ML18331A049
| ML18331A049 | |
| Person / Time | |
|---|---|
| Site: | Holtec |
| Issue date: | 10/31/2018 |
| From: | Holtec |
| To: | Office of Nuclear Material Safety and Safeguards |
| References | |
| 5014855, CoC 1014 | |
| Download: ML18331A049 (67) | |
Text
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 1-67 As stated earlier, the HI-STORM 100 confinement boundary is the MPC, which is a high integrity pressure vessel designed and constructed to be leak tight as discussed in Chapter 7. The HI-STORM 100 is a completely passive system with appropriate margins of safety; therefore, it is not necessary to deploy any instrumentation to monitor the cask in the storage mode. At the option of the user, temperature elements may be utilized to monitor the air temperature of the HI-STORM overpack exit vents in lieu of routinely inspecting the ducts for blockage. See Subsection 2.3.3.2 for additional details.
1.2.2.3.5 Maintenance Technique Because of their passive nature, the HI-STORM 100 System requires minimal maintenance over its lifetime. No special maintenance program is required. Chapter 9 describes the acceptance criteria and maintenance program set forth for the HI-STORM 100.
1.2.3 Cask Contents The HI-STORM 100 System is designed to house different types of MPCs. The MPCs are designed to store 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. A summary of the types of fuel authorized for storage in each MPC model is provided below. All fuel assemblies, non-fuel hardware, and neutron sources 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, except damaged fuel meeting the criteria specified below may be stored using Damaged Fuel Isolators (DFIs) (see Figure 2.1.10) when stored in allowed locations. DFIs may be used with damaged fuel assemblies that may have missing or partial fuel rods and/or fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks as long as the fuel assembly can be handled by normal means and where structural integrity is such that geometric rearrangement of fuel is not expected. Damaged fuel that does not meet these conditions must be stored in a DFC.
MPC-24 The MPC-24 is designed to accommodate up to twenty-four (24) PWR fuel assemblies classified as intact fuel assemblies, with or without non-fuel hardware.
MPC 24E and MPC-24EF The MPC-24E and MPC-24EF are designed to accommodate up to twenty-four (24) PWR fuel assemblies, with or without non-fuel hardware. Up to four (4) fuel assemblies may be classified as damaged fuel assemblies or fuel debris, with the balance being classified as intact fuel assemblies.
Damaged fuel assemblies and fuel debris must be stored in fuel storage locations 3, 6, 19, and/or 22 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 1-118 1.5 DRAWINGS The following HI-STORM 100 System drawings and bills of materials are provided on subsequent pages in this subsection:
Drawing Number/Sheet Description Rev.
3923 MPC Enclosure Vessel 30 3925 MPC-24E/EF Fuel Basket Assembly 9
3926 MPC-24 Fuel Basket Assembly 12 3927 MPC-32 Fuel Basket Assembly 16 3928 MPC-68/68F/68FF Basket Assembly 18 7195 MPC-68M Basket Assembly PR16 1495 Sht 1/6 HI-STORM 100 Assembly 13 1495 Sht 2/6 Cross Section "Z" - "Z" View of HI-STORM 18 1495 Sht 3/6 Section "Y" - "Y" of HI-STORM 12 1495 Sht 4/6 Section "X" -"X" of HI-STORM 13 1495 Sht 5/6 Section "W" -"W" of HI-STORM 15 1561 Sht 1/6 View "A" -"A" of HI-STORM 11 1561 Sht 2/6 Detail "B" of HI-STORM 15 1561 Sht 3/6 Detail of Air Inlet of HI-STORM 11 1561 Sht 4/6 Detail of Air Outlet of HI-STORM 12 3669 HI-STORM 100S Assembly 21 1880 Sht 1/10 125 Ton HI-TRAC Outline with Pool Lid 9
1880 Sht 2/10 125 Ton HI-TRAC Body Sectioned Elevation 10 1880 Sht 3/10 125 Ton HI-TRAC Body Sectioned Elevation "B" - "B" 9
1880 Sht 4/10 125 Ton Transfer Cask Detail of Bottom Flange 10 1880 Sht 5/10 125 Ton Transfer Cask Detail of Pool Lid 10 1880 Sht 6/10 125 Ton Transfer Cask Detail of Top Flange 10 1880 Sht 7/10 125 Ton Transfer Cask Detail of Top Lid 9
1880 Sht 8/10 125 Ton Transfer Cask View "Y" - "Y" 9
1880 Sht 9/10 125 Ton Transfer Cask Lifting Trunnion and Locking Pad 7
1880 Sht 10/10 125 Ton Transfer Cask View "Z" - "Z" 9
1928 Sht 1/2 125 Ton HI-TRAC Transfer Lid Housing Detail 11 1928 Sht 2/2 125 Ton HI-TRAC Transfer Lid Door Detail 10 2145 Sht 1/10 100 Ton HI-TRAC Outline with Pool Lid 8
2145 Sht 2/10 100 Ton HI-TRAC Body Sectioned Elevation 8
2145 Sht 3/10 100 Ton HI-TRAC Body Sectioned Elevation 'B-B' 8
2145 Sht 4/10 100 Ton HI-TRAC Detail of Bottom Flange 7
2145 Sht 5/10 100 Ton HI-TRAC Detail of Pool Lid 6
2145 Sht 6/10 100 Ton HI-TRAC Detail of Top Flange 8
2145 Sht 7/10 100 Ton HI-TRAC Detail of Top Lid 8
2145 Sht 8/10 100 Ton HI-TRAC View Y-Y 8
2145 Sht 9/10 100 Ton HI-TRAC Lifting Trunnions and Locking Pad 5
2145 Sht 10/10 100 Ton HI-TRAC View Z-Z 7
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 1.III-4 The MPC68M basket consists of homogeneously dispersed boron carbide (10% minimum by weight).
The B-10 areal density of the Metamic-HT panels which make up the basket is consistent with the areal density of the Metamic classic neutron poison panels in the MPC-68, therefore provides equivalent neutron shielding. From an overall shielding perspective the MPC-68M is expected to provide similar, if not better, shielding characteristics as the MPC-68. The MPC enclosure vessel, overpack, and transfer cask shielding properties are not modified.
1.III.2.3 Cask Contents The MPC-68M is designed to accommodate up to sixty-eight undamaged BWR fuel assemblies. Up to sixteen damaged fuel assemblies in DFCs/DFIs and/or up to eight DFCs containing fuel debris may be stored in the following fuel storage locations: 1, 2, 3, 8, 9, 16, 25, 34, 35, 44, 53, 60, 61, 66, 67, and/or 68 (Figure 1.III.2), with the remaining fuel storage locations filled with undamaged BWR fuel assemblies.
Damaged fuel assemblies in DFCs/DFIs and/or fuel debris in DFCs may be stored in fuel storage locations indicated in Figure 2.III.4 with the remaining fuel storage locations filled with undamaged BWR fuel assemblies. Fuel classified as damaged fuel assemblies will be loaded into damaged fuel containers (DFCs) or basket cell locations with DFIs installed at the upper and lower ends. Damaged fuel assemblies stored using DFIs may contain missing or partial fuel rods and/or fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks as long as the fuel assembly can be handled by normal means and whose structural integrity is such that geometric rearrangement of fuel is not expected. Damaged fuel that does not meet these conditions must be stored in a DFC. Table 1.2.2 as supplemented by Table 1.III.1 provides the key system data and parameters for the MPC-68M.
1.III.2.4 Qualification of Metamic-HT Metamic-HT is the designated neutron absorber in the MPC-68M baskets. It is also the structural material of the basket. The properties of Metamic-HT and key characteristics, necessary for ensuring nuclear reactivity control, thermal, and structural performance of the basket, are presented below.
(a) Overview Metamic-HT is a composite of nano-particles of aluminum oxide (alumina) and finely ground boron carbide particles homogeneously dispersed in the metal matrix of pure aluminum produced by an extrusion process that ensures a high level of isotropy. Metamic-HT is the constituent material of MPC-68M fuel basket. Metamic-HT neutron absorber is a successor to the Metamic (classic) product [1.III.1]
widely used in dry storage fuel baskets [1.III.8, 1.III.9] and spent fuel storage racks [1.III.2]. Metamic-HT is engineered to possess the necessary mechanical characteristics for structural application in spent nuclear fuel casks. The mechanical properties of Metamic-HT are derived from the strengthening of its aluminum matrix with ultra-fine grained (nano-particle size) alumina (Al2O3) particles that anchor the grain boundaries for high temperature strength (the HT designation is derived from this characteristic) and high creep resistance. The specific Metamic-HT composition utilized in this FSAR is defined in the MPC-68M drawings in Section 1.III.5. Metamic-HT was first certified by the USNRC in 2009 as the sole constituent material for the fuel basket types F-37 and F-32 in the HI-STAR 180 transport package 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 1.III-11 Table 1.III.1 Key Parameters for MPC-68M BWR MPC internal environment Helium fill (99.995% fill helium purity)
(all pressure ranges are at a reference temperature of 70oF)
(heat load < 28.19 kW)
(heat load >28.19 kW)
> 29.3 psig and < 48.5 psig OR 0.1218 +/-10% g-moles/liter
> 45.5 psig and < 48.5 psig Quarter Symmetric Heat Load (QSHL, Figure 2.III.1)
QSHL patterns in Figures 2.III.2 through 2.III.4
> 43.5 psig and < 46.5 psig
> 45.5 psig and < 48.5 psig B4C content in Metamic-HT (wt. %)
As specified on drawing in Section 1.5 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-38 determined to be acceptable for storage in the HI-STORM 100 System. Section 2.1.9 summarizes the authorized contents for the HI-STORM 100 System. Any fuel assembly that has fuel characteristics within the range of Tables 2.1.3 and 2.1.4 and meets the other limits specified in Section 2.1.9 is acceptable for storage in the HI-STORM 100 System. Tables 2.1.3 and 2.1.4 present the groups of fuel assembly types defined as array/classes as described in further detail in Chapter 6. Table 2.1.5 lists the BWR and PWR fuel assembly designs which are found to govern for three qualification criteria, namely reactivity, shielding, and thermal. Additional information on the design basis fuel definition is presented in the following subsections.
2.1.2 Intact SNF Specifications Intact fuel assemblies are defined as fuel assemblies without known or suspected cladding defects greater than pinhole leaks and hairline cracks, and which can be handled by normal means. The design payload for the HI-STORM 100 System is intact ZR or stainless steel (SS) clad fuel assemblies with the characteristics listed in Tables 2.1.17 through 2.1.24.
Intact fuel assemblies without fuel rods in fuel rod locations cannot be loaded into the HI-STORM 100 unless dummy fuel rods, which occupy a volume greater than or equal to the original fuel rods, replace the missing rods prior to loading. Any intact fuel assembly that falls within the geometric, thermal, and nuclear limits established for the design basis intact fuel assembly, as defined in Section 2.1.9 can be safely stored in the HI-STORM 100 System. If irradiated dummy stainless steel rods are present in the fuel assembly, the dummy/replacement rods will be considered in the site specific dose calculations.
The range of fuel characteristics specified in Tables 2.1.3 and 2.1.4 have been evaluated in this FSAR and are acceptable for storage in the HI-STORM 100 System within the decay heat, burnup, and cooling time limits specified in Section 2.1.9 for intact fuel assemblies.
2.1.3 Damaged SNF and Fuel Debris Specifications Damaged fuel and fuel debris are defined in the glossary.
Damaged fuel assemblies and fuel debris will be loaded into stainless steel damaged fuel containers (DFCs) provided with mesh screens having between 40x40 and 250x250 openings per inch, for storage in the HI-STORM 100 System (see Figures 2.1.1 and 2.1.2B, C, and D). The MPC-24, MPC-24EF, MPC-32 and MPC-32F are designed to accommodate PWR damaged fuel and fuel debris. The MPC-68, MPC-68F and MPC-68FF are designed to accommodate BWR damaged fuel and fuel debris. The appropriate structural, thermal, shielding, criticality, and confinement analyses have been performed to account for damaged fuel and fuel debris and are described in their respective chapters that follow. For damaged fuel assemblies that can be handled by normal means and whose structural integrity is such that geometric rearrangement of fuel is not expected, the use of Damaged Fuel Isolators (DFIs) (see Figure 2.1.10) can be substituted for the use of the DFC for storage in allowed locations. Damaged fuel that does not meet this criteria must be stored using DFCs. The DFI is a set of specially designed barriers at the top and bottom of a storage cell space used to prevent the 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-39 migration of fissile material from those cells. DFI storage locations are limited to the same locations allowed for DFCs in the MPC-68M. The limiting design characteristics for damaged fuel assemblies and restrictions on the number and location of damaged fuel containers authorized for loading in each MPC model are provided in Section 2.1.9. Dresden Unit 1 fuel assemblies contained in Transnuclear-designed damaged fuel canisters and one Dresden Unit 1 thoria rod canister have been approved for storage directly in the HI-STORM 100 System without re-packaging (see Figures 2.1.2 and 2.1.2A).
MPC contents classified as fuel debris are required to be stored in DFCs. The basket designs for the standard and F model MPCs are identical. The lid and shell designs of the F models are unique in that the upper shell portion of the canister is thickened for additional strength needed to qualify as a secondary containment, which used to be required under hypothetical accident conditions of transportation under 10 CFR 71. Figure 2.1.9 shows the details of the differences between the standard and F model MPC shells. These details are common for both the PWR and BWR series MPC models.
2.1.3.1 Damaged Fuel Isolator If the damaged fuel assembly can be handled by normal means and its structural integrity is such that geometric rearrangement of fuel is not expected, then the device known as the Damaged Fuel Isolator (DFI) can be used in place of the DFC. Like the DFC, the DFI prevents the migration of fissile material in bulk or coarse particulate form from the nuclear fuel stored in its cellular storage cavity.
The DFI can be used only if the fuel can be handled by normal means but is classified as damaged because of physical defect, viz., a breach in the fuel cladding or a structural failure in the grid strap assembly, etc., as explained in ISG-1. Damaged fuel stored utilizing the DFI may contain missing or partial fuel rods and/or fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks as long as the fuel assembly can be handled by normal means.
The DFI is made up of two end caps that, along with the four cell walls, comprise the fuel isolation space. The DFI is made of corrosion resistant alloy steel (eg. 304 stainless steel) and includes mesh screens or perforated plates at the top and bottom. (see Table 2.2.6 for component details). The essential attributes of the DFI are:
- 1. The bottom cap is a prismatic box with a flat baseplate which fits inside the storage cell space with a small clearance (for ease of installation). The sidewalls of the bottom cap have perforations or wire mesh to permit transmigration of gases but not fuel fragments or gross particulates and is equipped with a flexible permeable barrier against the storage cell walls for sequestration of coarse particulate matter.
- 2. The top cap is anatomically similar to the bottom cap as illustrated in Figure 2.1.10.
- 3. Both caps have engineered features to enable them to be remotely installed in any storage cell in which a fuel assembly needs to be isolated. Both caps are geometrically constrained to prevent their ejection from the storage cavity during a postulated accident event.
- 4. The design configuration of the DFI is common for all Light Water Reactor fuel.
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-42 As discussed earlier, the MPC-68, MPC-68F, MPC-68FF, MPC-32 and MPC-32F feature a basket without flux traps. In the aforementioned baskets, there is one panel of neutron absorber between two adjacent fuel assemblies. The MPC-24, MPC-24E, and MPC-24EF employ a construction wherein two neighboring fuel assemblies are separated by two panels of neutron absorber with a water gap between them (flux trap construction).
The minimum 10B areal density in the neutron absorber panels for each MPC model is shown in Table 2.1.15.
For all MPCs, the 10B areal density used for the criticality analysis is conservatively established below the minimum values shown in Table 2.1.15. For Boral, the value used in the analysis is 75%
of the minimum value, while for METAMIC, it is 90% of the minimum value. This is consistent with NUREG-1536 [2.1.5] which suggests a 25% reduction in 10B areal density credit when subject to standard acceptance tests, and which allows a smaller reduction when more comprehensive tests of the areal density are performed.
The criticality analyses for the MPC-24, MPC-24E and MPC-24EF (all with higher enriched fuel) and for the MPC-32 and MPC-32F were performed with credit taken for soluble boron in the MPC water during wet loading and unloading operations. Table 2.1.14 and 2.1.16 provide the required soluble boron concentrations for these MPCs.
