Text

Rev 1 to Calculation 22S-0-110M-0063, Max Cladding Temp for Uncovered Spent Fuel Rod
	ML20205R024
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Issue date:	12/30/1998
From:	Peterson R; COMMONWEALTH EDISON CO.
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ML20205Q999	List: ML20205Q997; ML20205R024; ML20205R050;
References
	22S--110M-63, 22S--110M-63-R1, 22S-0-110M-0063, 22S-0-110M-0063-R01, NUDOCS 9904220158
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Attachment 6 l

l Calculation Number 22S-0-110M-0063 Maximum Cladding Temperature for Uncovered Spent Fuel Rod December 30,1998 l

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Exhibit C N EP-12-02 Revision 6 CALCULATION TITLE PAGE Calculation No.: 22S-0-110M-0063 DESCRIPTION CODE: N00 DISCIPLINE CODE: N ZION NUCLEAR STATION SYSTEM CODE: _S F TITLE: Maximum Cladding Temperature for Uncovered Spent Fuel Rod

@ Safety Related C Augmented Quality 0 Non-Safety Related REFERENCE NUMBERS Type Number Type Number PROJ 10436-002. 10436-003 AEDV PSED COMPONENT EPN: DOCUMENT NUMBERS:

EPN Compt Type Doc Type /Sub Type Document Number REMARKS:

REV. REVISING APPROVED DATE NO. ORGANIZATION PRINT / SIGN 0 Sargent & Lundy Roben J. Peterson / Signature on file , _,

8/31/98 1 Sargent & Lundy Robert J. Peterson /((j(( g / @ fg)_fg

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Exhibit C NEP-12-02 COMMONWEALTIl EL)lSON COMPANY Revision 6 CALCULATION REVISION PAGE CALCULATION NO. 22S-0-110M-0063 PAGE NO.: 2 REVISION SUMMARIES REV: 0 REVISION

SUMMARY

Initial Issue: Main Text: 31 pages; Attachment pages as shown in Table of Contents I ELECTRONIC CALCULATION DATA FILES:

(Program Name, Version, File name ext / size /date/ hour /: min)

COBRA-SFS, Cycle 3 peakassm inp 29591 8/28/98 4:07:36pm peakassm.out 1491703 8/28/98 4:10:22pm radgen.inp 207 8/27/98 1:00.04pm radgen.out 857407 8/27/98 1:24:26pm SbW Prepared by: Mark Handrick/ Sie. nature on file 8/31/98 Print / Sign Date The reviewer's signature indicates compliance with SOP-0402 and the verification of the following minimum items: correctness of math for hand prepared calculations, appropriateness ofinput data, appropriateness of assumptions, and appropriateness of the calculation method.

Reviewed by: Kawei Chan/ Signature on file 8/31/89 Print / Sign Date Type of Review

@ Detailed O Alternate O Test DO ANY ASSUMPTIONS IN THIS CALCULATION REQUIRE LATER VERIFICATION O YES @ NO Tracked by:

Exhibit C N EP-12-02 COMMONWEALTil EDISON COMPANY Revision 6 CALCULATION REVISION PAGE CALCULATION NO. 22S-0-110Ni-0063 PROJECT NO. 10436-003 PAGE NO.: 2.1 REVISION

SUMMARY

REV:1 REVISION SUN 1 MARY: Revision 1 of this calculation is issued to incorporate a model of the Fuel Handling Building to determine the air space temperature following a postulated loss of all spent fuel pool cooling water (see Attachment D). This air space temperature is then used as a boundary condition to the COBRA-SFS model, used to calculate the maximum cladding temperature of the spent fuel assemblics. The decay heat load of the spent fuel assemblies is taken as that corresponding to a date of 2/1/99 for the postulated loss of spent fuel cooling water, as opposed to a date of 7/1/98 which was used in Revision 0. The computer files for the revised COBRA-SFS model are included in Attachment E.

The hardcopy printout of Revision 0 computer files (Attachments B and C) are deleted and replaced by microfiche in Revision 1. Changes to the calculation text are denoted by revision bars.

Revised Pages 1, 3-4, 6, 8-11,15,17-19, 24-28, 30, Al-A3, A5-A6, B 1, C l-C2 Added Pages 2.1,25.1,25.2, Dl-D16, El Deleted Pages B2-B405, C3-C172 Total Pages issued as Revision 1: 60 pages Electronic Calculation Data Files:

(Program Name, Version, File name ext' size /datchourh min)

COBRA-SFS, Cycle 3 zirc-rvl.inp 29591 kb 12/22/98 9:51:24am zirc-rvl .out 1491352 kb 12/22/98 10:00:56am KITTYlS zion-fhb.out 20811 kb 12/21/98 1:35:24pm Prepared bv: Mark Handrick .M744 /-/a.e d E / /. 3B cf Print / Sign Date The reviewer's signature indicates compliance with $8L Procedure SOP-0402 and the verification of, as a minimum, the following items: correctness of mathematics for manual calculations, appropriateness of input dab, appropriateness of assumptions, and appropriateness of the calculation method.

Reviewed by: Helmut Kopke [ } C [It Mhh?

Print / Sign / O Date Type of Review

[ x 1 Detailed [ ] Alternate [ ] Test DO ANY ASSUMPTIONS IN THIS CALCULATION REQUIRE LATER VERIFICATION [ ] YES [x ] NO Tracked by:

i This box is the responsibility of Comed -

Supplemental Review Required YES (NEP-12-05 documentation) NO Supenisor /

l l

1

S Exhibit D NEP-12-02 l

Revision 6 COMMONWEALTII EDISON COMPANY CALCULATION TABLE OF CONTENTS PROJECT # 10436-002 CALCULATION NO. 22S-0-110M-0063 REV.NO.1 PAGE NO. 3 SECTION PAGE NO. SUB-PAGE NO.

TITLE PAGE I Rev.I l REVISION

SUMMARY

2 2.1 TABLE OF CONTENTS 3 j

1. PURPOSE / OBJECTIVE 4

2. METHODOLOGY / ACCEPTANCE CRITERIA 4-8

3. ASSUMPTIONS AND LIMITATIONS 9 - 10

4. DESIGN INPUT 11 - 13 .

5. REFERENCES 14 - 15 l Rev.1 l 6. CALCULATIONS 16 - 25 25.1, 25.2

7.

SUMMARY

AND CONCLUSIONS 26 Tables and Figures 27 - 31 ATTACHMENTS:

Rev.I A Spreadsheet Cell Formulas for Tables Al - A6 B COBRA-SFS Input / Output Files from Revision 0 B1 Input File "peakassm.inp" (Microfiche)

Output File "peakassm.out" (Microfiche)

C RADGEN Input / Output Files from Revisions 0 and 1 Input File "radgen.inp" Cl Output File "radgen.out" (Microfiche) C2 D Computation of Fuel Handling Building Homogeneous D1 - D16

Air Temperature l

l E COBRA-SFS Input / Output Files from Revision 1 El Input File "zirc-rv1.inp"(Microfiche)

Output File "zirc-rvl.out"(Microfiche) l

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i Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 l

CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 4 1

1. PURPOSE / OBJECTIVE i

The purpose of this calculation is to determine the maximum zircaloy cladding temperature in the spent fuel pool at Zion Nuclear Station under postulated conditions where all of the cooling water is lost and the fuel is exposed to an air environment. It is desired that the maximum cladding temperature is kept below the critical temperature ,

necessary for oxidation of the cladding. The critical temperature is defmed in Reference 5.1 as 565 C (1049 F). I Rev.1 This calculation will determine the maximum cladding temperature at 710 days after the final reactor shutdown and subsequent fuel discharge to the spent fuel pool, assuming a sudden loss of cooling water. This time duration corresponds to a final reactor shutdown date of 2/21/97 (Reference 5.6), and a postulated loss of spent fuel pool cooling water on 2/1/99. If the calculated maximum cladding temperature does not exceed the critical value for the 2/1/99 reference date, then a postulated loss of spent fuel pool cooling water occurring after the reference date will not result in oxidation of the zircaloy cladding. ,

2. METHODOLOGY / ACCEPTANCE CRITERIA 2.1 Methodology Under normal conditions, the spent fuel in the spent fuel pool (SFP) is cooled by an active cooling water system.

This calculation postulates a sudden loss of all pool water, exposing the spent fuel rods to an air environment. This greatly reduces the mechanism for heat removal from the spent fuel rods, and consequently results in higher fuel rod temperatures as the decay heat of the spent fuel is dissipated to the air heat sink at steady-state conditions.

Under these conditions, the primary mechanism for heat transfer away from the spent fuel rods is by convective heat transfer to the ambient in the spent fuel pool room via air circulation currents established in the spent fuel pool cavity and Fuel Handling Building. The Auxiliary Building Ventilation system, which serves the Fuel Handling Building, provides a constant source of relatively cold air to the exposed spent fuel rods. The air is heated as it passes upwards along the fuel rods, then leaves the Fuel Handling Building via the return air ducts. A schematic representation of this flow pattern is shown in Figure 1.

l l

l REVISION NO. I

Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 ,

CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 5 l Figure 1 Schematic Representation of Convective Air Currents 1

