ML20052D792

Attachment 4 - Proposed FSAR Pages - (Non-Proprietary)

ML20052D792
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Site:	07201032
Issue date:	02/07/2020
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To:	Office of Nuclear Material Safety and Safeguards
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 1-57 1.2.3 Cask Contents This sub-section contains information on the cask contents pursuant to 10 CFR72, paragraphs 72.2(a)(1),(b) and 72.236(a),(c),(h),(m).

The HI-STORM FW System is designed to house both BWR and PWR spent nuclear fuel assemblies. Tables 1.2.1 and 1.2.2 provide key system data and parameters for the MPCs. A description of acceptable fuel assemblies for storage in the MPCs is provided in Section 2.1. This includes fuel assemblies classified as damaged fuel assemblies and fuel debris in accordance with the definitions of these terms in the Glossary. All fuel assemblies, non-fuel hardware, and neutron sources authorized for packaging in the MPCs must meet the fuel specifications provided in Section 2.1. All fuel assemblies classified as damaged fuel or fuel debris must be stored in damaged fuel containers (DFC) or fuel cell storage location equipped with a damaged fuel isolator (DFI) for damaged fuel that can be handled by normal means. Figure 2.1.7 shows a typical DFI.

As shown in Figure 1.2.1a (MPC-37) and Figure 1.2.2 (MPC-89), each storage location is assigned to one of three regions, denoted as Region 1, Region 2, and Region 3 with an associated cell identification number. For example, cell identified as 2-4 is Cell 4 in Region 2. A damaged fuel assembly in a DFC or using a DFI can be stored in the outer peripheral locations of the MPC-37/MPC-32ML and MPC-89 as shown in Figures 2.1.1 and 2.1.2, respectively. The permissible heat loads for each cell, region, and the total canister are given in Tables 1.2.3 and 1.2.4 for MPC-37/MPC-32ML and MPC-89, respectively. The sub-design heat loads for each cell, region and total canister are in Table 4.4.11.

As an alternative to the loading patterns discussed above, fuel storage in the MPC-37 and MPC-89 is permitted to use the heat load patterns shown in Figure 1.2.3 through Figure 1.2.5 (MPC-37) and Figures 1.2.6, 1.2.7 and 1.2.8 (MPC-89).

A minor deviation from the prescribed loading pattern in an MPCs permissible contents to allow one slightly thermally-discrepant fuel assembly per quadrant to be loaded as long as the peak cladding temperature for the MPC remains below the ISG-11 Rev 3 requirements is permitted for essential dry storage campaigns to support decommissioning.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 1 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 1-58 TABLE 1.2.1 KEY SYSTEM DATA FOR HI-STORM FW SYSTEM ITEM QUANTITY NOTES Types of MPCs 3

2 for PWR 1 for BWR MPC storage capacity:

MPC-37 Up to 37 undamaged ZR clad PWR fuel assemblies with or without non-fuel hardware, of classes specified in Table 2.1.1a. Up to 12 damaged fuel containers containing PWR damaged fuel and/or fuel debris may be stored in the locations denoted in Figure 2.1.1a with the remaining basket cells containing undamaged fuel assemblies, up to a total of 37. Alternative damaged fuel patterns are shown in Figures 1.2.3 through 1.2.5.

MPC storage capacity:

MPC-89 Up to 89 undamaged ZR clad BWR fuel assemblies. Up to 16 damaged fuel containers containing BWR damaged fuel and/or fuel debris may be stored in locations denoted in Figure 2.1.2 with the remaining basket cells containing undamaged fuel assemblies, up to a total of 89. Alternative damaged fuel patterns are shown in Figure 1.2.6 through 1.2.8 MPC storage capacity:

MPC-32ML Up to 32 undamaged ZR clad PWR fuel assemblies, of classes specified in Table 2.1.1b.

Up to 8 damaged fuel containers containing PWR damaged fuel and/or fuel debris may be stored in the locations denoted in Figure 2.1.1b with the remaining basket cells containing undamaged fuel assemblies, up to a total of 32.

Damaged fuel assemblies which can be handled by normal means can be stored in the designated locations for damaged fuel using DFIs or DFCs See Chapter 2 for a complete description of authorized cask contents and fuel specifications.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 2 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 1-59 TABLE 1.2.2 KEY PARAMETERS FOR HI-STORM FW MULTI-PURPOSE CANISTERS Parameter PWR BWR Pre-disposal service life (years) 100 100 Design temperature, max./min. (°F) 752/-40 752/-40 Design internal pressure (psig)

Normal conditions Off-normal conditions Accident Conditions 100 120 200 100 120 200 Total heat load, max. (kW)

See Table 1.2.3a/b See Table 1.2.4a/b Maximum permissible peak fuel cladding temperature:

Long Term Normal (°F)

Short Term Operations (°F)

Off-normal and Accident (°F) 752 752 or 1058 1058 752 752 or 1058 1058 Maximum permissible multiplication factor (keff) including all uncertainties and biases

< 0.95

< 0.95 B4C content (by weight) (min.) in the Metamic-HT Neutron Absorber (storage cell walls) 10%

10%

B10 Areal Density, g/cm2 (min.)

0.054 0.036 Basket cell wall thickness, in (min.)

0.57 (MPC-37 & MPC-32ML) 0.38 End closure(s)

Welded Welded Fuel handling Basket cell openings compatible with standard grapples Basket cell openings compatible with standard grapples Heat dissipation Passive Passive Maximum normal condition design temperatures for the MPC fuel basket. A complete listing of design temperatures for all components is provided in Table 2.2.3.

Temperature based on off-normal minimum environmental temperatures specified in Section 2.2.2 and no fuel decay heat load.

See Section 4.5 for discussion of the applicability of the 1058oF temperature limit during short-term operations, including MPC drying.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 3 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 1-64 TABLE 1.2.4b ALTERNATIVE LOADING PATTERNS FOR THE MPC-89 MPC Type Permissible Heat Load Per Storage Cell Maximum Heat Load, kW MPC-89 Undamaged, Damaged, and/or Fuel Debris Figure 1.2.6a 46.2 Figure 1.2.6b 44.92 Figure 1.2.7a 46.14 Figure 1.2.7b 44.98 Figure 1.2.8 48.48 ATTACHMENT 4 TO HOLTEC LETTER 5018075 4 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 1-89 1.665 (D/F) 0.20 1.665 (D/F) 0.12 (D/F) 0.19 0.425 1.44 0.425 0.19 0.12 (D/F) 0.35 (D/F) 0.16 1.06 0.315 0.35 0.315 1.06 0.16 0.35 (D/F) 0.18 1.06 0.265 0.285 0.34 0.285 0.265 1.06 0.18 1.665 (D/F) 0.595 0.315 0.27 0.65 0.165 0.65 0.27 0.315 0.595 1.665 (D/F) 0.20 1.445 0.385 0.34 0.17 0.125 0.17 0.34 0.385 1.445 0.20 1.665 (D/F) 0.595 0.315 0.27 0.65 0.165 0.65 0.27 0.315 0.595 1.665 (D/F) 0.18 1.06 0.265 0.285 0.34 0.285 0.265 1.06 0.18 0.35 (D/F) 0.16 1.06 0.315 0.35 0.315 1.06 0.16 0.35 (D/F) 0.12 (D/F) 0.19 0.425 1.44 0.425 0.19 0.12 (D/F) 1.665 (D/F) 0.20 1.665 (D/F)

Figure 1.2.8 Loading Pattern 89C1 for MPC-89 Containing Undamaged and Damaged Fuel in DFCs/DFIs, and/or Fuel Debris in DFC, per Cell Heat Load Limits (All Storage cell heat loads are in kW, Undamaged Fuel, or Damaged Fuel in DFCs and/or using DFIs, and/or Fuel Debris in a DFC may be stored in cells denoted by D/F.)

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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 2-3 To achieve compliance with the above criteria, certain design and operational changes are necessary, as summarized below.

i. The peak fuel cladding temperature limit (PCT) for long term storage operations and short term operations is generally set at 400ºC (752ºF). However, for MPCs containing all moderate burnup fuel, the fuel cladding temperature limit for short-term operations is set at 570ºC (1058ºF) because the nominal fuel cladding stress is shown to be less than 90 MPa [2.0.2]. Appropriate analyses have been performed as discussed in Chapter 4 and operating restrictions have been added to ensure these limits are met.

ii. A method of drying, such as forced helium dehydration (FHD) is used if the above temperature limits for short-term operations cannot be met.

iii. The off-normal and accident condition PCT limit remains unchanged at 570 ºC (1058ºF).

The MPC cavity is dried, either with FHD or vacuum drying (continuous or cyclic), and then it is backfilled with high purity helium to promote heat transfer and prevent cladding degradation.

The normal condition design temperatures for the stainless steel components in the MPC are provided in Table 2.2.3.

The MPC-37 and MPC-89 models allow for regionalized storage where the basket is segregated into three regions as shown in Figures 1.2.1a and 1.2.2. Decay heat limits for regionalized loading are presented in Tables 1.2.3a and 1.2.4 for MPC-37 and MPC-89 respectively. Specific requirements, such as approved locations for DFCs, DFIs and non-fuel hardware are given in Section 2.1.

As an alternative to the regionalized storage patterns, The MPC-37 and MPC-89 models allow for the use of the heat load charts shown in Figures 1.2.3 through 1.2.5 (MPC-37) and 1.2.6 through 1.2.8 (MPC-89).

Shielding The dose limits for an ISFSI using the HI-STORM FW System are delineated in 10CFR72.104 and 72.106. Compliance with these regulations for any particular array of casks at an ISFSI is necessarily site-specific and must be demonstrated by the licensee. Dose for a single cask and a representative cask array is illustrated in Chapter 5.

The MPC provides axial shielding at the top and bottom ends to maintain occupational exposures ALARA during canister closure and handling operations. The HI-TRAC VW bottom lid also contains shielding. The occupational doses are controlled in accordance with plant-specific procedures and ALARA requirements (discussed in Chapter 9).

ATTACHMENT 4 TO HOLTEC LETTER 5018075 6 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 2-28 2.1.6 Radiological Parameters for Design Basis SNF The principal radiological design criteria for the HI-STORM FW System are the 10CFR72 §104 and §106 operator-controlled boundary dose rate limits, and the requirement to maintain operational dose rates as low as reasonably achievable (ALARA). The radiation dose is directly affected by the gamma and neutron source terms of the assembly, which is a function of the assembly type, and the burnup, enrichment and cooling time of the assemblies. Dose rates are further directly affected by the size and arrangement of the ISFSI, and the specifics of the loading operations. All these parameters are site-dependent, and the compliance with the regulatory dose rate requirements are performed in site-specific calculations. The evaluations here are therefore performed with reference fuel assemblies, and with parameters that result in reasonably conservative dose rates. The reference assemblies given in Table 1.0.4 are the predominant assemblies used in the industry.