2.1.9 Summary of Authorized Contents Tables 2.1.3, 2.1.4, 2.1.12, and 2.1.17 through 2.1.27 together specify the limits for spent fuel and non-fuel hardware authorized for storage in the HI-STORM 100 System. The limits in these tables are derived from the safety analyses described in the following chapters of this FSAR. Fuel classified as damaged fuel assemblies or fuel debris must be stored in damaged fuel containers for storage in the HI-STORM 100 System, except for damaged fuel which can be handled by normal means whose structural integrity is such that geometric rearrangement of fuel is not expected can be stored using DFIs in allowed locations. For damaged fuel assemblies which can be handled by normal means and whose structural integrity is such that geometric rearrangement of fuel is not expected, the damaged fuel assemblies may be stored in basket cell locations with DFIs installed at the upper and lower ends, in place of storage in a DFC. Damaged fuel stored using DFIs may contain missing or partial fuel rods and/or fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks as long as the fuel assembly can be handled by normal means. Damaged fuel not meeting these requirements must be stored using DFCs. Fuel classified as fuel debris must be stored in damaged fuel containers for storage in all MPCs in the HI-STORM 100 System.
Tables 2.1.17 through 2.1.24 are the baseline tables that specify the fuel assembly limits for each of the MPC models, with appropriate references to the other tables in this section for certain other limits. Tables 2.1.17 through 2.1.24 refer to Section 2.1.9.1 for ZR-clad fuel limits on minimum cooling time, maximum decay heat, and maximum burnup for uniform and regionalized fuel loading.
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-92 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 7x7B, 10x10F (Page 1 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 5689.61 4491.21
-307.68
-181.84 2162.62
-140.54
-2761.10 1.25 6680.84 6378.87
-646.86
-187.68 2843.57
-326.36
-1535.72 1.5 9072.84 7764.60
-981.95
-192.20 3503.34
-554.78
-1248.10 1.75 11671.70 9494.03
-1549.05
-183.76 3984.48
-756.41
-1003.61 2.0 15761.10 10171.40
-1983.74
-180.41 4533.44
-1035.69
-1020.71 2.25 20683.90 10100.50
-2362.96
-171.37 4924.21
-1259.16
-1149.28 2.5 25710.50 9847.51
-2788.08
-162.18 5329.88
-1548.05
-1048.31 2.75 31858.60 7767.18
-2661.83
-154.93 5675.76
-1804.31
-992.87 3.0 38703.40 4333.22
-2101.88
-144.94 5898.42
-1990.59
-1030.87 4.0 65948.40
-16991.70 3924.57
-118.43 6390.16
-2406.62
-614.30 5.0 90881.20
-47264.90 16771.40
-112.75 6498.93
-2241.12
-192.49 6.0 111776.00
-79261.50 33399.20
-115.32 6416.04
-1620.07
-84.57 7.0 127348.00
-107023.00 50534.70
-139.25 6848.43
-1458.29
-14.89 8.0 140072.00
-130028.00 65223.10
-144.93 6836.24
-857.79
-99.75 9.0 150749.00
-150213.00 79005.50
-147.77 6773.51
-231.87
-331.15 10.0 158943.00
-167178.00 92612.70
-164.66 7287.36
-461.83
-382.12 11.0 165714.00
-179168.00 101557.00
-164.07 7241.92
-45.10
-521.50 12.0 171975.00
-190727.00 110548.00
-161.09 7166.19 380.43
-589.16 13.0 177624.00
-200947.00 118921.00
-158.82 7131.17 664.17
-667.75 14.0 182802.00
-210117.00 126526.00
-154.60 7016.50 1083.45
-747.88 15.0 186884.00
-214518.00 128584.00
-147.82 6809.36 1591.41
-783.35 16.0 191316.00
-221293.00 134071.00
-142.04 6646.92 2019.29
-841.16 17.0 195369.00
-231600.00 147624.00
-158.43 7404.40 946.55
-820.02 18.0 199404.00
-236224.00 150408.00
-148.69 7053.70 1655.35
-883.27 19.0 203726.00
-243272.00 157476.00
-143.31 6936.71 1903.09
-895.71 20.0 206861.00
-245479.00 159023.00
-137.13 6829.41 2091.47
-903.40 22.0 213325.00
-250875.00 163825.00
-127.55 6623.17 2500.20
-800.98 24.0 220063.00
-255065.00 166460.00
-114.40 6330.37 2896.83
-803.85 26.0 226903.00
-262541.00 177379.00
-115.77 6627.51 2189.72
-651.65 28.0 234964.00
-270961.00 187677.00
-102.37 6255.46 2595.08
-735.34 30.0 241796.00
-272482.00 188002.00
-88.80 5779.54 3315.93
-731.24 35.0 257457.00
-265751.00 183333.00
-71.68 5676.93 1648.24
-511.23 40.0 282525.00
-292276.00 240288.00
-43.47 4948.25 152.96
-833.96 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-93 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 8x8B (Page 2 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 5872.87 4876.54
-344.60
-182.28 2276.71
-160.55
-2884.11 1.25 7240.18 6669.72
-674.77
-191.67 3008.15
-265.92
-1737.98 1.5 9133.59 8738.93
-1206.18
-186.94 3586.01
-573.93
-1179.83 1.75 12212.40 10285.80
-1758.81
-183.81 4183.48
-839.00
-1095.35 2.0 15913.80 11664.70
-2480.99
-179.56 4694.73
-1100.00
-1003.87 2.25 20652.00 12023.80
-3025.66
-174.12 5204.92
-1412.29
-979.17 2.5 26986.10 10399.30
-3032.60
-163.94 5594.88
-1694.85
-1213.71 2.75 33074.30 8670.65
-3129.69
-156.84 5959.94
-1975.74
-1054.90 3.0 39987.50 5388.94
-2722.03
-146.15 6189.85
-2184.18
-1039.58 4.0 68821.60
-18071.10 4016.97
-119.21 6655.64
-2578.72
-677.77 5.0 95032.70
-50959.00 18228.50
-113.67 6737.08
-2341.46
-253.74 6.0 117864.00
-88879.60 39468.80
-128.75 6937.68
-1918.61
-203.01 7.0 133919.00
-117151.00 56431.30
-139.69 6960.80
-1212.83
-123.38 8.0 147621.00
-142952.00 73246.80
-143.67 6879.18
-441.73
-342.11 9.0 158036.00
-165478.00 90946.70
-167.32 7480.35
-551.45
-378.22 10.0 166796.00
-181378.00 101771.00
-165.98 7346.03 114.50
-504.04 11.0 174312.00
-195869.00 112810.00
-165.26 7291.07 642.48
-648.03 12.0 180736.00
-207916.00 122412.00
-163.34 7243.01 1055.04
-742.81 13.0 187002.00
-219945.00 132127.00
-159.70 7084.08 1641.84
-903.88 14.0 192382.00
-229413.00 139613.00
-156.32 7001.62 2085.84
-972.60 15.0 196087.00
-233618.00 142299.00
-151.48 6860.06 2570.55
-883.73 16.0 202268.00
-249608.00 159974.00
-162.80 7359.57 1999.93
-1048.13 17.0 206376.00
-256109.00 166401.00
-159.20 7309.03 2257.68
-1062.93 18.0 209117.00
-255071.00 162389.00
-151.82 7125.28 2596.49
-891.61 19.0 213124.00
-261295.00 168674.00
-146.82 7004.96 2966.11
-951.40 20.0 217047.00
-267281.00 175609.00
-141.96 6943.62 3118.99
-1012.59 22.0 223569.00
-268761.00 171389.00
-127.42 6436.52 4175.11
-877.23 24.0 233533.00
-291046.00 200512.00
-131.73 6830.33 3613.57
-988.74 26.0 238557.00
-284966.00 188216.00
-118.63 6424.02 4316.86
-862.50 28.0 245385.00
-285588.00 185055.00
-105.51 6116.61 4651.69
-844.39 30.0 254559.00
-295608.00 196106.00
-100.36 6027.39 4465.31
-886.90 35.0 272231.00
-295589.00 203313.00
-71.05 5259.94 4464.18
-744.47 40.0 290782.00
-286198.00 204311.00
-50.38 4868.38 2364.75
-614.59 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-94 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 8x8C/D/E (Page 3 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 5499.81 5105.94
-397.34
-189.28 2307.05
-157.92
-2489.82 1.25 6763.01 7033.27
-751.51
-192.95 3068.21
-286.64
-1564.51 1.5 9164.56 8675.52
-1179.12
-194.38 3726.15
-623.94
-1216.26 1.75 12495.50 10090.50
-1732.25
-187.02 4238.65
-847.44
-1136.06 2.0 16663.00 10889.80
-2211.52
-182.17 4831.25
-1175.27
-1260.49 2.25 21598.90 10980.20
-2691.18
-176.65 5300.72
-1453.46
-1219.04 2.5 27348.40 10071.30
-2967.33
-165.41 5680.31
-1735.86
-1252.79 2.75 33467.10 8232.39
-2999.52
-158.56 6061.56
-2033.93
-1086.98 3.0 40382.30 4849.42
-2525.53
-148.53 6314.10
-2257.89
-1075.95 4.0 68954.10
-18263.30 4048.93
-123.13 6850.62
-2734.70
-652.59 5.0 96324.30
-53730.10 19778.60
-114.90 6841.59
-2381.30
-353.71 6.0 118229.00
-89906.60 39997.30
-134.45 7190.60
-2120.86
-143.41 7.0 134948.00
-119919.00 58227.10
-143.18 7200.03
-1397.69
-170.37 8.0 149092.00
-147517.00 76590.50
-149.16 7110.00
-528.97
-313.19 9.0 159771.00
-170139.00 93968.00
-170.19 7649.69
-595.38
-403.04 10.0 168715.00
-187828.00 107088.00
-172.19 7651.82
-46.57
-555.81 11.0 176169.00
-201821.00 117349.00
-170.83 7550.84 552.84
-651.76 12.0 182662.00
-214445.00 127628.00
-169.36 7519.56 997.32
-756.73 13.0 189114.00
-227085.00 137699.00
-166.11 7388.07 1583.27
-844.97 14.0 195273.00
-239345.00 148361.00
-160.79 7228.22 2124.28
-1017.11 15.0 199939.00
-249862.00 159949.00
-174.10 7782.47 1566.35
-1026.32 16.0 204899.00
-258274.00 166856.00
-167.77 7534.06 2227.05
-1070.51 17.0 209356.00
-265290.00 173458.00
-161.96 7463.49 2386.89
-1040.14 18.0 213546.00
-272476.00 180667.00
-158.41 7387.49 2763.66
-1098.37 19.0 217506.00
-277100.00 183949.00
-150.21 7155.18 3240.82
-1107.07 20.0 219837.00
-275266.00 179705.00
-145.05 7009.96 3638.55
-1007.16 22.0 228092.00
-285272.00 186688.00
-133.55 6672.08 4473.64
-1122.87 24.0 237213.00
-304032.00 211958.00
-136.95 7000.92 4086.48
-1049.61 26.0 242060.00
-297359.00 199620.00
-125.83 6734.22 4465.79
-972.10 28.0 249432.00
-299622.00 196900.00
-111.26 6222.03 5440.43
-914.71 30.0 263307.00
-334844.00 247655.00
-111.83 6452.32 4775.31
-1191.53 35.0 273393.00
-291765.00 178985.00
-83.84 5736.80 4650.87
-621.35 40.0 293153.00
-283353.00 175255.00
-57.06 4937.79 3684.27
-559.25 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-95 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 9x9A (Page 4 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 6052.38 5071.10
-377.23
-189.39 2329.15
-150.71
-2461.49 1.25 7809.20 6767.24
-685.64
-190.94 3101.92
-385.33
-1625.56 1.5 9828.98 8472.94
-1182.05
-192.73 3783.44
-633.02
-1099.88 1.75 12766.50 10646.50
-1862.96
-187.66 4368.12
-896.69
-910.37 2.0 16564.30 12063.20
-2586.67
-184.87 4976.49
-1228.06
-894.91 2.25 22071.80 11834.70
-3015.91
-174.88 5443.22
-1518.94
-1014.33 2.5 27866.60 10993.50
-3286.54
-168.71 5965.88
-1909.06
-1027.88 2.75 34375.10 9004.62
-3367.62
-158.97 6305.05
-2182.06
-933.24 3.0 41566.50 5392.11
-2800.23
-149.79 6613.45
-2462.36
-904.38 4.0 72006.50
-20264.40 4921.01
-123.85 7211.86
-3004.62
-603.22 5.0 100197.00
-57315.80 21669.60
-118.72 7356.33
-2796.24
-243.52 6.0 124367.00
-99348.10 46264.80
-136.71 7648.05
-2394.38
-67.58 7.0 143009.00
-134740.00 68824.10
-143.35 7544.90
-1403.30
-173.80 8.0 157479.00
-165996.00 92255.30
-168.05 8114.30
-1315.88
-266.71 9.0 169636.00
-191379.00 110928.00
-172.50 8069.55
-500.37
-450.57 10.0 179282.00
-211202.00 125969.00
-172.12 7976.57 283.36
-617.13 11.0 187512.00
-228637.00 140325.00
-172.16 7928.03 894.69
-760.39 12.0 195321.00
-245580.00 154682.00
-170.38 7824.20 1596.02
-863.97 13.0 202110.00
-263050.00 173293.00
-187.18 8470.09 1003.55
-953.17 14.0 208171.00
-274758.00 183332.00
-179.75 8249.83 1717.21
-1103.07 15.0 213590.00
-284590.00 191650.00
-175.64 8098.33 2289.04
-1165.13 16.0 218091.00
-292503.00 199557.00
-171.84 8035.82 2659.38
-1119.03 17.0 223491.00
-302449.00 208733.00
-164.92 7833.36 3192.21
-1255.80 18.0 226523.00
-304524.00 209895.00
-162.71 7829.04 3410.57
-1091.33 19.0 231702.00
-312496.00 215730.00
-153.73 7552.13 4052.91
-1189.12 20.0 236531.00
-324776.00 232293.00
-164.72 8073.05 3368.73
-1233.57 22.0 244888.00
-335452.00 241932.00
-150.44 7566.26 4642.58
-1160.69 24.0 252171.00
-340795.00 244542.00
-141.18 7321.23 5355.16
-1142.40 26.0 259438.00
-343494.00 244340.00
-129.66 7094.56 5645.82
-1119.92 28.0 268823.00
-359239.00 266068.00
-130.16 7204.93 5605.85
-1064.30 30.0 277221.00
-363922.00 268930.00
-116.96 6799.84 6219.78
-1037.79 35.0 294285.00
-351643.00 245914.00
-99.35 6404.25 5923.44
-713.23 40.0 324174.00
-389397.00 319233.00
-77.68 5933.52 3992.56
-1188.62 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-96 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 9x9B (Page 5 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 5578.48 5425.38
-432.11
-186.86 2357.62
-150.45
-2405.55 1.25 7982.03 6676.90
-642.60
-194.99 3183.10
-399.01
-1919.75 1.5 9842.92 8943.64
-1222.01
-189.84 3845.90
-652.60
-1360.01 1.75 12927.50 10582.20
-1802.08
-190.88 4502.57
-944.36
-1169.95 2.0 17186.80 11657.20
-2441.58
-183.45 5049.98
-1246.51
-1156.40 2.25 21800.20 12295.50
-3074.77
-180.94 5660.86
-1631.58
-1064.82 2.5 28010.00 11198.70
-3349.88
-169.84 6074.18
-1943.73
-1220.46 2.75 34607.80 9092.75
-3327.98
-161.55 6476.70
-2279.47
-1090.70 3.0 41425.40 6300.12
-3202.59
-151.95 6782.84
-2566.85
-1000.46 4.0 71942.80
-18734.90 3920.65
-125.38 7367.52
-3119.27
-631.75 5.0 101151.00
-57291.00 21182.10
-118.05 7377.24
-2721.50
-361.88 6.0 125823.00
-99944.80 45636.60
-136.47 7588.00
-2124.69
-262.67 7.0 144638.00
-135378.00 67687.60
-143.88 7447.72
-995.76
-340.94 8.0 159872.00
-168383.00 91921.20
-168.66 7933.70
-673.04
-395.74 9.0 172305.00
-194121.00 110332.00
-172.16 7831.09 301.31
-634.37 10.0 181683.00
-213140.00 124418.00
-173.36 7740.03 1165.16
-753.12 11.0 190922.00
-232977.00 140095.00
-171.28 7581.53 2053.29
-1027.00 12.0 198213.00
-248066.00 152236.00
-170.70 7492.96 2781.03
-1087.99 13.0 205947.00
-268590.00 173240.00
-187.42 8096.44 2390.78
-1199.48 14.0 211867.00
-280583.00 184192.00
-183.14 8023.23 2903.27
-1325.04 15.0 217071.00
-289407.00 190649.00
-177.77 7760.30 3819.17
-1355.68 16.0 221340.00
-294404.00 193178.00
-173.59 7653.54 4235.81
-1282.26 17.0 227205.00
-306489.00 204027.00
-164.96 7309.81 5290.73
-1440.44 18.0 231085.00
-310612.00 206608.00
-160.03 7176.88 5715.32
-1383.11 19.0 236345.00
-320398.00 215697.00
-153.84 7020.00 6284.82
-1522.44 20.0 240125.00
-328538.00 227545.00
-170.25 7836.24 5008.11
-1382.77 22.0 245672.00