1 Fuel Handling Building I I

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l REVISION NO. I
Exhibit E NEP-12-02 l COMMONWEALTH EDISON COMPANY Revision 6 i
CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 6 Each fuel assembly stored in the spent fuel pool resides in a dedicated square storage cell, with each cell connected j to a lower plenum via a 5-inch diameter hole in the cell base-plate. The storage cells in the spent fuel pool are one of two types, called Region I and Region 11 type cells. There are no significant differences with respect to the interior of these cell types, as each type have an 8.94 inch inner dimension, 0.075 inch cell wall thickness, and a 5-Rev. I inch diameter flow hole in the base-plate (pages 3-2 through 3-5 of Reference 5.4). All cells are open to a common plenum above the fuel assemblies. Therefore, the spent fuel pool itselfis a system of parallel flow channels. The side walls of the storage cells provide a radial heat transfer path towards the heat sink associated with the spent fuel pool walls. A complete model of the entire spent fuel pool. with all fuel assemblics accounted for, would be prohibitively large, in terms of available computer resources.
The computer code COBRA-SFS Cycle 3, " Thermal Hydraulic Analysis of Spent Fuel Casks", will be used to determine the maximum cladding temperature following a postulated loss of cooling water. The COBRA-SFS computer program was developed at Pacific Northwest Laboratory for thermal-hydraulic analysis of multi-i assembly spent fuel storage and transportation systems. The code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Conunission for a generic license to apply the code to spent fuel storage system analysis. The companion computer code RADGEN is used to calculate the gray body 'iew factors employed in the radiation heat transfer module of the COBRA code. The RADGEN code validation is included as part of COBRA-SFS code package (Ref. 5.3).
To determine the maximum cladding temperature, the fuel assembly most recently discharged to the spent fuel pool, l with the highest decay heat, will be modeled explicitly. To conservatively maximize the calculated fuel / cladding temperature, this assembly is assumed to have an adiabatic boundary condition along the sides of the storage cell.
l This precludes any radial heat transfer towards the pool wall heat sink. This is representative of an assembly located in the center of the spent fuel pool, surrounded by other assemblies of equal decay heat. Under these
conditions, only heat transfer from the fuel rods to the air flowing axially upwards is applicable. Figure 2 illustrates the COBRA model of the storage cell, including the numbering scheme for the fuel rods and flow channels within the assembly.
Data on the history of fuel assembly discharge to the spent fuel pool is provided by Reference 5.6, with the latest discharge occurring on 2/21/97 and consisting of a full-core ofiload of 193 fuel assemblies. The decay heat l generation rate associated with these fuel assemblies is calculated in accordance with NRC Branch Technical l Rev.I Position ASB 9-2 (Ref. 5.8). To conservatively maximize the computed decay heat, a peaking factor of 1.369 is I applied to the average fuel assembly decay heat. This peaking factor is provided by Reference 5.24, and is associated with radial power peaking ofin-core fuel assemblies during full power operation. In this manner, the
' hottest' fuel assembly is modeled, conservatively maximizing the calculated cladding temperatures. A volumetric heat generation rate is computed based on the maximum decay heat and the active fuel length, and is used as input I
to the COBRA model. Details of these calculations are provided in Section 6.2 and Table 1.
To account for axial peaking with respect to decay heat, an axial power profile is applied to the average decay heat values for the most recent fuel assemblies discharged to die spent fuel pool. A sinusoidal distribution is employed, consistent with PWR axial power profiles, as cited in References 5.4 and 5.12. The axial profile is employed in the REVISION NO. I
Exhibit E j NEP-12-02 l COMMONWEALTH EDISON COMPANY Revision 6 l
CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 1 AGE NO. 7 i COBRA model to capture any localized peak fuel / cladding temperatures. Details of the axial power distribution are provided in Section 6.3. l
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Figure 2 l
COBRA Spent Fuel Assembly Model l l
l Stab No.1 Stab No.2 I
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-Slab 46 47 48 49 50 51 52 54 55 56 57 58 59 60 No.3 No.8 -i 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 61 62 63 64 66 67 68 69 70 72 73 74 75 81 82 83 64 85 86 87 88 89 90 91 92 93 94 95 96
- 76 77 79 80 81 82 83 84 85 86 87 89 90 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 91 92 93 94 95 96 97 98 99 100 01 02 03 1 105 113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 28 106 107 100 '* 110 111 112 i 114 115 116 118 119 120 - -
129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 44 121 122 123 124 125 126 127 128 129 130 131 132 133 134 136 1 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 180 l 136 37 - 139 140 141 142 143 144 145 146 47 149 150 l 161 162 163 164 165 166 167 168 189 170 171 172 173 174 175 76 151 52 53 154 156 157 158 159 160 162 183 164 186 gg 177 178 179 180 181 182~ 183 184 185 186 187 188 189 190 191 92 g No. 7 166 67 168 169 170 171 172 174 175 176 177 178 179 180 No.4 193 194 195 196 197 198 199 200 201 202 203 204 205 206 207 95
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81 82 184 185 187 188 189 191 192 194 209. 10 211 212 213 214 215 216 217 218 219 220 221 3 222 223 4 196 97 98 199 1 10 225 226 227 228 229 230 231 232 233 234 235 236 237 238 239 40 211 12 13 14 1 16 21 21 21 221 4 241 242 243 244 245 246 247 248 249 250 251 252 253 254 255 258 Stab No. 6 Stab No.5 g ControlRod REVISION NO. I
Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 8 l
t 2.2 Acceptance Criteria 1
In order to prevent the rapid oxidation of the zircaloy cladding following a loss of spent fuel pool cooling water, the j cladding temperature must be kept below the critical temperature in Reference 5.I as 565 C (1049 F). If the l Rev.I maximum calculated cladding temperature at the reference date of 2/1/99 is below the critical temperature, then the fuel rods are not expected to oxidize as a result of the postulated loss of spent fuel pool cooling water accident, provided the postulated accident occurs on or after 2/1/99.
l l
2.2.1 Computer Programs Used l l
l The following computer programs were used in the preparation of this calculation:
l l COBRA-SFS Cycle 3,"A Thermal-Hydraulic Analysis Code for Spent Fuel Storage And Transportation Casks",
Sargent & Lundy Program No. 03.7.672-1.0.
Microsoft Excel version 5.0c, Sargent & Lundy Program No. 03.1.138-5.0 Rev.I KITTYlS, " Thermal-Hydraulic Steady States in Arbitrary Solid and/or Fluid Channel Configurations", S&L I Program No. 03.7.171-5.11.
i l These computer programs are maintained by S&L's Software Center for use. The computer programs were run on l Rev.I g&L pC No. 5765, which is attached to S&L file server SNL2. COBRA-SFS Cycle 3 and KITTYIS have been validated per S&L's procedures. The Excel spreadsheet validation is implicit in the detailed review of the calculation and requires no additional documentation.
l l
l l
l l
1 REVISION NO. I l
Exhibii E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 9
~~3. ASSUMPTIONS AND LIMITATIONS '~~
3.1 Euct Assembiv/ Rod Assumptions )
3.1.1 The thermal conductivity ofirradiated UO is 2 assumed to be reduced by 60 percent from that of unirradiated UO 2per the statement in Table 5-1 of Reference 5.12. This is conservative with respect to calculated fuel / cladding temperatures and is representative of the fuel stored in the spent fuel pool.
3.1.2 The fuel rod grid spacers are assumed to be uniformly distributed along the fuel rod length. This !
assumption simplifies the COBRA model, with negligible impact on calculated cladding temperatures. !
3.1.3 The grid spacer frictional loss coefficients are assumed to be consistent with an abrupt area contraction followed by an abrupt area expansion (0.5 + 1.0 = 1.5). This assumption leads to conservatively high local ;
loss coefficients, resulting in reduced mass flow rates and consequently higher air temperatures for a specified decay heat load. Higher air temperatures will lead to conservatively higher calculated cladding temperatures.
3.1.4 The exit loss coefficient at the top of the fuel assembly is assumed to be 1.0, consistent with an abrupt area expansion into an inf*mitely large area, per page A-26 of Reference 5.21.
Rn1 3.1.5 The storage cell geometry is modeled after the Region 11 storage cells shown in Reference 5.4. This is representative of the vast majority of stored fuel assemblies, since Region Il cells comprise 2670 of the 3012 total storage locations in the spent fuel pool (Design Input 4.1). The flow hole at the bottom of each storage cell is 5 inches in diameter for both types of storage cells, per pages 3-1 through 3-5 of Reference 5.4. The dimensions of the storage cell (inner diameter and wall thickness) are identical for the Region I and Regiou Il cells. Since the storage cell is modeled as radially adiabatic, the minor differences in cell-to-cell pitch are not a factor in this analysis, and the Region I and Region II storage cells can be considered j identical for the purposes of this calculation.
3.1.6 The spent fuel assembly model is applicable to the VANTAGE 5 fuel design, which is the type of fuel used in fuel cycles 14 and higher, per Reference 5.18. The last batch of spent fuel discharged to the pool was Cycle 15, as shown in reference 5.6. Therefore, the modeling of VANTAGE 5 fuel is appropriate for determining the maximum cladding temperatures.
3.2 Heat Transfer Modelina Assumptions 3.2.1 .The fuel assembly storage cells are assumed to be adiabatic in the radial direction. This is conservative l with respect to calculated cladding temperature since the neglected heat transfer mechanism would act in parallel with convection to the air in the axial direction.
3.2.2 The air temperature at the inlet of the Fuel Handling Building, via the Auxiliary Building Ventilation I
system, is assumed to be 115 *F, consistent with the maximum Auxiliary Building area temperature under abnormal conditions, per Reference 5.19. Higher inlet air temperatures will result in higher calculated fuel / cladding temperaturas, and are therefore conservative.
