rolling from the billet; (e) For MMCs using a magnesium-alloyed aluminum matrix, changes in the billet formation process that could increase the likelihood of magnesium reaction with the

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-1 1 boron carbide, such as an increase in the maximum temperature or time at maximum temperature; (f) Changes in powder blending or melt stirring processes that could result in less uniform distribution of boron carbide, e.g., change in duration of powder blending; and (g) For MMCs with an integral aluminum cladding, a change greater than 25% in the ratio of the nominal aluminum cladding thickness (sum of two sides of cladding) and the nominal matrix thickness could result in changes in the mechanical properties of the final product.

A9.7.7 RADIAL NEUTRON SHIELDING TESTS The shielding performance of the radial polyester resin can be verified adequately by chemical analysis and verification of density. Uniformity is assured by installation process control.

Testing Requirements Chemical analysis shall be performed on the first batch mixed with a given set of components, and thereafter whenever a new lot of one of the major components is introduced. The acceptance values for the chemical composition of the polyester resin are listed in the following table. Note that the chemical composition used in the shielding models (i.e. listed in Table A7A.4-3) are included in the following table for comparison.

Table A7A.4-3 values Acceptance Testing Values Element nominal wt Element wt % acceptance

__________ _________  % range (wt %)

H 5.05 H 5.05 -10 / +20 B 1.05 B 1.05 +/- 20 C 35.13 C 35.13 +/- 20 Al 14.93 Al 14.93 +/- 20 O 41.73 O+Zn 43.84 +/- 20

_____________ ____________ (balance) ______________ ________________

Zn 2.11 ___________ ___________ ____________

Total 1 00 . 0% 100% ________________

A density measurement shall be performed on every mixed batch of the polyester resin.

The minimum polymer density measured shall be greater than 1.547 g/cm3.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.7-12 Process Controls Qualification tests of the personnel and procedure used for mixing and pouring the polyester resin shall be performed. Qualification testing shall include verification that the chemical composition and density is achieved, and the process is performed in such a manner as to prevent voids.

A9.7.8 FABRICATION LEAK TEST REQUIREMENTS A9.7.8.1 LID, VENT PORT COVER, AND DRAIN PORT COVER The lid, vent port cover, and drain port cover shall be helium leak tested at the fabricator in accordance with ANSI N14.5-1 997, with an acceptance criterion of 1E-7 ref cc/s.

A9.7.8.2 CONFINEMENT SHELL The confinement shell, bottom plate, flange, and associated welds, shall be helium leak tested at the fabricator in accordance with ANSI N14.5-1 997, with an acceptance criterion of 1 E-7 ref cc/s.

A9.7.8.3 LID, VENT, AND DRAIN PORT SEALS The main lid seal along with the vent and drain port seals shall be helium leak tested at the fabricator in accordance with ANSI N14.5-1997. The combined leak rate of all seals including inner and outer seals shall be less than 1 E-5 ref cc/s.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.8-1 A

9.8 REFERENCES

The references 1 through 3 listed in Section 9.7 are independent of cask design. The new references associated with Section A9 are:

4. American Society of Mechanical Engineers, ASME Boiler And Pressure Vessel Code, Sections II, III, V, and IX, 2004 edition including 2006 addenda.

5. SNT-TC-1A, American Society for Nondestructive Testing, Personnel Qualification and Certification in Nondestructive Testing,.

6. Aluminum Standards and Data, 2003 The Aluminum Association.

7 ASTM E1225, Thermal Conductivity of Solids by Means of the Guarded-Comparative-Longitudinal Heat Flow Technique

8. ASTM E1461, Thermal Diffusivity of Solids by the Flash Method

9. Natrella, Experimental Statistics, Dover, 2005.

10. Pyzak and Beaman, Al-B-C Phase Development and Effects on Mechanical Properties of B4C/Al Derived Composites, J. Am. Ceramic Soc., 78[2], 302-312 (1995)

11. Hydrogen Generation Analysis Report for TN-68 Cask Materials, Test Report No. 611 23-99N, Rev 0, Oct 23, 1998, National Technical Systems.