-325279.00 216287.00
-158.18 7517.98 5919.63
-1187.15 24.0 256479.00
-345503.00 236771.00
-144.07 6970.57 7508.12
-1317.75 26.0 260950.00
-331434.00 205388.00
-130.57 6497.58 8638.70
-1076.78 28.0 269984.00
-343628.00 218366.00
-134.58 6861.68 8165.52
-1062.58 30.0 278259.00
-348285.00 221391.00
-123.31 6538.19 8720.28
-1076.88 35.0 297697.00
-344053.00 202586.00
-105.06 6094.38 9194.58
-852.15 40.0 331243.00
-401432.00 313358.00
-81.82 5561.33 7636.50
-1470.42 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-97 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 9x9C/D (Page 6 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 5464.88 5428.10
-438.30
-184.75 2311.58
-146.87
-2296.51 1.25 7470.50 6970.15
-717.25
-187.88 3079.06
-374.49
-1640.55 1.5 9709.06 8904.78
-1225.72
-190.25 3758.72
-621.81
-1207.57 1.75 12213.40 11153.30
-1982.54
-189.94 4409.97
-917.68
-845.16 2.0 16691.80 11823.60
-2447.14
-185.99 5008.36
-1243.90
-1059.30 2.25 21740.60 12301.10
-3136.66
-173.22 5422.51
-1511.79
-1061.56 2.5 27709.70 11300.00
-3398.46
-167.10 5898.90
-1850.17
-1171.40 2.75 33988.10 9774.59
-3696.16
-158.15 6268.38
-2155.04
-974.14 3.0 41117.20 6515.41
-3381.03
-148.32 6548.78
-2413.74
-948.98 4.0 71428.60
-18297.80 3576.44
-123.51 7125.21
-2923.50
-632.21 5.0 100397.00
-56458.80 20611.70
-115.75 7125.58
-2528.06
-313.97 6.0 124283.00
-97234.10 43750.10
-135.36 7393.89
-2038.45
-178.07 7.0 142677.00
-131502.00 64937.90
-142.42 7276.64
-994.67
-255.89 8.0 158111.00
-164750.00 89150.00
-165.13 7682.79
-614.18
-382.56 9.0 169539.00
-187815.00 105688.00
-170.16 7701.54 95.21
-536.66 10.0 179168.00
-207560.00 120407.00
-172.05 7615.14 907.40
-757.15 11.0 187428.00
-224318.00 133228.00
-170.11 7472.64 1710.47
-885.30 12.0 195546.00
-241540.00 147050.00
-166.19 7281.30 2560.85
-1135.94 13.0 202256.00
-258699.00 164971.00
-182.40 7906.42 2044.37
-1182.19 14.0 207838.00
-268927.00 173192.00
-178.93 7770.91 2703.98
-1224.09 15.0 213979.00
-281611.00 184781.00
-172.75 7552.21 3409.13
-1276.86 16.0 217809.00
-285839.00 187221.00
-168.56 7458.11 3805.42
-1317.69 17.0 223749.00
-297214.00 196642.00
-160.86 7141.47 4676.19
-1362.21 18.0 226075.00
-295937.00 193130.00
-157.66 7127.19 4895.03
-1291.13 19.0 230997.00
-304670.00 201281.00
-150.53 6907.85 5558.32
-1353.07 20.0 238022.00
-324930.00 227066.00
-158.32 7284.25 5103.45
-1464.16 22.0 243676.00
-322706.00 217208.00
-147.77 6978.74 5979.30
-1239.05 24.0 251683.00
-332524.00 227486.00
-137.48 6744.91 6651.45
-1261.39 26.0 256408.00
-321812.00 204514.00
-125.79 6394.39 7373.18
-1135.32 28.0 264537.00
-330729.00 215269.00
-131.03 6864.20 6415.84
-1014.55 30.0 273958.00
-341208.00 225146.00
-115.29 6196.43 7947.39
-1073.39 35.0 292385.00
-333153.00 204415.00
-98.00 5956.86 7222.98
-860.79 40.0 329247.00
-419504.00 371883.00
-71.42 4943.73 7633.01
-1618.27 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-98 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 9x9E/F (Page 7 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 5714.25 5122.01
-403.48
-185.29 2270.44
-155.62
-2262.37 1.25 7108.39 6936.62
-728.46
-194.88 3064.64
-395.01
-1260.12 1.5 9860.40 8382.32
-1120.78
-188.06 3633.63
-600.75
-1040.69 1.75 12950.70 9922.89
-1671.50
-186.14 4214.90
-858.89
-884.98 2.0 16854.60 11084.70
-2322.04
-181.73 4769.99
-1147.12
-810.29 2.25 21630.80 11546.20
-2940.96
-172.49 5228.13
-1436.00
-839.61 2.5 27849.90 10029.20
-2985.66
-164.15 5650.51
-1736.59
-1040.92 2.75 34540.60 7548.11
-2786.62
-154.38 5990.92
-2013.50
-935.15 3.0 41307.10 4337.80
-2362.16
-146.82 6295.85
-2275.82
-884.96 4.0 70768.40
-20480.20 5197.61
-121.39 6876.47
-2797.83
-537.40 5.0 98180.80
-56583.30 21720.10
-115.24 7004.63
-2612.66
-168.15 6.0 120573.00
-94683.40 43765.30
-134.45 7390.91
-2400.88 20.85 7.0 138493.00
-128353.00 65326.00
-141.23 7368.45
-1657.87 2.12 8.0 151304.00
-154813.00 84923.70
-165.48 7997.42
-1799.73
-3.75 9.0 162835.00
-178601.00 102770.00
-169.20 8012.87
-1222.27
-178.21 10.0 173089.00
-200396.00 119704.00
-169.43 7906.04
-489.94
-481.35 11.0 180227.00
-213998.00 130552.00
-169.48 7924.61
-143.28
-537.04 12.0 188058.00
-230819.00 144797.00
-165.45 7782.15 482.35
-705.69 13.0 193490.00
-240795.00 153382.00
-163.80 7756.04 834.76
-753.66 14.0 199338.00
-255751.00 170303.00
-178.59 8424.78 16.81
-795.55 15.0 204471.00
-264530.00 177215.00
-172.61 8186.47 708.91
-873.25 16.0 209807.00
-275635.00 189071.00
-167.97 8087.71 1042.99
-936.73 17.0 214452.00
-282609.00 194830.00
-159.86 7819.12 1616.41
-906.17 18.0 217197.00
-283928.00 195786.00
-157.56 7869.81 1568.69
-890.15 19.0 221266.00
-288837.00 199363.00
-149.64 7592.40 2213.50
-965.82 20.0 225737.00
-295774.00 205279.00
-143.23 7337.40 2875.11
-876.23 22.0 234598.00
-314227.00 231133.00
-148.51 7825.76 2021.35
-879.15 24.0 242046.00
-320606.00 235951.00
-134.75 7367.58 2926.98
-913.50 26.0 247960.00
-318479.00 229552.00
-123.51 7133.33 3171.11
-783.22 28.0 261521.00
-352854.00 278305.00
-120.41 7120.21 3024.72
-1121.44 30.0 264913.00
-340198.00 263913.00
-111.92 6968.28 2888.33
-788.23 35.0 288082.00
-360268.00 293412.00
-86.40 6220.44 2894.70
-961.02 40.0 298948.00
-303570.00 215523.00
-55.72 5417.82 785.23
-415.39 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-99 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 9X9G (Page 8 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 6976.18 5184.24
-373.54
-181.54 2360.15
-124.29
-2514.26 1.25 9143.56 6689.33
-616.47
-187.09 3235.36
-396.20
-1745.46 1.5 11054.30 9053.26
-1217.15
-187.98 3963.72
-671.80
-1106.16 1.75 13609.90 11584.70
-2067.04
-188.84 4708.46
-1026.29
-782.65 2.0 18157.70 12664.10
-2736.69
-182.35 5344.31
-1383.11
-916.79 2.25 23646.70 12752.10
-3248.16
-178.95 5971.94
-1793.73
-925.78 2.5 29660.10 12309.80
-3821.64
-169.21 6473.09
-2183.65
-879.92 2.75 36525.80 10358.80
-3962.11
-162.46 6968.29
-2613.38
-863.49 3.0 44006.40 7030.85
-3698.49
-153.38 7336.54
-2971.63
-809.92 4.0 77288.30
-21207.50 4543.15
-125.70 8058.78
-3705.78
-537.87 5.0 110686.00
-69960.20 29062.30
-130.54 8442.77
-3626.36
-336.85 6.0 137786.00
-118830.00 58088.00
-136.52 8339.36
-2532.48
-201.40 7.0 160795.00
-169293.00 94340.50
-161.16 8672.27
-1671.25
-379.07 8.0 177763.00
-207034.00 122389.00
-170.18 8619.96
-400.24
-562.99 9.0 193108.00
-243101.00 150849.00
-171.94 8368.05 1156.18
-881.11 10.0 205042.00
-275555.00 181997.00
-195.35 9071.69 1098.87
-1083.51 11.0 215280.00
-300568.00 204362.00
-194.55 8934.09 2200.13
-1266.10 12.0 223585.00
-319189.00 220301.00
-191.69 8775.21 3201.84
-1325.62 13.0 230947.00
-335777.00 234994.00
-189.96 8659.97 4110.52
-1472.39 14.0 239135.00
-355478.00 253619.00
-183.93 8406.36 5194.67
-1726.13 15.0 245572.00
-374776.00 278406.00
-203.34 9278.36 4194.86
-1666.34 16.0 251881.00
-387322.00 288544.00
-193.80 8836.24 5557.89
-1689.56 17.0 257861.00
-401610.00 304798.00
-189.68 8737.81 6220.47
-1840.71 18.0 262232.00
-408488.00 311370.00
-185.11 8602.16 6925.67
-1728.75 19.0 265329.00
-406025.00 301388.00
-178.52 8347.70 7730.36
-1689.95 20.0 271234.00
-419055.00 315509.00
-171.72 8067.36 8751.47
-1705.40 22.0 283895.00
-451199.00 356261.00
-175.40 8389.72 8926.87
-1890.66 24.0 288388.00
-437401.00 323902.00
-164.80 8075.31 9968.86
-1575.02 26.0 299757.00
-459004.00 349014.00
-154.15 7793.16 11086.10
-1690.60 28.0 312233.00
-487890.00 389532.00
-156.41 8001.62 11248.70
-1695.28 30.0 317451.00
-470929.00 352843.00
-144.12 7616.90 12129.50
-1519.49 35.0 340908.00
-472938.00 320383.00
-126.33 6958.19 14189.40
-1265.87 40.0 355826.00
-406707.00 181832.00
-109.88 6567.54 13350.90
-690.33 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-100 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 10x10A/B/G (Page 9 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 5723.53 4982.96
-365.36
-189.50 2319.36
-173.73
-2587.84 1.25 7460.31 6576.80
-660.70
-190.93 3005.93
-368.19
-1663.24 1.5 8981.27 8950.03
-1364.94
-188.17 3629.28
-596.80
-988.51 1.75 12283.70 10322.70
-1785.67
-185.78 4201.56
-850.07
-879.86 2.0 16284.00 11316.60
-2373.42
-183.95 4757.49
-1129.72
-908.53 2.25 21494.10 11161.90
-2738.06
-174.87 5233.98
-1435.08
-1029.88 2.5 27378.90 10122.70
-3001.13
-163.37 5590.72
-1687.18
-1133.76 2.75 33997.50 7667.21
-2796.85
-154.59 5934.47
-1960.21
-1063.93 3.0 40669.30 4604.85
-2427.68
-146.64 6233.46
-2224.40
-1023.08 4.0 69456.60
-19048.60 4510.80
-121.07 6769.53
-2693.26
-595.32 5.0 96363.50
-53810.50 20060.80
-115.15 6852.01
-2455.28
-235.29 6.0 118075.00
-89649.00 40101.30
-135.03 7207.34
-2199.03
-31.82 7.0 135465.00
-121448.00 59891.00
-141.81 7176.22
-1464.52
-84.35 8.0 149172.00
-147759.00 77477.10
-146.29 7123.94
-720.75
-270.69 9.0 160098.00
-171854.00 96698.30
-168.49 7716.07
-861.33
-341.94 10.0 168703.00
-188210.00 108590.00
-170.65 7707.01
-369.98
-413.26 11.0 176895.00
-205123.00 122221.00
-167.56 7590.63 267.07
-597.28 12.0 183500.00
-217775.00 132403.00
-165.29 7503.92 748.16
-696.44 13.0 189527.00
-229054.00 141757.00
-162.77 7481.92 1050.96
-848.98 14.0 195892.00
-241671.00 152138.00
-155.37 7192.81 1854.09
-983.23 15.0 199561.00
-249322.00 161820.00
-172.75 7962.69 824.80
-863.19 16.0 204447.00
-258563.00 171271.00
-167.33 7839.02 1163.01
-928.77 17.0 209187.00
-266807.00 178586.00
-160.49 7588.94 1870.46
-983.28 18.0 212908.00
-270532.00 180865.00
-155.48 7487.99 2077.63
-955.84 19.0 216478.00
-274912.00 185127.00
-150.92 7417.63 2302.50
-949.30 20.0 219761.00
-276790.00 185299.00
-144.53 7207.71 2794.21
-860.04 22.0 230330.00
-297894.00 208958.00
-142.95 7317.84 2710.62
-1141.54 24.0 235204.00
-296597.00 207242.00
-136.96 7299.78 2658.68
-881.02 26.0 243035.00
-302622.00 210474.00
-120.72 6753.85 3686.66
-891.14 28.0 250446.00
-307503.00 216130.00
-107.51 6366.92 4185.55
-863.84 30.0 265199.00
-348982.00 280458.00
-107.22 6539.80 3562.03
-1192.36 35.0 273468.00
-298369.00 203934.00
-79.97 5875.23 3082.40
-627.85 40.0 292898.00
-285148.00 187876.00
-50.41 4835.07 2436.15
-509.94 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-101 Table 2.1.29 BWR FUEL ASSEMBLY COOLING TIME-DEPLETION COEFFICIENTS (ZR-CLAD FUEL)
ARRAY/CLASS 10x10C (Page 10 of 10)
Cooling Time (years)
A B
C D
E F
G 1.0 6190.24 5096.25
-384.43
-186.75 2340.62
-151.36
-2528.72 1.25 7815.16 6793.14
-675.23
-193.97 3174.53
-405.81
-1662.94 1.5 10010.30 8798.90
-1199.94
-193.60 3870.99
-669.85
-1247.50 1.75 13229.50 10326.80
-1757.80
-189.26 4471.71
-940.69
-1117.18 2.0 17325.30 11490.30
-2423.96
-183.30 5030.60
-1243.75
-1042.41 2.25 22130.00 11951.30
-2993.28
-179.73 5638.45
-1641.98
-1049.38 2.5 28141.40 10893.00
-3249.42
-171.97 6092.80
-1970.42
-1042.25 2.75 35001.90 8485.77
-3132.08
-161.49 6464.02
-2288.54
-1064.03 3.0 41817.40 5588.18
-2935.15
-152.37 6778.27
-2580.33
-960.42 4.0 72503.80
-20126.90 4676.40
-126.12 7389.26
-3161.51
-598.75 5.0 101686.00
-58844.80 22172.30
-118.88 7430.83
-2824.08
-314.90 6.0 125964.00
-100714.00 46115.40
-137.38 7670.65
-2280.40
-139.13 7.0 145279.00
-138063.00 69971.00
-145.81 7593.29
-1239.47
-240.17 8.0 160736.00
-171770.00 94922.90
-169.48 8074.18
-936.98
-413.14 9.0 173109.00
-198050.00 114195.00
-173.24 7952.04 107.22
-587.69 10.0 183348.00
-219689.00 130706.00
-174.38 7886.25 887.26
-747.19 11.0 192349.00
-239413.00 146643.00
-173.03 7738.68 1801.89
-960.79 12.0 198722.00
-251849.00 156661.00
-174.40 7779.41 2247.21
-1024.32 13.0 206317.00
-271870.00 177242.00
-191.21 8405.58 1825.60
-1138.70 14.0 212647.00
-284224.00 187282.00
-183.63 8103.28 2759.09
-1219.61 15.0 218920.00
-297923.00 200391.00
-179.50 7978.82 3335.37
-1313.57 16.0 223379.00
-304963.00 206476.00
-175.76 7922.23 3689.54
-1328.16 17.0 228676.00
-314595.00 214380.00
-168.29 7569.76 4728.35
-1384.57 18.0 233175.00
-321606.00 220636.00
-164.63 7582.84 4872.65
-1394.73 19.0 238334.00
-334048.00 236292.00
-170.69 7886.97 4618.40
-1403.78 20.0 242429.00
-340497.00 242818.00
-172.36 8094.92 4434.37
-1437.97 22.0 251428.00
-353397.00 253878.00
-156.59 7500.41 6060.21
-1412.04 24.0 257957.00
-354461.00 249954.00
-147.71 7305.10 6634.39
-1346.94 26.0 272010.00
-391459.00 299301.00
-145.25 7227.25 7258.81
-1619.05 28.0 273995.00
-368436.00 261102.00
-136.90 7071.78 7562.48
-1159.20 30.0 279666.00
-356857.00 232864.00
-125.34 6696.43 8273.08
-973.58 35.0 297242.00
-340805.00 191056.00
-108.66 6404.77 8127.91
-777.55 40.0 330405.00
-398218.00 299749.00
-84.01 5531.03 7980.06
-1232.79 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-117 FIGURE 2.1.10: DAMAGED FUEL ISOLATOR TOP/BOTTOM CAP ASSEMBLY (TYPICAL) 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-118 FIGURE 2.1.10 (CONTINUED): DAMAGED FUEL ISOLATOR IN BASKET CELL (TYPICAL) 5014855 Attachment 5
TABLE 2.2.6 MATERIALS AND COMPONENTS OF THE HI-STORM 100 SYSTEM MPC (1,2)
Notes:
1)
There are no known residuals on finished component surfaces 2)
All welding processes used in welding the components shall be qualified in accordance with the requirements of ASME Section IX. All welds shall be made using welders qualified in accordance with ASME Section IX. Weld material shall meet the requirements of ASME Section II and the applicable Subsection of ASME Section III.