REVISION NO. 1
i 1 Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.10 I
3.2.3 The emissivity of the stainless steel storage cell walls is assumed to be 0.2, consistent with the values listed in Table A.8 of Reference 5.22 for stainless steel. The emissivity of the zircaloy cladding is assumed to be 0.8, consistent with the value used in the COBRA validation case titled "TN24P", documenico in Reference 5.3. This validation case models a Transnuclcar, Inc. spent fuel storage cask loaded with standard Westinghouse 15 x 15 PWR spent fuel, with documented results showing that the COBRA model accurately predicts peak cladding temperatures with respect to the measured test temperatures.
3.2.4 It is assumed that the upper and lower plenum flows are one-dimensional, and that the flow mixes instantaneously in the plenum regions.
3.2.5 The environment in the Fuel Handling Building is assumed to be at atmospheric pressure, or 14.7 psia.
This pressure value is used to determine the inlet air density via the Auxiliary Building Ventilation system.
Variations of the ambient pressure are expected to be negligible, and have an insignificant efTect on the results of this calculation.
Ret i 3.2.6 The specific heat of air is assumed to be 0.253 Btu /lb.- F for the purpcse of determining the temperature rise along the heated length of the fuel rod, as shown in Table 3, " Boundary Conditions for COBRA Model". This specific heat value is consistent with an air temperature of 700 F as shown in Design input 4.5.1, and is representative of the average air temperature for flow along the heated length of the fuel rod.
1 REVISION NO. 1 i
Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY -
Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. I1
4. DESIGN INPUT 4.1 Reactor Core. Fuel Assembly. and Spent Fuel Storace Cell Parameters The following data is applicable to the reactor core, fuel assemblies, and spent fuel storage cells as Zion Station, with references as provided.
Parameter Value Reference Reactor Core and Soent Fuel Pool Rated Core Thermal Power 3250 MW Refs. 5.7, 5.14, 5.15 Number of Assemblies per Core Loading 193 assemblies Refs. 5.6, 5.14 Date of Last Discharge to Spent Fuct Pool 2/21/97 Ref. 5.6 Number of Assemblies in Last Discharge 193 Ref.5.6 Number of Days of Reactor Operation for 432 Ref.5.6 Last Batch Discharged to Pool Rev.1 Total Pool Heat Lo:d on Reference Date of 8.537586c6 (Btu /hr) Ref. 5.6 1/1/99 Fuel Assemblies Fuel Assembly type VANTAGE 5 Ref. 5. I 8 Fuel Rod Locations per Assembly 225; 15 x 15 array Refs. 5.4 (page 4-24),5.15 Active Fuel Rods per Assembly 204 Refs. 5.4 (page 4-24),5.15 l Fuel Rod Pitch 0.563 inches Refs. 5.4 (page 4-24),5.16 i Fill Gas Helium Ref. 5.14 Fuel Rod Length 152.17 inches Ref. 5.16 Active Fuel Length 144 inches Ref. 5.4 (page 5-25)
Fuel Pellet Diameter 0.3659 inches Refs. 5.4 (page 4-24), 5.16 Fuel Theoretical Density (TD) 10.97 g/cm' Ref. 5.12 (Table 4-2)
Fuel Pellet Density 95% of TD Ref. 5.4 (page 4-24) l Cladding Material Zircaloy-4 Refs. 5.4 (page 4-24),5.16 Cladding OD 0.422 inches Refs. 5.4 (page 4-24),5.16 Cladding ID 0.3734 inches Ref. 5.4 (page 4-24)
Cladding Thickness 0.0243 inches Refs. 5.4 (page 4-24),5.16 Rev.1 Radial Assembly Peaking Factor 1.369 Ref. 5.24 Number ofGrid Spacers per Assembly 10 Refs. 5.15 (page 4.2-8 ),5.17 Spent Fuel Pool Storage Cells Total Storage Cells in Spent Fuel Pool 3012 Ref. 5.4 (page 1-1)
Total Storage Cells in Region I 342 Ref. 5.4 (page 2-14)
Total Storage Cells in Region II 2670 Ref. 5.4 (page 2-15)
Storage Cell Material Stainless Steel Type 304 Ref. 5.4 (page 2-3) l Storage Cell Square Box ID 8.94 inches Ref. 5.4 (page 4-29)
Storage Cell Wall Thickness 0.075 inches Ref. 5.4 (page 4-29)
Base-Plate Flow Hole Diameter 5.0 inches Ref. 5.4 (page 3-5)
Rev.1 l Storage Cell Height above Baseplate 168 inches Ref. 5.4 (page 2-16)
REVISION NO. 1 L
Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.12 4.2 Auxiliary Buildine Ventilation System Pararneters 4.2.1 Per Reference 5.19, the Fuel Handling Building is served by the Auxiliary Building Ventilation system.
During normal operation, the system is designed to limit the discharge air temperature to 85 F, and under abnormal operation the main areas are expected to be limited to 115 F.
4.3 UO3 Fuel Physical Properties 1
References 5.12 and 5.13 provides the following physical properties of UO2 fuel. !
Temperature Thermal Conductivity Thermal Conductivity
~~Specific Heat l (Ref. 5.12) (Ref. 5.12) (Ref. 5.13)~~
(F) (Btu /hr-R- F) (Btu /hr-A- F) (Btu /lb - F) 75 0.06 200 4.5 1.80 0.063 400 3.5 1.40 500 0.07 6')0 2.8 1.12 800 2.5 1.00 1000 2.2 0.88 1200 2.0 0.8
Calculated UO 2thermal conductivity after 60% reduction following irradiation Note: The density and specific heat values are not used in the steady-state solution for maximum cladding temperature, but are required input to the COBRA code.
4.4 Physical Properties of Metals 4.4.1 Reference 5.13 provides the following physical propenies of Zircaloy cladding.
i Temperature Density Specific Heat Thermal Conductivity (F) (ib /ft') (Btu /lb.- F) (Btu /hr-R- F) 1 75 409 0.071 6.7 200 6.9
400 7.1 600 7.2 4.4.2 The thermal conductivity of Stainless Steel type 304 is given by Reference 5.10 as the following:
9.4 Btu /hr-R *F at 212 'F i 10.9 Btu /hr-ft *F at 572 F REVISION NO. 1
Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.13 4.5 Physical Properties of Gases 4.5.1 Air Properties Reference 5.10 (Table A-3) provides the following physical properties of air at atmospheric pressure. The enthalpy data is provided by Reference 5.11 (Table A-13E).
Temperature Enthalpy Thermal Specific Heat Specific Viscosity Conductivity Volume (F) (Btu /lbm) (Btu /hr-ft 'F) (Btu /lbm- F) (ft'/lbm) (Ibm /ft-hr) 0 109.9 0.0133 0.239 11.628 0.0400 100 133.9 0.0154 0.240 14.085 0.0463 200 157.9 0.0174 0.241 16.667 0.0518 300 182.1 0.01933 0.243 19.231 0.0580 400 206.5 0.0212 0.245 21.739 0.0630 500 231.1 0.0231 0.247 24.272 0.0680 I
600 256.0 0.0250 0.250 26.810 0.0720 700 281.1 0.0268 0.253 29.326 0.0770 800 306.7 0.0286 0.256 31.847 0.0810 900 332.5 0.0303 0.259 34.364 0.0850 1000 358.6 0.0319 0.262 36.900 0.0889 1500 493.6 0.040 0.276 49.505 0.1080 The molecular weight of air is given as 28.97, per Table A-lE of Reference 5.11.
4.5.2 Helium Properties Reference 5.10 (Table A-3) provides the following physical properties of helium.
Temperature Thermal Conductivity l (F) (Btu /hr-ft- F) l 200 0.097 l 400 0.115 i 600 0.129 l 800 0.138 4.6 Miscellaneous 1
4.6.1 The Universal g7s constant is given as 10.73 psi-ft'/lbmole- R per Reference 5.22. The gas constant for air is obtained using the molecular weight of 28.97, and has a value of 53.34 fi-lbr/lbm- R. I I
REVISION NO. I f
f i
Exhibit E NEP-12-02 i
COMMONWEALTH EDISON COMPANY Revision 6 l
CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.14
~~5. REFERENCES 5.1 NUREG/CR-6451, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Power Plants", August 1997.~~
I 5.2 NUREG/CR-6411 BNL-NUREG-52494, " Analysis of Spent Fuel Heatup Following Loss of Water in a l Spent Fuel Pool", Draft Report for Comment, May 1998.
5.3 COBRA-SFS, "A Thermal-Hydraulic Analysis Code for Spent Fuel Storage And Transportation Casks",
September 1995, Prepared for the U.S. Department of Energy by Pacific Northwest Laboratory, Sargent &
Lundy Computer Program No. 03.7.672-1.0.
5.4 Commonweali Edison Letter from S. F. Stimac, Nuclear Licensing Administrator, to Dr. Thomas E.
Murley, Office of Nuclear Regulatory Commission, dated January 15,1992; Attachment B, " Zion Station Spent Fuel Pool Modification Licensing Report for Proposed Changes to Facility operating Licenses DPR-39 and DPR-48".
5.5 Safety Evaluation by the Oflice of the Nuclear Reactor Regulation Related to Amendment No.142 to Facility Operating license No. DPR-39 and Amendment No.131 to Facility Operating License No. DPR-48, Commonwealth Edison Company Zion Nuclear Power Station, Units 1 and 2, Docket Nos. 50-295 and 50-304, dated Febmary 23,1993.
5.6 Calculation No. 22N-0-110M-0058, Revision 0, " Zion Decommissioning Spent Fuel Pool Heat Load and Time to saturation Calculation",4/20/98.
5.7 Calculation No. 22N-0-110X-0057, Revision 0, " Fuel Handling Accident Offsite Dose Calculation with Extended Radioactive Decay and no AB Filtration",3/24/98.
5.8 NUREG-0800, Branch Technical Position ASB 9-2, Residual Decay Energy for Light-Water Reactors for Long-Term Cooling, Revision 2, July 1981.
5.9 Zion Station UFSAR, Figure 4.3-1, " Typical Normalized Power Density Distribution", July 1993.
5.10 Princioles of Heat Transfer, Frank Kreith,3rd Edition,1973.
5.11 Fundamentals of Engineering Thermodynamics. Michael J. Moran, Howard N.
Shapiro,1988.
5.12 Nuclear Heat Transoort, M. M. El-Wakil,1978.
5.13 Nuclear Reactor Engineering. Samuel Glasstone & Alexander Sesonke,1963.
5.14 Zion Station UFSAR, Section 4.1, July 1993.
REVISION NO. I
I Exhibit E
.NEP-12-02 COMMONWEALTil EDISON COMPANY Revision 6 l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.15 5.15 Zion Station UFSAR, Section 4.2, July 1993.
5.16 Zion Station UFSAR, Table 4.2-2, July 1993.
! 5.17 Zion Station UFSAR, Figure 4.2-10, July 1993.
5.18 Zion Station UFSAR, Section 4.3.4.1, July 1993.
5.19 Zion Station UFSAR, Section 9.4.3, May 1996.
5.20 Comed Drawing No. M-78, Rev.. Y, Zion Station Unit 1 & 2 Diagram of Auxiliary Building Vent System.
5.21 Flow of Fluids Through Valves. Fittinas. and Pine, Crane Technical Paper No. 410, Twenty Fifth Printing,1991.
5.22 Lniroduction to Heat Transfer, Frank P. Incropera & David P. DeWitt, Second Edition,1990.
i Rev.1 5.23 Deleted 5.24 Zion Station UFSAR, Figure 4.3-2, July 1993.
I 4
REVISION NO. I
Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.16
~~6. CALCULATIONS 6.1 Nomenclature A ' Cross-sectional flow area (ft')~~
c, Specific heat (Btu /lb.- F)
D Fuel rod outer diameter (inches) 2 g Gravitational acceleration (32.2 ft/sec )
2
~~g. Conversion factor (32.2 lb.-ft/lbr-sec )~~
Gr Grashofnumber h Fluid enthalpy (Btu /lb.), per context h Heat transfer coefficient (Btu /hr-ft' *F), per context k Thermal conductivity (Btu /hr-ft 'F)
K Frictionalloss coefficient L Length (ft) di Mass flow rate (Ib./sec)
MW Molecular weight N Number of assemblies Pr Prandtl number P Fuel rod pitch (inches) q Decay heat for a single assembly (Btu /hr)
G Total decay heat from a group of assemblies (Btu /hr)
R Universal gas constant t Time (sec)
