ML20154N223

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CSAR Changed Pages (Revisions) Indicated Relative to Rev 4A of SAR
ML20154N223
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Site: 07201004
Issue date: 10/09/1998
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ML20138L335 List:
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NUDOCS 9810210216
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{{#Wiki_filter:. OV ATTACHMENT C i CSAR Channed Panes (revisions indicated relative to Revision 4A of SAR. Inserted text shown in italic, deleted text shown in strikeout) Table 1.2-2 (Key Design Parameters for the Standardized NUHOMS* System e Section 3.1.1 (Material to be Stored) Table 3.2-1 (Summary of NUHOMS* Component Design Loadings) Section 3.3.7.1 (Cladding Temperature Limits) e Table 7.3-2 (Shielding Analysis Results) e Section 7.4.1 (Operational Dose Assessment) e Section 8.1.l(B) (Design Basis Internal Pressure) e Table 8.1-6 (NUHOMS*-24P DSC Maximum Normal, Off-Normal and e Accident Pressures) -Table 8.1-7 (NUHOMS*-52B DSC Maximum Normal, Off-Normal and e Accident Pressures). Figure 8.1-43 (NUHOMS*-24P DSC Internal Pressure) Figure 8.1-44 (NUHOMS*-52B DSC Internal Pressure) e Section 8.2.7.2 (Accident Analysis) last two paragraphs only e Section 8.2.8.3 (DSC Leakage Accident Dose Calculations) Section 8.2.9.1 (Accident Analysis)

  • ' Section 10.3.1 (Fuel Specifications) 9810210216 981009 PDR ADOCK 0720 4

i I L October 1998 72-1004 Amendment Page C.1 l Revision 2 .R 7 \\ o 11 o W',gdgq'

. - ~ _ ( Table 1.2-2 \\ Key Desien Parameters for the Standardized NUHOMS* System Category Criteria or Parameter Value PWR BWR N Fuel Assembly Criteria: Initial Uranium 475472 198 l Content (kg/ assembly) initial Enrichment 4.0% 4.0% (U-235 equivalent) Fuel Bumup 50,000 6 45,00(E,000 l (MWD /MTU) Gamma Radiation 4.48E15 1.55E15 Source (photons / (10 year cooled) A sec/ assembly) 8407.45E15 2442.63E15 l (5 year cooled) Neutron Radiation 1.55E8 8.40E7 Source (neutron / (10 year cooled)* g seclassembly) 2.23E8 1.01E8 kj\\ + (5 year cooled) Decay Heat Power 1.00 0.37 (kW/ assembly) Dry Shielded Fuel Assemblies 24 52 Canister: per DSC Size: Overall Length 4.72m 4.97m (186.0 in.) (196.0 in.) Outside Diameter 1.71m (67.25 in.) l l ShellThickness 16mm (0.625 in.) Heat Rejection (kW) 24.0 19.2 Intemal Atmosphere Helium Helium (1) Enveloping design basis fuel. l (2) 10 year cooled fuel data provided for information only. l October 1998 72-1004 Amendment Page C.2 Revision 2 l l

. _ _ _ -. _. ~. - _.. _... _.. -. _.. _ _ _. _.. _ _.. _. _ - _.. - _ _ _ _ _ _. _ _ _.. - _ l ( Table 1.2-2 Key Design Parameters for the Standardized NUHOMS* System (continued) Category Criteria or Parameter Value Dry Shielded Maximum Design Conservatively Based on 100% Release Canicter: Pressure of Fill Gas and 30% Release of Fission (Concluded) Gas from 1% of Rods (normal),10% of Rods (off-normal) and 100% of Rods (accident) Equivalent Cask 75g Vertical (End) and Horizontal (Side), Drop Deceleration 25g Oblique (Comer) Materials of Carbon Steel Intemals, Carbon Steel or Construction. Steel Encased Lead Shield Plugs, and Stainless Steel Shell Assembly Service Life 50 Years' fh iD) On-Site Payload Capacity 36,300 kg Transfer Cask: (80,000 lbs.)(dry) 40,900 kg (90,000 lbs.) (wet) Gross Weight 90,700 kg (200,000 lbs.) (handling) 86,200 kg (190,000 lbs.)(transport) Surface Dose Rate ALARA Equivalent Cask 75g Vertical (End) and Horizontal (Side) Drop Deceleration 25g Oblique (Comer) Materials of Carbon Steel, Stainless Steel, Lead, and Construction Neutron Absorbing Material Service Life 50 Years I Expected life is much longer (hundreds of years), however, (,,)i for the purpose of this generic SAR, the service life is taken as 50 years. L October 1998 72-1004 Amendment Page C.3 Revision 2 l l l -.-

.~.....-... -. l-p Table 1.2 2

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Key Design Parameters for the Standardized NUHOMS* System (concluded) Category Criteria or Parameter Value l l Horizontal Capacity One DSC per HSM Storage Module: Array Size Single Module to 2xL Module Array. L may be any value. l-HSM Size: Length PWR: 5.8m (19.0 ft.) BWR: 6.0m (19.8 ft.) i Height 4.6m (15 ft.) l Width 2.9m (9.7 ft.) Surface Dose Rate ALARA Heat Rejection 24.0 kW p) Capacity (5 yr. cooled) 's.J t Heat Removal Natural Circulation l Materials of Reinforced Concrete Construction and Structural Steel [ Service Life 50 years i l l l l l l i l C 4;x October 1998 72-1004 Amendment Page C.4 Revision 2 l l