3)
Component nomenclature taken from Bill of Materials in Chapter 1.
4)
A, B, and C denote important to safety classifications as described in the Holtec QA Program. NITS stands for Not Important to Safety.
5)
For details on Alloy X material, see Appendix 1.A. It is also noted that duplex stainless steel shall not be used for the fabrication of MPC baskets and internal components..
6)
Must be Type 304, 304LN, 316, or 316 LN with tensile strength > 75 ksi, yield strength > 30 ksi and chemical properties per ASTM A554.
7)
Corrosion resistant alloy steel eg. 304 stainless steel, Monel, etc.
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-147 Primary Function Component (3)
Safety Class (4)
Codes/Standards (as applicable to component)
Material Strength ( ksi)
Special Surface Finish/Coating Contact Matl.
( if dissimilar)
Structural Integrity Vent Shield Block Spacer C
Non-code Alloy X See Appendix 1.A NA NA Operations Vent and Drain Tube C
Non-code S/S Per ASME Section II Thread area surface hardened NA Operations Vent & Drain Cap C
Non-code S/S Per ASME Section II NA NA Operations Vent & Drain Cap Seal Washer NITS Non-code Aluminum NA NA Aluminum/SS Operations Vent & Drain Cap Seal Washer Bolt NITS Non-code Aluminum NA NA NA Operations Reducer NITS Non-code Alloy X See Appendix 1.A NA NA Operations Drain Line NITS Non-code Alloy X See Appendix 1.A NA NA Operations Damaged Fuel Container C
ASME Section III; Subsection NG S/S (Primarily 304 S/S)
See Appendix 1.A NA NA Operations Damaged Fuel Isolator C
ASME Section II; Subsection IX S/S(7)
NA NA NA Operations Drain Line Guide Tube NITS Non-code S/S NA NA NA 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2-162 Table 2.2.7 HI-STORM 100 ASME BOILER AND PRESSURE VESSEL CODE APPLICABILITY HI-STORM 100 Component Material Procurement Design Fabrication Inspection Overpack steel structure Section II,Section III, Subsection NF, NF-2000 Section III, Subsection NF, NF-3200 Section III, Subsection NF, NF-4000 Section III, Subsection NF, NF-5350, NF-5360 and Section V Anchor Studs for HI-STORM 100A Section II,Section III, Subsection NF, NF-2000*
Section III, Subsection NF, NF-3300 NA NA MPC confinement boundary Section II,Section III, Subsection NB, NB-2000 Section III, Subsection NB, NB-3200 Section III, Subsection NB, NB-4000 Section III, Subsection NB, NB-5000 and Section V MPC fuel basket Section II,Section III, Subsection NG, NG-2000; core support structures (NG-1121)
Section III, Subsection NG, NG-3300 and NG-3200; core support structures (NG-1121)
Section III, Subsection NG, NG-4000; core support structures (NG-1121)
Section III, Subsection NG, NG-5000 and Section V; core support structures (NG-1121)
HI-TRAC TrunnionsSection II, Section III, Subsection NF, NF-2000 ANSI N14.6 Section III, Subsection NF, NF-4000 See Chapter 9 MPC basket supports (Angled Plates)
Section II,Section III, Subsection NG, NG-2000; internal structures (NG-1122)
Section III, Subsection NG, NG-3300 and NG-3200; internal structures (NG-1122)
Section III, Subsection NG, NG-4000; internal structures (NG-1122)
Section III, Subsection NG, NG-5000 and Section V; internal structures (NG-1122)
HI-TRAC steel structure Section II,Section III, Subsection NF, NF-2000 Section III, Subsection NF, NF-3300 Section III, Subsection NF, NF-4000 Section III, Subsection NF, NF-5360 and Section V Damaged fuel container Section II,Section III, Subsection NG, NG-2000 Non-Code Section III, Subsection NG, NG-4000 Section III, Subsection NG, NG-5000 and Section V Damaged fuel isolator Section II Section III, Subsection NG, NG-3300 and NG-3200 Section IX Section V Overpack concrete ACI 349 as specified by Appendix 1.D ACI 349 and ACI 318.1-89(92) as specified by Appendix 1.D ACI 349 as specified by Appendix 1.D ACI 349 as specified by Appendix 1.D Table 2.2.8
- Except impact testing shall be determined based on service temperature and material type.
Section V applies to Code welds only unless otherwise noted.
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-2 non-mechanistic). The basis for the lateral deflection limit in the active fuel region,, is provided in
[2.III.6.1] as w
=
where is defined as the maximum total deflection sustained by the basket panels under the loading event and w is the nominal inside (width) dimension of the storage cell. The limiting value of is provided in Table 2.III.4. The above deflection-based criterion has been used previously in the HI-STAR 180 Transportation Package [2.III.6.2] to qualify similar Metamic-HT fuel baskets.
ii.
Thermal The design and operation of the HI-STORM 100 System with the MPC-68M must meet the intent of the review guidance contained in ISG-11, Revision 3 [2.0.8] as described in Subsection 2.0.1.
All applicable material design temperature limits in Section 2.2 and 4.3 continue to apply to the MPC-68M. Temperature limits of MPC-68M fuel basket and basket shim materials are specified in Table 4.III.2.
The MPC-68M is designed for both uniform and regionalized fuel loading strategies as described in Subsection 2.0.1. The regions for the MPC-68M are given in Table 2.III.1. Additionally, four quarter-symmetric heat load patterns have been defined for MPC-68M as shown in Figures 2.III.1 through 2.III.4. The same temperature limits apply to these configurations.
iii.
Shielding Same as Subsection 2.0.1.
iv.
Criticality Same as Subsection 2.0.1 with the clarifications herein.
Criticality control is maintained by the geometric spacing of the fuel assemblies and spatially distributed B-10 isotope in the Metamic-HT. No soluble boron is required in the MPC-68M water.
The minimum specified boron concentration in the Metamic-HT purchasing specification must be met in every lot of the material manufactured. No credit is taken for burnup. Enrichment limits are delineated in Table 2.III.2.
v.
Confinement Same as Subsection 2.0.1 vi.
Operations 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-3 Same as Subsection 2.0.1. Generic operating procedures for the HI-STORM 100 System with MPC-68M are provided in Chapter 8 with certain limitations and clarifications provide in Supplement 8.III.
vii.
Acceptance Tests and Maintenance Same as Subsection 2.0.1. The acceptance criteria for the HI-STORM 100 System with MPC-68M are provided in Chapter 9 and Supplement 9.III.
vi.
Decommissioning The MPC is designed to be transportable in a HI-STAR overpack and is not required to be unloaded prior to shipment off-site. Decommissioning of the HI-STORM 100 System is addressed in Section 2.III.4.
2.III.1 SPENT FUEL TO BE STORED Table 2.1.22 and the limitations/clarification in this supplement provide the limits for material to be stored in the MPC-68M. All BWR fuel assembly array/classes which are authorized for the MPC-68 are authorized in the MPC-68M except fuel assembly array/classes 6x6A, 6x6B, 6x6C, 7x7A, 8x8A, 10x10D, and 10x10E. Table 2.1.4 in Chapter 2 provides the acceptable fuel characteristics for the fuel array/class authorized for storage in the MPC-68M, however fuel with planar-average initial enrichments up to 4.8 wt.% U-235 are authorized in the MPC-68M. The maximum planar-average initial enrichments acceptable for loading in the MPC-68M, for each fuel assembly array/class given in Table 2.1.4, are provided in Table 2.III.2. Table 2.III.3 provides the description of two new fuel assembly array/classes which are added as acceptable contents to the MPC-68M only, 10x10F and 10x10G. No credit is taken for fuel burnup or integral poisons such as gadolinia for any fuel assembly array/class. The maximum allowable initial enrichment for fuel assemblies are consistent with the criticality analysis described in Supplement 6.III.
Fuel classified as damaged fuel assemblies will be loaded into damaged fuel containers (DFCs) or basket cell locations with DFIs installed at the upper and lower ends. Fuel debris will be loaded into DFCs for storage in the MPC-68M. Damaged fuel assemblies stored with DFIs may contain missing or partial fuel rods and/or fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks as long as the fuel assembly can be handled by normal means and whose structural integrity is such that geometric rearrangement of fuel is not expected. Damaged fuel that does not meet these conditions must be stored in a DFC. The appropriate thermal and criticality analyses have been performed to account for damaged fuel and fuel debris and are described in Supplements 4.III and 6.III, respectively. Figures 2.III.1 through 2.III.4 contain loading patterns for storage of fuel in the MPC-68M. The loading pattern in Figure 2.III.4, allows damaged fuel to be stored in inner locations. Non-fuel hardware is not applicable to all the BWR fuel classes/
arrays.
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-6 Table 2.III.2: BWR FUEL ASSEMBLY INITIAL ENRICHMENTS FOR LOADING IN MPC-68M (Note 1)
Fuel Assembly Array and Class Maximum Planar-Average Initial Enrichment (wt.%235U)
(Note 3, 4) 7x7 B 4.8 8x8 B 4.8 8x8 C 4.8 8x8 D 4.8 8x8 E 4.8 8x8 F 4.5 (Note 2) 9x9 A 4.8 9x9 B 4.8 9x9 C 4.8 9x9 D 4.8 9x9 E 4.5 (Note 2) 9x9 F 4.5 (Note 2) 9x9 G 4.8 10x10 A 4.8 10x10 B 4.8 10x10 C
4.8 Notes
1.
All other fuel assembly array/class specifications from Table 2.1.4 apply.
2.
Fuel assemblies classified as damaged fuel assemblies are limited to 4.0 wt.% U-235 except when loaded in the pattern as shown in Figure 2.III.4. Fuel assemblies classified as damaged fuel assemblies are limited to 4.5 wt.% U-235when loaded in the pattern as shown in Figure 2.III.4.
3.
For MPC-68M loaded with both undamaged fuel assemblies and damaged fuel assemblies or fuel debris, the maximum planar average initial enrichment for the undamaged fuel assemblies is limited to the enrichment of the damaged assembly.
4.
In accordance with the definition of undamaged fuel assembly, certain fuel assemblies may be limited to 3.3 wt.% U-235. When loading these fuel assemblies, all fuel assemblies in the MPC are limited to 3.3 wt.% U-235.
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-7 Table 2.III.3: BWR FUEL ASSEMBLY CHARACTERISTICS FOR LOADING IN MPC-68M (Note 1)
Fuel Assembly Array and Class 10x10F 10x10G Clad Material (Note 2)
Zr Zr Design Initial U (kg/assy.)
(Note 3) 192 188 Maximum Planar-Average Initial Enrichment (wt.%
235U) (Note 8, 9) 4.7 (Note 7) 4.75 (Note 10)
Initial Rod Maximum Enrichment (wt.% 235U) 5.0 5.0 No. of Fuel Rod Locations 92/78 (Note 4) 96/84 Fuel Clad O.D. (in.)
0.4035 0.387 Fuel Clad I.D. (in.)
0.3570 0.340 Fuel Pellet Dia. (in.)
< 0.3500
< 0.334 Fuel Rod Pitch (in.)
0.510 0.512 Design Active Fuel Length (in.)
< 150
< 150 No. of Water Rods (Note 6) 2 5
(Note 5)
Water Rod Thickness (in.)
> 0.030
> 0.031 Channel Thickness (in.)
0.120 0.060 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-8 Table 2.III.3 (continued)
BWR FUEL ASSEMBLY CHARACTERISTICS NOTES:
1.
All dimensions are design nominal values. Maximum and minimum dimensions are specified to bound variations in design nominal values among fuel assemblies within a given array/class.
2.
See glossary for the definition of ZR.
3.
Design initial uranium weight is the nominal uranium weight specified for each assembly by the fuel manufacturer or reactor user. For each BWR fuel assembly, the total uranium weight limit specified in this table may be increased up to 1.5 percent for comparison with users fuel records to account for manufacturer tolerances.
4.
This assembly contains 92 total fuel rods; 78 full length rods and 14 partial length rods.
5.
One diamond-shaped water rod replacing the four center fuel rods and four rectangular water rods dividing the assembly into four quadrants.
6.
These rods may also be sealed at both ends and contain ZR material in lieu of water.
7.
Fuel assemblies classified as damaged fuel assemblies are limited to 4.6 wt.%
U-235 for the 10x10F array/class.
8.
For MPC-68M loaded with both undamaged fuel assemblies and damaged fuel assemblies or fuel debris, the maximum planar average initial enrichment for the undamaged fuel assemblies is limited to the enrichment of the damaged assembly.
9.
In accordance with the definition of UNDAMAGED FUEL ASSEMBLY, certain assemblies may be limited to 3.3 wt.% U-235. When loading these fuel assemblies, all other undamaged fuel assemblies in the MPC are limited to enrichments specified in this table and Table 2.III.2.
10.