T Temperature ( F) 9 Volumetric flow rate (cfm)
Vol Volume (ft')
z Axial coordinate in the z-direction c Emissivity a Stephan-Boltzmann Constant (Btu /hr-ft2 - R)
Dynamic viscosity of fluid (ib./hr-ft) 6.2 Spent Fuel Decay Heat Generation The decay heat load of the last spent fuel batch discharged to the spent fuel pool is calculated per the guidance of the NRC Branch Technical Position ASB 9-2 (Ref. 5.8). A power fraction for a particular spent fuel batch is determined based on the time at operating power, and the time after reactor shutdown. This power fraction is then applied to the rated thermal power of a fuel assembly, giving the decay heat generation rate for that assembly. The i
REVISION NO. 1 L.
Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.17 decay heat power fraction is comprised of two components; one due to fission product decay and another due to heavy element decay.
l 6.2.1 Fission Product Decay Heat The fraction of operating power, P/P,, due to fission product decay is calculated according to Equations 1 and 2 of Reference 5.8 as:
l (t , t,) = (1 + K) (e, t,)- (e,t, + t,)
0 0 O l !
l vtt
}
i j) (e, t,) = A, exp(-a,t,)
where:
to = cumulative operating power (seconds) t, =
time after reactor shutdosm (seconds)
I Rev.1 l K = uncertainty factor; 0.2 s t, <10',0.1 for 10's t, < 10', and 0.0 for t, > 10'
~~A , a. = coefficients having the following values~~
, n A, ^
an (sec ')
l l 1 0.598 1.772 l 4 2 1.650 5.774 x 10 3 3.100 6.743 x 10 2 i 4 3.870 6.214 x 10-2 5 2.330 4.739 x 10-'
6 1.290 4.810 x 104 4
7 0.462 5.344 x 10 I 8 0.328 5.716 x 10~7 ,
9 0.170 1.036 x 10'7 I 4
10 0.0865 2.959 x 10 I i1 0.I140 7.585 x 10" l l q
6.2.2 Heavy Element Decay Heat l
2 i The decay heat generation due to the heavy elements U and Np is calculated according to Equations 3 and 4 of I l Reference 5.8 as:
l
' = 2.28xl 0-'
~~0.7~~
~~1 - exp(-4.91x10" t,)fexp(-4.9 lx10"t,)~~
l i
REVISION NO. 1 i
q Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. I8 p(y pu9) 1.007 1 - exp(-3.41x10-'t,)' exp(-3.41x10-'t,) '
= 2.17x10-'
~~0.7 * < -~~
P. -0.007 1 - exp(-4.91x10"t,). exp(-4.91x10"t,)
l These calculations are shown in Table i for the last batch of fuel discharged to the spent fuel pool. Also included in Table 1 is the volumetric decay heat generation rate required as input to the COBRA code. This value is based on Rn I the fuel heated length of 144 inches, and a radial peaking factor of 1.369. Use of this large radial peaking factor, in conjunction with the last fuel batch discharged to the spent fuel pool, ensures that a conservative decay heat value is used in this calculation.
l 6.3 Decay Heat Generation Axial Profils I
The decay heat generation axial profile is assumed to follow a sinusoidal distribution along the active fuel length.
The peak-to-average weighting factor, X, is given as the following function of axial position z and active fuel length L:
y(:) = EsinE 2 L An appropriate distribution obeys the following normalization:
L y(:)d: = 1.0
'O When this distribution is integrated over the active fuel length, the correct normalization is obtained.
l '- xL fic' y(:)d: = 1 --- cos 1'L ' L' L-
= - = 1.0 L ., L 2x <L>o - L_2 <2s_ L l
The decay heat generation at any location : is then the average decay heat per fuel rod times the peak-to-average weighting factor y(z). Values of the peak-to-average weighting factor are shown in Table 2 for a nodalization of 24 axial segments over the 144 inches of active fuel. The value ofy(r) at the axial centerpoint is 1.571.
6.4 Fuel Assembly Geometry 6.4.1 Parameters for Square Array Fuel Rod Assemblies The COBRA code requires information on the geometry of each channel modeled for a given assembly. The required subchannel data is derived from the following assembly parameters:
REVISION NO. I
r i
Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO.19 D= Rod diameter = 0.422 inches P= Rod pitch = 0.563 inches N= Number of rods in one row = 15 W= Rod diameter plus gap between the rod and assembly wall; for an assembly inner dimension of 8.94 inches (Design input 4.1), W = 0.422 + 0.5*[8.94-(14*P+D)] = 0.422 + 0.318 = 0.740 inches Per Table 4.2 of Reference 5.2, subchannel parameters are calculated as shown below.
Subchannel Number of Channel Flow Area (in') Channel Wetted Perimeter (in)
Type Channels '
Central (N - 1)2 = 196 P, - x D = 0.1771 2 xD =1.326 4
Wall 4*(N - !) = 56 /
x W - D' P - x D = 0.2279 2
- D + P = 1.2259 s 2s 8 2 Comer 4 r 3 x /
y_3 # 2 D = 0.2449 - D + 2 W - D' = 1.3894 s 2s 16 4 s 2s Rev.1 The hydraulie, or equivalent, diameter is defined as four times the cross-sectional area divided by the wetted perimeter. The assembly-average equivalent diameter, weighted by each subchannel type, is computed as:
196(.1771) + 56(.2279) + 4(.2449) 48.4536
_4 = 0.580 inches = 0.04834 feet 196(1.326) + 56(1.2259)+ 4(1.3894) 334.104 The heated perimeter is the perimeter of all fuel rods seen by each channel, with values given below:
Subchannel Type Heated Perimeter (in)
Central xD =1.326 Wall x
-D = 0.6629 ;
2
-D = 0.3314 4
6.4.2 Fuel Assembly Wall Geometry l The fuel assembly walls are modeled using the slab solid structure component of the COBRA code, with slab i identification numbers as shown in Figure 2. Dimensions of the slab nodes are given in the following table.
1 REVISION NO. I
f Exhibit E NEP-12-02 COMMONWEALTII EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M 0063 PROJECT NO. 10436-002 PAGE NO. 20 2
Slab Node Node Length (in) Node Width (in) Node Axial Area (in )
Number 1 4.545 0.075 0.3409 2 4.545 0.075 0.3409 3 4.47 0.075 0.3353 4 4.47 0.075 0.3353 5 4.545 0.075 0.3409 6 4.545 0.075 0.3409 7 4.47 0.075 0.3353 8 4.47 0.075 0.3353 Geometry factors are specified in the COBRA code input file for node-to-node thermal connections pairs. The geometry factors, defined per Equation 6.2 of Reference 5.3, are shown below for the corresponding node-to-node thermal connections pairs.
w =
distance from the solid node center to the edge facing the adjacent node I = length of the solid node at the face in contact with the adjacent node, perpendicular to the l
axial direction FO,A = gCometry factor for slab node A of the pair = (w/l)4
=
Fo3 geometry factor for slab node B of the pair = (w/l)u i
I
, I
! Node A Node B
! Node Pair w I Fa.4 w 1 Fo3 '
(A.B) (in) (in) (N/A) (in) (in) (N/A)
I 1,2 2.2725 0.075 30.30 2.2725 0.075 30.30 l 2,3 0.0375 0.075 0.500 2.235 0.075 29.800 3,4 2.235 0.075 29.800 2.235 0.075 29.800 1 4,5 2.235 0.075 29.800 0.0375 0.075 0.500 5,6 2.2725 0.075 30.30 2.2725 0.075 30.30 l 6,7 0.0375 0.075 0.500 2.235 0.075 29.800 7,8 2.235 0.075 29.800 2.235 0.075 29.800 8,1 2.235 0.075 29.800 0.0375 0.075 0.500 6.4.3 Miscellaneous Fuel Assembly Parameters REVISION NO. I
1 Exhibit E NEP-12-02 COMMONWEALTIf EDISON COMPANY Revision 6 l
l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO 21 The thermal conductivity of the UO 2fuel is assumed to be 0.8 Btu /hr-ft- F. This is a lower bound on the values given in Design input 4.3 for irradiated fuel. Lower fuel thermal conductivities result in higher fuel temperatures, which is conservative with respect to calculated cladding temperatures.
The UO2fuel density is computed at 95 percent of theoretical density, per Design input 4.1.
pra = 0.95*10.97 = 10.422 g/cm' *(62.428 g/cm' per Ibdft') = 650.6 lb /ft' The fuel / cladding gap conductance is computed by dividing the thermal conductivity of helium by the gap width.
For helium at 800 F, the thermal conductivity is 0.138 Btu /hr-ft- F per Design input 4.5.2. The gap width is calculated by subtracting the cladding thickness and fuel pellet diameter from the total cladding outer diameter as shown:
d
[0.422 - 2*(0.0243) - 0.3659] / 2 = 0.00375 inches = 3.125 x 10 ft 4
gap conductance = 0.138 / 3.125 x 10 = 441.6 Bru/hr-ft2 ,op It is conservative to maximize the gap conductance for the purposes of calculating the maximum cladding temperature. Therefore, a gap conductance of 500 Btu /hr-ft2 - F will be used in the COBRA model.
l 6.5 Frictional Pressure Droo 6.5.1 Friction Factor for Subchannel Flow l The friction factor is given as a function of Reynolds Number according to the expression (see Equation 4.55 of )
Ref. 5.2):
C' f = Re" where Cr is a friction constant as defined below, per Equation (4.54) of Reference 5.2 n = 1 for laminar flow n = 0.18 for turbulent flow r 8 r m2 Cf = a + b, p -I +b 2 p -I
<D > <D >
For interior subchannels with P/D = 1.334, Table 4.1 of Reference 5.2 provides the following coefficients for computing both laminar and turbulent friction constant Cr.
a bi b2 i
l Laminar 35.55 263.7 -190.2 l Turbulent 0.1339 0.09059 -0.09926 REVISION NO. I
Exhibit E NEP-13-02 COMMONWEALTII EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 22 Substituting these coefficients into the expre.ssion for the friction constant gives:
Laminar Flow: Cr = 102.4 Turbulent Flow: Cr = 0.153 6.5.2 Local Frictional Loss Coefficients 6.5.2.1 Base-plate Flow Hole Loss Coefficient The loss coefficient associated with the 5-inch diameter flow hole it the base-plate of the storage cell, K , is given by Equation (4.59) of Reference 5.2.
K* = ^'*A, C
where: A. is the cell cross sectional area An is the base-plate hole area Co is the discharge coeflicient The areas are computed as:
2 2 A. = 8.94 - 225*n/4*(.422)2 = 48.453 in 2
An = n/4*(5.0)2 = 19.635 in With the discharge coefficient taken as 0.6 per Reference 5.2, the value of K, is 14.14.
6.5.2.2 Fuel Assembly Grid Spacer Loss Coeflicient The loss coefficients for the grid spacers are conservatively assumed to have a value of 1.5, consistent with r.n abrupt area contraction followed by an abrupt area expansion per Reference 5.21 (page A-26). The ten grid spacers are assumed to be uniformly distributed along the heated length of the fuel, as shown below.
Description Axial Coordinate Non-dimensional Loss Coefficient Axial Coordinate (inches) (N/A)
Base Plate Flow Hole, Ko 0.0 0.0 14.14 Bottom of Active Fuel 0.0 0.0 Grid Spacer 7.2 0.05 1.5 Grid Spacer 21.6 0.15 1.5 Grid Spacer 36.0 0.25 1.5 IMF Grid Spacer 50.4 0.35 1.5 f
REVISION NO. 1 I
1 Exhibit E NEP-12-02 COMMONWEALTil EDISON COMPANY Revision 6 l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 23 l
Grid Spacer 64.8 0.45 1.5 IMF Grid Spacer 79.2 0.55 1.5 Grid Spacer 93.6 0.65 1.5 l
IMF Grid Spacer 108.0 0.75 1.5 4 Grid Spacer 122.4 0.85 1.5 Grid Spacer 136.8 0.95 1.5 Top of Active Fuel 144.0 1.0 6.6 Heal _ Transfer Correlations for Assembly Subchannels Convective heat transfer coefficients for heat exchange between the fuel rods and the fluid are specified in the COBRA code using the following relation (see Equation 6.8 of Reference 5.3):