The design basis dose rates can be met by a variety of burnup levels and cooling times. Table 2.1.1 provides the acceptable ranges of burnup, enrichment and cooling time for all of the authorized fuel assembly array/classes. Table 2.1.5 and Figures 2.1.3 and 2.1.4 provide the axial distribution for the radiological source terms for PWR and BWR fuel assemblies based on the axial burnup distribution. The axial burnup distributions are representative of fuel assemblies with the design basis burnup levels considered. These distributions are used for analyses only, and do not provide a criterion for fuel assembly acceptability for storage in the HI-STORM FW System.

Non-fuel hardware, as defined in the Glossary, has been evaluated and is also authorized for storage in the PWR MPCs as specified in Table 2.1.1.

2.1.6.1 Radiological Parameters for Spent Fuel and Non-fuel Hardware in MPC-32ML, MPC-37 and MPC-89 MPC-32ML is authorized to store 16x16D spent fuel with burnup - cooling time combinations as given in Table 2.1.9. Spent fuel with burnup - cooling time combinations authorized for storage according to the alternative storage patterns shown in Figures 1.2.3 through 1.2.5 (MPC-37) and 1.2.6 through 1.2.8 (MPC-89) are given in Table 2.1.10.

The burnup and cooling time for every fuel assembly loaded into the MPC-32ML, MPC-37 and MPC-89 must satisfy the following equation:

where, Ct = Minimum cooling time (years),

Bu = Assembly-average burnup (MWd/mtU),

A, B, C, D = Polynomial coefficients listed in Table 2.1.9 or Table 2.1.10 ATTACHMENT 4 TO HOLTEC LETTER 5018075 7 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 2-34 Table 2.1.1a (continued)

MATERIAL TO BE STORED PARAMETER VALUE MPC-37 MPC-89 Other Limitations Quantity is limited to 37 undamaged ZR clad PWR fuel assemblies with or without non-fuel hardware.

Up to 12 damaged fuel containers or damaged fuel isolators containing PWR damaged fuel and/or fuel debris may be stored in the locations denoted in Figure 2.1.1a with the remaining basket cells containing undamaged ZR fuel assemblies, up to a total of

37. Alternative damaged fuel patterns are shown in Figures 1.2.3b, 1.2.3c, 1.2.4b, 1.2.4c, 1.2.5b, and 1.2.5c.

One NSA.

Up to 30 BPRAs.

BPRAs, TPDs, WABAs, water displacement guide tube plugs, orifice rod assemblies, and/or vibration suppressor inserts, with or without ITTRs, may be stored with fuel assemblies in any fuel cell location.

CRAs, RCCAs, CEAs, NSAs, and/or APSRs may be stored with fuel assemblies in fuel cell locations specified in Figure 2.1.5a.

Quantity is limited to 89 undamaged ZR clad BWR fuel assemblies. Up to 16 damaged fuel containers or damaged fuel isolators containing BWR damaged fuel and/or fuel debris may be stored in locations denoted in Figure 2.1.2 with the remaining basket cells containing undamaged ZR fuel assemblies, up to a total of 89.

Alternative damaged fuel patterns are shown in Figures 1.2.6b, 1.2.7b and 1.2.8.

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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 2-53 TABLE 2.1.10 BURNUP AND COOLING TIME FUEL QUALIFICATION REQUIREMENTS FOR MPC-37 AND MPC-89 Cell Decay Heat Load Limit (kW)

Polynomial Coefficients, see Paragraph 2.1.6.1 A

B C

D (Note 1)

MPC-37 0.85 1.68353E-13

-9.65193E-09 2.69692E-04 2.95915E-01 0.85 < decay heat 3.5 1.19409E-14

-1.53990E-09 9.56825E-05

-3.98326E-01 MPC-89 0.32 1.65723E-13

-9.28339E-09 2.57533E-04 3.25897E-01 0.32 < decay heat 0.39 1.61236E-13

-1.24069E-08 3.67868E-04

-7.11229E-01 0.39 < decay heat 0.5 3.97779E-14

-2.80193E-09 1.36784E-04 3.04895E-01 0.5 < decay heat 0.65 2.16783E-14

-1.83483E-09 1.02512E-04 3.41558E-01 0.65 < decay heat 0.75 1.44353E-14

-1.21525E-09 8.14851E-05 3.31914E-01 0.75 < decay heat 1.1

-7.45921E-15 1.09091E-09

-1.14219E-05 9.76224E-01 1.1 < decay heat 1.45 3.10800E-15

-7.92541E-11 1.56566E-05 6.47040E-01 1.45 < decay heat 1.6

-8.08081E-15 1.23810E-09

-3.48196E-05 1.11818E+00 1.6 < decay heat 1.67

-2.69360E-15 6.46465E-10

-1.83165E-05 1.01818E+00 Notes:

1. For BLEU fuel, coefficient D is increased by 1.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 9 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 2-57 3-1 3-2 3-3 3-4 3-5 3-6 2-1 3-7 3-8 3-9 3-10 3-11 2-2 2-3 2-4 2-5 2-6 3-12 3-13 3-14 2-7 2-8 2-9 2-10 2-11 2-12 2-13 3-15 3-16 3-17 2-14 2-15 1-1 1-2 1-3 2-16 2-17 3-18 3-19 3-20 2-18 2-19 2-20 1-4 1-5 1-6 2-21 2-22 2-23 3-21 3-22 3-23 2-24 2-25 1-7 1-8 1-9 2-26 2-27 3-24 3-25 3-26 2-28 2-29 2-30 2-31 2-32 2-33 2-34 3-27 3-28 3-29 2-35 2-36 2-37 2-38 2-39 3-30 3-31 3-32 3-33 3-34 2-40 3-35 3-36 3-37 3-38 3-39 3-40 Figure 2.1.2 Location of DFCs/DFIs for Damaged Fuel or Fuel Debris in the MPC-89 (Shaded Cells)

See Figures 1.2.6, 1.2.7 and 1.2.8 for the locations of DFCs/DFIs for alternative damaged fuel and fuel debris loading patterns.

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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-2 compliance with ISG-11 and with NUREG-1536 guidelines, subject to the exceptions and clarifications discussed in Chapter 1, Table 1.0.3.

As explained in Section 1.2, the storage of SNF in the fuel baskets in the HI-STORM FW system is configured for a three-region storage system under regionalized storage and uniform storage.

Figures 1.2.1a, 1.2.1b and 1.2.2 provide the information on the location of the regions and Tables 1.2.3a, 1.2.3b and 1.2.4a provide the permissible specific heat load (heat load per fuel assembly) in each region for the PWR and BWR MPCs, respectively. The Specific Heat Load (SHL) values under regionalized storage are defined for two patterns that in one case maximizes ALARA (Table 1.2.3a, Pattern A and Table 1.2.4a) and in the other case maximizes heat dissipation (Table 1.2.3a, Pattern B). The ALARA maximized fuel loading is guided by the following considerations:

Region 1: Located in the core region of the basket is permitted to store fuel with medium specific heat load.

Region 2: This is the intermediate region flanked by the core region (Region I) from the inside and the peripheral region (Region III) on the outside. This region has the maximum SHL in the basket.

Region 3: Located in the peripheral region of the basket, this region has the smallest SHL. Because a low SHL means a low radiation dose emitted by the fuel, the low heat emitting fuel around the periphery of the basket serves to block the radiation from the Region II fuel, thus reducing the total quantity of radiation emanating from the MPC in the lateral direction.

Thus, the 3-region arrangement defined above serves to minimize radiation dose from the MPC and peak cladding temperatures mitigated by avoiding placement of hot fuel in the basket core.

To address the needs of cask users having high heat load fuel inventories, fuel loading Pattern B is defined in Table 1.2.3a to maximize heat dissipation by locating hotter fuel in the cold peripheral Region 3 and in this manner minimize cladding temperatures. This has the salutary effect of minimizing core temperature gradients in the radial direction and thermal stresses in the fuel and fuel basket.

As an alternative to the loading patterns discussed above, fuel storage in the MPC-37 and MPC-89 is permitted to use the heat load charts shown in Figures 1.2.3a, 1.2.4a, 1.2.5a (MPC-37) and Figures 1.2.6a, 1.2.7a (MPC-89) or Figures 1.2.3b, 1.2.3c, 1.2.4b, 1.2.4c, 1.2.5b, 1.2.5c (MPC-

37) and Figures 1.2.6b, 1.2.7b, 1.2.8 (MPC-89) for damaged fuel and fuel debris in certain locations.

The salutary consequences of all regionalized loading arrangements become evident from the computed peak cladding temperatures in this chapter, which show margin to the ISG-11 limit discussed earlier.

The safety analyses summarized in this chapter demonstrate acceptable margins to the allowable limits under all design basis loading conditions and operational modes. Minor changes to the ATTACHMENT 4 TO HOLTEC LETTER 5018075 11 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-5 stresses due to restraint on basket periphery thermal growth is eliminated by providing adequate basket-to-canister shell gaps to allow for basket thermal growth during all operational modes.

[

PROPRIETARY INFORMATION WITHHELD PER 10CFR2.390

]

The MPCs uniform & regionalized fuel storage scenarios are defined in Figures 1.2.1a, 1.2.1b and 1.2.2 in Chapter 1 and design maximum decay heat loads for storage of zircaloy clad fuel are listed in Tables 1.2.3a, 1.2.3b, 1.2.3d, 1.2.4a and in Figures 1.2.3a thru 1.2.3c, 1.2.4a thru 1.2.4c, 1.2.5a thru 1.2.5c, 1.2.6a thru 1.2.6b, 1.2.7a thru 1.2.7b and 1.2.8. The axial heat distribution in each fuel assembly is conservatively assumed to be non-uniformly distributed with peaking in the active fuel mid-height region (see axial burnup profiles in Figures 2.1.3 and 2.1.4). Table 4.1.1 summarizes the principal operating parameters of the HI-STORM FW system.

The fuel cladding temperature limits that the HI-STORM FW system is required to meet are discussed in Section 4.3 and given in Table 2.2.3. Additionally, when the MPCs are deployed for storing High Burnup Fuel (HBF) further restrictions during certain fuel loading operations (vacuum drying) are set forth herein to preclude fuel temperatures from exceeding the normal temperature limits. To ensure explicit compliance, a specific term short-term operations is defined in Chapter 2 to cover all fuel loading activities. ISG-11 fuel cladding temperature limits are applied for short-term operations.

The HI-STORM FW system (i.e., HI-STORM FW overpack, HI-TRAC VW transfer cask and MPC) is evaluated under normal storage (HI-STORM FW overpack), during off-normal and accident events and during short-term operations in a HI-TRAC VW. Results of HI-STORM FW thermal analysis during normal (long-term) storage are obtained and reported in Section 4.4.

Results of HI-TRAC VW short-term operations (fuel loading, on-site transfer and vacuum drying) are reported in Section 4.5. Results of off-normal and accident events are reported in Section 4.6.