Fuel assemblies classified as damaged fuel assemblies are limited to 4.6 wt.%
U-235 for the 10x10G array/class except when loaded to Figure 2.III.4. Fuel assemblies classified as damaged fuel assemblies are limited to 4.5 wt.% U-235 for the 10x10G array/class when loaded to Figure 2.III.4 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-10 1
0.5*
2 0.5*
3 0.5*
4 0.5 5
1.2 6
1.2 7
0.5 8
0.5*
9 0.5*
10 0.5 11 1.2 12 0.4 13 0.4 14 1.2 15 0.5 16 0.5*
17 0.5 18 1.2 19 0.4 20 0.4 21 0.4 22 0.4 23 1.2 24 0.5 25 0.5*
26 1.2 27 0.4 28 0.4 29 0.4 30 0.4 31 0.4 32 0.4 33 1.2 34 0.5*
35 0.5*
36 1.2 37 0.4 38 0.4 39 0.4 40 0.4 41 0.4 42 0.4 43 1.2 44 0.5*
45 0.5 46 1.2 47 0.4 48 0.4 49 0.4 50 0.4 51 1.2 52 0.5 53 0.5*
54 0.5 55 1.2 56 0.4 57 0.4 58 1.2 59 0.5 60 0.5*
61 0.5*
62 0.5 63 1.2 64 1.2 65 0.5 66 0.5*
Cell ID 67 0.5*
68 0.5*
Heat Load (kW)
- Note: This figure provides per cell allowable heat loads for MPC-68M with all UNDAMAGED FUEL assemblies. For MPC-68M with DAMAGED FUEL and/or FUEL DEBRIS stored in this location (in a DFC), the per cell allowable heat load of the cell is limited to 0.35 kW. When DFIs are utilized for DAMAGED FUEL the value in the figure applies.
Figure 2.III.1 QSHL Pattern Per Cell Allowable Heat Loads (kW) - MPC-68M 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-11
- DFCs/DFIs are allowed in shaded cells. When DAMAGED FUEL or FUEL DEBRIS (in a DFC) is stored in this location, the allowable heat load of the cell is reduced by 25%. When DFIs are utilized for DAMAGED FUEL the value in the figure applies.
Figure 2.III.2 QSHL-1 Pattern Per Cell Allowable Heat Loads (kW) - MPC-68M 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-12
- DFCs/DFIs are allowed in shaded cells. When DAMAGED FUEL or FUEL DEBRIS (in a DFC) is stored in this location, the allowable heat load of the cell is reduced by 25%. When DFIs are utilized for DAMAGED FUEL the value in the figure applies.
Figure 2.III.3 QSHL-2 Pattern Per Cell Allowable Heat Loads (kW) - MPC-68M 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 2.III-13
- DFCs/DFIs are allowed in shaded cells. Cells 19, 22, 47 and 50 must remain empty.
Figure 2.III.4 QSHL-3 Pattern Per Cell Allowable Heat Loads (kW) - MPC-68M 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-1 SUPPLEMENT 4.III1 THERMAL EVALUATION OF THE MPC-68M 4.III.0 OVERVIEW The MPC-68M is a 68 cell BWR canister engineered with a high B10 containing Metamic-HT basket for enhanced criticality control. The MPC-68M is evaluated for storage in the aboveground family of HI-STORM overpacks. For a bounding evaluation an MPC-68M emplaced in the most flow resistive HI-STORM 100S Version B overpack2 is analyzed under normal, off-normal and accident conditions. The evaluations described herein parallel those of the aboveground HI-STORM cask contained in the main body of Chapter 4 of this FSAR. In addition, MPC-68M permits a heat load layout defined as the "Quarter Symmetric Heat Load" (QSHL) pattern. In this pattern, the maximum permissible heat load in each storage cell, q, is specific to its location within the quadrant and is limited to a unique prescribed value given in Figure 2.III.1. This QSHL pattern seeks to minimize the large temperature differences between cladding temperatures in proximate fuel assemblies and is especially suited for canisterizing of fuel with widely varying specific heat loads such as at a plant undergoing decommissioning. To maximize fuel loading flexibility additional heat load patterns QSHL-2, QSHL-3 and QSHL-4 are defined in Figures 2.III.2, 2.III.3 and 2.III.4 and evaluated herein.
It should be noted that the QSHL patterns are special cases of regionalized loading..
To ensure readability, the section in the main body of the chapter to which each section in this supplement corresponds is clearly identified. All tables in this supplement are labeled sequentially.
4.III.1 INTRODUCTION The information presented in this supplement is intended to serve as a complement to the information provided in the main body of Chapter 4. Except for the fuel basket and basket support materials, the information in Chapter 4 that remains applicable to the MPC-68M analysis is not repeated herein. Specifically the following information in the main body of Chapter 4 is not repeated:
- 1. The thermal properties of materials in Section 4.2 applicable to the MPC-68M.
- 2. The specifications for components in Section 4.3 applicable to the MPC-68M.
- 3. The descriptions of the thermal modeling of the MPC and its internals, including fuel assemblies, in Section 4.4 which are applicable in their entirety to the MPC-68M.
1 For ease of supplement review the sections are numbered in parallel with the main Chapter 4.
2 This approach is identical to the HI-STORM thermal analysis in Section 4.4.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-4 in the Holtec-proprietary benchmarking report [4.1.6]. This is the same modeling approach used in the HI-STORM 100 cask analyses. The underside of the HI-STORM 100 concrete pad is assumed to be supported on a subgrade at 77ºF. This is the same boundary condition applied to the bottom of the ISFSI pad for the HI-STORM 100 modeling in Section 4.4.
4.III.4.2 Thermal Analysis The MPC-68M has been designed to permit storage under the array of uniform and regionalized heat loads defined in Chapter 2 as a function of the regionalization parameter X. As shown in Chapter 4 the highest cladding temperatures are reached under regionalized storage at X = 0.5. This scenario is co-incident with the maximum permissible MPC heat load and therefore temperatures of other sub-systems (such as fuel basket, MPC shell and overpack) also reach their highest values. The fuel cladding temperature under long term storage in HI-STORM is presented in Table 4.III.3.a The QSHL patterns defined in Supplement 2.III are also analyzed using the same FLUENT model previously used in this FSAR: no changes were made to the existing thermal model. The selected heat loads in Figure 2.III.1, 2.III.2, 2.III.3 and 2.III.4 are suitably limited to ensure that the peak cladding temperature in the MPC remains below that in the governing MPC analyzed in this FSAR (MPC-32) under all thermal scenarios. The results are tabulated in Table 4.III.3a. Thus the peak cladding temperature for the QSHL patterns are limited by a previously analyzed and licensed MPC (See Table 4.4.6).
Other important safety aspects of the QSHL patterns are:
- 1. The hottest fuel assemblies are located in-board of the peripheral locations in the basket so that the colder fuel in the peripheral cells helps block the radiation emitted by the hottest fuel assemblies.
- 2. The cell specific heat load, q, provided in Figures 2.III.1 thru 2.III.4 are the maximum value permitted for that location. In virtually every case, the actual heat load in every cell will be lower than the allowed limit, thus resulting in a lower cladding temperature field overall than that computed herein.
- 3. The fuel cladding temperature for QSHL patterns under long term storage in HI-STORM are presented in Table 4.III.3.a. The predicted PCTs are higher than that for the scenario with decay heat based on regionalized parameter X defined in Chapter 2. Within these patterns the QSHL pattern defined in Figure 2.III.1 is bounding. For this reason, the QSHL pattern is adopted as the licensing basis pattern for MPC-68M.
- 4. The PCT and basket temperatures under the QSHL pattern is lower than that in the thermally governing case (MPC-32).
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-5 The QSHL scenario is adopted for demonstration of compliance with the temperature and pressure limits set forth in this Supplement and Chapter 2. The limiting scenario is analyzed and maximum temperatures and pressures under normal storage tabulated in Tables 4.III.3b and 4.III.4. The results are below the Chapter 2 and Supplement 4.III normal temperature and pressure limits. In accordance with NUREG-1536 MPC-68M pressures are computed assuming 1% (normal), 10% (off-normal) and 100% (accident) rod ruptures with 100% rods fill gases and fission gases release in accordance with NUREG-1536 release fractions. The pressures are computed and tabulated in Table 4III.4. The 100% rods rupture pressure is below the accident design pressure (Table 2.2.1).
4.III.4.3 Engineered Clearances to Eliminate Thermal Interferences To minimize thermal stresses in load bearing members, the MPC-68M is engineered with adequate gaps to permit free thermal expansion of the fuel basket and MPC in axial and radial directions. In this subsection, differential thermal expansions are evaluated to ensure the adequacy of engineered gaps. The following gaps are evaluated:
- a. Fuel Basket-to-MPC Radial Gap
- b. Fuel Basket-to-MPC Axial Gap
- c. MPC-to-Overpack Radial Gap
- d. MPC-to-Overpack Axial Gap The FLUENT thermal model articulated above provides the temperature field in the HI-STORM overpack and MPC-68M from which the changes in the above gaps are directly computed. The nominal cold gaps are presented on the drawings in Section 1.5 and the corresponding differential expansions under normal storage conditions are presented in Table 4.III.8. The calculations show significant margins against restraint to free-end expansion are available in the design.
4.III.4.4 Evaluation of Fuel Debris Storage Fuel debris is permitted for storage in up to eight peripheral cells under the permitted heat load pattern shown in Figure 2.III.1. Although fuel debris is not required to meet cladding temperature limits, its effect on fuel stored in the interior cells must be assessed. Fuel debris in the canister is thermally conservatively evaluated assuming a bounding debris configuration and design heat load in all storage cells. The following assumptions are adopted to maximize the computed cladding temperatures:
- 1.
The fuel debris is assumed to be completely pulverized and compacted into a square prismatic bar enclosed by the damaged fuel canister (DFC) with open helium space above it. In this manner the height of the prismatic bar emitting heat is minimized resulting in the maximization of lineal thermal loading (kw/ft) of the DFC and co-incident local heating of the fuel basket and neighboring storage cells.
- 2.
Fuel debris assumed to be completely composed of UO2. As UO2 has a lower conductivity relative to cladding, heat dissipation is understated.
- 3.
The fuel debris is assumed to block through flow of helium inside the DFC.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-6
- 4.
All 16 peripheral storage locations (not just the 8 permitted by CoC) are assumed to contain fuel debris emitting maximum heat permitted by Technical Specifications (CoC Appendix B, Figure 2.4-1) and all interior cells are emitting design basis heat under the applicable heat load scenario.
- 5.
The MPC operating pressure is understated to minimize internal convection heat transfer The results of the analysis are tabulated in Table 4.III.11. The results support the following conclusions:
Cladding temperature is substantially below the ISG-11, Rev. 3 limit.
MPC basket is below the design limit (Table 4.III.2) by large margin.
MPC shell and Overpack metal temperatures are below design limits (Table 2.2.3).
Overpack body and lid concrete are well below design limits (Table 4.3.1).
4.III.4.5 Evaluation of Fuel in DFI Fuel assemblies classified as damaged fuel which can be handled by normal means may be stored with Damaged Fuel Isolators (DFI) installed at the top and bottom of fuel as defined in Chapter 2, Figure 2.1.101. DFIs are designed with perforated side walls to minimize resistance to helium circulation. In this manner fuel temperatures are not affected and allow penalty free accommodation of design basis heat loads2.
4.III.5 THERMAL EVALUATION OF SHORT TERM OPERATIONS 4.III.5.1 HI-TRAC Thermal Model The HI-TRAC thermal model presented in Section 4.5 is adopted for the evaluation of MPC-68M under short term operations.
4.III.5.2 Maximum Time Limit During Wet Transfer Operations Time-to-boil is calculated using the same methodology described in Section 4.5.2. Table 4.III.13 summarizes the thermal inertia of the constituent components in the loaded HI-TRAC transfer cask.
Using the methodology described in Section 4.5.2, the time-to-boil is provided at representative initial temperatures for maximum QSHL in Table 4.III.14. This is an example calculation for the maximum design basis heat load. The same methodology can be adopted to determine the time-to-boil for canisters loaded at lower heat loads. An alternate method using the FLUENT thermal model described in Section 4.III.5.1 can be adopted to evaluate the time for water within the MPC to boil for site-specific conditions. Principal modeling steps and acceptance criteria are defined in Table 4.5.11.
1 Combination storage of DFCs and DFIs in authorized locations is acceptable.
2 Fuel storage in DFIs supported by suitably bounding calculations archived in supporting Calculation Package
[4.III.5].
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-7 4.III.5.3 MPC Temperature During Moisture Removal Operations 4.III.5.3.1 Vacuum Drying Prior to helium backfill the MPC-68M must be drained of water and demoisturized. At the start of draining operation, both the HI-TRAC annulus and the MPC are full of water. The presence of water in the MPC ensures that the fuel cladding temperatures are lower than design basis limits by large margins. As the heat generating region is uncovered during the draining operation, the fuel and basket mass will undergo a monotonic heat up from the initially cold conditions when the heated surfaces were submerged under water. To limit fuel temperatures demoisturization of the MPC-68M by the vacuum drying method is permitted provided the HI-TRAC annulus remains water filled during vacuum drying operations. To support vacuum drying operations two limiting scenarios are defined below:
Scenario A: The MPC-68M is loaded with Moderate Burnup Fuel assemblies generating heat at the maximum permissible rate defined in Chapter 2 under the bounding regionalized storage scenario X = 0.5.
Scenario B: The MPC-68M is loaded with one or more High Burnup Fuel assemblies and the MPC-68M decay heat is less than a conservatively defined threshold heat load Q =
29 kW1.
Scenario C: The MPC-68M is loaded with Moderate Burnup Fuel assemblies generating heat at the maximum permissible rate defined under QSHL pattern (Figure 2.III.1)2.
To evaluate the above scenarios the vacuum drying analysis methodology presented in Section 4.5 is adopted and an MPC-68M specific thermal model constructed. The principal features of the thermal model are as follows:
- i.
A bounding steady-state analysis is performed under the heat loads defined in the scenarios above.
ii.
The water in the HI-TRAC annulus is conservatively assumed to be boiling under the hydrostatic head of water at the annulus bottom (232oF).
iii.
The bottom surface of the MPC is insulated.
The thermal model articulated above is used to compute the maximum cladding temperature under the vacuum drying scenarios defined above. The results tabulated in Table 4.III.5 are in compliance with the ISG-11 temperature limits of Moderate Burnup Fuel (Scenario A) and High Burnup Fuel (Scenario B).
1 Threshold heat load is defined as the product of maximum loaded assembly heat load rmax and the number of fuel storage cells (n=68). Under this stipulation rmax must not exceed 0.426 kW.
2 Limiting pattern in the QSHL array of patterns evaluated in Table 4.III.3a.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-8 At heat loads greater than threshold heat load defined above, the peak cladding temperature cannot be maintained below the ISG-11, Revision 3 limit of 400oC for HBF under a vacuum condition of infinite duration. Under this scenario, cycles of vacuum drying resulting in heatup followed with cooling by helium are performed until drying criteria is achieved. The thermal model described above is used for heatup/cooldown cycles for site-specific canister heat load maps. It must be ensured per ISG-11 Rev 3 that the repeated thermal cycling is limited to less than 10 cycles, with cladding temperature variations less than 65oC (117oF) each.
4.III.5.3.2 Forced Helium Dehydration Evaluation of Forced Helium Dehydration in Section 4.5 is applicable to MPC-68M.
4.III.5.4 Cask Cooldown and Reflood During Fuel Unloading Operations Evaluation of cask cooldown and reflood operation in Section 4.5 is applicable to MPC-68M.
4.III.5.5 HI-TRAC Onsite Transfer Operation A 3D FLUENT thermal model of an MPC-68M emplaced in a HI-TRAC transfer cask is constructed to evaluate the thermal state of fuel under onsite transport in the vertical orientation1. A bounding analysis is performed under the following conditions:
(i) Steady state maximum temperatures have reached.
(ii) The MPC-68M is loaded with fuel generating heat at the maximum permissible level under the limiting Quarter Symmetric Heat Load (QSHL) pattern (See Table 4.III.3a).
(iii) The HI-TRAC annulus is air filled.