h = (a, Re2 Pr"> + a.)I D,
where: ai, a2, a3, a.i = empirical coefficients Re = Reynolds Number Pr = Prandtl Number k = thermal conductivity of the fluid D, = channel hydraulic diameter The heat transfer correlations are entered in pairs, with one set of coefficients for laminar flow and one set for turbulent flow. The COBRA code then employs the maximum value obtained with the laminar and turbulent equations, where the local Reynolds number and Prandtl number are calculated internally by the code.
Reference 5.3 (page 7.5) cites the following correlations for subchannel flow.
hu,,,,,, = (0.83 Re " Pr*")I D,
hm,,,,
= (0.33 Re" D. Pr"')I These correlations were used in the PNL validation case entitled Single-Assembly Heat Transfer Test (SAlfiT) as documented in Ref. 5.3. The use of these correlations resulted in very close agreement between the calculated temperatures by COBRA and the measured test data. The temperatures predicted by the COBRA code conservatively bound the test data, per the validation documented in Reference 5.3.
REVISION NO. 1
Exhibit E N EP-12-02 COMMONWEALTII EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 24 )
6.7 Boundary Conditions for COBRA Model Boundary conditions required for the COBRA model include the following: !
Rev.I
~~1) Inlet air temperature at the base of the spent fuel storage cell, T,,~~
~~2) Inlet average mass flux at the base of the spent fuel storage cell, G The air temperature at the inlet to the spent fuel storage cell is taken as the Fuel handling Building air temperature.~~
This temperature is computed as 445.6 F, as documented in Attachment D of this calculation. This value is conservatively rounded up to 450 F for the COBRA model.
The average inlet mass flux is selected by an iterative procedure, as described below and shown in Table 3. At steady-state conditions, natural circulation patterns will be established in the Fuel Handling Building as shown in l Figure 1, with the inlet air being drawn towards the spent fuel pool cavity as the heated air rises past the fuel assemblies, heating the Fuel Handling Building environment. Under these conditions, the total pressure drop for upward flow along the fuel rods will match the pressure gain for the return flow as the air returns to the bottom of the spent fuel pool. Equating these pressure differences provides a means of determining the mass flux for flow in the modeled spent fuel storage cell.
At steady-state, the one-dimensional conservation of momentum equation for upwards flow along the heated length of the fuel assembly is given below, with positive taken as the upwards direction.
OpVV a' fpV'
=--- -- pg
& & 2D, Substituting for the mass flux G, where G = pV results in:
D 'G =
a' JG' - pg
&<ps & 2pD, Integrating the momentum equation over the control volume from the bottom of the fuel assembly, subscript o, to the top of the fuel assembly, subscript L, yields the following equation.
'G = -(P - P,)- h JG' h ' h
<ps, t
, 2p(:)D, K"G' 2p, -[ K'G , 2p, - 2pt K'G' -p(:)gd:
Defining AP = (P,- P ), tthe above equation is rearranged as AP = G, 1 1' 6
- fG 2
h + g,,g2 + [ g,g2 +
g,g 2 4
+ + p(:)gd:
spt p,) , 2p(:)D, 2p, , 2p, 2p, ,
l where the total pressure drop for upflow from o to L , AP, is comprised of the following components:
AP = APmeimo. + AP,aru + AP ,t ,p, + APya.m + AP,x,o<,,, + APy mun.
l I REVISION NO. I l
Exhibit E NEP-12-02 COMMONWEALTil EblSON COMPANY Revision 6 l
CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 25 l
Rev.1 Since the precise nature of the temperature and density profile along the fuel rod heated length is not known a priori, the following approximations are used in place of the integrations.
For the friction component along the fuel rod surface:
JG* 3 JLG'
', 2p(:)D, 2 p,,D, For the gravitational term:
L, p(:)gd: 2 p ,gL 0
An average air density is also used for the local form losses at the grid spacers, gising:
[K'G s G* ZK, n 2,P 2Pa., o Substituting these approximations into the total pressure loss equation yields the following expression for upwards flow from point a to point L:
l'
+ K' 1
AP = G* - -
+ JLG' + G' K" + * + p,'gL
<Pt p,, 2p ,D, 2p, 2p,, 2pt l
The return flow path is from point L back to point o, through the plenum region above the spent fuel pool cavity, with recirculation back to the base of the spent fuel storage cell. Frictional losses are neglected, since the return flow is through the relatively large open areas of the Fuel Handling Building and open regions of the spent fuel pool. Under these conditions, the momentum equation for the return flow has the following form.
jp r g2 jp
&\ps & #
Integrating from point L to pe;r+ > gives r g28
~~o, l~~
= -(P, - Pt )- p(:)gd:
\P>t 't which simplifies to:
'] 1'
~~AP = G' p(:)gd:~~
\ Pt Pos .t i i
l Upon exiting the top of the spent fuel assembly, the air mixes with the other parallel flows from the other spent fuel storage cells, reaching the homogeneous air temperature of the Fuel Handling Building documented in Attachment i
D. Therefore, the density of the retum air flow can be considered as constant, consistent with the Fuel Handling REVISION NO. I i
l Exhibit E N EP-12-02 COMMONWEALTil EDISON COMPANY Revision 6
( CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 25.1 Rev.I Building air temperature of 450 F. The gravitational term simplifies as a result of the constant density, with the appropriate length scale for downwards flow taken as the height of the spent fuel storage cell, H. Horizontal flow along the bottom of the spent fuel pool cavity does not contribute to the gravitational integral, as no elevation changes occur. Therefore, the pressure difference for return flow from point L to point o is expressed as:
r; ;3 .
AP = G* + pniaRH
\Pt Poi i
\
Equating the expressions of pressure difference for the upwards flow and the retum flow gives:
G' JL + K*+['K*+ K' + p,gL = pniegH 2p,D, 2p, 2p , 2pt l
l Solving for the mass flux G results in the following. j v2 J
g (pninH - p,L)g 1
fL + K* + [' K* + K' L2p,D, 2p, 2p, 2pt_
All of the quantitics in the above expression are known with the exception of the friction factorf, the average air density pm, and the outlet air density at the top of the heated fuel assembly, pt. For a given guess value of assembly inlet air mass flux, the outlet and average air temperatures (and densities) can be determined. He friction factor can be evaluated from the correlation presented in Section 6.5.1. The iterative procedure used to determine the appropriate assembly inlet air mass flux is shown in Table 3, and follows the algorithm described below.
2
~~1) Choose guess value of assembly inlet mass flux, G (ib./sec-ft )~~
2) Compute outlet air temperature, based on an inlet air temperature of 450 F. He peak decay heat for a fuel assembly, q,m,,, (8tu/hr), is obtained from Table 1. The outlet air temperature can then be computed according to the equation:
l',,' = T"' + 4*""'
3600sec/ hr
G
~~c, where the specific heat, c,(Btu /lbm 'F) corresponds to a temperature of 700 F, which is representative of the average air temperature along the heated length, per Assumption 3.2.6.~~
~~3) The average air temperature for upwards flow in the assembly is then obtained.~~
~~7. = 0.5 * ( T,, + T,,, )~~
I REVISION NO. I
I Exhibit E NEP-12-02 COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 25.2 Rev.1
~~4) The outlet and average air densities are obtained using the Ideal Gas law at the respective air temperatures:~~
_ P,,,
RT
5) The friction factor is determined by the correlation given in Section 6.5.1, where the Reynolds number is based on the guess value of the assembly inlet air mass flux, calculated as shown below.
i l Re = GD' l lI l The air viscosity is evaluated from a curve fit to the data provided in Design Input 4.5.1. The curve fit is l shown in Figure 3 as a ftmetion of air temperature, and matches well with the viscosity data.
l
6) Using the air densities and friction factor determined in steps 4 and 5 respectively, the expression for G derived earlier by equating the pressure differences along the upwards and return flow path, is evaluated
~~and denotea as G'.~~
~~l 1/2~~
~~c. , (pmaH - p ,L)g~~
\
l s + x. + [,K" + x.
l _ 2p, D, 2p, 2 p,, 2pt_
1 The guess value of assembly inlet air mass flux, G, is adjusted until convergence is obtained with the calculated value of G'.
The calculation details of the iterative procedure are shown in Table 3, with the boxed values being used as input to the COBRA-SFS model. For an inlet air temperature of 450 *F, the inlet air mass flux for the modeled spent fuel 2
assembly is 1.1212E-04 Mlb./hr-ft , as shown in Table 3. Also shown in Table 3 is the total pressure drop for upwards air flow along the spent fuel assembly, AP, as well as each of the pressure loss components. The total pressure drop computed in Table 3,0.00424 psia, will be compared to the COBRA-SFS output as a check on the accuracy of the modeling described in this section of the calculation.
l l
i REVISION NO. I
f Exhibit E NEP-12-02 !
COMMONWEALTH EDISON COMPANY Revision 6 l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-002 PAGE NO. 26
7.
~~SUMMARY~~
AND CONCLUSIONS l
Rev.1 The COBRA-SFS computer code is used to determine the maximum fuel rod cladding temperatures following a ,
l. postulated loss of cooling water from the Zion Station spent fuel pool. This COBRA-SFS input file is called I z irc-rv1. inp, with a corresponding output file titled zirc-rvi . out. The results of the COBRA-SFS analysis for the bounding fuel assembly are documented in Attachment E, with the maximum calculated cladding i temperature of 974.5 F. Also documented in the COBRA-SFS cutput are the pressure drop for upwards flow in l
, the spent fuel assembly,0.004094 psia, and the assembly average outlet air temperature of 972.8 'F. Both of these i values compare favorably with the values shown in Table 3,0.00424 psia and 975.2 'F respectively.
It is shown that the peak cladding temperature following loss of cooling water is below the critical oxidation temperature of 1049 'F for those cases where the spent fuel pool has been cooled for approximately 710 days or longer since the final reactor shutdown. This cooling duration corresponds to a final reactor shutdown date of 2/21/97, and a postulated reference date of 2/1/99 for the loss of spent fuel pool cooling water. Therefore, rapid oxidation of the zirconium cladding will not occur as a result of a sudden loss of spent fuel pool cooling water, i provided the postulated accident occurs after 2/1/99.