l l l ' C 3.1.1 Material to be Stored The inventory of PWR fuel types which currently resides in spent fuel pools in the U.S. is sitown i in Figure 3.1-1. B&W 15x15 fuel is selected as the enveloping fuel design for a wide range of PWR fuel types as it is the most reactive and has the most limiting physical characteristics. Table 3.1-1 lists the principal design garameters for the B&W 15x15 fuel selected as the design basis for the standardized NUHOMS -24P system documented in this SAR. Table 3.1 la lists the PWRfuel assembly designs which have currently been demonstrated to be suitablefor storage in the standardized NUHOMS*-24P system provided they meet the requirements of Section 10.3. Similarly, the inventory of BWR fuel types residing in spent fuel pools in the U.S. is shown in Figure 3.1-2. GE 7x7 fuel is selected as the enveloping fuel design for a wide range of BWR fuel types. Table 3.1-2 lists the principal design parameters for the GE 7x7 fuel selected as the design basis for the standardized NUHOMS*-52B system documented in this SAR. Table 3.1-2a lists the BWRfuel designs which have currently been demonstrated to be suitablefor storage in the standardized NUHOMS*-52B system provided they meet the requirements of Section 10.3. The following acceptance criteria is established for BWR and PWR fuels other than the SAR design basis fuels. A. For shielding, the gamma and neutron source strengths and resulting HSM contact roof doses must be less than or equal to the limits set forth by this SAR. B. For therraal, the total decay heat power per DSC and the resulting temperatures must be less , d than or equal to the limits set forth by this SAR. C. For criticality, the initial enrichment and resulting reactivity must be less than or equal to the l limits set forth by this SAR. D. For structural, the fuel weight and the total weight of the DSC and transfer cask must be less thr.n or equal to the limits set forth by this S AR. The operating controls and limits for PWR and BWR fuel qualified for dry storage in the standardized NUHOMS* system are specified in Chapter 10. The parameters for acceptable candidate fuel assemblies for dry storage are described further in the subsections which follow. l l >( t ,L October 1998 72-1004 Amendment Page C.5 l. Revision 2 l n

m l l l lr r' l Table 3.1 la i PWR Fuel Assembly Designs Suitable for Storare Assembly Unirradiated Heavy Unirradiated Length with Assembly Weight w/ Metal Width Length Components Weight Components Weight Cladding l N l Type (in) (in) (in) (Ibs) (Ibs) (kg-U) Material B&W 15x15 8.536 165.75 170.875 1550.0 1682.0 475.0 Zircaloy-4 CE 14x14 Fort Calhoun" 8.100 147.00 152.000 1220.0 1286.5 365.6 Zircaloy-4 CE 15x15 Palisades

  • 8.250 149.00 n/a*

1360.0 n/a 412.4 Zircaloy-4 CE 14x14 Standard /Ger.aric 8.100 157.00 161.000 1270.0 1346.8 382.2 Zircaloy-4 Westinghouse 14x14* 7.763 160.13 164.800 1302.0 1425.0 405.0 Zircaloy-4 Westinghouse 15x15* 8.434 160.10 165.250 1472.0 1637.0 460.0 Zircaloy-4 Westinghouse 17x17" 8.434 160.10 167.220 1482.0 1663.2 461.0 Zircaloy-4 l Limit: 8.536 165.75 171.710 1682.0 1682.0 475.0 l l (1) Each fuel assembly must be qualified for storage per Section 10.3. (2) Includes Exxon /ANF FT. CALHOUN 14 X 14 ANF (3) Includes Exxon /ANF 15x15 CE (4) CE 15x15 Control Components exceed the NUHOMS* dimensional parameters and are not acceptable for storage (5) includes Exxon /ANF 14x14 Westinghouse (6) includes Exxon /ANF 15x15 Westinghouse l (7) includes Babcock and Wilcox WE 17 X 17 B&W Mark BW l (8) Excludes Westinghouse 15x15 reload fuel for B&W 15x15 reactors l !O lO l i l l i i l l 1 l l O NJ October 1998 72-1004 Amendment Page' C.6 Revision 2 l

1 m Table 3.12a (d \\ BWR Fuel Assembly Desiens Suitable for Storage Heavy Channeled Unirradiated Assembly Metal Width Length Weight Weight Cladding Type") (in) (in) (Ibs) (kg U) Material i23 GE 6x6 Dresden-1 4.850 136.00 400 111.4 Zircaloy-2 GE 7x7' ) 5.438 175.87 690 194.9 Zircaloy-2 GE 8x8") 5.440 176.05 690 186.7 Zircaloy-2 Limit: 5.454 176.16 725 198.0 (1) Each fuel assembly must be qualified for storage per Section 10.3. (2) includes Exxon /ANF DRESDEN 16 X 6 ANF (3) includes Exxon /ANF GE BWR 7 X 7 ANF (4) includes Exxon /ANF GE BWR 8 X 8 ANF l I l l I ( ')