The scenario defined above represents upper bound temperatures reached in the HI-TRAC without the aid of any auxiliary cooling such as the Supplemental Cooling System (SCS) defined in Section 4.5. The maximum cladding temperatures computed using the thermal model articulated above are tabulated in Table 4.III.6. As the cladding temperatures are below the limiting High Burnup Fuel temperature limits mandated by ISG-11 [4.1.4] SCS cooling is not necessary for ensuring cladding safety under onsite transfer operations for the MPC-68M canister. Accordingly SCS cooling is not mandated in the MPC-68M Technical Specifications. Additionally, the peak fuel cladding temperatures are bounded by MPC-32 (Section 4.5).
4.III.5.6 Sensitivity Study In lieu of anodization of the extruded shims used in the MPC-68M, they are passivated in water to form a thin oxide layer. The emissivity of extruded shim surfaces is therefore reduced and requires a thermal evaluation.
1 In accordance with Section 4.5 onsite transfer in the horizontal orientation is not permitted.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-16 As required by NUREG-1536 (4.0,IV,3), the maximum internal pressure of the cask remains within its design pressure for normal, off-normal, and accident conditions. Design pressures are specified in Table 2.2.1.
As required by NUREG-1536 (4.0,IV,4), all cask materials and fuel cladding are maintained within their temperature limits under normal, off-normal and accident conditions to enable them to perform their intended safety functions. Material temperature limits are specified in Tables 2.2.3 and 4.III.2.
As required by NUREG-1536 (4.0,IV,5), the cask system ensures a very low probability of cladding breach during long-term storage. For long-term normal conditions, the maximum CSF cladding temperature is below the ISG-11 limit of 400oC (752oF).
As required by NUREG-1536 (4.0,IV,7), the cask system is passively cooled. All heat rejection mechanisms described in this supplement, including conduction, natural convection, and thermal radiation, are passive.
As required by NUREG-1536 (4.0,IV,8), the thermal performance of the cask is within the normal storage design criteria specified in Chapters 2 and 4. All thermal results are within the limits under normal conditions of storage.
4.III.8 REFERENCES
[4.III.1] Aluminum Alloy 2219 Material Data Sheet, ASM Aerospace Specification Metals, Inc.,
Pompano Beach, FL.
[4.III.2] United States Code of Federal Regulations, Title 10, Part 71.
[4.III.3] Gregory, J.J. et. al., Thermal Measurements in a Series of Large Pool Fires, SAND85-1096, Sandia National Laboratories, (August 1987).
[4.III.4] Jakob, M. and Hawkins, G.A., Elements of Heat Transfer, John Wiley & Sons, New York, (1957).
[4.III.5] HI-STORM Thermal Hydraulic Analysis Supporting Upto 36.9 kW High Heat Load Amendment, HI-2043317, Latest Revision.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 164.C REPORT HI-2002444 4.III-19 Table 4.III.3a: Fuel Loading Patten Screening Evaluations Loading PatternNote 4 Total Decay Heat, kW Peak Cladding Temperature, oF Case 1: X=0.5 (Note 1) 36.9 598 Case 2: QSHL (Note 2) 42.8 708Note 3 Case 3: QSHL-2 (Note 2) 38.92 703 Case 4: QSHL-3 (Note 2) 38.92 701 Case 5: QSHL-4 (Notes 2, 5) 38.52 704 Note 1: The decay heat distribution is described in Section 2.1.9.
Note 2: Quarter symmetric heat load pattern is defined in Figures 2.III.1, 2.III.2, 2.III.3 and 2.III.4.
Note 3: Since the highest PCT is reached for the QSHL pattern, it is adopted for all the licensing basis evaluations of fuel storage in MPC-68M.
Note 4: Cases 1 and 2 are defined in Section 4.III.4.1.
Note 5: Pattern evaluated under bounding scenario wherein all DFC/DFI locations store fuel in DFCs.
Table 4.III.3b: Maximum Temperatures Under Normal Long-Term Storage Component Temperature (oF)
Fuel Cladding 708 Basket 674 Basket Shims 563 MPC Shell 499 Overpack Inner Shell 3581 Overpack Body Concrete2 252 Overpack Lid Concrete2 257 Overpack Outer Shell 190 Area Averaged Air Outlet3 244 1 Nominal exceedance of temperature limits has no risk on its structural integrity.
2 Maximum thru thickness section average temperature reported.
3 Reported herein for the option of outlet ducts air temperature surveillance set forth in the Technical Specifications.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 4.III-20 Table 4.III.4: Maximum Pressures Under Normal Long Term StorageNote 3 Condition Pressure (psig)
Initial maximum backfillNote 1 (at 70F) 46.5 NormalNote 4 intact rods 1% rods ruptureNote 2 98.7 99.2 Off-Normal (10% rods rupture)Note 4 104.0 Accident (100% rods rupture)Note 4 152.0 Note 1 Conservatively assumed at the Tech. Spec. maximum value (see Table 1.III.1).
Note 2 Per NUREG-1536, pressure analysis with ruptured fuel rods is performed with release of 100% of the ruptured fuel rod fill gas and 30% of the significant radioactive gaseous fission products.
Note 3: Results tabulated under QSHL pattern.
Note 4: QSHL tabulated pressures bound QSHL 2, 3 & 4 computed pressures
[4.III.5].
Table 4.III.5: Maximum MPC-68M Temperatures Under Vacuum Drying Scenarios Component Scenario A (oF)
Scenario B (oF)
Scenario C (oF)
Fuel Cladding 754 732 896 Fuel Basket 729 698 854 Basket Shims 522 482 592 MPC Shell 325 307 343 Notes:
(1) The peak cladding temperatures are below the ISG 11 temperature limits of Moderate Burnup Fuel (Scenarios A and C) and High Burnup Fuel (Scenario B).
(2) The component temperatures are below the Chapter 2 and Supplement III temperature limits.
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 5.III-2 higher decay heat, which reduces the allowable assembly burnup. In summary, no new analyses are necessary to qualify those additional array classes.
Therefore, the main body of this chapter remains fully applicable for the HI-STORM 100 System using an MPC-68M and the new assembly classes.
As discussed in Subsection 2.0.1, each MPC basket, except MPC-68F, allows for two loading strategies, namely the uniform fuel loading and the regionalized loading with two regions. The additional 3-region and 5-region loading patterns, shown in Figures 2.III.1 through 2.III.4, are evaluated to determine acceptability as approved contents in the MPC-68M only. This evaluation performs a dose rate comparison between
- a uniform loading pattern (results shown in Table 5.4.9) and a 3-region pattern (Table 5.III.1) in which one region contains 2-year cooled spent nuclear fuel;
- a uniform and a bounding 5-region loading patterns (Table 5.III.3) with a minimum cooling time in some basket locations as low as 1.0 years cooled spent nuclear fuel.
It should be noted that the design basis GE 7x7 source term calculations, discussed in Section 5.2 of the main part of this chapter, are performed using the SCALE 4.3 system [5.1.2, 5.1.3]. The evaluation in this Supplement 5.III is performed with an updated version of SCALE (SCALE 5.1) which is consistent with other approved Holtec applications [5.III.3]. (SCALE 4.3 is no longer maintained by Oak Ridge National Laboratory, and does not work on contemporary computer operating systems.) To ensure that the dose rate comparison is not affected by the SCALE code version, a comparison between results generated using SCALE 4.3 and SCALE 5.1 was performed. There were no significant differences in the neutron and fuel gamma source terms between the two SCALE versions. The Cobalt-60 photon source calculated with SCALE 5.1 were substantially higher than the Cobalt-60 source calculated using SCALE 4.3 which is encompassed in the first two dose rate results columns of Table 5.III.2. The remaining comparisons shown in Table 5.III.2 are performed using the updated version of SCALE (SCALE 5.1), which compared the uniform loading pattern (50,000 MWD/MTU, 3-year cooling time) against the 3-region loading pattern presented in Table 5.III.1 for both the MPC-681 and MPC-68M.
Table 5.III.2 shows the dose rates for the 3-region loading pattern (Figure 2.III.1 and Table 5.III.1) are bounded by the uniform loading pattern (50,000 MWD/MTU and 3-year cooling). For this reason, the 3-region loading pattern shown in Figure 2.III.1 is added as approved contents of the MPC-68M.
1 The 3-region loading pattern shown in Figure 2.III.1 and Table 5.III.1 is approved for the MPC-68M only. All results related to the 3-region loading pattern (in Figure 2.III.1 and Table 5.III.1) in the MPC-68 are for the purposes of comparison only.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 5.III-3 Also, the evaluation of the 5-region loading pattern in this Supplement 5.III is performed with an updated version of MCNP (MCNP5-1.51) [5.III.5] which is consistent with other approved Holtec applications [5.III.6]. To ensure that the dose rate comparison is not affected by the MCNP code version, a comparison between results for the uniform loading pattern generated using MCNP4a and MCNP5 was performed. There were no significant differences in the dose rates between the two MCNP versions. Hence the remaining comparisons shown in Table 5.III.4 are performed using the updated version of MCNP (MCNP5-1.51), which compared the uniform loading pattern (50,000 MWD/MTU, 3-year cooling time) against the 5-region loading pattern presented in Table 5.III.3 for MPC-68M.
Table 5.III.4 shows the dose rates for the 5-region loading pattern (Figures 2.III.2 through 2.III.4 and Table 5.III.3) are bounded by the uniform loading pattern (50,000 MWD/MTU and 3-year cooling). For this reason, the 5-region loading patterns shown in Figures 2.III.2 through 2.III.4 are added as approved contents of MPC-68M. The MPC-68M minimum cooling time limits are provided in Figure 2.III.5.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 5.III-4 5.III.1 REFERENCES
[5.III.1]
I.C. Gauld and O.W. Hermann, "SAS2: A Coupled One-Dimensional Depletion and Shielding Analysis Module," ORNL/TM-2005/39, Revision 5.1,, Oak Ridge National Laboratory, November 2006.
[5.III.2]
I.C. Gauld, O.W. Hermann, and R.M. Westfall, "ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms," ORNL/TM-2005/39, Version 5.1, Oak Ridge National Laboratory, November 2006.
[5.III.3]
Holtec International Report HI-2114830, Final Safety Analysis Report on the HI-STORM FW System, USNRC Docket 72-1032, latest revision.
[5.III.4]
Holtec International Report HI-2146423 Revision 1, HI-STAR 190 Source Terms and Loading Patterns.
[5.III.5]
X-5 Monte Carlo Team, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5, LA-UR-03-1987, Los Alamos National Laboratory, April 2003 (Revised 2/1/2008).
[5.III.6]
Holtec International Report HI-2146214, Safety Analysis Report HI-STAR 190 Cask System, USNRC Docket 71-9373, latest revision.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 5.III-7 Table 5.III.3 5-REGION LOADING PATTERN FOR MPC-68M (Sheet 1 of 2)
Region (Assemblies)
Burnup (MWD/MTU)
Initial Enrichment (wt%)
Cooling Time (years)
Calculated Decay Heat (kW)
Bounding Decay Heat for Shielding Analyses (kW) 1 (12) 65000 4.4 1.8 1.80 1.66 60000 3.7 1.8 1.73 55000 3.6 1.6 1.78 50000 3.0 1.6 1.70 47500 3.0 1.4 1.82 40000 3.0 1.2 1.81 35000 2.6 1.2 1.68 30000 2.2 1.0 1.80 25000 2.1 1.0 1.59 20000 1.6 1.0 1.41 2
(8) 65000 4.4 3.0 1.14 1.10 60000 3.7 2.8 1.16 55000 3.6 2.6 1.14 50000 3.0 2.4 1.16 45000 3.0 2.2 1.15 40000 3.0 2.0 1.13 35000 2.6 1.8 1.15 25000 2.1 1.4 1.17 20000 1.6 1.0 1.41 3
(16) 65000 4.4 10 0.51 0.51 60000 3.7 8
0.52 55000 3.6 6.0 0.55 50000 3.0 5.0 0.57 45000 3.0 4.5 0.55 40000 3.0 4.0 0.53 35000 2.6 3.5 0.55 25000 2.1 2.8 0.53 20000 1.6 2.4 0.55 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 5.III-8 Table 5.III.3 5-REGION LOADING PATTERN FOR MPC-68M (Sheet 2 of 2)
Region (Assemblies)
Burnup (MWD/MTU)
Initial Enrichment (wt%)
Cooling Time (years)
Calculated Decay Heat (kW)
Bounding Decay Heat for Shielding Analyses (kW) 4 (8) 65000 4.4 30 0.31 0.31 60000 3.7 26 0.31 55000 3.6 21 0.31 50000 3.0 16 0.31 47500 3.0 13 0.32 45000 3.0 11 0.31 40000 3.0 7
0.33 35000 2.6 6.0 0.32 25000 2.1 4.0 0.34 20000 1.6 3.5 0.33 5
(24) 65000 4.4 42 0.25 0.25 60000 3.7 37 0.25 55000 3.6 32 0.25 50000 3.0 26 0.25 47500 3.0 23 0.25 45000 3.0 20 0.25 40000 3.0 13 0.25 35000 2.6 9.0 0.25 25000 2.1 5.0 0.26 20000 1.6 4.0 0.28 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 5.III-9 Table 5.III.4 DOSE RATES COMPARISON BETWEEN UNIFORM1 AND 5-REGION2 LOADING PATTERN FOR 100-TON HI-TRAC Dose Point Location Totals 5-Region MPC-68M Totals Uniform MPC-68M (mrem/hr)
(mrem/hr)
ADJACENT TO THE 100-TON HI-TRAC 1
1600.95 1966.8 2
1682.93 2557.0 3
1252.80 1445.2 4
680.63 827.4 5 (pool lid) 7812.79 9061.9 ONE METER FROM THE 100-TON HI-TRAC 1
351.95 511.4 2
731.26 1083.9 3
278.02 354.4 1 Design Basis GE 7x7 spent fuel with a burnup of 50,000 MWD/MTU, cooling time of 3 years, and initial enrichment of 3.6 wt% U-235.
2 As shown in Figures 2.III.2 through 2.III.4 and Table 5.III.3.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.A-3 1) n(n n
/
)
k
(
k 2
n 1
i i
n 1
i 2
i 2
k (6.A.2) k K
)
k (1
Bias (6.A.3) where ki are the calculated reactivities for n critical experiments; k is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); and K is the one-sided multiplier for 95% probability at the 95% confidence level (NBS Handbook 91
[6.A.18]).
Formula 6.A.3 is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented on page 6.A-2. The first portion of the equation, (1-k ), is the actual bias which is added to the MCNP4a and KENO5a results. The second term, K k, which corresponds to B in Section 6.4.3, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95% probability at the 95% confidence level.
The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively.
The larger of the calculational biases (truncated bias) was used to evaluate the maximum keff values for the cask designs.
6.A.1.1 Summary of Benchmark Calculations for the MCNP5-1.51 Code with the ENDF/B-VII Library The same methodology as discussed in Section 6.A.1 is applied in the determination of the bias and standard error of the bias (95% probability at the 95% confidence level) for the MCNP5-1.51 code with the ENDF/B-VII library. These calculations are documented in Appendix C of
[6.A.19] for MCNP5-1.51 with the ENDF/B-VII library; the total bias (systematic error, or mean of the deviation from a keff of exactly 1.000) are shown in the table below.
Calculational Bias of MCNP5-1.51 with the ENDF/B-VII Library Total Truncated MCNP5-1.51 with the ENDF/B-VII Library
-0.0024 +/- 0.0008 0.0004 +/- 0.0003 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.A-4 For calculations made in this report using MCNP5-1.51 with the ENDF/B-VII library, the larger of the calculational biases (truncated bias) was used to evaluate the maximum keff values.
6.A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46% to 5.74% and therefore span the enrichment range for the MPC designs. Figures 6.A.3 and 6.A.4 show the calculated keff values (Table 6.A.1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO5a). Thus, there are no corrections to the bias for the various enrichments.
As further confirmation of the absence of any trends with enrichment, the MPC-68 configuration was calculated with both MCNP4a and KENO5a for various enrichments. The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 6.A.2 and Figure 6.A.5, confirm no significant difference in the calculated values of keff for the two independent codes as evidenced by the 45º slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.
6.A.3 Effect of 10B Loading Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the cask designs. Of these critical experiments, those performed by B&W are the most representative of the cask designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment), the reactivity worth of the absorbers in the PNL tests is very low and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.