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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-6 Table 4.1.1 HI-STORM FW OPERATING CONDITION PARAMETERS Condition Value MPC Decay Heat, max.

Tables 1.2.3a, 1.2.3b and 1.24a Figures 1.2.3a thru 1.2.3b, Figures 1.2.4a thru 1.2.4c, Figures 1.2.5a thru 1.2.5c, Figures 1.2.6a thru 1.2.6b and Figures 1.2.7a thru 1.2.7b and Figure 1.2.8.

MPC Operating Pressure Note 1 Normal Ambient Temperature Table 2.2.2 Helium Backfill Pressure Table 4.4.8 Note 1: The MPC operating pressure used in the thermal analysis is based on the minimum helium backfill pressure specified in Table 4.4.8 and MPC cavity average temperature.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 13 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-18 4.4 THERMAL EVALUATION FOR NORMAL CONDITIONS OF STORAGE The HI-STORM FW Storage System (i.e., HI-STORM FW overpack and MPC) and HI-TRAC VW transfer cask thermal evaluation is performed in accordance with the guidelines of NUREG-1536 [4.4.1] and ISG-11 [4.1.4]. To ensure a high level of confidence in the thermal evaluation, 3-dimensional models of the MPC, HI-STORM FW overpack and HI-TRAC VW transfer cask are constructed to evaluate fuel integrity under normal (long-term storage), off-normal and accident conditions and in the HI-TRAC VW transfer cask under short-term operation and hypothetical accidents. The principal features of the thermal models are described in this section for HI-STORM FW and Section 4.5 for HI-TRAC VW. Thermal analyses results for the long-term storage scenarios are obtained and reported in this section. The evaluation addresses the design basis thermal loadings defined in Section 1.2.3. Based on these evaluations the limiting thermal loading condition is defined in Subsection 4.4.4 and adopted for evaluation of on-site transfer in the HI-TRAC (Section 4.5) and off-normal and accident events defined in Section 4.6.

4.4.1 Overview of the Thermal Model As illustrated in the drawings in Section 1.5, the basket is a matrix of interconnected square compartments designed to hold the fuel assemblies in a vertical position under long term storage conditions. The basket is a honeycomb structure of Metamic-HT plates that are slotted and arrayed in an orthogonal configuration to form an integral basket structure. [

PROPRIETARY INFORMATION WITHHELD PER 10CFR2.390

]

Thermal analysis of the HI-STORM FW System is performed for all heat load scenarios defined in Chapter 1 for regionalized storage (Figures 1.2.1a and 1.2.2) and uniform storage (Figures 1.2.1b). Each fuel assembly is assumed to be generating heat at the maximum permissible rate (Tables 1.2.3a, 1.2.3b, 1.2.4a, Figures 1.2.3a thru 1.2.3c, 1.2.4a thru 1.2.4c, 1.2.5a thru 1.2.5c, 1.2.6a thru 1.2.6b, 1.2.7a thru 1.2.7b and 1.2.8). While the assumption of limiting heat generation in each storage cell imputes a certain symmetry to the cask thermal problem, it grossly overstates the total heat duty of the system in most cases because it is unlikely that any basket would be loaded with fuel emitting heat at their limiting values in each storage cell. Thus, the thermal model for the HI-STORM FW system is inherently conservative for real life applications. Other noteworthy features of the thermal analyses are:

i.

While the rate of heat conduction through metals is a relatively weak function of temperature, radiation heat exchange increases rapidly as the fourth power of absolute temperature.

ii.

Heat generation in the MPC is axially non-uniform due to non-uniform axial burnup profiles in the fuel assemblies.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 14 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-24 f) The air flow in the HI-STORM FW/MPC annulus is simulated by the k-turbulence model with the transitional option enabled. The adequacy of this turbulence model is confirmed in the Holtec benchmarking report [4.1.6]. The annulus grid size is selected to ensure a converged solution.(See Section 4.4.1.6).

g) A limited number of fuel assemblies defined in Table 1.2.1 classified as damaged fuel are permitted to be stored in the MPC inside Damaged Fuel Containers (DFCs) or Damaged Fuel Isolators (DFIs). A DFC or DFI can be stored in the outer peripheral locations of MPC-37, MPC-32ML, and MPC-89 as shown in Figures 2.1.1a, 2.1.1b and 2.1.2, respectively. Additionally, a DFC or DFI can be stored in outer peripheral locations or in certain interior locations as shown in Figures 1.2.3a thru 1.2.3c, 1.2.4a thru 1.2.4c, 1.2.5a thru 1.2.5c, 1.2.6a thru 1.2.6b and 1.2.7a thru 1.2.7b and 1.2.8. DFC or DFI emplaced fuel assemblies have a higher resistance to helium flow because of the debris screens. DFC/DFI fuel storage under peripherally permitted scenarios does not affect temperature of hot fuel stored in the core of the basket because DFC DFI storage is located away from hot fuel. For these scenarios the thermal modeling of the fuel basket under the assumption of all storage spaces populated with intact fuel is justified. Interior permitted DFC/DFI storage scenarios are addressed under item m below.

h) As shown in HI-STORM FW drawings in Section 1.5 the HI-STORM FW overpack is equipped with an optional heat shield to protect the inner shell and concrete from radiation heating by the emplaced MPC. The inner and outer shells and concrete are explicitly modeled. All the licensing basis thermal analyses explicitly include the heat shields. A sensitivity study is performed as described in paragraph 4.4.1.9 to evaluate the absence of heat shield on the overpack inner shell and overpack lid.

i) To maximize lateral resistance to heat dissipation in the fuel basket, 0.8 mm full length inter-panel gaps are conservatively assumed to exist at all intersections.

This approach is identical to that used in the thermal analysis of the HI-STAR 180 Package in Docket 71-9325. The shims installed in the MPC peripheral spaces (See MPC-37, MPC-32ML and MPC-89 drawings in Section 1.5) are explicitly modeled. For conservatism bounding as-built gaps are assumed to exist and incorporated in the thermal models.

j) The thermal models incorporate all modes of heat transfer (conduction, convection and radiation) in a conservative manner.

k) The Discrete Ordinates (DO) model, previously utilized in the HI-STAR 180 docket (Docket 71-9325), is deployed to compute radiation heat transfer.

l) Laminar flow conditions are applied in the MPC internal spaces to obtain a lowerbound rate of heat dissipation.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 15 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-25 m) A limited number of fuel assemblies classified as damaged or fuel debris placed in Damaged Fuel Containers (DFCs) or damaged fuel in Damaged Fuel Isolators (DFIs) are permitted to be stored in certain interior locations of MPC-37 and MPC-89 under heat load charts defined in Figures 1.2.3b thru 1.2.3c, 1.2.4b thru 1.2.4c, 1.2.5b thru 1.2.5c, 1.2.6b, 1.2.7b and 1.2.8. These scenarios are evaluated herein.

The 3-D model described above is illustrated in the cross-section for the MPC-89, MPC-32ML and MPC-37 in Figures 4.4.2a, 4.4.2b and 4.4.3, respectively. A closeup of the fuel cell spaces which explicitly include the channel-to-cell gap in the 3-D model applicable to BWR fueled basket (MPC-89) is shown in Figure 4.4.4. The principal 3-D modeling conservatisms are listed below:

1) The storage cell spaces are loaded with high flow resistance design basis fuel assemblies (See Table 2.1.4).

2) Each storage cell is generating heat at its limiting value under the regionalized storage scenarios defined in Chapter 2, Section 2.1.

3) Axial dissipation of heat by conduction in the fuel pellets is neglected.

4) Dissipation of heat from the fuel rods by radiation in the axial direction is neglected.

5) The fuel assembly channel length for BWR fuel is overstated.

6) The most severe environmental factors for long-term normal storage - ambient temperature of 80F and 10CFR71 insolation levels - were coincidentally imposed on the system.

7) Reasonably bounding solar absorbtivity of HI-STORM FW overpack external surfaces is applied to the thermal models.

8) To understate MPC internal convection heat transfer, the helium pressure is understated.

9) No credit is taken for contact between fuel assemblies and the MPC basket wall or between the MPC basket and the basket supports.

10) Heat dissipation by fuel basket peripheral supports is neglected.

11) Conservatively specified fuel basket emissivity in Table 1.2.8 adopted in the thermal analysis.

12) Lowerbound stainless steel emissivity obtained from cited references (See Table 4.2.1) are applied to MPC shell.

13) The k-model used for simulating the HI-STORM FW annulus flow yields uniformly conservative results [4.1.6].

14) Fuel assembly length is conservatively modeled equal to the height of the fuel basket.

The effect of crud resistance on fuel cladding surfaces has been evaluated and found to be negligible [4.1.8]. The evaluation assumes a thick crud layer (130 m) with a bounding low conductivity (conductivity of helium). The crud resistance increases the clad temperature by a very small amount (~0.1oF) [4.1.8]. Accordingly this effect is neglected in the thermal evaluations.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 16 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-30 (vii) MPC-37 under heat load Figures 1.2.3a/b/c, 1.2.4a/b/c and 1.2.5a/b/c (viii) MPC-89 under heat load Figures 1.2.7a, 1.2.7b and 1.2.8.

To evaluate the above scenarios, 3D FLUENT screening models of the HI-STORM FW cask are constructed, Peak Cladding Temperatures (PCT) computed and tabulated in Table 4.4.2. The results of the calculations yield the following:

(a) Fuel storage in MPC-37 produces a higher peak cladding temperature than that in MPC-89 (b) Fuel storage in the minimum height MPC-37 is limiting (produces the highest peak cladding temperature).

To bound the HI-STORM FW storage temperatures the limiting scenario ascertained above is adopted for evaluation of all normal, off-normal and accident conditions.

4.4.1.6 Grid Sensitivity Studies To achieve grid independent CFD results, a grid sensitivity study is performed on the HI-STORM FW thermal model. The grid refinement is performed in the entire domain i.e. for both fluid and solid regions in both axial and radial directions. Non-uniform meshes with grid cells clustered near the wall regions are generated to resolve the boundary flow near the walls.

A number of grids are generated to study the effect of mesh refinement on the fuel and component temperatures. All sensitivity analyses were carried out for the case of MPC-37 with minimum fuel length under the bounding heat load pattern A. Following table gives a brief summary of the different sets of grids evaluated and PCT results.

Mesh No Total Mesh Size PCT (oC)

Permissible Limit (oC)

Clad Temperature Margin (oC) 1 (Licensing Basis Mesh) 1,536,882 373 400 27 2

3,354,908 372 400 28 3

7,315,556 372 400 28 Note: Because the flow field in the annulus between MPC shell and overpack inner shell is in the transitional turbulent regime, the value of y+ at the wall-adjacent cell is maintained on the order of 1 to ensure the adequate level of mesh refinement is reached to resolve the viscosity affected region near the wall.