c i

l i t l l l l Ln l l October 1998 72-1004 Amendment Page C.7 Revision 2

. ~.. l l I O Table 3.21 ( Summary of NUHOMS' Component Desian Loadinas l. SAR Design Load Section - Component Type Reference Design Parameters Applicable Codes l Horizontal ACI 349-85 and Storage ACl 349R-85 Module: (design) ACI l 318-83 (con-struction only) Design Basis 3.2.1 Max. wind pressure : 397 psf NRC Reg. Guide Tornado Max. speed: 360 1,76 and ANSI A58.11982 DBT Missile 3.2.1 Max speed:126 NUREG-0800, l. Section 3.5.1.4 i Types: Automobile 3967 lb., 8 in, diam. shell 276 lb., j 1 in. diam. solid steel sphere i ('* ( Flood 3.2.2 Maximum water height: 50 feet 10CFR72.122(b) Maximum velocity: 15 ft/sec. l x i Seismic 3.2.3 Hor, ground acceleration: 0.25g NRC Reg. Guides (both directions) 1.60 & 1.61 l Vert, ground acceleration:0.17g Snow and Ice 3.2.4 Maximum load: 110 psf ANSI A58.11982 (included in live loads) l Dead Loads 8.1.1.5 Dead weightincluding loaded ANSI 57.91984 DSC (concrete density of 150 pcf) Normal and 8.1.1.5 DSC with spent fuel rejecting ANSI 57.9-1984 Off normal 24.0 kW of decay heat for 5 yr. Operating cooling time. Ambient air Temperatures temperature range of -40*F to 125 F 4 O b i l October 1998 72-1004 Amendment Page C.8 i Revision 2

Table 3.21 pv) Summary of NUHOMS' Component Deslan Loarlinas -( (continued) SAR Design Load Section Component Type Reference Design Parameters Applicable Codes Accident 8.2.7.2 Same as off-normal conditions ANSI 57.9-1984 Condition with HSM vents blocked for Temperatures 5 days orless Normal 8.1.1.1 Hydraulic ram load of 20,000 lb. ANSI 57.91984 Handling Loads Off-norma! 8.1.1.4 Hydraulic ram load of 80,000 lb ANSI 57.9-1984 Handling Loads Live Loads 8.1.1.5 Design load: 200 psf ANSI 57.91984 (includes snow and ice loads) Fire and 3.3.6 Enveloped by other design 10CFR72.122(c) ( Explosions basis events Dry ' ASME Code, Shielded Section lil, Canister: Subsection NB, I Class 1 Component Flood 3.2.2 Maximum water height: 50 ft. 10CFR72.122(b) Seismic 3.2.2 Horizontal ground acc.: 0.25g NRC Reg. Guides Vertical round acc.: 0.17g 1.60 & 1.61 t Dead Loads 8.1.1.2 Weight of loaded DSC: 80,000 ANSI 57.9-1984 lb. enveloping. Normal and 8.1.1.2 Enveloping intemal pressure of 10CFR72.122(h) Of1-Normal s10.0 psig (normal) ands 41.0 Pressure psig (off-normal) Test Pressure 8.1.1.2 Enveloping intemal pressure of 10CFR72.122(h) 8 psig applied w/o DSC outer top cover plate C( October 1998 72-1004 Amendment Page C.9 Revision 2

Table 3.21 ( Summary of NUHOMS' Component Deslan Loadinas (continued) SAR Design Load Section Component Type Reference Design Parameters Applicable Codes Normal and 8.1.1.2, DSC with spent fuel rejecting ANSI 57.9-1984 Off-normal 8.1.2.2. 24.0 kW (PWR) or 19.2 kW Operating (BWR) decay heat for 5 year Temperature cooling time. Ambient air temperature -40*F to 125*F Normal 8.1.1.2 Hydraulic ram load of ANSI 57.9-1984 Handling 20,000 lb. Loads Off-norma! 8.1.2.1 Hydraulic rnm load of 80,000 lb. ANSI-57.9-1984 Handling Loads Accidental 8.2.5 Equivalent static deceleration 10CFR72.122(b) Cask Drop of 75g for vertical end drop Loads and horizontal side drops, and 25g oblique corner drop Accident 8.2.7 Enveloping intemal 10CFR72.122(h) Internal 8.2.9 pressure of 560604 psig based l Pressure on 100% fuel cladding rupture and fill gas release,30% fission gas release, and ambient air temperature of 125'F Dry AISC Specifi-Shielded cation for Steel Structural Canister Steel Buildings l Support Structure: Dead Weight 8.1.1.4 Loaded DSC plus self weight ANSI-57.9-1984 Seismic 3.2.3 DSC reaction loads with hori-NRC Reg. Guides zontal ground acc. of 0.25g 1.60 & 1.61 j. and vertical ground acc. of l 0.179 I l 4 ~ f i October 1998 72-1004 Amendment Page C.10 I Revision 2-

.m l l l ' p Table 3.2-1 g Summary of NUHOMS' Component Desian Loadinas ~\\ (continuea) SAR Des'j.i Load Section Component Type Reference Design Parameters Applicable Codee Normal 8.1.1.4 DSC reaction loads with ANSI-57.9-1984 Handling hydraul'c ram load of Loads 20,000 lb. Off-normal 8.1.1.4 DSC reaction loads with ANSI-57.9-1984 Handling hydraulic ram load of Loads 80,000 lb On-site") ASME Code Transfer. Section lit, Cask: Subsection NC, Class 2 Component

  • Design Basis 3.2.1 Max. wind pressure: 397 psf NRC Reg. Guide Tornado Wind Max. wind speed: 360 mph 1.76 and i