12. ASTM B557, Standard Test Methods of Tension Testing Wrought and Cast Aluminum- and Magnesium-Alloy Products

13. ASTM E290, Standard Test Methods for Bend Testing of Material for Ductility

14. ASTM B311, Test Method for Density Determination for Powder Metallurgy (P/M) Materials Containing Less Than Two Percent Porosity

15. ASTM E94, Recommended Practice for Radiographic Testing

16. ASTM E142, Controlling Quality of Radiographic Testing

17. ASTM E545, Standard Method for Determining Image Quality in Thermal Neutron Radiographic Testing

18. Thermal Conductivity Measurements of Boron Carbide/Aluminum Specimens, Oct 1998, testing by Precision Measurements and Instruments Corp. for Transnuclear, Inc., Purchase Order Number 98037

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A9.8-2

19. Eagle Picher Report AAQR06, Qualification of Thermal Conductivity, Borated Aluminum 1100, May 2001

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Table A9.7-1 THERMAL CONDUCTIVITY FOR SAMPLE NEUTRON ABSORBERS Temperature Material _____________

°C 1 2 3 4 20 193 170 194 194 100 203 183 207 201 200 208 - -

250 - 201 218 206 300 211 204 220 203 314 - - - 202 342 - - - 202 Units: W/mK Materials:

1) Boralyn MMC, aluminum 1100 with 15% B 4 C

2) Borated aluminum 1100, 2.5% boron as TiB 2

3) Borated aluminum 1100, 2.0% boron as TiB 2

4) Borated aluminum 1100, 4.3% boron as AlB 2 Sources:

References 18 and 19

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 TABLE A9.7-2 SAMPLE DETERMINATION OF THERMAL CONDUCTIVITY ACCEPTANCE CRITERION Single Plate Model n Al 1100 absorber total thickness (inch) 0 0.437 0.437 conductivity at 70°F (B in-° n/a 9.11 n/a conductance (Btu/hr-° 0 3.98 3.98*

Dual Plate Construction n Al 1100 absorber total thickness (inch) 0.312 0.125 0.437 as modeled conductivity at 70°F (Btu/h-.in-°F 11.09 4.17 n/a conductance (Btu/hr-°F) 3.46 0.52 3.98 thicker neutron thickness (inch) 0.187 0.250 0.437 absorber conductivity at 70°F (Btu/hr-in-°F) 11.09 7.62 n/a conductance (Btu/hr-°F) 2.07 1.91 3.98 thinner neutron thickness (inch) 0.359 0.078 0.437 absorber conductivity at 70°F (Btu/hr-in-°F) 11 .09 0 n/a conductance (Btu/hr-°F) 3.98 0 3.98 The acceptance criterion is identified by boldface type for each thickness.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A10-1 SECTION A10 OPERATING CONTROLS AND LIMITS The operating controls and limits associated with the TN-40HT cask are located in the Prairie Island ISFSI Technical Specifications.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A11.1-1 SECTION A11 QUALITY ASSURANCE A1 1.1 QUALITY ASSURANCE PROGRAM DESCRIPTION 10 CFR 72.140 requires that a quality assurance (QA) program be established and implemented. The previously approved QA program described in Section 11 will be applied to activities, systems and components of the TN-40HT cask commensurate with their importance to safety. This is the same previously approved NSPM QA Program which satisfies applicable criteria of 10 CFR 50, Appendix B.

Since NSPM is currently licensed under 10 CFR 50 to operate nuclear power facilities, a quality assurance program meeting the requirements of 10 CFR 50, Appendix B, is already in place. The governing document for this program is Northern States Power Company-Minnesota, Quality Assurance Topical Report, (QATR) (Reference 1 in Section 11.3) which has been reviewed and approved by the NRC. The QATR is applied to the important to safety activities associated with the TN-40HT cask, as allowed by 10 CFR 72.140.d. This program is implemented through directives, instructions, and procedures. The objective of the company QATR is to comply with the criteria as expressed in 10 CFR 50, Appendix B, as amended, and with the quality assurance program requirements for nuclear power plants as referenced in the Regulatory Guides and industry standards. This program will be applied to those activities associated with the TN-40HT cask.