As can be seen from the above table, the PCT is essentially the same for all the meshes. The solutions from the different grids used are in the asymptotic range. Therefore, it can be concluded that the Mesh 1 is reasonably converged. To provide further assurance of convergence, the sensitivity results are evaluated in accordance with the ASME V&V 20-2009 ATTACHMENT 4 TO HOLTEC LETTER 5018075 17 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-36 The following observations can be derived by inspecting the temperature field obtained from the thermal models:

The fuel cladding temperatures are below the regulatory limit (ISG-11 [4.1.4]) under all uniform and regionalized storage scenarios defined in Chapter 1 (Figures 1.2.1a, 1.2.1b, and 1.2.2) and thermal loading scenarios defined in Tables 1.2.3a, 1.2.3b, 1.2. 4a, Figures 1.2.3a/b/c, 1.2.4a/b/c, 1.2.5a/b/c, 1.2.6a/b, 1.2.7a/b and 1.2.8.

The limiting fuel temperatures are reached under the Pattern A thermal loading condition defined in Table 1.2.3a in the MPC-37. Accordingly, this scenario is adopted for thermal evaluation under on-site transfer (Section 4.5) and under off-normal and accident conditions (Section 4.6).

The maximum temperature of the basket structural material is within its design limit.

The maximum temperatures of the MPC pressure boundary materials are below their design limits.

The maximum temperatures of concrete are within the guidance of the governing ACI Code (see Table 2.2.3).

The calculated fuel temperature for the 15x15I short fuel assembly (Table 4.4.12) is bounded by the thermal evaluations for the minimum MPC-37 for short fuel (Table 4.4.3). The temperatures of other cask components are similar. It is reasonable to conclude that the temperatures and pressure for the minimum height MPC-37 (short fuel) bounds all scenarios.

The above observations lead us to conclude that the temperature field in the HI-STORM FW System with a loaded MPC containing heat emitting SNF complies with all regulatory temperature limits (Table 2.2.3). In other words, the thermal environment in the HI-STORM FW System is in compliance with Chapter 2 Design Criteria.

Also, all the licensing basis thermal evaluations documented in this chapter are performed for the most limiting thermal scenarios i.e. minimum MPC-37 with heat load pattern A.

4.4.4.2 Minimum Temperatures In Table 2.2.2 of this report, the minimum ambient temperature condition for the HI-STORM FW storage overpack and MPC is specified to be -40F. If, conservatively, a zero decay heat load with no solar input is applied to the stored fuel assemblies, then every component of the system at steady state would be at a temperature of -40F. Low service temperature (-40oF) evaluation of the HI-STORM FW is provided in Chapter 3. All HI-STORM FW storage ATTACHMENT 4 TO HOLTEC LETTER 5018075 18 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-37 overpack and MPC materials of construction will satisfactorily perform their intended function in the storage mode under this minimum temperature condition.

4.4.4.3 Effect of Elevation The reduced ambient pressure at site elevations significantly above the sea level will act to reduce the ventilation air mass flow, resulting in a net elevation of the peak cladding temperature. However, the ambient temperature (i.e., temperature of the feed air entering the overpack) also drops with the increase in elevation. Because the peak cladding temperature also depends on the feed air temperature (the effect is one-for-one within a small range, i.e., 1F drop in the feed air temperature results in ~1F drop in the peak cladding temperature), the adverse ambient pressure effect of increased elevation is partially offset by the ambient air temperature decrease. The table below illustrates the variation of air pressure and corresponding ambient temperature as a function of elevation.

Elevation (ft)

Pressure (psia)

Ambient Temperature Reduction versus Sea Level Sea Level (0) 14.70 0oF 2000 13.66 7.1oF 4000 12.69 14.3oF A survey of the elevation of nuclear plants in the U.S. shows that nuclear plants are situated near about sea level or elevated slightly (~1000 ft). The effect of the elevation on peak fuel cladding temperatures is evaluated by performing calculations for a HI-STORM FW system situated at an elevation of 1500 feet. At this elevation the ambient temperature would decrease by approximately 5oF (See Table above). The peak cladding temperatures are calculated under the reduced ambient temperature and pressure at 1500 feet elevation for the limiting regionalized storage scenario evaluated in Table 4.4.2. The results are presented in Table 4.4.9.

These results show that the PCT, including the effects of site elevation, continues to be well below the regulatory cladding temperature limit of 752oF. In light of the above evaluation, it is not necessary to place ISFSI elevation constraints for HI-STORM FW deployment at elevations up to 1500 feet. If, however, an ISFSI is sited at an elevation greater than 1500 feet, the effect of altitude on the PCT shall be quantified as part of the 10 CFR 72.212 evaluation for the site using the site ambient conditions.

4.4.4.4 Evaluation of Overpack Heat Shields As discussed in Sub-section 4.4.1.9 above, a thermal evaluation is performed to evaluate the effect of removal of heat shields from a HI-STORM overpack. The predicted temperatures from this sensitivity study of normal condition of storage are summarized in Table 4.4.14. The peak ATTACHMENT 4 TO HOLTEC LETTER 5018075 19 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-45 Table 4.4.2 RESULTS OF SCREENING CALCULATIONS UNDER NORMAL STORAGE CONDITIONS Storage Scenario Peak Cladding Temperature, oC (oF)

MPC-37 (Note 2)

- regionalized storage Table 1.2.3a Minimum Height1 Reference Height Maximum Height MPC-37 (Note 4)

- heat load Figure 1.2.3a

- heat load Figure 1.2.4a

- heat load Figure 1.2.5a

- heat load Figure 1.2.3bNotes 5,6,7 353 (667) 342 (648) 316 (601) 371 (700) 368 (694) 367 (693) 364 (687)

MPC-32ML (Note 3) 349 (660)

MPC-89 (Note 2)

- regionalized storage Table 1.2.4a MPC-89 (Note 4)

- heat load Figure 1.2.6a

- heat load Figure 1.2.6bNote 5, 7

- heat load Figure 1.2.7a

- heat load Figure 1.2.7bNote 5, 7

- heat load Figure 1.2.8Note 5, 7 333 (631) 366 (691) 360 (680) 365 (689) 358 (676) 370 (698)

Notes:

(1) The highest temperature highlighted above is reached under the case of minimum height MPC-37 designed to store the short height Ft. Calhoun 14x14 fuel. This scenario is adopted in Chapter 4 for the licensing basis evaluation of fuel storage in the HI-STORM FW system. See Note 4.

(2) All the screening calculations for MPC-37 and MPC-89 were performed using a reference coarse mesh

[4.1.9] and flow resistance based on the calculations in Holtec report [4.4.2].

(3) Screening calculations for MPC-32ML performed using a mesh with similar density as the licensing basis converged mesh adopted for MPC-37 in Section 4.4.1.6.

(4) Screening evaluation used the same mesh as licensing basis mesh adopted in Section 4.4.1.6. The computed temperatures are bounded by the licensing basis minimum height temperatures tabulated in Table 4.4.3.

(5) PCT of intact fuel assemblies in the loading patterns with fuel debris in the DFCs is bounded by that with damaged fuel in the DFCs as justified next. It is conservatively assumed that the damaged fuel assemblies inside DFCs/DFIs have the same axial heat distribution as the intact fuel assemblies to maximize the PCT of intact fuel assemblies. Fuel debris consistent with its physical condition is modeled as packed towards bottom of the DFCs. This yields less impact on the PCT of intact fuel assemblies.

(6) The computed temperature under short length Damaged Fuel Storage is bounded by undamaged fuel temperatures computed above in heat load Figure 1.2.3a. This reasonably supports the conclusion that Damaged Fuel Storage under standard and long fuel storage in Figures 1.2.4b/c, 1.2.5b/c is bounded by undamaged fuel heat load scenarios evaluated in Figure 1.2.4a and 1.2.5a above.

(7) Peak temperatures including damaged fuel in DFC/DFI tabulated herein.

1 Bounding scenario adopted in this Chapter for all thermal evaluations.

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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-47 Table 4.4.4 MINIMUM MPC FREE VOLUMES Item Lowerbound Height MPC-37 (ft3)

MPC-89 (ft3)

Net Free Volume*

211.89 203.58 MPC-32ML (ft3)

Net Free Volume*

278.7

• Net free volumes are obtained by subtracting basket, fuel, aluminum shims, spacers, basket supports and DFCs metal volume from the MPC cavity volume.

Table 4.4.5

SUMMARY

OF MPC INTERNAL PRESSURES UNDER LONG-TERM STORAGE*

Condition MPC-37 ***

(psig)

Pattern A/Pattern B MPC-89***

(psig)

Initial maximum backfill** (at 70F) 45.5/46.0 47.5 Normal:

intact rods 1% rods rupture 96.6/97.9 97.7/99.0 98.4 99.0 Off-Normal (10% rods rupture) 107.5/108.9 104.0 Accident (100% rods rupture) 191.5/194.4 156.9

• Per NUREG-1536, pressure analyses with ruptured fuel rods (including BPRA rods for PWR fuel) is performed with release of 100% of the ruptured fuel rod fill gas and 30% of the significant radioactive gaseous fission products.

• • Conservatively assumed at the Tech. Spec. maximum value (see Table 4.4.8).

• • • Tabulated pressures bound storage under heat load Figures 1.2.3a/b/c, 1.2.4a/b/c, 1.2.5a/b/c, 1.2.6a/b, 1.2.7a/b and 1.2.8.

(continued next page)

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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-50 Table 4.4.7 THEORETICAL LIMITS* OF MPC HELIUM BACKFILL PRESSURE**

MPC Minimum Backfill Pressure (psig)

Maximum Backfill Pressure (psig)

MPC-37 Pattern A 41.0 47.3 MPC-37 Pattern B 40.8 47.1 MPC-37 Figures 1.2.3a Figures 1.2.4a Figure 1.2.5a 43.9 43.6 44.1 50.6 50.3 50.8 MPC-89 Table 1.2.4a Figure 1.2.6a Figure 1.2.7a Figure 1.2.8 41.9 41.7 41.8 41.1 48.4 48.2 48.3 47.5 MPC-32ML 39.7 50.6

• The helium backfill pressures are set forth in the Technical Specifications with a margin (see Table 4.4.8).