These results reficct the significant conservatisms inherent in the heat transfer modeling as described in this
, calculation. Primarily, these include no credit %ving taken for radial heat transfer from the fuel storage cell to neighboring heat sinks. Also, the decay heat generation values used in the calculation are based on continuous reactor operation at 100 percent rated power. This idealized condition is conservative with respect to probable fuel cycle histories, anu ads to higher calculated fuel cladding temperatures.
l i
Tables e Table 1 Decay Heat Generation for Spent Fuel Rods e Table 2 Axial Power Profile e Table 3 Boundary Conditions for COBRA Model i
I t
REVISION NO. 1
Comed Calc. No. 22S-0-110M-0063 Zion Str. tion Unita 1 cnd 2 Table 1 Revision 1
& het No.10436@3 Decay Heat Generation for Spent Fuel Rods Page No. 27 Parameter Units Value Basis Rated Core 'Ihermal Power (MW) 3250 Design Input 4.1 Rated Core nermal Power (Btu /hr) 1.109E+10 Number of Assemblies per Full Core 193 Design Input 4.1 Rated nennal Power per Assembly (Btu /hr) 5.75E+07 l Number of Assemblies in Last Discharge 193 Design Input 4.1 Operating time (days) 432 Design Input 4.1 Operating time, to (seconds) 3.73E+07 ,
Shutdown Date 2/21/97 Design Input 4.1 i Analysis Reference Date 2/1/99 l Cooling Duration (days) 710 l l
l Time aner shutdown, t. (seconds) 6.13E+07 l
Einnien. Product Decav IIcat d
An a.(sec ) Aoexp(-ao .t.) Section 6.2 0.598 1.7720E+00 0.000E+00 1.65 5.7740E-01 0.000E+00 3.1 6.7430E-02 0.000E+00 3.87 6.2140E-03 0.000E+00 2.33 4.7390E-04 0.000E+00 1.29 4.8100E-05 0.000EH)0 0.462 5.3440E-06 1.964E-143 0.328 5.7160E-07 1.939E-16 0.17 1.0360E-07 2.954E-04 0.0865 2.9590E-08 1.408E-02 0.114 7.5850E-10 1.088E-01 Power Fraction, P/P (x,t.) 6.160E-04 4
An an (sec ) Anexp[-an *(t.+t.)] Section 6.2 0.598 1.7720E+00 0.000E+00 1.65 5.7740E-01 0.000E+00 3.1 6.7430E-02 0.000E+00 3.87 6.2140E-03 0.000E+00 2.33 4.7390E-04 0.000E+00 1.29 4.8100E-05 0.000E+00 0.462 5.3440E-06 4.647E-230 0.328 5.7160E-07 1.052E-25 0.17 1.0360E-07 6.181 E-06 0.0865 2.9590E-08 4.667E-03 0.114 7.5850E-10 1.058E-01 Power Fraction, P/P (x,t.+t.) 5.523E-04 Uncertainty Factor, K 0.0 Section 6.2 Ileavy Element Decav Ileat i
P/P.(U-239) 0.00E+00 Section 6.2 P/P.(Np-239) 2.18E-94 Section 6.2 Total Decay IIcat Power Fraction, P/P.(t.,t.) +
, P/P4U-239) + P/P.(Np-239) 6.37E-05 i l
7JOZIRCI.XLS Decay llent Generation
I Comed Calc. No. 22S-0-11OM-0063 Zion Station Units I cad 2 Table 1 Revision I lY ject No. M3m3 Page No. 28 Decay Heat Generation for Spent Fuel Rods Parameter Units Value Basis Average Decay lleat per Assembly for Last Discharge (Bru/hr) 3.6615E+03
(
Radial Peaking Factor 1.369 Design input 4.1 Peak Decay IIcat per Assembly for Last Discharge, q ,,, (Btu /hr) 5.013E+03 i Total number ofrods per Assembly _ 225 Design Input 4.1 Number of Unheated Rods 21 Design Input 4.1 Number ofIleated Rods 204 Design input 4.1 Peak Decay Ileat per ifeated Rod, qma (Btu /hr) 24.57 Rod Outer Diameter (inches) 0.422 Design Input 4.1 Rod IIcated Length (inches) 144 Design Input 4.1 Fuel Rod Volume based on lleated Length (A') 0.011656 Peak Rod Volumetric IIcat Generation (Btu /hr-ft') 2108.1 Peak Rod Vohunetric IIcat Generation (MBtu/hr ft')l 0.002108 l l
ZIOZIRCl.XLS Decay llent Generation
I l
Comed Calc. No. 22S-0-110M-0063 l Zion Station Units 1 and 2 Revision 1 Table 2 l Project No.1043W2 Page No. 29 Axial Power Profile i
Active Fuel Length, L (inches) 144 Number of axial segments, N 24 Axial Segment licight in Active Fuct Region, Az (inches) 6 Nondimensional Relative Power l Description Elevation Elevation Factor Power Fraction l z z/L x(z) x(z) * (Az/L) l (inches) (N/A)
Bottom of Active Fuel 0.0 0.000 0.000 0.000 6.0 0.042 0.205 0.009 12.0 0.083 0.407 0.017 18.0 0.125 0.601 0.025 24.0 0.167 0.785 0.033 l 30.0 0.208 0.956 0.040 36.0 0.250 1.I11 0.046 42.0 0.292 1.246 0.052 48.0 0.333 1.360 0.057 54.0 0.375 1.451 0.060 1 60.0 0.417 1.517 0.063 ,
l 66.0 0.458 1.557 0.065 l 72.0 0.500 1.571 0.065 78.0 0.542 1.557 0.065 l 84.0 0.583 1.517 0.063 90.0 0.625 1.451 0.060 96 0 0.667 1.360 0.057 i 102.0 0.708 1 ' 46 0.052 l 108.0 0.750 1.Ii1 0.046 114.0 0.792 0.956 0.040 i 120.0 0.833 0.785 0.033 126.0 0.875 0.601 0.025 132.0 0.917 0.407 0.017 138.0 0.958 0.205 0.009 Top of Active Fuci 144.0 1.000 0.000 0.000 l Total of Segment Power Fractions 1.00 l
l l
ZIOZlRCl.XLS Axial Power Distribution b
)
ComFA Calc. No. 22S4-110M 0063 l Zion Station Units I cnd 2 Revision 1 Table 3 l Project No. 10436 403 Page No. 30 Boundary Conditions for COBRA Model i
Description Units Value Basis Peak Decay lleat Generation Rate per Assembly, q, (Utu'hr) 5,012.57 Table 1 l Assembly-Average Ilydraulic Diameter, D. (ft) 0.04834 Section 6.4.1 l Fuel Rod Length, L (ft) 12.68 Design input 4.1
( Total Fuel Assembly Flow Area, A (ff) 0.33648 Section 6.4.1 Storage Cell licight above Baseplate, II (11) 14.0 Design Input 4.1 Air Inlet Conditkins Fuel llandling Duilding Air Temperature, T. (F) l 450 l Attachment D
Pressure, P (psia) 14.7 Assumption 3.2.5 Air Gas Constant, R (ft-lbrib. *R) 53.34 Design input 4.6.1 Density, p. (Ib,,,/ff) 0.0436 = P.* 144/W(T.+459.67) 2 Total Maas Flu < for Assembly, G (ib./sec-ft ) 0.03114 Selected by Iteration Total Mass Flux for Assembly, G . (Mlbdhr-ft') l 1.1212E-04 l=G'3600/IE6 Air Outlet Conditions Outlet Air Temperature, T (F) 975.2 -T.+q /(3600'G'A ,,,*c,)
Density, p (Ib,,,/ft') 0.0277 =P.*144/R/(T, +459.67)
Air Average Conditions Temperature, T. ('F) 712.6 -(T. + T )/2 Density, p., (Ibdff) 0.0339 =P.*144/W(Tm+459.67)
Dynamic Viscosity, p (Ib./ft-sec) 2.13E-05 Curve fit (Figure 3)
Specific Ilcat, c, (Bra'Ib. *F) 0.253 Assumption 3.2.6 Pressure Drop for Upwards Flow Along Fuel Rod Reynolds Number, Reo, 70.6 -G*D,'
Friction Factor based on Reynolds Number, f 1.450 -102.4.11eo.
Coefficient for Base-Plate lloie Losses, K. 14.14 Section 6.5.2.1 Coeflicient for Spacer Losses, K.,- 1.50 Assumption 3.1.3 Number of Spacer 10 Design input 4.1
, Channel Loss Coefficient, Ka = f'UD. + EK 395 -f*l/D. + 10*K ,
Coeflicient for Exit Losses, K. 1.0 Assumption 3.1.4 Solve for Total Assembly Mass Flux, G' (Ibdsec-ff) 0.03114 Section 6.7 Check Ibr Convernence Difference in Mass Fluxes (G G') 0.0000 -G - O' Pressure Dron Components Acceleration Pressure Drop, AP.,. (psia) 0.000003 -G'*(l'po. - 1/p.y(144'32.2) l Fuel Rod Frictional Pressure Drop, APr (psia) 0.00118 =(f'UD.)*G'/(144'2*pm*32.2)
Local Form Loss Pressure Drop, APg (psia) 0.00008 =G'*[K.'(2*p.)+10'K,/(2'p., )+ K /(2*p )}'(144'32.2)
Gr:vitational Pressure Drop, AP, (psia) 0.00298 =p *32.2'!/(144'32.2)
Total Pressure Drop for Upwards Flow along Fuel Rod, AP (psia) 0.00424 =AP.. + APr + APg + AP, ZIOZlRCl.XLS BoundaryConditions
l 3 1 a 6 nn 0 oi 0iF
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j Comed Calc. No.2254110M 0063 l
Zion Station Units 1 and 2 TtbliAl Revision 1 Project No. 10436-003 Attachment A Page No. Al Spreadsheet Cell Formulas for Table 1 A l H l C l D l E l F L
1 Parameter I! nits Value Basis 1
1 Rated Core Thermal Power (MW) 3250 Design input 4.1 1 Rated Core Thermal Power (Btuhr) -C4* 1000000*3.4127 l 1 Nwnbar of Assemblies per Full Core 193 Design Input 4.1 l
l 1 Rated Thermal Power por Assembly (Bluhr) =CS/C6 L
Nanbar of Aasemblies in Last l 1 Discharge 193 Design Input 4.1
)
3 Operating time (days) 432 Design Input 4.1 11 Operating time, t. (seconds) *C10*3600*24 3 Shutdown Dats 35482 Dcaign input 4.1 2 Analysis Reference Date 36192 3 Cooling Duration (days) =Cl3-Cl2 E
3 Time aner shutdown, t, (seconds) =Cl4*24*3600 E
3 Fission Product Deca ot {gt 3 A. a. (sec) A exp(-a t.) Section 6.2
=C20*EXP(-
2 0.598 1.772 D20*$C$16)
=C21*EXP(-
1 1.65 0.5774 D21*$C$16)
=C22*EXP(-
3 3.1 0.06743 D22*$C$16)
=C23 *EXP(-
2 3.87 0.006214 D23*$C$16)
=C24*EXP(-
3 2.33 0.0004739 D24*$C$16)
=C25 *EXP(-
2 1.29 0.0000481 D25*$C$16)
=C26*EXIN-3 0.462 0.000005344 D26*$C$16)
=C27*EXP(-
2 0.328 0.0000005716 D27*$C$16)
=C28'EXP(-
28 0.17 0.0000001036 D28*$C516) l -C29'EXP(-
2 0 0865 0.00000002959 D29*$CS16)
=C30*EXP(-
2 0.1I4 0.0000000007585 D30*$CS16) 31 Power Fraction. P P.(s.t.) =SUME20 E30)'200 7J023RCl.XLS Decay lisat Generatism (Egns)
I l-
r Comed Calc. No.225-0-110M4063 Zion St: tion Unita i and 2 R*d'iaa 1 Tzbis Al Pricct No 10436 @ 3 Spreadsheet Cell Formulas for Table 1 A tachment A Page No. A2 A l H l C l D l E l F E
3 A. a. (sec) A.exp/-a.*(t.+t.)] Section 6.2
=C34*EXP(-
3 0 598 1.772 D34*($C$11+$C$16))
1 -C35'EXP(-
2 1.65 0.5774 D35*($C$11+5C$16))
=C36*EXP(-
36 3.1 0.06743 D36*($C$ll+$C$16))
C37'EXP(-
I E 3.87 0.006214 D37'($C$11+$C$16))
=C38*EXP(-
3 2.33 0.0004739 D38*($C$11+5C$16)) l I
-C39'EXP(-
E I.29 0.0000481 D39*($C$l t +$C$16))
=C40*EXP(-
3 0.462 0.000005344 D40*($C$11 +$C$16))
=C41*EXP(-
1 0.328 0.0000005716 D41*($C$11+ $C$16))
=C42*EXP(-
1 0.17 0.0000001036 D42*(SC$11 +$C$16))
=C43 *EXP(-
1 0.0865 0.00000002959 D43*($C$11+ $C$16))
=C44*EXP(-
3 0.I14 0.0000000007585 D44*(SCS11+$CS16))
l 3 Power Fraction, P,P.(z,t.+t.) = SUM (E34.E44y200 3 Uncertainty Factor, K 0 Section 6.2 1
l 3licavv Element Deca gilgt 4 00228*0.7*(1 EXP(- ;
0.000491'$C$11))*E l 3P.P.(U-239) XP( 0 000491*$CS16) Section 6.2 0.00217*0.7*(1.007*
(1-EXP(-
l 0.00000341*$C$11))*
i EXP(-
l 0.00000341*$C116).
i 0.007'(1-EXP(-
, 0.000491*$C$11;*E I XP(-
Se P P.(Np.239) 0 000491'$C$16)) Section 6.2 1
Total Decay llcat Poact Fraction, P.P.(t t.) + P/P.(U-219) + P.P.(Np- =(l +C46)*E31-p 239) E45 +C49+C50 53
' ZIO2.IRC1.XLS Decay lical Generation (Eqns)
y.
Comed- Calc. No. 225-0-Il0M-0063 Zion St: tion Units I and 2 Tabli Al Revision 1 1%ect Na 104p3 Attac t A Page Na A3 l Spreadsheet Cell Formulas for Table 1
! A l B l C l D l E l F E
X Paranwter Units Value Basis Average Decay licat por Assembly fer E Last Discharge (Blu,hr) =C7'C$2 E
E g Radial Peaking Factor 1.369 Design input 4.1 Peak Decay IIcal por Assembly for Last 1 Discharge. % (Bluhr) =C57'C60 E
R Total number of rods per Assembly 22$ Desip Input 4.1 g Number of Unheated Rods 21 Desip Input 4.1 g Number ofiteated Rods =C63464 Desip input 4.1 E
i g Peak Decay lleat perIIcated Rod, % (Btuhr) -C61/C65 E
f Rod Outer Diameter (inches) 0.422 Desigr, Input 4.1 2 Rod licated Length (inches) 144 Desip input 4.1 Fuct Rcd Volume based on llcated =PI(y4*C69?C70/l j 1 Length (A') 2^3 !
E 2 Peak Rod Volumetric IIcat Generation (Dtuhr-A') =C67/C71 74 Peak Rod Volumetric IIcal Generation (hUltahr-ft') l=C73/1000000 l l