ANSI 58.1-1982 LO Flood 3.2.2 Not included in design batis 10CFR72.122(b) ! ' d' due to infrequent short duration use of cask Seismic 3.2.3 Horizontal ground acc.: 0.25g NRC Reg. Guides Vertical ground acc.: 0.179 1.60 & 1.61 Snow and Ice 3.2.4 Extemal surface temp. and 10CFR72.122(b) smooth circular section will preclude build-up of snow and l ice loads when cask is in use l (1) The transfer cask is not part of the cask storage system which for NUHOMS' l . consists of the canister and module. l (2) ASME Subsection NCA does not apply. l l I f , r 1 l October 1998 72-1004 Amendment Page C.11 l Revision 2 l

p Table 3.2-1 Summary of NUHOMS' Component Desian Loadinas 'N (continued) SAR Design Loed Section Appilcable Codes Component Type Reference Design Parameters Dead Weight 8.1.1.8 a. Vertical orientation, self ANSI 57.9-1984 weight with loaded DSC and water in cavity: 200,000 lbs. enveloping b. Horizontal orientation self ANSI 57.9-1984 weight with loaded DSC on transfer skid: 200,000 lbs enveloping Normal and 8.1.1.8, Loaded DSC rejecting 24.0 kw ANSI 57.9-1984 . Off-normal 8.1.2.2 decay heat with 5 yr. cooling Operating time. Ambient air temperature Temperatures range: 40*F to 125'F w/ solar shield, 40*F to 100'F w/o solar shield. \\\\ Normal 8.1.1.8 a. Upper lifting trunnions - ANSI N14.6-1993"I Handling in fuel / reactor building: Loads Stresses s yield with 6 x load and s ultimate with 10 x load b. Upper lifting trunnions - ASME Section 111 on-site transfer c. Lower support trunnions: ASME Section til proportional weight of loaded cask during down loading and transit to HSM d. Hydraulic ram load of ANSI 57.91984 40,000 lb. U3 The trunnion design stress allowables are consistent with that of lifting devices governed by N14.6. 's) October 1998 72-1004 Amendment Page C.12 i Revision 2

Table 3.21 O Summarv of NUHOMS* Component Deslan Loadinas (continued) i l l SAR Design Load Section Component Type Reference Design Parameters Applicable Codes Off-normal 8.1.2.1 Hydraulic ram load of ANS! 57.9-1984 Handling 80,000 lb. Loads Accidental 8.2.5 Equivalent static deceleration Cask Drop of 75g for vertical end drops 10CFR72.122(b) Loads and horizontal side drops, and i 25g for oblique comer drop Fire and 3.3.6 Enveloped by other design 10CFR72.122(c) Explosions basis events N/A - DSC provides pressure 10CFR72.122(h) Internal Pressure boundary r i l l l-i' ! \\ October 1998 72-1004 Amendment Page C.13 Revision 2

-. - -..~ _ 3.3.7.1.1 Cladding Temperature Limits Maximum allowable cladding temperature limits are determined for both BWR and PWR design basis fuel according to the methodology presented in Reference 3.21. The maximum allowable average cladding temperature for long term storage is based on the end of life hoop stress in the cladding and the cladding temperature at the beginning of dry storage. The method is estimated to calculate a storage temperature limit that will result in a probability of cladding breach of less than 0.5% in the peak rod during storage. Using this methodology produces cladding temperature limit of 384*C for design basis PWR fuel and 421 C for the design basis BWR fuel - cooled for five years or more. Since the damage mechanism in this methodology is thermal creep, the temperature limits are based on an average long term ambient temperature during storage of 70 F. 384*C (724*F) and 421*C (790*F) are the cladding temperature limits calculatedfor design basis 5-year cooled PWR and BWRfuel, respectively. Three steps were taken to extend the same methodology to the range ofcooling times in the Fuel Quahfication Table shown in Section .10.3.1. First, the same thermal computer models used to perform the design basis cladding temperature calculation.were run parametrically to determine cladding temperature vs. heat inputfor the PWR and BWR baskets. Second, the methodology ofReference 3.21 was used to develop a relationship between the maximum cladding temperature limit vs. cooling times y beyond 5 years. This relationship is shown as afunction offuel burnup in Figure 3.317for PWR fuel and in Figure 3.318for BWRfuel. Third, these twofunctions were combined to obtain {} maximum heat input vs. cooling time. In this way, each cell of the Fuel Quahfication Table has its own unique cladding temperature limit based on the same methodology as was usedfor the design basisfuelassemblies. Higher cladding temperatures may be sustained for brief periods without affecting cladding integrity, however. During short term conditions such as DSC drying, transfer of the DSC to and from the HSM, and off-normal and accident temperature excursions, the maximum fuel cladding temperature is limited to 570 C (1,058 F) or less. This value is based on the results of experiments which have shown that Zircaloy clad rods subjected to short term temperature excur-sions below 760 C did not show indications of failure (3.20). October 1998 72-1004 Amendment Page C.14 Revision 2

V I i ] l \\ Figare 3.317 Maximum PWR Claddins Temocrature Limit vs. Coolins Time t j 720 i x' < 40 GWd/MTU l-700 40-45 GWd/MTU i-45-50 GWd/MTU 1 l-C t \\.. 680 i .h 0 k l-

k.,.

l 's ~. '. 640 i f s

t s s.;...

s s._.,'** s 620 j s..N...... a 600 0 5 -10 15 20 25 30 Cooling Time (years) i l l l 1 1 i i i 1 1 1 ,/ l i I l l October 1998 - 721004 Amendment Page C.15 i i Revision 2 l

... _.... _ _.. _ _.. _.. =.... _ _ _... _.. _ ___.___._.__. _. i l~

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(

f Figure 3.318 )- ' Maximum BWR Claddins Temnerature Limit vs. Coolins Time d i i-1 l-740 < L 'e i 720 ~ .i. l F 7M M0 g 1 660 .\\ i 640 \\ 1 ij-P'/1 i l 1 i 600 l 0 5 to 15 20 25 30 Cooling Time (years) 1 s-P if I i 1 1 2 1 1 - October 1998 72-1004 Amendment Page C.16 1 . Revision 2