• • The pressures tabulated herein are at 70oF reference gas temperature.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 22 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-51 Table 4.4.8 MPC HELIUM BACKFILL PRESSURE SPECIFICATIONS MPC Item Specification MPC-37 Pattern A Minimum Pressure 42.0 psig @ 70oF Reference Temperature Maximum Pressure 45.5 psig @ 70oF Reference Temperature MPC-37 Pattern B Minimum Pressure 41.0 psig @ 70oF Reference Temperature Maximum Pressure 46.0 psig @ 70oF Reference Temperature MPC-89 Table 1.2.4a Minimum Pressure 42.5 psig @ 70oF Reference Temperature Maximum Pressure 47.5 psig @ 70oF Reference Temperature MPC-32ML Minimum Pressure 41.5 psig @ 70oF Reference Temperature Maximum Pressure 45.5 psig @ 70oF Reference Temperature MPC-37 Figures 1.2.3a/b/c Minimum Pressure 45.5 psig @ 70oF Reference Temperature Maximum Pressure 49.0 psig @ 70oF Reference Temperature MPC-37 Figures 1.2.4a/b/c Minimum Pressure 44.0 psig @ 70oF Reference Temperature Maximum Pressure 47.5 psig @ 70oF Reference Temperature MPC-37 Figures 1.2.5a/b/c Minimum Pressure 44.5 psig @ 70oF Reference Temperature Maximum Pressure 48.0 psig @ 70oF Reference Temperature MPC-89 Figures 1.2.6a/b Minimum Pressure 42.0 psig @ 70oF Reference Temperature Maximum Pressure 47.0 psig @ 70oF Reference Temperature MPC-89 Figures 1.2.7a/b Minimum Pressure 42.0 psig @ 70oF Reference Temperature Maximum Pressure 47.0 psig @ 70oF Reference Temperature MPC-89 Figure 1.2.8 Minimum Pressure 42.0 psig @ 70oF Reference Temperature Maximum Pressure 46.0 psig @ 70oF Reference Temperature ATTACHMENT 4 TO HOLTEC LETTER 5018075 23 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-58 Table 4.4.15 DESIGN OPERATING ABSOLUTE PRESSURESNote 1 MPC-37 Loading Pattern A Loading Pattern B 7.1 atm 7 atm MPC-32ML 6.5 atm MPC-89 Table 1.2.4a 7 atm MPC-37 load Figure 1.2.3a 7.0 atm MPC-37 heat load Figure 1.2.4a 6.9 atm MPC-37 heat load Figure 1.2.5a 6.8 atm MPC-89 heat load Figure 1.2.6a 6.8 atm MPC-89 heat load Figure 1.2.7a 6.8 atm MPC-89 heat load Figure 1.2.8 7.0 atm Note 1: Table 4.4.8 helium backfill specifications ensure MPC operating pressures meet or exceed design values tabulated herein.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 24 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-97 Table 4.5.19 MPC DRYING OPERATIONS MPC Type Fuel Heat Load Limit (kW)

Method of Drying MPC-32ML MBF 44.16 (Note 1)

FHD/Vacuum Drying without Time Limit HBF 44.16 (Note 1)

FHD/Vacuum Drying with Time Limit 28.704 FHD/Vacuum Drying without Time Limit MPC-37 MBF 44.09 (Pattern A) 45.0 (Pattern B) 37.4/39.95/44.85 (Figures 1.2.3a, 1.2.4a, 1.2.5a) 34.4 (Figures 1.2.3b/c) 36.65 (Figures 1.2.4b/c) 40.95 (Figures 1.2.5b/c)

(Note 1)

FHD/Vacuum Drying without Time Limit HBF 44.09 (Pattern A) 45.0 (Pattern B)

(Note 1)

FHD/Vacuum Drying with Time Limit 29.6 FHD/Vacuum Drying without Time Limit MPC-89 MBF 46.36 (Table 1.2.4a) 46.2 (Figure 1.2.6a) 44.92 (Figure 1.2.6b) 46.14 (Figure 1.2.7a) 44.98 (Figure 1.2.7b)

(Note 1) 48.48 (Figure 1.2.8)

FHD/ Vacuum Drying without Time Limit HBF 46.36 (Note 1) 48.48 (Figure 1.2.8)

FHD/Vacuum Drying with Time Limit 30 FHD/Vacuum Drying without Time Limit Note 1: Design Basis heat load.

Note 2:Cyclic drying under time limited vacuum drying operations is permitted in accordance with ISG-11, Rev. 3 requirements by limiting number of cycles to less than 10 and cladding temperature variations to less than 65oC (117oF). Suitable time limits for these cycles shall be evaluated based on site specific conditions and thermal methodology defined in Section 4.5.

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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-98 Table 4.5.20 MAXIMUM TEMPERATURES OF MPC-37 DURING VACUUM DRYING CONDITIONS UNDER SHORT FUEL HEAT LOAD FIGURE 1.2.3a (Note 3)

Component Temperature oC (oF)Note 1 Fuel Cladding 465 (869)

MPC Basket 412 (774)

Basket Periphery 335 (635)

Aluminum Basket Shims 272 (522)

MPC Shell 165 (329)

MPC LidNote 2 100 (212)

Note 1: Addresses vacuum drying of Moderate Burnup Fuel Note 2: Section temperature reported Note 3: Bounding scenario of short fuel from Table 4.4.2 is evaluated.

Table 4.5.21 MAXIMUM TEMPERATURES OF MPC-89 DURING VACUUM DRYING CONDITIONS UNDER HEAT LOAD FIGURE 1.2.6a (Note 3)

Component Temperature oC (oF)Note 1 Fuel Cladding 469 (876)

MPC Basket 449 (840)

Basket Periphery 362 (684)

Aluminum Basket Shims 305 (581)

MPC Shell 181 (358)

MPC LidNote 2 119 (246)

Note 1: Addresses vacuum drying of Moderate Burnup Fuel Note 2: Section temperature reported Note 3: Bounds heat load Figure 1.2.7a.

Note 4: The difference in PCT (Table 4.4.2) between the heat load pattern in Figures 1.2.6a and 1.2.8 is significantly lower than the margins computed above. Therefore, vacuum drying of MBF is permitted for the loading pattern Figure 1.2.8 also per the evaluation presented above.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 26 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 4-124

4.8 REFERENCES

[4.1.1] Deleted

[4.1.2] FLUENT Computational Fluid Dynamics Software, Fluent, Inc., Centerra Resource Park, 10 Cavendish Court, Lebanon, NH 03766.

[4.1.3] The TN-24P PWR Spent-Fuel Storage Cask: Testing and Analyses, EPRI NP-5128, (April 1987).

[4.1.4] Cladding Considerations for the Transportation and Storage of Spent Fuel, Interim Staff Guidance - 11, Revision 3, USNRC, Washington, DC.

[4.1.5] Topical Report on the HI-STAR/HI-STORM Thermal Model and its Benchmarking with Full-Size Cask Test Data, Holtec Report HI-992252, Revision 1, Holtec International, Marlton, NJ, 08053.

[4.1.6] Identifying the Appropriate Convection Correlation in FLUENT for Ventilation Air Flow in the HI-STORM System, Revision 1, Holtec Report HI-2043258, Holtec International, Marlton, NJ, 08053.

[4.1.7] Performance Testing and Analyses of the VSC-17 Ventilated Concrete Cask, EPRI TR-100305, (May 1992).

[4.1.8] Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System, Holtec Report No. 2002444, Revision 7, NRC Docket No. 72-1014.

[4.1.9] Thermal Evaluation of HI-STORM FW, Holtec Report HI-2094400, Latest Revision.

[4.1.10] Not Used

[4.1.11] Safety Analysis Report on the HI-STAR 180 Package, Holtec Report HI-2073681, Latest Revision.

[4.1.12] Effective Properties of PWR and BWR Fuel assemblies to Support Thermal Evaluations of HI-STORM FW, HI-2167474, Rev. 0.

[4.2.1] Baumeister, T., Avallone, E.A. and Baumeister III, T., Marks Standard Handbook for Mechanical Engineers, 8th Edition, McGraw Hill Book Company, (1978).

[4.2.2] Rohsenow, W.M. and Hartnett, J.P., Handbook of Heat Transfer, McGraw Hill Book Company, New York, (1973).

[4.2.3] Creer et al., The TN-24P Spent Fuel Storage Cask: Testing and Analyses, EPRI NP-ATTACHMENT 4 TO HOLTEC LETTER 5018075 27 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-6 10CFR72 contains two sections that set down main dose rate requirements: §104 for normal and off-normal conditions, and §106 for accident conditions. The relationship of these requirements to the analyses in this Chapter 5, and the burnup and cooling times selected for the various analyses, are as follows:

10CFR72.104 specifies the dose limits from an ISFSI (and other operations) at a site boundary under normal and off-normal conditions. Compliance with §104 can therefore only be demonstrated on a site-specific basis, since it depends not only on the design of the cask system and the loaded fuel, but also on the ISFSI layout, the distance to the site boundary, and possibly other factors such as use of higher density concrete or the terrain around the ISFSI. The purpose of this chapter is therefore to present a general overview over the expected or maximum dose rates, next to the casks and at various distances, to aid the user in applying ALARA considerations and planning of the ISFSI.

For the accident dose limit in 10CFR72.106 it is desirable to show compliance in this Chapter 5 on a generic basis, so that calculations on a site-by-site basis are not required. To that extent, a burnup and cooling time calculation that maximizes the dose rate under accident conditions needs to be selected.

It is recognized that for a given heat load, an infinite number of burnup and cooling time combinations could be selected, which would result in slightly different dose rate distributions around the cask. For a high burnup with a corresponding longer cooling time, dose locations with a high neutron contribution would show higher dose values, due to the non-linear relationship between burnup and neutron source term. At other locations dose rates are more dominated by contribution from the gamma sources. In these cases, short cooling time and lower burnup combinations with heat load comparable to the higher burnup and corresponding longer cooling time combinations would result in higher dose rates. However, in those cases, there would always be a compensatory effect, since for each dose location, higher neutron dose rates would be partly offset by lower gamma dose rates and vice versa. This is further complicated by the regionalized loading patterns qualified from a thermal perspective and shown in Figure 1.2.3 through Figure 1.2.5 for MPC-37 and Figures 1.2.6 through 1.2.8 for MPC-89. These contain cells with substantially different heat load limits, and hence substantially different ranges of burnup, enrichment and cooling time combinations. The approach to cover all those variations in a conservative way is outlined below.

To prescribe radiological limits for the fuel to be loaded, loading curves are defined in Tables 2.1.9 and 2.1.10, where a loading curve specifies the minimum cooling time as a function of fuel burnup. Different loading curves are defined for the different heat load limits, so that the thermal and radiological requirements for the fuel in each cell are approximately aligned.

However, it should be noted that thermal and radiological limits for each assembly are applied completely independent from each other. The uniform and regionalized loading curves for the fuel to be loaded in the MPC-37, MPC-32ML or MPC-89 canisters are discussed in Subsection 5.2.7.

As it is discussed in Subsection 5.1.2, a site-specific shielding evaluation may be required for accident-condition of MPC-32ML.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 28 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-7 To determine dose rates consistent with both the uniform and regionalized thermal loading, it is necessary to consider the ranges of burnup and cooling times from all loading curves. For that, 8 burnup values between 5 and 70 GWd/mtU are selected, and corresponding minimum required cooling times are established and used in the dose analyzes. The heat load patterns in Figures 1.2.3 through 1.2.8 contain from 5 to 20 regions each, i.e. from 5 to 20 principal locations with different heat load limit. Applying 8 burnup and cooling time combinations to each location would result up to 820 = 1.15E+18 different burnup and cooling time loading arrangements per pattern. Analyzing and comparing those many arrangements would be excessive. Therefore, for the radiological evaluations, some regions and loading patterns (MPC-37) are combined using the highest heat load limit (source term) of each group. For MPC-37, the heat loads for each cell are based on the Long fuel heat loads in Figure 1.2.5a.