ZK)ZIRCI.XLS Decay lical Gancration (Eqns) l
Comed Calc. No. 22S-0-110M-0063 Zion Station Units I cnd 2 Table A2 Revision 1 Project Na 10436-002 Anchment A Page Na A4 Spreadsheet Cell Formulas for Table 2 Active Fuel Length, L (inches) 144 Number of axial segments, N 24 Axial Segment Ileight in Active Fuel Region, Az (inches) =D3/D4 l
Nondimensional Description Elevation Elevation Relative Power Factor Power Fraction z z/L x(z) x(z) * (Az/L)
(inches) (N/A)
Bottom of Active Fuel 0 =B12/$D$3 =PI()/2* SIN (PI()*C12) =D12*1D$5/$D$3
=B12+5D$5 =B13/$D$3 =PI()/2* SIN (PI()*Cl3) =D13*$D$5/$D$3
=B13+$D$5 =B14/$D$3 =PI()/2* SIN (PI()*Cl4) =D14*$D$5/$D$3
=B14+5D$5 =iil5/$D$3 =PI()/2* SIN (PI()*C15) =D15*$D$5/$D$3
=B15+$D$5 =B16/$D$3 =PI()/2* SIN (PI()*C16) =D16*$D$5/$D$3
=B16+$D$5 =B17/$D$3 - =PI()/2* SIN (PI()*C17) =D17*$D$5/$D53
=B17+$D$5 =B18/$D$3 =PI()/2* SIN (PI()*C18) =D18*$D$5/$D$3 ,
=B18+$D$5 =B19/$D$3 =PI()/2* SIN (PI()*C19) =D19*$D$5/$D$3
=B19+$D$5 =B20/$D$3 -PI(y2* SIN (PI()*C20) =D20*$D$5/$D$1 -
=B20+$D$5 =B21/$D$3 =PI()/2* SIN (PI()*C21) =D21 *$D$5/$D$3 l
=B21+$D55 =B22/$D$3 =PI()/2* SIN (PI()*C22) =D22*$D$5/$D13 I
=B22+$D$5 =B23/$D$3 =PI(y2* SIN (PI()*C23) =D23*$D$5/$D$3
=B23+5D$5 =B24/$D$3 =PI()/2* SIN (PI()*C24) =D24*$D$5/$D$3
=B24+$D$5 =B25/$D$3 =PI(y2* SIN (PI()*C25) =D25*$D$5/$D$3
=B25+$D$$ =B26/$D$3 =PI()/2* SIN (PI()*C26) =D26*$D$5/$D$3
=B26+$D$5 =B27/$D$3 =PI()/2* SIN (PI()*C27) =D27*$D$5/$D$3
=B27+$D$5 =B28/$D$3 =PI(y2* SIN (PI()*C28) =D28*$D$5/$D$3
=B28+$D$5 =B29/$D$3 =PI()/2* SIN (PI()*C29) =D29*$D$5/$D$3
=B29+$D$5 =B30/$D$3 =PI(y2* SIN (PI()*C30) =D30*$D$5/$D$3
=B30+$D$5 =B31/$D$3 =PI(y2* SIN (PI()*C31) =D31*$D$5/$D$3
=B31+$D$5 - =B32/$D53 =PI(y2* SIN (PI()*C32) =D32*$D$5/$D$3
=B32+$D$5 =B33/$D53 =PI(y2* SIN (PI()*C33) =D33*$D$5/$D$3 '
=B33+5D$5 =B34/$D$3 =PI(y2* SIN (PI()*C34) =D34*$D$5/$D$3
=B34+$D$5 =B35/$D$3 =PI(y2* SIN (PI()*C35) =D35*$D$5/$D$3 Top ofActive Fuel =B35+$D$5 =B36/$D$3 =PI(y2* SIN (PI()*C36) =D36*$D$5/$D$3 Total of Segment Power Fractions = SUM (E13:E37)
I i
ZIOZIRC1.XLS Axial Power Distribution (Eqns)
r Caned Calc No 22 Sol 10M 0063 Zion Station Umts 1 and 2 Revision I Tchle A3
"" ' U""3 Attachment A Page No A5 Spreadsheet Cell Formulas for Table 3 A l B l e i 13 1
1 Description Units Value Basis l
l 4 Peak Decay liest Generation Rate per Assembly, q (Blutir) =' Decay liest Generation "$C561 Table 1 I 3 Assembly AverageIlydraulce Diameter D. (A) 0 04834 Section 641 ;
1 Fuel Rod Length, L (A) 12 68 DesignInput 41 l l
3 Total Fuel Assembly Flow Area, A (A') 0.33648 Section 6 41 l 1 Storage Celllleight above Baseplate II - (ft) 14 Design input 4 i l
] Air talet Cond6tions j l
3 Fuel Handling Buildmg Air Temperature. T. (*F) 450 Attachment D 3 Pressure, P (psia) 14 7 Assumption 3 2.5 2 Air Oas Constant, R (ft-Ibrik'R) 5334 Design input 4 6.1 j 3 Density, pi. (Ib'ft') =Cl2*l44'(Cl3*(Cil+459 67)) *P.* 144W(T.+459 67)
E 3 Total Mass Flux for Assembly O (ib./sec-ft') 0 oli143138135242i Selected by lu: ration 2 Total Mass Flux (m Assembly, O (AGb'hr ft )
8
=C16* %00/1000000 =0*1600/IE6 Air Outist CondMAons i to Outlet AirTemperature.T (*F) acil+C4/(3600*C16*C7'C27) *T.+q, ,43600*O*A *c,)
3 nensity, p., (Ib'ft') =Cl2* 144'(Cl3*(C20+459 67)) =P.*144R/(T +459 67)
Air AverageConditions 24 Temperature, T (*F) =(Cil+C20y2 =(T.+ T y2 l
l l
3 Density, p, (Ib.'ft') =Cl2*144/(Cl3*(C24+459 67)) =P.*1441LTT +45967)
,26, Dynanue Yiscosity, p (Ib'ft sec) =0 000012+0.00000001308'C24 Curve fit (Figure 3) 3 Specific llent, c, (Dtu'Ib*F) 0 253 Assumption 3 2 6
~
2e 29 Pressure Drop for Upwards Flow Along Fuel Rod 5
g Reynolds Number, Reo . =C16*CS/C26 =G*D,/p 3 Fnction Facta bened on Reynolds Number, f =102 4'C31 =102 4Reo, 3Coc5cient for Base-Plate } lote Losses, K. 14 14 Section 6 5 21
! 34 Coefncient (w Spacer Lesses. K, t5 Assumption 31.3 g Number of Spacers 10 Design inptit 41 36 Channel Loss Coemcient. Kch '= f*l)De + SK =C32*C6/C5+C35'C34 -f*1JD, + 10*K.,
3Coe5cient for Exit Losses, K. I Assumption 314 i
~
se 1
~~SQRT((Cl4*C8-C25 *C6)* 32 17/(C3&(2 *C25)+C33/(2~~
~~8 2 Solve fa Total Assembly Mass Flux, O' (Ib./sec-A ) Cl4)+C37/(2*C21))) Section 6.7 5~~
4P Ched for convernence ZKEIRCI XLS Boundary Conditions (Eqns)
j- ,
l l
l Comed Calc. No 2254110MM3 I Zum Station Umts I and 2 Table A3 Revision i Pr ja1Na IN3 Attacht A Page N A6 Final Spreadsheet Cell Formulas for Table 3 A I e I e I n e DifTerence m Mass Fluxes (O 0) =C16439 =0 , 0
.S.,
,,te,,, bnawtDrppf4!mp90nola l
3 Acceleration Presswe Drop, AP (psia) al(144*3217)*Cl6^2*(1/C21 1/Cl4) =O'*(iip. 1/pg)/(144*32 2)
= l (144
~~32.17)*C32~~
C6
~~C 16^2/(2 *C5~~
~~M Fuel Rod Fnctional Presswe Drop. APs (psia) C25) -(f*L/D.)*O'/(144*2*p.,.*32.2)~~
=l/(144*3217)*Cf 6^2*(C33/(2*Cl4p =Oa*[K./(2*rgy10*K,,/(2*p PK./(2*p.))/
M Local Form Loss Presswe Drop, APs (psia) C35'C34t2*C25PC37/(2*C21)) (144*32 2) 48 Gravitsuonal Presswe Drop, AP, , (psia) =l/(144*3217)*C25'3217*C6 =p *32.2*U(144*32.2)
.32.
Se Total Pressure Drop for Upwards Flow along Fuel Rod, AP (psia) =C45+C4G C47+C48 =AP + AP, + APs + AP , )
i 1
l l
4 l
I l
1 ZIOZ1RCLXL3 BoumlaryC<mditions (Fes)
Exhibit E NEP-12-02 COMMONWEALTII EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. B1 Final ATTACIIMENT B )
COBRA-SFS Input / Output Files from Revision 0 1
Loss of Spent Fuel Pool Cooling Water on Reference Date of 7/1/98 (Microfiche)
Files included: l "peakassm. inp" (1 sheet of microfiche)
"peakassm. out- (2 sheets of microfiche) i j
REVISION NO. l1
r-Exhibit E NEP-12-02 COMMONWEALTil EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. Cl ATTACllMENT C
\1 RADGEN Input / Output Files from Revisions 0 and 1 RADGEN Input File film "radgen.inp" 0
RADGEN input; Calculation No. 22S-0-110M-0063 Rev. 0 01.334 15 15 0 0.4220.3180.3180.3180.318 0.8000.200 8
0.2000.2000.2000.2000.2000.2000.2000.200 J 1 !
l 1
4 l
l l
1 REVISION NO. I
i Exhibit E NEP-12-02 l
COMMONWEALTH EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. C2 Final ATTACilMENT C RADGEN Input / Output Files from Revisions 0 and 1 I
! RADGEN Output File l
(Microfiche) l File:
"radgen.out" (1 sheet ofmicrofiche) l l
l l
l 1
i l
l l
REVISION NO. I
N EP-12-02 Revision 6 i
COMMONWEALTII EDISON COMPANY 1
l CALCULATION NO. 22S-0-Il0M-0063 PROJECT NO. 10436-003 PAGE NO. D1 l Attachment D l Computation of Fuel llandling Building Ilomogeneous Air Temperature DI PURPOSE / OBJECTIVE The purpose of this attachment is to compute the air temperature of the Fuel Handling Building (FHB) l following a loss of all water in the spent fuel pool. The loss of cooling water is postulated to occur on 2/1/99, roughly two years afler the final reactor shutdown at Zion Station,2/21/97. The FIIB air temperature determined in this attachment is used as the inlet air temperature boundary condition for the COBRA-SFS model described in the main body of this calculation.
D2 METHODOLOGY / ACCEPTANCE CRITERIA Methodology The air temperature in the FHB following a loss of spent fuel pool cooling water is computed using the Sargent & Lundy computer program KITTYlS," Thermal-Hydraulic Steady States in Arbitrary Solid and/or Fluid Channel Configurations", S&L Program No. 03.7.171-5.11. This computer program calculates node temperatures and path heat /enthalpy transfer rates for user-defined configurations under steady-state conditions. The spent fuel pool heat load is modeled as a heat source to the FHB air space.
Enthalpy transport to/from the FHB air space via the HVAC system is modeled by specifying the . mass flow rate of the HVAC inlet flow and the inlet air temperature. Preserving this mass flow rate at the outlet, the KITTYlS program will iterate on the FHB air space temperature until the net energy balance on the FHB is satisfied. Natural convection heat transfer paths from the FHB air space to the vertical interior walls of the FHB are included, as well as convection heat transfer paths from the exterior FHB walls to the ambient. Conduction heat transfer paths through the concrete walls are also included in the KITTYlS model. The FHB concrete walls are ofvarying thickness, with values of 1.5,2.83, and 3.0 feet as shown in the Zion Station FHB Foundation drawings (Reference D5.3). Each vertical wall type (1.5,2.83, or 3.0 feet total thickness) is partitioned into three nodes of equal surface area. The surface nodes facing the FHB air space and the ambient are modeled as 1 inch in thickness. Natural convection heat transfer to/from these surface nodes is modeled, and conduction heat transfer through the central node is modeled to establish the temperature gradient through the concrete wall.
A schematic diagram of the Fuel Handling Building is shown as Figure D1 on the following page.
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l NEP-12-02 l l Revision 6 j l COMMONWEALTII EDISON COMPANY l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. D2 l Figure Di Fuel llandling Building l
Bottom of Roof; Elev. 663' i I
l Elev. 617'
, ---- y e Spent /I
/ Fuel /l ,
/ Pool / 1 l
p_----_q l l l )
l l ll/
l V.------
l/
v l
Elev. 592' North The KITTYlS model conservatively neglects any heat transfer to the concrete floor or ceiling slabs.
Also, radiative heat transfer from the air space to the vertical wall surfaces is neglected.
A schematic diagram of the FHB KITTYlS modelis shown on the following page.
l REVISION NO.1 I
T NEP-12-02 Revision 6 l COMMONWEALTII EDISON COMPANY l
l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. D3 l Figure D2 KITTYIS Model of FIIB Air Temperature l
l Conduction i 3 1.5 feet thick walls
{
Convection Convection HVAC 4 5 6 Outlet Air --
W-, @ @ @
2.83 feet thick walls 1
13 FHB Air :
@ 7 8
9 Space h h h: Ambient a
@ 3.0 feet thick walls i
@ 10 11 12 2 -- !
HVAC h Ov Inlet Air j Nodes i l
O eeths Acceptance CriterL.
The are no specific acceptance criteria applicable to this attachment. The FHB air space temperature computed in this attachment is used as input the COBRA-SFS model.
l REVISION NO.1 l
F NEP-12-02 Revision 6 COMMONWEALTII EDISON COMPANY l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. D4 l l D3.0 ASSUMPTIONS D3.1 The FHB inlet air temperature, via the Auxiliary Building HVAC system, is assumed to be 115 l
F, per Design Input 4.2.1 (see main body of the calculation).
D3.2 The ambient air temperature is assumed to be 95 F. This conservatively bounds the summer design dry bulb temperature of 92 F for Waukegan, IL, per Chapter 24 (Table 1) of the 1993 ASHRAE Handbook, Reference D5.1.