Table 6.A.3 lists the subset of experiments using thin neutron absorbers (from Table 6.A.1) and shows the reactivity worth ( k) of the absorber.
The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental ( k) change in reactivity due to the absorber.
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.A-9
[6.A.15]
B.M. Durst et al., Critical Experiments with 4.31 wt % 235U Enriched UO2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.
[6.A.16]
S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1986.
[6.A.17]
E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.
[6.A.18]
M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
[6.A.19]
HI-revision, Holtec International.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-1 SUPPLEMENT 6.III1: CRITICALITY EVALUATION OF THE MPC-68M 6.III.1 DISCUSSION AND RESULTS In conformance with the principles established in NUREG-1536 [6.III.1.1], 10CFR72.124
[6.III.1.2], and NUREG-0800 Section 9.1.2 [6.III.1.3], the results in this supplement demonstrate that the effective multiplication factor (keff) of the HI-STORM 100 System with the MPC-68M, including all biases and uncertainties evaluated with a 95% probability at the 95% confidence level, does not exceed 0.95 under all credible normal, off-normal, and accident conditions.
Criticality safety of the HI-STORM 100 System with the MPC-68M depends on the following principal design parameters:
The inherent geometry of the fuel basket design of the MPC-68M; The incorporation of spatially distributed B-10 isotope in the Metamic-HT fuel basket structure.
Based on the tests for the neutron absorber content in Metamic-HT (see Supplement 1.III and Supplement 9.III), and consistent with the approach taken for Metamic (see Section 9.1.5.3.2),
90% of the minimum B-10 (B4C) content is credited in the analysis. With a specified minimum B4C content of 10 wt%, the concentration credited in the analysis is therefore 9 wt%.
The off-normal and accident conditions defined in Section 2.2 are applicable to the HI-STORM System using the MPC-68M. These accidents are considered in Supplement 11.III and have no adverse effect on the design parameters important to criticality safety, except for the non-mechanistic tip-over event, which could result in limited plastic deformation of the basket.
However, a bounding basket deformation is already included in the criticality models for normal conditions, and thus, from the criticality safety standpoint, the off-normal and accident conditions are identical to those for normal conditions.
Results of the design basis criticality safety calculations for a single internally flooded HI-TRAC transfer cask with full water reflection on all sides (limiting cases for the HI-STORM 100 System),
loaded with undamaged fuel assemblies (see definition in Supplement 1.III) are listed in Table 6.III.1.1 and 6.III.1.4, conservatively evaluated for the worst combination of manufacturing tolerances (as identified in Section 6.III.3), and including the calculational bias, uncertainties, and calculational statistics. Table 6.III.1.1 provides the information for undamaged fuel without known or suspected cladding defects larger than pinhole leaks or hairline cracks, while Table 6.III.1.4 provides information for low-enriched, channeled BWR undamaged fuel without known or suspected grossly breached fuel rods. In addition, a result for a single internally dry (no moderator)
HI-STORM storage cask with full water reflection on all external surfaces of the overpack, including the annulus region between the MPC and overpack, is listed in Table 6.III.1.2 to confirm the low reactivity of the HI-STORM 100 System with an MPC-68M in storage. The maximum keff for an MPC-68M loaded with undamaged fuel and up to 16 cells containing damaged fuel/fuel debris in DFCs and/or damaged fuel in damaged fuel isolators (DFIs) is listed in Table 6.III.1.3. The 1 Evaluations and results presented in this chapter are supported by documented calculation package(s)
[6.III.1.4].
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-2 maximum keff for an MPC-68M loaded with undamaged fuel, 4 empty cells, and up to 8 cells containing damaged fuel/fuel debris in DFCs and/or damaged fuel in DFIs is listed in Table 6.III.1.5.
6.III.1.1 Calculations Performed with the MCNP5-1.51 Code and the ENDF/B-VII Library All results in this supplement are obtained using the MCNP4a code with the ENDF/B-V library unless otherwise indicated. The results listed in the following tables are obtained using the MCNP5-1.51 code [6.III.1.5] with the ENDF/B-VII library [6.III.1.6]:
TABLE 6.III.1.5: BOUNDING MAXIMUM keff VALUES FOR THE MPC-68M WITH UNDAMAGED FUEL, 4 EMPTY CELLS, AND UP TO 8 CELLS CONTAINING DAMAGED FUEL/FUEL DEBRIS IN DFCs AND/OR DAMAGED FUEL IN DFIs TABLE 6.III.4.11: MAXIMUM keff VALUES FOR THE MPC-68M WITH UNDAMAGED FUEL, 4 EMPTY CELLS, AND 8 CELLS CONTAINING DAMAGED FUEL/FUEL DEBRIS IN DFCs TABLE 6.III.4.12: REACTIVITY STUDY FOR UNDAMAGED FUEL AND 4 EMPTY CELLS WITH DAMAGED FUEL/FUEL DEBRIS IN 8 DFCs VERSUS UNDAMAGED FUEL AND 4 EMPTY CELLS WITH DAMAGED FUEL IN 8 DFIs 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-4 TABLE 6.III.1.2 REPRESENTATIVE keff VALUES FOR MPC-68M IN THE HI-STORM 100 OVERPACK Fuel Assembly Class Maximum Allowable Planar-Average Enrichment (wt% 235U)
Maximum keff 10x10A 4.8 0.3754 TABLE 6.III.1.31 BOUNDING MAXIMUM keff VALUES FOR THE MPC-68M WITH UNDAMAGED FUEL AND UP TO 16 CELLS CONTAINING DAMAGED FUEL/FUEL DEBRIS IN DFCs AND/OR DAMAGED FUEL IN DFIs Fuel Assembly Class Maximum Allowable Planar-Average Enrichment (wt% 235U)
Maximum keff All BWR Classes except 8x8F, 9x9E/F, 10x10F and 10x10G 4.8 0.9408 8x8F and 9x9E/F 4.0 0.9028 10x10F and 10x10G 4.6 0.9453 TABLE 6.III.1.4 BOUNDING MAXIMUM keff VALUES FOR THE MPC-68M WITH LOW ENRICHED, CHANNELED BWR FUEL Fuel Asembly Class Maximum Allowable Planar-Average Enrichment (wt% 235U)
Maximum keff All BWR Classes 3.3 0.9269 1 Configuration shown in Figure 6.III.1.1 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-5 TABLE 6.III.1.51 BOUNDING MAXIMUM keff VALUES FOR THE MPC-68M WITH UNDAMAGED FUEL, 4 EMPTY CELLS, AND UP TO 8 CELLS CONTAINING DAMAGED FUEL/FUEL DEBRIS IN DFCs AND/OR DAMAGED FUEL IN DFIs Fuel Assembly Class Maximum Allowable Planar-Average Enrichment (wt% 235U)
Maximum keff All BWR Classes except 8x8F, 9x9E/F, 10x10F and 10x10G 4.8 0.9093 8x8F, 9x9E/F and 10x10G 4.5 0.9039 10x10F 4.6 0.9024 Note: The results presented in Tables 6.III.1.2, 6.III.1.3, 6.III.1.4, and 6.III.1.5 above have an additional bias of 0.0021 applied to the 10x10 fuel assembly classes to conservatively account for any potential distributed enrichment effects. See Section 6.III.2.
1 Configuration shown in Figure 6.III.1.2. Results from this table are obtained using MCNP5-1.51 with the ENDF/B-VII library. See Subsection 6.III.1.1.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-6 Figure 6.III.1.1: MPC-16 Cells Containing either Damaged Fuel/Fuel Debris in DFCs and/or Damaged Fuel in X
X X
S S
S S
X X
S S
S S
S S
X S
S S
S S
S S
S X
S S
S S
S S
S S
X X
S S
S S
S S
S S
X S
S S
S S
S S
S X
S S
S S
S S
X X
S S
S S
X X
X 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-7 Figure 6.III.1.2: MPC-l S
S S
S S
S S
S S
S X
S S
X S
S S
X E
S S
E X
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
X E
S S
E X
S S
S X
S S
X S
S S
S S
S S
S S
S 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-12 of B4C). The evaluation provided in Section 6.3.2, which concludes that 10B depletion is negligible, is therefore directly applicable to the MPC-68M, and no additional evaluations are required in that respect.
Composition of the Metamic-HT is listed in Table 6.III.3.4.
6.III.3.1 Model Specification with the MCNP5-1.51 Code and the ENDF/B-VII Library Calculational models for the MPC-68M using MCNP5-1.51 with the ENDF/B-VII library are generally the same as those described in Section 6.III.3 except for the following changes:
The number of cycles and number of simulated histories per cycle are changed, see Subsection 6.III.4.6 for details; A lot of cross-sections in the ENDF/B-V and -VI libraries are provided by elements (e.g.
natural Zirconium with the nuclide identification number (ZAID) of 40000.66c, while the ENDF/B-VII library provide cross-sections by isotope (e.g. 90Zr, 91Zr with the following ZAIDs: 40090.70c, 40091.70c, etc.). The material definitions have been updated to reflect this change as shown in Table 6.III.3.5; and DFIs may contain fuel that is considered damaged fuel but not fuel debris. DFIs are modeled such that the MPC basket cell ID is assumed as the fuel boundary. The analysis of the fuel in the DFI is like that for damaged fuel in Subsection 6.2.5 of [6.III.3.1] where the fuel rod arrays are cladded fuel rods in arrays of varying sizes.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-17 TABLE 6.III.3.5 COMPOSITION OF THE MAJOR COMPONENTS OF THE HI-STORM 100 PACKAGE WITH THE ENDF/B-VII LIBRARY HI-STORM 100 Nuclide MCNP ZAID
[6.III.1.5]
Wt. Fraction UO2 4.8% ENRICHMENT, DENSITY 10.686 g/cm3 16O 8016.70c 1.1850E-01 235U 92235.70c 4.2310E-02 238U 92238.70c 8.3919E-01 METAMIC-HT, DENSITY 2.60 g/cm3, 9% B4C 10B 5010.70c 1.3031E-02 11B 5011.70c 5.7405E-02 C
6000.70c 1.9565E-02 27Al 13027.70c 9.1000E-01 COMMON MATERIALS ZR CLAD, DENSITY 6.55 g/cm3 90Zr 40090.70c 5.0706E-01 91Zr 40091.70c 1.1181E-01 92Zr 40092.70c 1.7278E-01 94Zr 40094.70c 1.7891E-01 96Zr 40096.70c 2.9438E-02 MODERATOR (H2O), DENSITY 1.00 g/cm3 1H 1001.70c 1.1189E-01 2H 1002.70c 2.5717E-05 16O 8016.70c 8.8580E-01 17O 8017.70c 2.2932E-03 ALUMINUM, DENSITY 2.70 g/cm3 27Al 13027.70c 1.0000E+00 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-18 TABLE 6.III.3.5 (continued)
COMPOSITION OF THE MAJOR COMPONENTS OF THE HI-STORM 100 PACKAGE WITH THE ENDF/B-VII LIBRARY COMMON MATERIALS Nuclide MCNP ZAID
[6.III.1.5]
Wt. Fraction LEAD, DENSITY 11.34 g/cm3 204Pb 82204.70c 1.3781E-02 206Pb 82206.70c 2.3955E-01 207Pb 82207.70c 2.2074E-01 208Pb 82208.70c 5.2592E-01 HOLTITE-A, DENSITY 7.82 g/cm3 10B 5010.70c 1.410E-03 11B 5011.70c 6.420E-03 27Al 13027.70c 2.129E-01 1H 1001.70c 5.920E-02 16O 8016.70c 4.237E-01 C
6000.70c 2.766E-01 14N 7014.70c 1.980E-02 STAINLESS STEEL, DENSITY 7.84 g/cm3 50Cr 24050.70c 7.9050E-03 52Cr 24052.70c 1.5853E-01 53Cr 24053.70c 1.8322E-02 54Cr 24054.70c 4.6467E-03 55Mn 25055.70c 2.0010E-02 54Fe 26054.70c 3.8983E-02 56Fe 26056.70c 6.3458E-01 57Fe 26057.70c 1.4917E-02 58Fe 26058.70c 2.0200E-03 58Ni 28058.70c 6.7198E-02 60Ni 28060.70c 2.6776E-02 61Ni 28061.70c 1.1834E-03 62Ni 28062.70c 3.8348E-03 64Ni 28064.70c 1.0082E-03 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-21 6.III.4 CRITICALITY CALCULATIONS The calculations in this supplement use the same computer codes and methodologies that are used in the main part of Chapter 6, except for those listed in Subsection 6.III.1.1 where the MCNP5-1.51 code with the ENDF/B-VII library is used (see Subsection 6.III.4.6). Specifically, the conservative approach to model damaged fuel and fuel debris in DFCs, using arrays of bare fuel rods, is the same.
However, the approach to model damaged fuel in DFIs uses arrays of cladded fuel rods because there is no fuel debris (see discussion in Subsection 6.III.4.1 below).
The basket design of the MPC-68M is essentially identical to that of the MPC-68, in respect to the characteristics important to criticality safety. Specifically, The bounding number and configuration of the cells for undamaged and damaged fuel/fuel debris are unchanged; The basket dimensions are essentially the same; and The same poison material (B4C) is used, but a larger 10B content in the basket walls.
The content is also the same, except for the following Higher enrichments are qualified, consistent with the higher 10B content in the basket walls; Two additional fuel assembly types are analyzed, that are variations of existing types with slightly different dimension; and DFIs are analyzed, which may contain damaged fuel but not fuel debris.
To verify that the bounding fuel parameter variations analyzed in the MPC-68 are also applicable to the MPC-68M, additional studies are performed and discussed in Subsection 6.III.4.2 below.
Due to the strong similarity in the basket design, the conclusions of the various studies presented in the main part of this Chapter on the MPC-68 are directly applicable to the MPC-68M. Nevertheless, to confirm this is also applicable to the MPC-68M, numerous studies with various moderation conditions that conclude that the fully flooded basket is the bounding case are re-analyzed and discussed in subsection 6.III.4.3. All analyses are therefore performed under the following condition:
Basket, and DFCs or DFIs as applicable, are fully flooded with pure water at the maximum density; and Pellet-to-clad gaps of undamaged assemblies are assumed flooded (see also discussion in Subsection 6.III.4.2 below)
All assemblies and DFCs are located eccentrically in the basket, closest to the center of the basket.
Results for all design basis calculations are listed in Subsection 6.III.1. All maximum keff values are below the regulatory limit of 0.95.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-22 6.III.4.1 Damaged Fuel and Fuel Debris For damaged fuel and fuel debris within a DFC in the MPC-68M the same conservative approach is used as in the main part of this chapter, see discussion in Section 6.4.4, specifically 6.4.4.2. For damaged fuel within a DFI in the MPC-68M, the same modeling approach described in Subsection 6.2.5 of [6.III.3.1] is used. Important aspects of this approach that ensure its conservatisms are as follows:
All damaged fuel must be in damaged fuel containers (DFCs) or damaged fuel isolators (DFIs).
All fuel debris must be in DFCs. All DFCs or DFIs must be located in specifically designated cells of the basket as discussed in Subsection 2.III.1.
o The allowable configurations of DFCs and DFIs as specified in Subsection 2.III.1 are briefly described as follows:
Undamaged fuel with up to 16 cells containing damaged fuel/fuel debris in DFCs and/or damaged fuel in DFIs as shown in Figure 6.III.1.1; and Undamaged fuel with 4 empty cells and up to 8 cells containing damaged fuel/fuel debris in DFCs and/or damaged fuel in DFIs as shown in Figure 6.III.1.2.
For damaged fuel/fuel debris in DFCs, a conservative model is used that bounds both damaged fuel and fuel debris. In other words, damaged fuel is always conservatively modeled as fuel debris in a DFC. The model consists of regular arrays of fuel rods without cladding. The rod pitch (array size) is varied to determine the optimum moderation condition.
For damaged fuel in DFIs, Thus, the model consists of regular arrays of fuel rods with cladding. The rod pitch (array size) is varied to determine the optimum moderation condition. Additionally, DFIs are modeled such that the MPC basket cell ID is assumed as the fuel boundary to maximize the area of moderated fuel.