The established bounding heat load limits are provided in Tables 5.0.3 and 5.0.4.

This then results in effectively only 2 or 5 regions to be independently varied for the considered bounding MPC-37 and MPC-89 patterns, and hence 82 = 64 or 85 = 32,768 different burnup and cooling time arrangements per pattern is to be analyzed, which is manageable. The selected burnup, enrichment and cooling time combinations for the uniform and regionalized loading patterns are listed in Tables 5.0.3, 5.0.4a, 5.0.4b, 5.0.4c and 5.0.5. The dose rates in the various important locations are calculated for each of these combination arrangements and the maximum is determined for each dose rate location. It should be noted that this maximum can be from a different loading arrangement in different locations.

Based on this approach, the source terms used in the analyses of MPC-37, MPC-32ML or MPC-89 are reasonably bounding for all realistically expected assemblies. All dose rates in this chapter are developed using this approach, unless noted otherwise. Also, as discussed in Section 5.2, the design basis BPRA activities are considered for MPC-37 and MPC-32ML in this chapter, unless noted otherwise.

All dose rates in Section 5.1 are developed using the approach discussed above. Some dose rates in Section 5.4 were retained from previous versions of the FSAR and that are based on a representative (while still conservative) uniform loading pattern, as discussed in that Section.

It should also be noted that most results provided in this chapter are based on conservative assumptions (such as the minimum concrete density for HI-STORM FW, and the minimum lead and neutron absorber thicknesses for HI-TRAC VW), with the hypothetical upper bound source terms that cover the worst possible allowed MPC content (except when it is explicitly mentioned). Such a hypothetical loading scenario is not considered credible and significantly lower dose rates are expected on a site-specific basis.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 29 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-14 Table 5.0.4c SELECTED BURNUP, ENRICHMENT, COOLING TIME COMBINATIONS FOR THE MPC-89 LOADING PATTERNS BASED ON FIGURE 1.2.8 AND TABLE 2.1.10 Region Burnup (MWd/mtU)

Enrichment (wt% 235U)

Cooling Time (years)

Reference Decay Heat (kW)

Low Heat Load Basket Regions (Region 1) 5000 0.7 1.4 0.32 10000 0.9 2.0 20000 1.6 3.0 30000 2.4 4.0 40000 3.0 6.0 50000 3.3 10.0 60000 3.7 18.0 70000 4.0 29.0 Low Heat Load Basket Regions (Region 2) 5000 0.7 1.0 0.39 10000 0.9 1.8 20000 1.6 2.8 30000 2.4 3.5 40000 3.0 4.0 50000 3.3 6.0 60000 3.7 11.0 70000 4.0 19.0 Low Heat Load Basket Regions (Region 3) 5000 0.7 1.0 0.65 10000 0.9 1.2 20000 1.6 1.8 30000 2.4 2.2 40000 3.0 2.8 50000 3.3 3.5 60000 3.7 4.5 70000 4.0 5.0 ATTACHMENT 4 TO HOLTEC LETTER 5018075 30 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-15 Table 5.0.4c (continued)

SELECTED BURNUP, ENRICHMENT, COOLING TIME COMBINATIONS FOR THE MPC-89 LOADING PATTERNS BASED ON FIGURE 1.2.8 AND TABLE 2.1.10 Region Burnup (MWd/mtU)

Enrichment (wt% 235U)

Cooling Time (years)

Reference Decay Heat (kW)

High Heat Load Basket Regions (Region 4) 5000 0.7 1.0 1.1 10000 0.9 1.0 20000 1.6 1.0 30000 2.4 1.4 40000 3.0 1.6 50000 3.3 2.2 60000 3.7 2.6 70000 4.0 2.8 High Heat Load Basket Regions (Region 5) 5000 0.7 1.0 1.67 10000 0.9 1.0 20000 1.6 1.0 30000 2.4 1.0 40000 3.0 1.0 50000 3.3 1.2 60000 3.7 1.6 70000 4.0 1.8 NOTE:

To simplify the dose analyses in Chapter 5 that show bounding conditions, for some cells, burnup and cooling time combinations are selected for the dose analyses that may correspond to a higher decay heat than is permitted for that cell. The decay heat limits and burnup/cooling time limits remain independent of each other, so this does not impact the decay heat limit for a cell. The cell decay heat limits are given in Figures 1.2.8.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 31 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-17 5.1.1 Normal and Off-Normal Operations Chapter 12 discusses the potential off-normal conditions and their effect on the HI-STORM FW system. None of the off-normal conditions have any impact on the shielding analysis. Therefore, off-normal and normal conditions are identical for the purpose of the shielding evaluation.

The 10CFR72.104 criteria for radioactive materials in effluents and direct radiation during normal operations are:

1.

During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area, must not exceed 25 mrem to the whole body, 75 mrem to the thyroid and 25 mrem to any other critical organ.

2.

Operational restrictions must be established to meet as low as reasonably achievable (ALARA) objectives for radioactive materials in effluents and direct radiation.

10CFR20 Subparts C and D specify additional requirements for occupational dose limits and radiation dose limits for individual members of the public. Chapter 11 specifically addresses these regulations.

In accordance with ALARA practices, design objective dose rates are established for the HI-STORM FW system and presented in Table 2.3.2.

Figure 5.1.1 identifies the locations of the dose points referenced in the dose rate summary tables for the HI-STORM FW overpack. Dose Point #2 is located on the side of the cask at the axial mid-height. Dose Points #1 and #3 are the locations of the inlet and outlet air ducts, respectively.

The dose values reported for these locations (adjacent and 1 meter) were averaged over the duct opening. Dose Point #4 is the dose location on the overpack lid. The dose values reported at the locations shown on Figure 5.1.1 are averaged over a region that is approximately 1 foot in width.

Figure 5.1.2 identifies the location of the dose points for the HI-TRAC VW transfer cask. Dose Point Locations #1 and #3 are situated below and above the water jacket, respectively. In the case of the HI-TRAC VW Version V2, Dose Point Locations #1 and #3 are situated below and above the neutron shield, respectively. Dose Point #4 is the dose location on the HI-TRAC VW lid and dose rates below the HI-TRAC VW are estimated with Dose Point #5. Dose Point Location #2 is situated on the side of the cask at the axial mid-height.

Since the MPC-89 heat load pattern in Figure 1.2.8 contains the high heat load assemblies in the peripheral basket cells, it represents the bounding pattern with the highest side dose rates adjacent to and at a distance from the overpack. Specifically, it affects the dose rate at mid-height (Dose Point #2) and below the neutron shield (Dose Point #1) of HI-TRAC VW (standard) and HI-TRAC VW Version V2 as well as the dose rate at mid-height (Dose Point #2) and adjacent to the inlet vents (Dose Point #1) of HI-STORM FW, while all the other dose rate locations remain comparable with the other loading patterns or reduced.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 32 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-18 The total dose rates presented are presented for two cases: with and without BPRAs. The dose from the BPRAs was conservatively assumed to be the maximum calculated in Subsection 5.4.4.

Tables 5.1.1, 5.1.2 and 5.1.13 provide dose rates adjacent to and one meter from the HI-TRAC VW during normal conditions for the MPC-37, MPC-89 and MPC-32ML. The dose rates listed in Tables 5.1.1, 5.1.2 and 5.1.13 correspond to the normal condition in which the MPC is dry and the HI-TRAC water jacket is filled with water. It should be noted that the minimum lead thickness of HI-TRAC VW with MPC-32ML is more than that of HI-TRAC with MPC-37.

Tables 5.1.10 provides dose rates adjacent to and one meter from the HI-TRAC VW Version V2 during normal conditions for the MPC-89. The dose rates listed in Table 5.1.10 correspond to the normal condition in which the MPC is dry and the Gamma Shield Cylinder and Neutron Shield Cylinder are present.

Tables 5.1.5, 5.1.6 and 5.1.11 provide the design basis dose rates adjacent to the HI-STORM FW overpack during normal conditions for the MPC-37, MPC-89 and MPC-32ML. Tables 5.1.7, 5.1.8 and 5.1.12 provide the design basis dose rates at one meter from the HI-STORM FW overpack containing the MPC-37, MPC-89 and MPC-32ML, respectively.

The dose to any real individual at or beyond the controlled area boundary is required to be below 25 mrem per year. The minimum distance to the controlled area boundary is 100 meters from the ISFSI. Table 5.1.3 presents the annual dose to an individual from a single HI-STORM FW cask and various storage cask arrays, assuming an 8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> annual occupancy at the dose point location. The minimum distance required for the corresponding dose is also listed. The results are based on the bounding source terms, which is the MPC-89 loading pattern in Figure 1.2.8. It is noted that these data present the hypothetical upper bound annual dose values and they are provided for illustrative purposes only. A detailed site-specific evaluation of dose at the controlled area boundary must be performed for each ISFSI in accordance with 10CFR72.212.

The site-specific evaluation will consider dose from other portions of the facility and will consider the actual conditions of the fuel being stored (burnup and cooling time).

Figure 5.1.3 is an annual dose versus distance graph for the HI-STORM FW cask array configurations provided in Table 5.1.3. Figure 5.1.4 is an annual dose versus distance graph for the HI-STORM FW cask array configurations provided in Table 5.4.21. These curves, which are based on an 8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> occupancy, are provided for illustrative purposes only and will be re-evaluated on a site-specific basis.

Subsection 5.2.3 discusses the BPRAs, TPDs, CRAs and APSRs that are permitted for storage in the HI-STORM FW system. Subsection 5.4.4 discusses the increase in dose rate as a result of adding non-fuel hardware in the MPCs.

The analyses summarized in this section demonstrate that the HI-STORM FW system is in compliance with the radiation and exposure objectives of 10CFR72.106. Since only representative dose rate values for normal conditions are presented in this chapter, compliance with 10CFR72.104 is not being evaluated. This will be performed as part of the site specific evaluations.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 33 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-20 through 5.1.4d with MPC-89 for the HI-TRAC VW at a distance of 1 meter and at a distance of 100 meters. The normal condition dose rates are provided for reference. The dose for a period of 30 days is shown in Table 5.1.9, where 30 days is used to illustrate the radiological impact for a design basis accident. Based on this dose rate and the short duration of use for the loaded HI-TRAC transfer cask, it is evident that the dose as a result of the design basis accident cannot exceed 5 rem at the controlled area boundary for the short duration of the accident, even in case of the bounding loading pattern (see Figure 1.2.8) with the highest side dose rates.

The HI-TRAC VW Version V2 shielding accident case where potentially the Holtite-A is lost from fire is bounded by the standard HI-TRAC VW for accident cases since the total radial through thickness of steel and lead is slightly more for the HI-TRAC VW Version V2 than the standard HI-TRAC VW.