j D3.3 The FHB is assumed to be at a pressure of 14.7 psia. This pressure, as well as the FHB air temperature, is used to compute air density via the Ideal Gas Law. Small deviations in pressure have a negligible effect on the computed density, relative to the large changes in air temperature. ;
Also, the presence of the operational HVAC system is expected to limit pressure changes in the i FHB.
1 D3.4 The total spent fuel pool heat load due to decay heat is conservatively based on a date of 1/1/99, rather than 2/1/99. Reference 5.6 (see main body of the calculation) provides heat load data points for 1/1/99 and 7/1/99. Rather than interpolate between these points for the value at 2/1/99, the conservative date of 1/1/99 is chosen.
l l
l REVISION NO.1 l
I l N EP-12-02 Revision 6 l COMMONWEALTII EDISON COMPANY l
l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. DS l D4.0 DESIGN INPUT l
l D4.1 Dimension for the vertical walls of the Fuel Handling Building are listed below, with appropriate l Zion Station FHB Foundation drawings (Reference D5.3) as cited.
Thickness Length Height Area Reference North Wall i
Elev. 592' to 617', 3.0 feet 71.17 feet 25 'cet 1779.3 ft' Dwg. Nos. B-107, Columns X - W B 12), B-122, B-123 Shield Door at N/A 18.0 feet 20.0 feet -360.0 ft 2
Dwg. No. B-654 Elev. 592' Elev. 617' to 663', 1.5 feet 71.17 feet 46 feet 3273.8 ft 2
Dwg. No. B-420 Columns X - W j 2
l Elev. 617' to 663', 1.5 feet 35.0 feet 46 feet 1610.0 ft Dwg. No. B-421 l Colunms R - X South Wall j 2
Elev. 592' to 617', 3.0 feet 71.17 feet 25 feet 1779.3 ft Dwg. Nos. B-107, l Columns X - W B-121, B-122, B-123 2
Elev. 617' to 663', 1.5 feet 71.17 feet 46 feet 3273.8 ft Dwg. No. B-420 Columns X - W 2
Elev. 617' to 663', 1.5 feet 35.0 feet 46 feet 1610.0 ft Dwg No. B-421 l Columns R - X West Wall 2
l Elev. 592' to 617', 2.83 feet 84 feet 25 feet 2100.0 ft Dwg. Nos. B-107, Columns 17 - 23 B-118, B-119, B-120 2
l Elev. 617' to 663', 2.83 feet 87 feet 46 feet 4002.0 ft Dwg. No. B-420 l Columns 17 - 23 Total Wall Surface Areas 2
i 1.5 feet Wall Thickness 9767.6 fl 2
l 2.83 feet Wall Thickness 6102.0 f1 2
3.0 feet Wall Thickness 3198.6 ft D4.2 The physical properties of concrete are taken from Reference D5.4, and are shown below.
Density 145 lb /ft 8 i Specific Heat 0.156 Btu /lbm- F Thermal Conductivity 0.92 Btu /hr-f1- F l REVISION NO.1 I t
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NEP-12-02 Revision 6 COMMONWEALTil EDISON COMPANY l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. D6 l i
D4.3 The following thermophysical properties of air at atmospheric pressure are taken from Reference D5.2. Additionally, the air propedies cited as Design Input 4.5.1 (see main body of the calculation) are also used in the KITTYlS model.
k 2 2 Temperature gppfp Pr 1 8
( F) (Btu /hr-ft- F) (1/F-R) (N/A) 100 0.0154 1.76 0.72 200 0.0174 0.85 0.72 300 0.0193 0.444 0.71 j 400 0.0212 0.258 0.689 )
500 0.0231 0.159 0.683 I D4.4 The total spent fuel pool heat load due to spent fuel, as a function of calendar date, is taken from Table B1 of Reference 5.6 (see main body of the calculation).
6 Date: January 1,1999 Heat Load: 8.537586 x 10 BTU / hour D4.5 Per Reference 5.20 (see main body of the calculation), the Auxiliary Building HVAC System I supplies 21,000 cfm of air to the Fuel Handling Building Space.
I REVISION NO.1 l l
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NEP-12-02 Revision ti COMMONWEALTil EDISON COMPANY l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. D7 l D
~~5.0 REFERENCES~~
D5.1 1993 ASHRAE Handbook Fundamentals, I-P Edition.
D5.2 Principles of Heat Transfer, Frank Kreith, Third Edition,1976 Impression D5.3 Zion Station Fuel Handling Building Foundation Drawings B-107, Rev. K, FHB Foundation Plan El. 617'-0" West Area B-108, Rev. P, FHB Foundation Plan El. 617'-0" East Area B-109, Rev. J, FHB Foundation Plan El. 602'-0" West Area B-118, Rev. L, FHB Foundation Section A-A B-119, Rev. R, FHB Foundation Section B-B l B-120, Rev. N, FHB Foundation Section C-C B-121, Rev. H, FHB Foundation Section D-D B-122, Rev. E, FHB Foundation Section E-E ,
B-123, Rev. M, FHB Foundation Section F-F l B-420, Rev. A, FHB Wall Plan El. 617'-0" West Area B-421, Rev. A, FHB Wall Plan El. 617'-0" East Area B-654, Rev. G, FHB Ground Fl. Plan El. 592'-0" D5.4 NUREG 0800, Standard Review Plan, Rev.1, Section 6.2.1.5 l
l REVISION NO.1 l l
l NEP-12-02 l Revision 6 l COMMONWEALTII EDISON COMPANY l
l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. D8 l l
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D6.0 CALCULATIONS '
l D6.1 Air Mass Flow Rate of Auxiliary Building HVAC System l To compute the mass flow rate ofinlet air (required by the KITTYlS model), the inlet air density must be determined using the Ideal Gas Law as:
P 1 IIi 1
where P = 14.7 psia, per Assumption D3.3 R = 1545(ft-lb/lbmole- R) / 28.97(Ibm /lbmole) = 53.33(R-lb/lbm- R)
T= 115 + 460 = 575 R, per Assumption D3.1 .
1 Solving for the air density gives the following: l l
2 p = 14.7(psi)
~~144(in ffg2) / [ 53.33(R-lb/lbm- R)~~
575( R) ] = 0.069 lbm/fl3 I The volumetric flow rate of FHB inlet air,21,000 cfm, is convened to a mass flow rate. To conservatively minimize the mass flow rate, the density corresponding to 115 F air is used. l th = 21,000 (R'/ min)
~~0.069 (ibm /ft')~~
~~60(min / hour) = 86940 lbm/ hour This value is used as input to the KITTYlS model for heat transfer paths 1 and 2. ,~~
l 1
D6.2 Volume of Vertical Concrete Walls A required KITTYl S input parameter is the volume of solid nodes. The concrete wall node volumes are determined by multiplying the node thickness by the surface area, for each respective wall type. The surface nodes are 1 inch in thickness, with the central node thickness set to the total minus 2 inches as shown in the table below.
Wall Thickness Area Surface Nodes Central Node Thickness Volume Thickness Volume i
2
! 1.5 feet 9767.6 R 1 inch 814.0 ft' l.33 feet 13020.2 d' I 2 3 2.83 feet 6102.0 R 1 inch 508.5 R 2.67 feet 16274.0 n' 2 3 3.0 feet 3198.6 ft 1 inch 266.6 R 2.83 feet 9052.0 ft' l REVISION NO.1 l l
r NEP-12-02 I Revision 6 l COMMONWEALTII EDISON COMPANY l cal CULATION NO. 22S-0-Il0M-0063 PROJECT NO. 10436-003 PAGE NO. D9 l D6.3 Natural Convection for Vertical Surfaces Reference DS.2 provides the following correlation for turbulent natural convection along vertical planes or cylinders, where the local heat transfer coemcient is nearly constant over the surface:
Fut = = 0.13(Gr t Pr)"'
for Gr > 10' Justification for the use of the turbulent natural convection heat transfer correlation will be provided in Section D7.0.
Substituting for Grt and Pr gives the following expression for the heat transfer coemeient for natural convection to/from the vertical walls of the FHB E, = 0.13 b Pr"' # '
(7; - 7;,)"'
L < p >
with the air properties evaluated at the mean temperature, T,,,, between the air and the wall surface.
The heat transfer coemcient can be simply expressed as:
E, = 0.13k Pr (7;- T,,)"' = C(7;- 7;,)"'
\ p The value of C is computed for a range of air temperature as shown in the table below, with property j data repeated from Design Input D4.3. The temperature-dependent C coemcients are used in the !
KITTYlS model for natural convection at the vertical wall surfaces. !
)
2 2 Temperature k gppfp Pr C = 0.13*k*[(gpp2/ 2)*Pr]"3 l (F) (Btu /hr-fi- F) (1/F-n') (N/A) (Btu /hr-fl2 opo) !
6 100 0.0154 1.76 x 10 0.72 0.217 6
l 200 0.0174 0.85 x 10 0.72 0.192 6
300 0.0193 0.444 x 10 0.71 0.171 400 0.0212 0.258 x 10' O.689 0.155 6
500 0.0231 0.159 x 10 0.683 0.143
[ REVISION NO.1 l
NEP-12-02 Revision 6 ,
COMMONWEALTIl EDISON COMPANY l CALCULATION NO. 22S-0-110M-0063 PROJECT NO. 10436-003 PAGE NO. DIO l 1
D7.0
~~SUMMARY~~
AND CONCLUSIONS j The results of the KITTYlS model of the Fuel Handling Building are included on the following pages.
The model predicts a homogeneous FHB air space temperature of 445.6 F. This value will be conservatively rounded up to 450 F for use in the COBRA-SFS model.
Justification for the use of the turbulent natural convection heat transfer coeflicient for the vertical walls is provided by computing the Grashof number, Gr. The heat transfer correlation provided in Section D6.3 is applicable for Gr > 10' The Grashof number is defined as /gpp'/ 'l
L'
AT. A lower bound on Gr is computed using the smallest value of/gpp'/ f shown in Section D6.3, the smallest height of the vertical walls listed in Design Input D4.1, and the minimum temperature difference between the wall surface and the air. The minimum value of ATis found at the exterior surface of the 3.0 feet thick concrete wall, per examination of the KITTYIS output. This value is 164.7 F - 95 F = 69.7 F. l l
Computing the lower bound value of Gr gives:
Gr = [0.159 x 10'(1/ F-ff)] * [25.0 fl]' * [69.7 F] = 1.73 x 10" This lower bound value of Gr is above the required Grashof number necessary for turbulent natural convection along vertical surfaces.
l REVISION NO.1 l
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(KITTY 1S-51 Varsion 5.1 -
Sargsnt & Lundy Program No. 03.7.171-5.1 lUsor OK0212 on PC5765 Monday, Dsccmbar 21, 1998 Tims: 13:36:24
~~C:ntrolled Files:~~
Driva V: =.SNL2\SYS3: \
l Volum3 in drive V is SYS3 (B203)\OUT1S 04-09-1992 15:06 (B2co)\SRS.WP5 03-16-1992 15:03 (Cacs)\ KITTY 1S.EXE 03-19-1992 15:18 (B2co]\ KITTY 1S.WP5 02-27-1992 16:44 (BacG)\ DATA 1S 10-03-1991 10:25 (Baso)\SDD.WP5 04-09-1992 13:40 (Baco)\ ABSTRACT.WP5 02-27-1992 15:51 (B2ca)\SVVR.WP5 04-09-1992 13:44 Activo Data Directory:
Voluze in drive C is PC5765 Volume Serial Number is 35AB-783F Directory of C:\ WORK \ ZION \ZIRC\REVl\ KITTY
. <DIR> 12-21-98 1:28p
.. <DIR> 12-21-98 1:28p kittyout tmp 6,109 12-21-98 1:35p zion-fhb inp 4,303 12-21-98 1:35p zion-fhb out 20,811 12-21-98 1:35p 5 file (s) 31,223 bytes 381,173,760 bytes free End of Controlled File Information scope l
t l
Calculation. No. 22S-0-110M-0063 Resision 1 Project No. 10436-003 Attachment D Page No. D16 Final
l Exhibit E NEP-12-02 l COMMONWEALTil EDISON COMPANY Revision 6 CALCULATION NO. 22S-0-110M-0063 PKOJECT NO. 10436-003 PAGE NO. El Final ATTACHMENT E COBRA-SFS Input / Output Files from Revision 1 Loss of Spent Fuel Pool Cooling Water on Reference Date of 2/1/99 l
(Microfiche) ;
Files included:
a zirc-rv1. inp" (1 sheet of microfiche)
"zirc-rv1.out" (2 sheets ofmicrofiche) l REVISION NO. I
Attachment 7 l l
Calculation Number 22N-0-110M-0067 Gamma Shine Dose at the Exclusion Area Boundary Resulting from Spent Fuel Pool Dewatering Event December 9,1998

ML20205R024

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