Those major components of the TN-40HT cask which are important to safety are listed in Table A4.5-1. As such, the QATR delineates the requirements for engineering, procurement, fabrication, and inspection of this equipment.

The procurement documents (specifications, requisitions, etc.) of the TN-40HT casks will be reviewed technically prior to use to ensure that the proper criteria have been specified. During the cask design phase, vendor information (drawings, specifications, procedures, etc.) will be reviewed to ensure compliance with technical requirements.

During cask fabrication, NSPM representatives will visit the vendors shop to ensure compliance with requirements and to witness parts of the cask fabrication and testing.

Until NSPM is satisfied that the cask meets the technical requirements, the vendor may not ship the cask.

Each of the 18 criteria of 10 CFR 50, Appendix B and their applicability to the TN-40HT storage casks and associated activities are described in Sections A1 1.1.1 through A11.1.18.

A11.1.1 ORGANIZATION The information in Section 11.1.1 is independent of cask design.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A11.1-2 A1 1.1.2 QUALITY ASSURANCE PROGRAM The information in Section 11.1.2 is independent of cask design.

A11.1.3 DESIGN CONTROL The information in Section 11.1.3 is independent of cask design.

A11.1.4 PROCUREMENT DOCUMENT CONTROL The information in Section 11.1.4 is independent of cask design.

A11.1.5 INSTRUCTIONS, PROCEDURES AND DRAWINGS The information in Section 11.1.5 is independent of cask design.

A11.1.6 DOCUMENT CONTROL The information in Section 11.1.6 is independent of cask design.

A11.1.7 CONTROL OF PURCHASED MATERIALS, EQUIPMENT AND SERVICES The information in Section 11.1.7 is independent of cask design.

A11.1.8 IDENTIFICATION AND CONTROL OF MATERIALS, PARTS AND COMPONENTS The information in Section 11.1.8 is independent of cask design.

A11.1.9 CONTROL OF SPECIAL PROCESSES The information in Section 11.1.9 is independent of cask design.

A11.1.10 INSPECTION The information in Section 11.1.10 is independent of cask design.

A11.1.11 TEST CONTROL The information in Section 11.1.11 is independent of cask design.

A11.1.12 CONTROL OF MEASURING AND TEST EQUIPMENT The information in Section 11.1.12 is independent of cask design.

A11.1.13 HANDLING, STORAGE AND SHIPPING The information in Section 11.1.13 is independent of cask design.

A11.1.14 INSPECTION, TEST AND OPERATING STATUS The information in Section 11.1.14 is independent of cask design.

PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A11.1-3 A11.1.15 NON-CONFORMING MATERIALS, PARTS OR COMPONENTS The information in Section 11.1.15 is independent of cask design.

A11.1.16 CORRECTIVE ACTION The information in Section 11.1.16 is independent of cask design.

A11.1.17 QUALITY ASSURANCE RECORDS The information in Section 11.1.17 is independent of cask design.

A11.1.18 AUDITS The information in Section 11.1.18 is independent of cask design.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A11.2-1 A1 1.2 QUALITY ASSURANCE PROGRAM - CONTRACTORS A1 1.2.1 ARCHITECT-ENGINEER The information in Section 11.2.1 is independent of cask design.

A1 1.2.2 CASK SUPPLIER As described in the QATR, NSPM has the ultimate responsibility for ensuring that the manufacture of important to safety components is done in accordance with the plan. In accordance with the plan, the cask manufacturer must do work under the approved NSPM QA Program.

A11.2.3 CONCRETE STORAGE PAD CONTRACTOR The references listed in Section 11.2.3 are independent of cask design and there are no additional references associated with Section A1 1.

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PRAIRIE ISLAND INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Revision: 13 Page A11.3-1 A

11.3 REFERENCES

The references listed in Section 11.3 are independent of cask design and there are no additional references associated with Section A1 1.

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