Analyses summarized in this section demonstrate that the HI-STORM FW system, including the HI-TRAC VW transfer cask, is in compliance with the 10CFR72.106 limits. It should be noted that, as a defense in depth, site-specific shielding evaluation shall be performed if there is any fuel to be loaded into MPC-32ML with a burnup more than the design basis accident-condition burnup for MPC-32ML in Table 5.0.2.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 34 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-24 Table 5.1.2c MAXIMUM DOSE RATES FROM THE HI-TRAC VW FOR NORMAL CONDITIONS MPC-89 DESIGN BASIS FUEL REGIONALIZED LOADING BASED ON FIGURE 1.2.8 Dose Point Location Fuel Gammas (mrem/hr)

(n,)

Gammas (mrem/hr) 60Co Gammas (mrem/hr)

Neutrons (mrem/hr)

Totals (mrem/hr)

ADJACENT TO THE HI-TRAC VW 1

641.9 34.7 2737.1 78.7 3492.3 2

9286.5 18.2

<0.1 44.6 9349.4 3

17.6 6.3 701.7 7.2 732.8 4

52.4 1.9 503.8 302.5 860.6 5

162.4 4.5 1979.6 1674.6 3821.2 ONE METER FROM THE HI-TRAC VW 1

1588.6 7.3 263.8 16.1 1875.8 2

4824.6 5.1 16.9 17.3 4863.9 3

401.0 6.5 344.0 9.4 760.9 4

43.6 0.6 351.5 81.8 477.5 5

181.2 1.2 1386.4 387.7 1956.5 Notes:

Refer to Figure 5.1.2 for dose locations.

Dose rates are based on no water within the MPC, an empty annulus, and a water jacket full of water. For the majority of the duration that the HI-TRAC bottom lid is installed, the MPC cavity will be flooded with water. The water within the MPC greatly reduces the dose rate.

Streaming may occur through the annulus. However, during handling/operations the annulus is filled with water and lead snakes are typically present to reduce the streaming effects. Further, operators are not present on top of the transfer cask.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 35 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-25 Table 5.1.3 MAXIMUM DOSE RATES FOR ARRAYS OF HI-STORM FWs CONTAINING MPC-89 WITH REGIONALIZED LOADING BASED ON FIGURE 1.2.8 Array Configuration 1 cask 2x2 2x3 2x4 2x5 HI-STORM FW Overpack Annual Dose (mrem/year) 18 17 25 14 18 Distance to Controlled Area Boundary (meters) 400 500 500 600 600 Notes:

Values are rounded up to nearest integer.

8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> annual occupancy is assumed.

Dose location is at the center of the long side of the array.

The bounding regionalized loading source term, consistent with Table 5.1.8c for dose point location 2, is used.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 36 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-29 Table 5.1.4d MAXIMUM DOSE RATES FROM HI-TRAC VW WITH MPC-89 FOR ACCIDENT CONDITIONS AT REGIONALIZED LOADING BURNUP AND COOLING TIMES BASED ON FIGURE 1.2.8 Dose Point Location Fuel Gammas (mrem/hr)

N, Gamma (mrem/hr)

Co-60 Gamma (mrem/hr)

Neutrons (mrem/hr)

Total (mrem/hr) 1 meter from HI-TRAC VW 2 (Accident Condition) 6574.7 4.0 36.5 3644.3 10259.6 2 (Normal Condition) 4824.6 5.1 16.9 17.3 4863.9 100 meters from HI-TRAC VW 2 (Accident Condition) 2.5 0.0 0.2 1.8

4.5 Notes

Refer to Figure 5.1.2 for dose locations.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 37 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-33 Table 5.1.6c MAXIMUM DOSE RATES ADJACENT TO HI-STORM FW OVERPACK FOR NORMAL CONDITIONS MPC-89 REGIONALIZED BURNUP AND COOLING TIME BASED ON FIGURE 1.2.8 Dose Point Location Fuel Gammas (mrem/hr)

(n,)

Gammas (mrem/hr) 60Co Gammas (mrem/hr)

Neutrons (mrem/hr)

Totals (mrem/hr) 1 614.8 0.3 21.6 0.8 637.5 2

456.8 0.4

<0.1 0.4 457.8 3 (surface) 14.2 0.3 20.2 4.4 39.1 3 (overpack edge) 5.7 0.1 10.0 1.5 17.3 4 (center) 0.4 2.5 2.2 2.1 7.2 4 (mid) 19.5 1.0 14.5 3.6 38.6 4 (outer) 0.4

<0.1 2.0 0.1

2.6 Notes

Refer to Figure 5.1.1 for dose locations.

Dose location 3 (surface) is at the surface of the outlet vent. Dose location 3 (overpack edge) is in front of the outlet vent, but located radially above the overpack outer diameter.

Dose location 4 (center) is at the center of the top surface of the top lid. Dose location 4 (mid) is situated directly above the vertical section of the outlet vent.

Dose location 4 (outer) is extended along the top plane of the top lid, located radially above the overpack outer diameter.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 38 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-37 Table 5.1.8c MAXIMUM DOSE RATES AT ONE METER FROM HI-STORM FW OVERPACK FOR NORMAL CONDITIONS MPC-89 REGIONALIZED BURNUP AND COOLING TIME BASED ON FIGURE 1.2.8 Dose Point Location Fuel Gammas (mrem/hr)

(n,)

Gammas (mrem/hr) 60Co Gammas (mrem/hr)

Neutrons (mrem/hr)

Totals (mrem/hr) 1 140.5 0.1 4.5 0.2 145.3 2

224.9 0.2 0.7 0.2 225.9 3

15.3

<0.1 5.5 0.2 21.0 4 (center) 2.6 0.9 3.8 1.9

9.2 Notes

Refer to Figure 5.1.1 for dose locations.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer girds.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 39 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-38 Table 5.1.9 MAXIMUM DOSE FROM HI-TRAC VW FOR ACCIDENT CONDITIONS AT 100 METERS Dose Point Location Dose Rate (rem/hr)

Accident Duration (days)

Total Dose (rem)

Regulatory Limit (rem)

Time to Reach Regulatory Limit (days) 2 (Accident Condition) 4.5E-03 30 3.27 5

45 Notes:

Refer to Figure 5.1.2 for dose locations.

Values are rounded to nearest integer where appropriate.

Dose rate used to evaluated Total Dose (rem) is the maximum from Tables 5.1.4a through 5.1.4d.

Regulatory Limit is from 10CFR72.106.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 40 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-39 Table 5.1.10a DOSE RATES FROM THE HI-TRAC VW VERSION V2 FOR NORMAL CONDITIONS, DRY MPC-89 WITH NEUTRON SHIELD CYLINDER PRESENT, BASED ON FIGURE 1.2.7 Dose Point Location Fuel Gammas (n,)

Gammas 60Co Gammas Neutrons Totals (mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

ADJACENT TO THE HI-TRAC VW 1

1020 2

5794 362 7178 1*

104 2

675 384 1164 2

1412 19

< 1 154 1585 3

4 2

144 137 286 4

134 3

515 427 1078 5

174 6

1799 2123 4102 ONE METER FROM THE HI-TRAC VW 1

345 3

63 31 441 1*

339 3

52 30 425 2

653 6

4 51 715 3

35 1

54 11 100 4

140 1

490 103 734 5

186 1

1430 462 2080 Notes:

• Location 1* uses a steel shield ring pedestal for the Neutron Shield Cylinder, which may be present for ALARA purposes. The critical shielding dimensions of the optional steel shield ring pedestal are as follows: Outer Diameter is 8 feet; radial thickness is 2.5 inches; Axial bottom of shield ring is 3 inches below MPC baseplate bottom surface; top of shield ring is in contact with Neutron Shield Cylinder.

Refer to Figure 5.1.2 for dose locations.

Values are rounded to nearest integer.

Dose rates are based on no water within the MPC, an empty annulus, and the Neutron Shield Cylinder present. For the majority of the duration that the HI-TRAC bottom lid is installed, the MPC cavity will be flooded with water. The water within the MPC greatly reduces the dose rate.

Streaming may occur through the annulus. However, during handling/operations the annulus is filled with water and lead snakes are typically present to reduce the streaming effects.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 41 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-40 Table 5.1.10b DOSE RATES FROM THE HI-TRAC VW VERSION V2 FOR NORMAL CONDITIONS, DRY MPC-89 WITH NEUTRON SHIELD CYLINDER PRESENT, BASED ON FIGURE 1.2.8 Dose Point Location Fuel Gammas (n,)

Gammas 60Co Gammas Neutrons Totals (mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

ADJACENT TO THE HI-TRAC VW 1

1267.1 2.1 6933.1 433.2 8635.4 1*

134.0 1.6 864.7 322.3 1322.7 2

2265.8 10.4

<0.1 92.0 2368.1 3

2.8 1.4 154.1 110.7 269.0 4

140.5 2.2 606.6 327.1 1076.4 5

170.9 4.7 1975.4 1696.9 3847.9 ONE METER FROM THE HI-TRAC VW 1

610.5 4.3 84.2 40.7 739.6 1*

616.0 3.5 61.9 34.2 715.6 2

1200.7 3.3 4.8 31.8 1240.7 3

55.6 1.1 55.4 11.0 123.1 4

212.2 0.7 510.6 100.7 824.3 5

204.2 1.3 1398.3 423.3 2027.1 Notes:

• Location 1* uses a steel shield ring pedestal for the Neutron Shield Cylinder, which may be present for ALARA purposes. The critical shielding dimensions of the optional steel shield ring pedestal are as follows: Outer Diameter is 8 feet; radial thickness is 2.5 inches; Axial bottom of shield ring is 3 inches below MPC baseplate bottom surface; top of shield ring is in contact with Neutron Shield Cylinder.

Refer to Figure 5.1.2 for dose locations.

Values are rounded to nearest integer.

Dose rates are based on no water within the MPC, an empty annulus, and the Neutron Shield Cylinder present. For the majority of the duration that the HI-TRAC bottom lid is installed, the MPC cavity will be flooded with water. The water within the MPC greatly reduces the dose rate.