Table 7.3 2 p Shieldian Analysis Results for 5 Year Cooled FuelNUHOMS* 24P System l Gamma Dose Rate Neutron Dose Rate (mrom/hr) Total Dose Location (mram/hr.) Primary and Secondary Rate Direct Reflected Direct Reflected (mrem /hr.) DSC in HSM

1. HSM Surface 1.1 HSM Wall or Roof 0.4 (1) 48.2 (1) 48.6 1.2 HSM Front Bird Screen 0.8 15.7 75.2 234.2 325.9 1.3 HSM Roof Bird Screen 0.8 35.2 103.6 427.8 567.4 1.4 Center of Door 28.2 (1) 22.2 (1) 62.3*50,4 l

- (exterior) 1.5 Center of Door 966.1 (1; 1135.2 (1) 2329*a194,3 l Opening (door romoved) i DSC IN CASK

1. Centerline p 0.4 (1) 79.4 (1) 79.8 Shield Plug
2. Top Cover Plate (cavity drained with water in annulus and 3 inches of temporary neutron and

. n' 1* of gamma shielding) 2.1 Centerline 10.2 (1) 31.1 (1) 41.3 2.2 Outer Edge" 8.1 (1) 24.9 349.0 382.0 (1) The reflected dose at these locations is negligible. (2) The DSC/ cask annulus is filled with water and additional neutron shielding material is utilized as required. In addition, all but the top six inches of the DSC inner cavity is assumed to be filled with water for this operation. (3) - The same gap dose rate applies for cases where only the top shield plug is in place. The dose rates reported are with water in the DSC/ cask annulus (however, no water is assumed to be in the DSC). (4)- Represents the greatest totaldose rate calculatedforfuelassemblies suitablefor storage per Table 10.3 3. i l 10 LQ L October 1998 72-1004 Amendment Page C.17 l Revision 2

_.... _.. _. ~. 1 Table 7.3-2 i Shh!!.iss Analysis Results for 5 Year Cooled Fuel NUHOMS*-24P System (Concluded) Gamma Dose Rate Neutron Dose Rate (mrem /hr) Total Dose Location (mrem /hr.) Primary and Secondary Rate Direct Reflected Direct Reflected (mrernhr.) 4 DSC IN CASK (continuedl

3. Transfer Cask j

3.1 Radial 163.9 (1) 427.9 (1) 591.8 1 - a. Surface 75.4 (1) 211.8 (1) 287.2

b. 3 Ft. from Surface 3.2 Top axial 16.8 (1) 4.4 (1) 33 4 % 4,2 f

3.3 Bottom axial 28.3 (1) 39.1 (1) 69 4 % 7,4 j i 9 i i i q-4. t 4 4 ? i i e s 5 i i 1 7 v-October 1998 72-1004 Amendment Page C.18 i I Revision 2 -

__._____.__r 7.4.1 Operational Dose Assessment This S AR section establishes the anticipated cumulative dose exposure to site personnel during the fuel handling and transfer activities associated with utilizing one NUHOMS* HSM for storage of one DSC. Chapter 5 describes in detail the NUHOMS* operational procedures, a l number of which involve potential radiation exposure to personnel. A summary of the operational procedures which result in radiation exposure to personnel is given in Table 7.4-1. The cumulative dose can be calculated by estimating the number of individuals per-forming each task and the amount of time associated with the operation. The resulting man-hour figures can then be multiplied by appropriate dose rates near the transfer cask surface, the exposed DSC top surface, or the HSM front wall. Dose rates can be obtained from the Section 7.3 results of dose rate versus distance from the cask side, DSC top end (with and without the top cover plate and cask lid in place) and HSM front wall. Every operational aspect of the NUHOMS* system, from canister loading through drying, sealing, transport, and transfer is designed to assure that exposure to occupational personnel is as low as reasonably achievable (ALARA).' In addition, many engineered design features are j incorporated into the NOHOMS* system which minimize occupational exposure to plant personnel during placement of fuel in dry storage as well as off-site dose to the nearest neighbor during long-term storage. The resulting dose at the ISFSI site boundary is to be within the limits specified by 10CFR72 and 40CFR190. 3 lV Based on the experience for an operating NUHOMS* system, the occupational dose for placing a L canister of spent fuel into dry storage for the operational steps listed in Table 7.4-1 is less than 1.2&O man-rem. "With the use of effective procedures and experienced ISFSI personnel, the total l l accumulated dose can be reduced further below one man-rem per canister. l. l l l-l !O 1 October 1998 72-1004 Amendment Page C.19 Revision 2

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m. .m... u...... _. m e... _ m r i.,. C O C v..-... m-y..~ ....y .....y m...... v. We "..-- *. y^ '. y -."m '.. . ~ ~. ~.. '..... ~ ". ~ -. ~. '--.;"v_' ;. c "~ *