Undamaged and damaged fuel/fuel debris in DFCs or damaged fuel in DFIs in the same basket have the same enrichment limit, which may be different from the enrichment limit for undamaged fuel only.
The results for loading with undamaged fuel only in Table 6.III.1.1 utilize different enrichment limits for different assembly classes, to ensure that the maximum keff is always below 0.95. It is therefore not possible to establish a single bounding assembly class/enrichment combination to be used in all analyses with damaged fuel/fuel debris in DFCs or damaged fuel in DFIs.
Therefore, and in order to optimize the enrichment for the loading of undamaged and damaged fuel/fuel debris in DFCs or damaged fuel in DFIs for each assembly class, undamaged assemblies are grouped by enrichment limit, and the undamaged assembly with the highest maximum keff in each group is used for the calculations together with damaged fuel/fuel debris in DFCs or damaged fuel in DFIs. These are:
o For the calculations with undamaged fuel and 16 cells containing damaged fuel/fuel debris in DFCs and/or damaged fuel in DFIs (presented in Table 6.III.1.3):
Undamaged assemblies of 4.5 wt%: Assembly class 9x9E/F. For the calculations with undamaged and damaged fuel, an enrichment of 4.0 wt% is used.
Undamaged assembly of 4.7 and 4.75 wt%: Assembly class 10x10G. For the calculations with undamaged and damaged fuel, an enrichment of 4.6 wt% is used.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-23 Undamaged assembly of 4.8 wt%: Assembly class 10x10A. For the calculations with undamaged and damaged fuel, an enrichment of 4.8 wt% is used.
o For the calculations with undamaged fuel, 4 empty cells, and 8 cells containing damaged fuel/fuel debris in DFCs and/or damaged fuel in DFIs (presented in Table 6.III.1.5):
Undamaged assemblies of 4.5 (8x8F, 9x9E/F) and 4.75 (10x10G) wt%: Assembly class 10x10G. For the calculations with undamaged and damaged fuel, an enrichment of 4.5 wt% is used.
Undamaged assembly of 4.7 wt%: Assembly class 10x10F. For the calculations with undamaged and damaged fuel, an enrichment of 4.6 wt% is used.
Undamaged assembly of 4.8 wt%: Assembly class 10x10A. For the calculations with undamaged and damaged fuel, an enrichment of 4.8 wt% is used.
Consistent with the results in the main part of this chapter for the MPC-68, array sizes of 9x9, 10x10 and 11x11 show the optimum moderation condition for the configuration shown in Figure 6.III.1.1. This is confirmed for undamaged assembly classes 9x9E/F, 10x10A and 10x10G by evaluating all arrays from 3x3 to 17x17 rods (see Table 6.III.4.1).
Array sizes of 10x10 and 11x11 show the optimum moderation condition in DFCs for the confiugration shown in Figure 6.III.1.2. This is confirmed for undamaged assembly classes 10x10A, 10x10G, and 10x10F by evaluating all arrays from 3x3 to 17x17 rods (see Table 6.III.4.11).
With respect to reactivity, damaged fuel in DFIs are bounded by damaged fuel/fuel debris in DFCs. This is shown for the MPC-68M by evaluating the configuration shown in Figure 6.III.1.2 with 4 empty cells and 8 DFIs containing damaged fuel at optimum moderation (see Table 6.III.4.12). Therefore, damaged fuel in DFIs may be stored in any location approved for damaged fuel/fuel debris in DFCs.
6.III.4.2 Fuel Parameters and Parameter Variations In the main part of the FSAR, extensive analyses of fuel dimensional variations have been performed. These calculations demonstrate that the maximum reactivity corresponds to:
maximum active fuel length, maximum fuel pellet diameter, maximum fuel rod pitch, minimum cladding outside diameter (OD),
maximum cladding inside diameter (ID),
minimum guide tube/water rod thickness, and maximum channel thickness (for BWR assemblies only) part length rods (if present) removed.
The reason that those are bounding dimensions, i.e. that they result in maximum reactivity is directly based on, and can be directly derived from the three main characteristics affecting reactivity, namely
- 1) characteristics of the fission process; 2) the characteristics of the fuel assemblies and 3) the characteristics of the neutron absorber in the basket. These affect the reactivity as follows:
5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-27 essentially independent from the external water density. Nevertheless, all further evaluations are performed with full external water density. In a definitive study, Cano, et al. [6.4.2] have demonstrated that the phenomenon of a peak in reactivity at low moderator densities (sometimes called "optimum" moderation) does not occur in the presence of strong neutron absorbing material or in the absence of large water spaces between fuel assemblies in storage. All calculations are therefore performed with full water density inside the MPCs.
Partial Flooding: The partial flooding of the basket, either in horizontal or vertical direction, reduces the amount of fuel that partakes effectively in the thermal fission process, while essentially maintaining the fuel-to-water ratio in the volume that is still flooded. This will therefore result in a reduction of the reactivity of the system (similar to that of the reduction of the active length), and due to the similarity of the fuel assemblies is not dependent on the specific fuel type. The reactivity changes during the flooding process were evaluated in both the vertical and horizontal positions for all MPC designs. For these calculations, the cask is partially filled (at various levels) with full density (1.0 g/cm3) water and the remainder of the cask is filled with steam consisting of ordinary water at a low partial density (0.002 g/cm3 or less), as suggested in NUREG-1536. Results of these calculations are shown in Table 6.III.4.8. In all cases, the reactivity increases monotonically as the water level rises, confirming that the most reactive condition is fully flooded. Note that the studies for partial flooding are performed with the design basis model for the assembly class 10x10A that has the partial length rods removed for added conservatism, while the calculations in the main part of the chapter for the MPC-68 were performed for an assembly class that did not include partial length rods. This shows that the conclusion from partial flooding, i.e. that the fully flooded condition is bounding, applies equally to assemblies with and without partial lengths rods.
Pellet-to-clad Gap Flooding: As demonstrated by the studies shown in Table 6.III.4.2, all assemblies are undermoderated. Flooding the pellet-to-clad gap will therefore improve the moderation and therefore increase reactivity for all assembly types.
Preferential Flooding: The only preferential flooding situation that may be credible is the flooding of the bottom section of the DFCs while the rest of the MPC internal cavity is already drained. In this condition, the undamaged assemblies have a negligible effect on the system reactivity since they are not flooded with water. The dominating effect is from the damaged fuel model in the DFCs. However, the damaged fuel/fuel debris model in DFCs is conservatively based on an optimum moderated array of bare fuel rods in water, and therefore representative of all fuel types and therefore the fully flooded condition is bounding of the preferential flooding condition.
6.III.4.4 Low Enriched, Channeled BWR fuel The calculations in this subsection show that low enriched, channeled BWR fuel with indeterminable cladding condition is acceptable for loading in all storage locations of the MPC-68M without placing those fuel assemblies into DFCs or DFIs, hence classifying those assemblies as undamaged. The main characteristics that must be assured are:
The channel is present and attached to the fuel assembly in the standard fashion; and 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-28 The channel is essentially undamaged; and The maximum planar average enrichment of the assembly is less than or equal to 3.3 wt%
235U This analysis covers older assemblies, where the cladding integrity is uncertain, and where a verification of the cladding condition is prohibitive. An example of this type of fuel is the so-called CILC (Copper Induced Localized Corrosion) fuel, which has potential corrosion-induced damaged to the cladding but does not have grossly breached spent fuel rods.
The presence of the essentially undamaged and attached channel confines the fuel rods to a limited volume and the low enrichment, required for all assemblies in the MPC, limits the reactivity of the fuel even under optimum moderation conditions. Due to the uncertain cladding condition, the analysis of this fuel follows essentially the same approach as that for the Damaged Fuel and Fuel Debris in DFCs, i.e. bare fuel rod arrays of varying sizes are analyzed within the confines of the channel. This is an extremely conservative modeling approach for this condition, since reconfiguration is not expected and cladding would still be present. The results of this conservative analysis are listed in Table 6.III.4.9 and show that the system remains below the regulatory limit with these assemblies in all cells of the MPC-68M, without DFCs or DFIs.
These results confirm that even with unknown cladding condition the maximum keff values are below the regulatory limit when fully flooded and loaded with any of the BWR candidate fuel assemblies, therefore if the cladding is not grossly breached and the fuel assembly is structurally sound it can be considered undamaged when loaded in an MPC-68M.
6.III.4.5 Thoria Rod Canister The criticality evaluation of thoria rod canister was performed for MPC-68 or MPC-68F and results presented in Section 6.4.6 show that it is permissible to load the Thoria Rod Canister together with any approved content in a MPC-68 or MPC-68F. While only a single canister is qualified for storage, the analysis assumes such a canister in every basket cell, and calculates a very low reactivity of less than 0.2 for this condition, based on a UO2 content of 1.8 wt%. The conversion of Th-232 to U-233 during depletion may results in a slight increase in reactivity for the hypothetical case of a MPC-68 or MPC-68F entirely filled with Thoria Rod Canisters, however, the real condition of a single canister loaded together with spent fuel would still be bounded by the design basis case with fuel assemblies only. Since the MPC-68M has equal or better criticality performance than the MPC-68 due to the basket itself being made from the neutron absorber, Metamic-HT.
Without any further evaluations it can therefore be concluded that, from a criticality perspective, the thoria rods with the actual composition can be safely stored in the HI-STORM 100 system in an MPC-68M canister.
6.III.4.6 Calculational Method and Results for the MCNP5-1.51 Code with the ENDF/B-VII library 5014855 Attachment 5
HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-29 For the additional calculations presented in Subsection 6.III.1.1, a newer version of the Monte Carlo N-Particle code, namely MCNP5-1.51 [6.III.1.5], and the continuous energy cross-section data based on ENDF/B-VII [6.III.1.5], are used.
To ensure the problem convergence and low calculation uncertainty, the calculations used a minimum of 10,000 simulated histories per cycle, a minimum of 400 cycles were skipped before averaging, and a minimum of 400 cycles were accumulated. The convergence of the power iteration process is ensured using the Shannon entropy, as implemented in MCNP5 [6.III.1.5]. Since the eigenvalue (keff) converges faster than the fission source distribution, the convergence of the keff is assured by the convergence of the source distribution. The Shannon entropy convergence has been checked for each calculation.
In calculating the maximum reactivity for results obtained using MCNP5-1.51 with the ENDF/B-VII library, the same methodology as presented in Subsection 6.4.3 is used, except the bias and standard error of the bias (95% probability at the 95% confidence level) for MCNP5-1.51 with the ENDF/B-VII library is used.
The critical experiment benchmarking and the derivation of the bias and standard error of the bias (95% probability at the 95% confidence level) for MCNP5-1.51 with the ENDF/B-VII library are presented in Subsection 6.A.1.1.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-30 TABLE 6.III.4.1 MAXIMUM keff VALUES IN THE MPC-68M WITH UNDAMAGED FUEL AND 16 CELLS CONTAINING DAMAGED FUEL/FUEL DEBRIS IN DFCs Maximum keff Bare Rod Array inside the DFC Assembly Classes 8x8F and 9x9E/F (4.0 wt%)
Assembly Class 10x10F and 10x10G (4.6 wt%)
All other assembly classes (4.8 wt%)
3x3 0.8926 0.9310 0.9267 6x6 0.8942 0.9338 0.9295 8x8 0.8986 0.9395 0.9344 9x9 0.9028 0.9414 0.9371 10x10 0.9024 0.9432 0.9387 11x11 0.9024 0.9420 0.9381 12x12 0.9018 0.9412 0.9373 13x13 0.9007 0.9397 0.9353 14x14 0.8993 0.9385 0.9352 16x16 0.8985 0.9376 0.9335 17x17 0.8976 0.9366 0.9328 Note: The results do not include the bias for distributed enrichments discussed in Section 6.III.2.
Configuration shown in Figure 6.III.1.1.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-40 TABLE 6.III.4.111 MAXIMUM keff VALUES FOR THE MPC-68M WITH UNDAMAGED FUEL, 4 EMPTY CELLS, AND 8 CELLS CONTAINING DAMAGED FUEL/FUEL DEBRIS IN DFCs Maximum keff, Bare Rod Array inside the DFC Assembly Classes 8x8F, 9x9E/F and 10x10G (4.5 wt%)
Assembly Class 10x10F (4.6 wt%)
All other assembly classes (4.8 wt%)
3x3 0.8680 0.8670 0.8702 6x6 0.8775 0.8757 0.8795 8x8 0.8906 0.8902 0.8951 9x9 0.8970 0.8966 0.9012 10x10 0.9013 0.9002 0.9059 11x11 0.9018 0.9003 0.9072 12x12 0.8995 0.8985 0.9026 13x13 0.8961 0.8939 0.8991 14x14 0.8918 0.8896 0.8947 16x16 0.8862 0.8839 0.8888 17x17 0.8836 0.8822 0.8866 Note: The results do not include the bias for distributed enrichments discussed in Section 6.III.2.
1 Configuration shown in Figure 6.III.1.2. Results from this table are obtained using MCNP5-1.51 with the ENDF/B-VII library. See Subsection 6.III.1.1.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-41 Table 6.III.4.121 REACTIVITY STUDY FOR UNDAMAGED FUEL AND 4 EMPTY CELLS WITH DAMAGED FUEL/FUEL DEBRIS IN 8 DFCs VERSUS UNDAMAGED FUEL AND 4 EMPTY CELLS WITH DAMAGED FUEL IN 8 DFIs Assembly Class Configuration with 4 Empty Cells Array Inside DFC or DFI for Optimum Moderation Calculated keff keff keff Assembly Classes 8x8F, 9x9E/F and 10x10G (4.5 wt%)
8 DFCs With Damaged Fuel/Fuel Debris 11x11 Bare Rod Array 0.9003 0.0004 Reference Assembly Classes 8x8F, 9x9E/F and 10x10G (4.5 wt%)
8 DFIs With Damaged Fuel 10x10 Cladded Rod Array 0.8936 0.0004
-0.0067 Assembly Class 10x10F (4.6 wt%)
8 DFCs With Damaged Fuel/Fuel Debris 11x11 Bare Rod Array 0.8988 0.0004 Reference Assembly Class 10x10F (4.6 wt%)
8 DFIs With Damaged Fuel 10x10 Cladded Rod Array 0.8926 0.0004
-0.0062 All other assembly classes (4.8 wt%)
8 DFCs With Damaged Fuel/Fuel Debris 11x11 Bare Rod Array 0.9057 0.0004 Reference All other assembly classes (4.8 wt%)
8 DFIs With Damaged Fuel 10x10 Cladded Rod Array 0.8973 0.0004
-0.0084 1 Configuration shown in Figure 6.III.1.2. Results from this table are obtained using MCNP5-1.51 with the ENDF/B-VII library. See Subsection 6.III.1.1.
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HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL HI-STORM 100 FSAR Proposed Rev. 16 REPORT HI-2002444 6.III-43 6.III.5 CRITICALITY BENCHMARK EXPERIMENTS Same as in Section 6.5, except for calculations performed with the MCNP5-1.51 code and the EDNF/B-VII library as discussed in Subsection 6.III.1.1. Benchmark calculations with the MCNP5-1.51 code with the EDNF/B-VII library are presented in Subsection 6.A.1.1. No significant trends were evident in the benchmark calculations or the derived bias.
6.III.6 REGULATORY COMPLIANCE Same as in Section 6.6 6.III.7 REFERENCES
[6.III.1.1] NUREG-1536, Standard Review Plan for Dry Cask Storage Systems, USNRC, Washington, D.C., January 1997.
[6.III.1.2] 10CFR72.124,
[6.III.1.3] USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev.
2 - July 1981.
HI-STAR 100 AND HI-STORM 100 ADDITIONAL CRITICALITY CALCULATIONS
-2012771 Rev.20 (proprietary)
[6.III.1.5]
X-5 Monte Carlo Team, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5, LA-UR-03-1987, Los Alamos National Laboratory, April 2003 (Revised 2/1/2008).
[6.III.1.6]
ENDF/B-VII.0 Evaluated Nuclear Data Library, U.S. Cross Section Evaluation Working Group (CSEWG), Release December 15, 2006
[6.III.3.1]
HI-STAR 190 SAR, HI-2146214, latest revision, Holtec International 5014855 Attachment 5