Streaming may occur through the annulus. However, during handling/operations the annulus is filled with water and lead snakes are typically present to reduce the streaming effects.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 42 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-46 1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04 0

100 200 300 400 500 600 700 mrem/yr meters SingleCask 2x2Array 2x3Array 2x4Array 2x5Array Figure 5.1.3 ANNUAL DOSE VERSUS DISTANCE FOR VARIOUS CONFIGURATIONS OF MPC-89 FOR REGIONALIZED LOADING BASED ON FIGURE 1.2.8 (8760 HOUR OCCUPANCY ASSUMED)

ATTACHMENT 4 TO HOLTEC LETTER 5018075 43 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-117 Table 5.4.4c MAXIMUM DOSE RATES FOR THE HI-TRAC VW FOR THE FULLY FLOODED MPC CONDITION WITH AN EMPTY NEUTRON SHIELD MPC-89 DESIGN BASIS ZIRCALOY CLAD FUEL REGIONALIZED LOADING BASED ON FIGURE 1.2.8 Dose Point Location Fuel Gammas (mrem/hr)

(n,)

Gammas (mrem/hr) 60Co Gammas (mrem/hr)

Neutrons (mrem/hr)

Totals (mrem/hr)

ADJACENT TO THE HI-TRAC VW 1

595.8 0.5 1909.2 93.6 2599.1 2

9297.8 1.7

<0.1 291.5 9591.0 3

2.5

<0.1 432.4 6.6 441.5 4

3.5

<0.1 222.6

<0.1 226.2 5 (bottom lid) 65.9

<0.1 1535.4 7.4 1608.7 ONE METER FROM THE HI-TRAC VW 1

1704.6 0.3 212.7 62.4 1980.0 2

5229.7 0.4 10.9 90.7 5331.8 3

410.0 0.2 200.2 33.3 643.6 4

2.9

<0.1 154.7 0.2 157.8 5

52.9

<0.1 999.2 3.6 1055.8 Notes:

Refer to Figure 5.1.2 for dose point locations.

MPC internal water level is 5 inches below the MPC lid.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 44 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-120 Table 5.4.5 MAXIMUM DOSE RATES FOR THE HI-TRAC VW FOR THE FULLY FLOODED MPC CONDITION WITH A FULL NEUTRON SHIELD MPC-89 DESIGN BASIS ZIRCALOY CLAD FUEL REGIONALIZED LOADING BASED ON FIGURE 1.2.8 Dose Point Location Fuel Gammas (mrem/hr)

(n,)

Gammas (mrem/hr) 60Co Gammas (mrem/hr)

Neutrons (mrem/hr)

Totals (mrem/hr)

ADJACENT TO THE HI-TRAC VW 1

347.9 0.8 1163.7 3.1 1515.5 2

5610.9 2.7

<0.1 13.5 5627.1 3

0.8

<0.1 235.6

<0.1 236.4 4

3.5

<0.1 222.7

<0.1 226.2 5 (bottom lid) 66.2

<0.1 1535.0 7.0 1608.1 ONE METER FROM THE HI-TRAC VW 1

1027.3 0.4 118.3 1.9 1148.0 2

3147.9 0.9 6.1 6.1 3160.9 3

249.9 0.3 106.2 1.1 357.6 4

2.9

<0.1 154.7

<0.1 157.6 5

53.0

<0.1 999.1 2.8 1054.9 Notes:

Refer to Figure 5.1.2 for dose point locations.

MPC internal water level is 5 inches below the MPC lid.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 45 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-121 Table 5.4.6 ANNUAL DOSE AT 400 METERS FROM A SINGLE HI-STORM FW OVERPACK WITH THE XL LID DESIGN CONTAINING MPC-89 WITH DESIGN BASIS ZIRCALOY CLAD FUEL Dose Component Regionalized Loading Based on Figure 1.2.8 (mrem/yr)

Fuel gammas 16.59 60Co Gammas 0.43 Neutrons 0.02 Total 17.04 Notes:

Gammas generated by neutron capture are included with fuel gammas.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 46 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-122 Table 5.4.7 DOSE VALUES USED IN CALCULATING ANNUAL DOSE FROM VARIOUS HI-STORM FW ISFSI CONFIGURATIONS WITH THE XL LID DESIGN ZIRCALOY CLAD FUEL (REGIONALIZED BURNUP AND COOLING TIME COMBINATIONS BASED ON FIGURE 1.2.8)

Distance A

Side of Overpack (mrem/yr)

B Top of Overpack (mrem/yr)

C Side of Shielded Overpack (mrem/yr) 100 meters 990.9 110.1 198.2 200 meters 154.2 17.1 30.8 300 meters 42.7 4.7 8.5 400 meters 15.3 1.7 3.1 500 meters 5.7 0.6 1.1 600 meters 2.4 0.3

0.5 Notes

8760 hour0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> annual occupancy is assumed.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 47 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-131 Table 5.4.17a DOSE RATES FOR THE HI-TRAC VW VERSION V2 FOR THE FULLY FLOODED MPC AND FLOODED ANNULUS CONDITION WITHOUT NEUTRON SHIELD CYLINDER PRESENT, BASED ON FIGURE 1.2.7 MPC-89 DESIGN BASIS ZIRCALOY CLAD FUEL Dose Point Location Fuel Gammas (n,) Gammas 60Co Gammas Neutrons Totals (mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

ADJACENT TO THE HI-TRAC VW VERSION V2 1

277.3

<1 3383.0 14.7 3675.1 2

4890.2 2.1

<1 227.7 5120.1 3

1.2

<1 331.3

<1 332.5 4

4.3

<1 253.8

<1 258.1 5 (bottom lid) 92.1

<1 1486.5 8.9 1587.5 ONE METER FROM THE HI-TRAC VW VERSION V2 1

1175.5

<1 209.3 15.6 1400.6 2

2274.8

<1 9.7 31.8 2316.7 3

91.7

<1 106.4 6.3 204.5 4

3.1

<1 210.6

<1 213.8 5

65.9

<1 1060.2 3.3 1129.4 Notes:

Refer to Figure 5.1.2 for dose point locations.

MPC internal water level is 5 inches below the MPC lid.

Annulus water level is 2 inches above bottom surface of the MPC lid.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 48 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-132 Table 5.4.17b DOSE RATES FOR THE HI-TRAC VW VERSION V2 FOR THE FULLY FLOODED MPC AND FLOODED ANNULUS CONDITION WITHOUT NEUTRON SHIELD CYLINDER PRESENT, BASED ON FIGURE 1.2.8 MPC-89 DESIGN BASIS ZIRCALOY CLAD FUEL Dose Point Location Fuel Gammas (n,) Gammas 60Co Gammas Neutrons Totals (mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

ADJACENT TO THE HI-TRAC VW VERSION V2 1

486.7 0.2 4278.7 18.2 4783.8 2

9400.0 0.9

<0.1 94.2 9495.0 3

1.1

<0.1 339.1

<0.1 340.2 4

4.3

<0.1 263.4

<0.1 267.7 5 (bottom lid) 59.0

<0.1 1538.0 9.8 1606.8 ONE METER FROM THE HI-TRAC VW VERSION V2 1

2580.4 0.2 295.1 28.0 2903.7 2

5043.1 0.3 13.7 39.6 5096.6 3

180.3

<0.1 129.4 8.2 317.9 4

2.3

<0.1 219.3

<0.1 221.5 5

61.8

<0.1 1009.1 2.7 1073.6 Notes:

Refer to Figure 5.1.2 for dose point locations.

MPC internal water level is 5 inches below the MPC lid.

Annulus water level is 2 inches above bottom surface of the MPC lid.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 49 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-133 Table 5.4.18a DOSE RATES FOR THE HI-TRAC VW VERSION V2 FOR THE FULLY FLOODED MPC AND FLOODED ANNULUS WITH NEUTRON SHIELD CYLINDER PRESENT, BASED ON FIGURE 1.2.7 MPC-89 DESIGN BASIS ZIRCALOY CLAD FUEL Dose Point Location Fuel Gammas (n,)

Gammas (mrem/hr) 60Co Gammas Neutrons Totals (mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

ADJACENT TO THE HI-TRAC VW VERSION V2 1

169.6

<1 2058.4 2.0 2230.0 1*

22.0

<1 255.1 1.3 278.4 2

678.2

<1

<1 2.7 681.0 3

<1

<1 41.4

<1 41.7 4

4.3

<1 253.4

<1 257.8 5 (bottom lid) 91.1

<1 1485.9 8.8 1585.9 ONE METER FROM THE HI-TRAC VW VERSION V2 1

165.6

<1 21.2

<1 187.2 1*

164.8

<1 16.1

<1 181.4 2

320.9

<1 1.0

<1 322.9 3

15.4

<1 10.8

<1 26.3 4

3.1

<1 210.5

<1 213.6 5

66.2

<1 1060.2 3.4 1129.8 Notes:

• Location 1* uses a steel shield ring pedestal for the Neutron Shield Cylinder, which may be present for ALARA purposes. The critical shielding dimensions of the optional steel shield ring pedestal are as follows: Outer Diameter is 8 feet; radial thickness is 2.5 inches; Axial bottom of shield ring is 3 inches below MPC baseplate bottom surface; top of shield ring is in contact with Neutron Shield Cylinder.

Refer to Figure 5.1.2 for dose point locations.

MPC internal water level is 5 inches below the MPC lid.

Annulus water level is 2 inches above bottom surface of the MPC lid.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 50 of 51

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2114830 Proposed Rev. 6.J 5-134 Table 5.4.18b DOSE RATES FOR THE HI-TRAC VW VERSION V2 FOR THE FULLY FLOODED MPC AND FLOODED ANNULUS WITH NEUTRON SHIELD CYLINDER PRESENT, BASED ON FIGURE 1.2.8 MPC-89 DESIGN BASIS ZIRCALOY CLAD FUEL Dose Point Location Fuel Gammas (n,)

Gammas (mrem/hr) 60Co Gammas Neutrons Totals (mrem/hr)

(mrem/hr)

(mrem/hr)

(mrem/hr)

ADJACENT TO THE HI-TRAC VW VERSION V2 1

239.2

<0.1 2692.3 2.3 2933.9 1*

30.6

<0.1 334.4 1.3 366.4 2

1202.5 0.4

<0.1 5.6 1208.5 3

0.1

<0.1 44.7

<0.1 44.9 4

4.3

<0.1 264.1

<0.1 268.5 5 (bottom lid) 57.2

<0.1 1535.1 9.5 1601.9 ONE METER FROM THE HI-TRAC VW VERSION V2 1

351.5

<0.1 30.6 1.2 383.3 1*

350.0

<0.1 22.7 1.2 373.9 2

684.0 0.2 1.2 2.7 688.1 3

29.9

<0.1 14.8 0.3 45.0 4

2.4

<0.1 219.6

<0.1 222.0 5

59.7

<0.1 1008.3 2.8 1070.9 Notes:

• Location 1* uses a steel shield ring pedestal for the Neutron Shield Cylinder, which may be present for ALARA purposes. The critical shielding dimensions of the optional steel shield ring pedestal are as follows: Outer Diameter is 8 feet; radial thickness is 2.5 inches; Axial bottom of shield ring is 3 inches below MPC baseplate bottom surface; top of shield ring is in contact with Neutron Shield Cylinder.

Refer to Figure 5.1.2 for dose point locations.

MPC internal water level is 5 inches below the MPC lid.

Annulus water level is 2 inches above bottom surface of the MPC lid.

The Fuel Gammas category includes gammas from the spent fuel and 60Co from the spacer grids.

ATTACHMENT 4 TO HOLTEC LETTER 5018075 51 of 51