  • The maximum internal

-~~ '~ ^^ ~ y ~ -~~. ...... ~.. v... pressuresfor the NUHOMS.24P DSCfor normal, off. normal, and accident operating conditions are shown in Table 8.1-6. The maximum DSC internalpressure during normal, off-normal and accident conditions is based onfailure of1%,10%, and 100% of thefuel rodse respectively. 100% of thefuel rodfill gas and 30% of thefission gases are assumed to be released in the DSC cavityfrom thefailedfuel rods. The DSC internalpressurefor normal operating conditions is a maximum of 21.7 psia (7.0 psig)for the seasonal normal operating temperature range ofO T to 100 T. This pressure corresponds to the DSC in the transfer cask at 100 T ambient condition which bounds the DSC in the HSM storage case at 100 T ambient. The DSC internalpressurefor off. normal operating conditions is a maximum of 24.8 psia (10.1 psig) over a seasonal off-normal temperature range of-40 T to 125 T. This pressure corresponds to the DSC in the transfer cask at 100 T ambient condition which bounds the DSC in the HSM storage case at 125 T ambient. The maximum DSC internalpressureforpostulated accident conditions is a maximum of 73.4 psia (58.7 psig) over a seasonal temperature range of-40 T o 125 T. The bounding design basis accident pressure is determined by assuming claddingfailure in 100% of the spentfuel rods stored in the DSC. The claddingfailure is p assumed to release all thefuel rodfill gas and 30% of thefission gas generated in PWRfuel \\ assemblies irradiated to 50,000 MWD /MTU. This postulated worst case condition is included in ( I October 1998 72-1004 Amendment Page C.20 l Revision 2 i

the design basisfor the NUHOMS*-24P DSC as a conservative means ofproviding overpressure p} t protectionfor the DSC containment boundary. For normal operating conditions, the gas inside the DSCis assumed to be brought to thermodynamic equilibrium with the maximum normal ambient tc merature of1007. These calculations are repeatedfor the NUHOMS* S2B DSC with the results listed in Table 8.1-7. Similar assumptions are madefor of normal and accident conditions to determine the DSC internalpressure occurring with a gas temperature at an extreme ambient temperature of12S T. The effects ofpostulated accident pressures are c': scribed in Section 8.2. l O (- r . October 1998 72-1004 Amendment Page C.21 Revision 2 -

i Table 8.1-6 NUHOMS'.24P DSC Maximum Normal. Off-Normal and Accident Pressures Operating Limiting Case Max DSC Max TotalDSC Design Basis Condition Description Pressurf' (psia) Pressure (psia) Pressure (psia) Normal DSC in Cask, 21.4

21. Y

24.7 1009 Of1-Normal DSC in Cask, 21.4 24.8 55.7 100*F Accident Blocked HSM 26.8 73.4 74.7 vents,12ST (1) Normal operating totalpressure with 1% of fuel rod cladding failure. (2) Off-normal operating totalpressure with 10% of fuel rod cladding failure (3) Enveloping accident totalpressure with 100% of fuel rod cladding failure. j (4) Total DSC Intemalpressure without any fuel rod cladding failure. 1 4

  • p I

f3 V October 1998 72-1004 Amendment Page C.22 Revision 2

..... ~. - -. -,. _ -. -. f Table 8.17 \\ NUHOMS* 52B DSC Maximum Normal. Off-Normal and Accident Pressures Operating Limiting Case Max DSC Max TotalDSC Design Basis Condition Description Pressure * (psia) Pressure (psia) Pressure (psia) Normal DSC in Cask, 21.0 21.1"' 24.7 100T Off-Normal DSC in Cask, 21.0 22.8 55.7 100*F Accident Blocked HSM 24.7 45.d 74.7 vents.125*F (1) Normal operating totalpressure with 1% of fuel rod cladding failure. (2) Off-normal operating totalpressure with 10% of fuel rod cladding failure (3) Enveloping accident totalpressure with 100% of fuel rod cladding failure. l (4) Total DSC intemalpressure without any fuel rod cladding failure. l 1< 1 1 ( LU ' October 1998 72-1004 Amendment Page C.23 Revision 2 l

- - -... ~ - ... ~. ~,... -. _... -........ _ 1 t -{N .b 50 - 40 - l _ j o .N 20 - 6 10 I 0 i Normal off. Normal Accident Operating Condition I-4-Without Rod Failure o-With Rod Failure Design Basis Pressures l Figure 8.143 i NUHOMS*-24P DSC Internal Pressure b " v Cctot.r1998 72-1004 Amendment Page C.24 - Revision '.

m. _..... - s._._._.... I~ 60 m 50 - 40 - 1 a p 30 - 1 I N t a i e 'N I l 20 -- 1 O Normal Off Normal Accident Operating Condition l-*-Without Rod Failure -o-With Rod Failure Design Basis Pressures l Figure 8.144 ~; f NUHOMS*-52B DSC Internal Pressure U October 1998 72-1004 Amendment Page C.25

Revision 2

l l 8.2.7.2 Accident Analysis (last two paragraphs only) l i These temperatures are below the levels that safety impairing damage would occur to the HSM or DSC. The short time exposure of the DSC and the spent fuel assemblies to the elevated temperatures will not cause any damage or result in the release of radioactivity. The maximum DSC internalpressure during this event is 12.1 psigfor PWRfuel and 10.0 psigfor BWRfuel(assuming L that nofission andfuel rodfill gas is released). If thefission andfill gases are releasedfrom 100% l. of the spentfuel rodsfrom all assemblies, the DSC internalpressure would be 58.7 psigfor PWR l fuel and 31.1 psigfor BWRfuel assuming the 30% offission gases and 100% offuel rodfill gas is l released. The design basis pressure consideredfor this accident analysis of the DSC is 60 psig. I The thermal-induced stresses for the blocked vent case are calculated using the HSM structural i models shown in Figure 8.1-23 and 8.1-24 as discussed in Section 8.1.1.5, paragraphs C and E. l The non-linear transient thermal gradients are converted to equivalent linear gradients in l-accordance 'with the guidelines of ACI 349 Appendix A. The worst case equivalent linear l. thermal gradients are then applied to the Figure 8.1-23 and 8.1-24 computer models to calculate l the elastic forces and mornents induced. The resulting elastic forces and moments are modified to account for the concrete cracked section properties in accordance with ACI 349 Appendix A, and combined with the calculated forces and moments from other loads, i s l f ( October 1998 72-1004 Amendment Page C.26 Revision 2

b 8.2.8.3 Accident Dose Calculations k The postulated accident assumes that one DSC is ruptured and that all the spent fuel rod cladding fails simultaneously such that 35 ef the fission gasses in the spent fuel assemblies, (m&lyincluding 30% of the Kr-85,10% of the noble gases other than Kr-85 (primarily H 3), and 10% of the radioactive iodine are instantaneously released to the atmosphere. The whole body dose and skin dose at 1000 feet from the storage site under the worst meteorologic conditions were calculated and are listed in Table 8.2-10. From this table it is seen that the resultant accident dose is well within the 10CFR72.106 limit, which restricts the maximum whole body or organ dose beyond the owner controlled area from any design basis accident to be less than five rem. Doses (on-site and off site) must be assessed on a site specific basis by the licensee. Table 8.2-10 shows that typical site boundary doses are below the 10CFR72 limits. o O l l i l l-i l October 1998 ' 72-1004 Amendment Page C.27 . Revision 2

l 8.2.9.1 Accident Analysis The bounding intemal pressurization of the DSC for this conservative generic evaluation is postulated to result from cladding failure of the spent fuel, and the consequent release of spent fuel rod fill gas and free fission gas. Fission gas release fractions are not so easily estimated however. A recent report on ISFSI facility evaluations (8.43) uses a release fraction of 8% as a nominal case and 30% as an upper bound. For design purposes, and as a means of providing over pressure protection for the DSC, it is conservatively assumed that all fuel rods in the DSC suffer cladding failure, as discussed in Section 8.1.1.1, paragraph B. It is further assumed that the fission gas release fraction is 30%. For the i PWR (BWR: noted parenthetically) DSC, this results in release of 40;00013,050 in.3 (BWR: 4700 l 3,5") in ?) of fission gas per assembly (interpolated from the data provided in Reference 8.42)

==-! g the cdgina' fu:! red F.!! g= p== = i: 180 p:!g (B?'R: 30 p !g)..^.: d:en 1: Td!: ~ 8.! 6 =d Figu= 8. ! 13 (BV'R: Td!: 8.1 =d Figu= 8.! Ai), th:==!!!=g DSC p==== !: 37 A p:!g (BV'R: 19.2 p;ig) "t: $: cc::!& :=S!=: 2!r ::=p:=:u= i; 125 F The limiting postulated accident for DSC pressurization is the HSM blocked vent conditionfor 5 days. As shown n Table 8.1-6 and Figure 8.1-43 (BWR: Table 8.1-7 and Figure 8.1-44) $: h:!!um g= ::=p=== i=ide $: DSC ri=c :c 667 F (BV'R: 719 S pred::ing thea maximum DSC intemal pressure with the i fission andfill gas release is -ef-60.0-58.7 psig (BWR: M:931.1 psig). The stress analysis of the DSC shell assembly for an intemal pressure of 50-060.0 psig is described in Section 8.1.1.1.-The

==imum DSC de!! =:=br=: ::== in:=2y =!=! ::d i 2.7 k:i. 1 ( l I I l-t ( October 1998 72-1004 Amendment Page C.28 i Revision 2 l -c + T r 7 7 -+ v' w v'

. [] Table 8.210 l Dose at 300m from ISFSI Site Due to Ke-86-Release from Postulated DSC Ruoture ) 1 l Dose Type X /OW (Sec/m ) Dose (Rem) Whole Body 5.0E-3 0.2264E-2 Skin 5.0E-3 6.7 (1) Taken from Reg. Guide 1.4 Ground level Release Data. 1 l October 1998 72-1004 Amendment Page C.29 Revision 2

10.3.1 Fuel Specifications Limit / Specification: The characteristics of the spent fuel which is allowed to be stored in the standardized NUHOMS' system are limited by those included in Tables 10.3-1 and 10.3-2, and Figure 10.31. l Applicability: The specification is applicable to all fuel to be stored in the standardized NUHOMS* system. Objective: The specification is prepared to ensure that the peak fuel rod temperatures, maximum surface doses, and nuclear criticality effective neutron multiplication factor are below the design valuestimits. Furthermore, the l fuel weight and type ensures that structural conditions in the SAR bound those of the actual fuel being stored. Action: Each spent fuel assembly to be loaded into a DSC shall have the parameters listed in Tables 10.3-1 and Figure 10.3-1for PWRfuel, endor Table 10.3 2 for BWRfuel verified and documented. Fuel not meeting this specification shall not be stored in the standardized NUHOMS* system. Surveillance: Immediately, before insertion of a spent fuel assembly into an DSC, the identity of each fuel assembly shall be independently verified and documented. Bases: The specification is based on consideration of the design basis parameters included in the SAR and limitations imposed as a result of the staff review. Such parameters stem from the type of fuel analyzed, structural limitations, criteria for criticality safety, criteria for heat removal, and criteria for radiological protection. The standardized NUHOMS' system is designed for dry, horizontal storage of irradiated light water reactor (LWR) fuel. The principal design parameters of the fuel to be stored can accommodate standard PWR and BWR fuel designs as listed in Chapter 3. The analyses l presented in the SAR are based on non-consolidated, zircaloy-clad fuel with no known or suspected gross breaches. (See Tables 10.3-1 and 10.3-2.) L The physical parameters that def'me the mechanical and structural design of the HSM and the DSC are the fuel assembly dimensions and weight. The calculated stresses are based on the physical parameters given in Tables 10.3-1 and 10.3-2 and represent the upper bound. l The design basis for nuclear criticality safety is based on the standard L Babcock & Wilcox 15x15/208 pin fuel assemblies with initial enrichments up to 4.0 wt. % U-235, and General Electric 7x7 fuel assemblies with initial enrichments up*to 4.0 wt. % U-235, for the standardized NUHOMS*-24P and NUHOMS -52B designs, respectively. The DSC is designed to permit storage of irradiated fuel such that the irradiated fuel reactivity is less than or October 1998 721004 Amendment Page C.30 Revision 2

equal to 1.45 wt. % U-235 equivalent unirradiated fuel for the NUHOMS*- 24P design (as defined in Figure 10.3-1), and less than or equal to 4.0 wt. % l ,\\ U-235 initial enrichment fuel for the NUHOMS*-52B design. The thermal design criterion of the fuel to be stored is that the maximum heat generation rate per assembly be such that the fuel cladding temperature is maintained within established limits during normal and off-normal conditions. Fuel cladding temperature limits were established based on methodology in PNL-6189 and PNL-4835 (References 10.5.1,10.5.2). B:=d c $i: ::Sede!cgy, i:==i=== '.=: g==::10: =:: cf ! k"' p=

.....,.. :... u.... _2..: _.._. i... c_.. u... n u r o.. c.... _ u.... _ __ a., __._..a.

n..,,. i., u. r

___ u p====bly in : Sc= ding v.' : f= $: B"'R ft:! ;c ': ::c=d. The radiological design criterion is that the g== =d :::===== ... _ _ _.u.. c. u.. a.._._ a. c... .- _ _. u......... u_ u..._ 2 _ a. u.,...._..__ _ r. u...

ten =d g==: =y=== :t=gis t=d by 1: v=d= i- $:

Wi:! ding = 'y i:fuelstored in the NUHOMS* system must not increase the average calculated HSM or transfer cask surface dose rates beyond those calculatedfor a canisterfull ofdesign basisfuel assemblies.-The d=ign bri:===: ::==g$ =: driv:d f== bu=up = 'yi f= (!) P"'R ft:! wit d.0 w:!ght pe==: U 235 !-!:!:! :=id :::, I- :di:::d b.e

==!= = cf 10,^^^ M"/D/"TU, =d : p=: !=:dia:!c :!=: cf 'iv: y:=:; m_, o..u..r.o. c.... :. u. a..n.... : _ u...,........ n m_... :, _ _ _: t _. __2 s g' ir:dia::d :: :==!== cf 35,^^^ ""!D.^4TU, =d a p=: ir:dia:i= t'.: 0f 5 y==. The design value average HSM and cask surface dose rates were calculated to be 48.6 mrem /hr and 591.8 mrem /hr, respectively, for twentyfour (24) Babcock and Wilcox 15x15 PWR assemblies with 4.0 wt.% U-235 initial enrichment, irradiated to 40,000 mwd /MTU, and having a post irradiation time offive years. i 1 1 0 .t October 1998 72-1004 Amendment Page C.31 Revision 2

-~ .-.~- -. Table 10.31 PWR Fuel Snecif1 cations of Fuel to be Stored in the Standardized NUHOMS* 24P DSC Title or Parameter Spc.:ifications Fuel Only intact, unconsolidated PWR fuel assemblies, with or without control components, with the following requirements Physical Parameters Fuel Design See SAR Table 3.1-la l Assembly Length See SAR Chapter 3 Nominal Cross-Sectional Envelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter 3 No. of Assemblies per DSC s 24 intact assemblies Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches S.=m ' Ch==:==::= D=:y !k:: P := p= F :! ^.=:=5!y c !.0 h"' ($!: ='= i:==::: '= =y g!== ==r5!y, ::d =:y ::: 5:== ged f= '! 21

== r '!!=' ".:di:!:;5='. Ch==::i:: ", i E== p c 10,^J^ "!D/TU P=: h;di:S T :: 15y== 'A= :: !:&!Enih =: c 4.0; e U 235 [ =Imur ! ':::' Unni = C ::=' . m..t,'--...'.'.,.. =::: ' :&! 'L;==!=: Erih =' c 345. 5 U 2M 'k :::: S:== P= ^.=:25!y c 3,3333 7jn; ; :g 37=:;= g;=g3 3 $;;:_ 7 r'L - a. 'T m r O A D C;r = Sc== P= ^.=:r5!y c ? tSE!! ph::crb$b': [pN: ;m 5:=&d by $2: H Ch:;:=cfS.^R Nuclear Parameters FuelInitial Enrichment $4.0 wt. % U-233 Fuel Burnup and Cooling Time Per Table 10.3 3 Alternate Nuclear Parameters initial Enrichment $4.0 wt. % U-233 Burnup S40,000 mwd /MTU andper Figure 10.31 Decay Heat s1.0 kWper assembly Neutron Source s2.23 x id n/sec per assy with spectrum bounded by that in Chapter 7 ofSAR Gamma Source 17.45 x 10 g/sec per assy with spectrum bounded by that in Chapter 7 ofSAR s October 1998 72-1004 Amendment Page C.32 Revision 2

1 l l f3 (V) - Table 10.3 2 BWR Fuel Specifications of Fuel to be Stored in the Standardized NUHONIS* 52B DSC l Title or Parameter Specifications Fuct Only intact, unconsolidated BWR fuel assemblies with the following requirements Physical Parameters Fuel Design See SAR Table 3.12a l Assembly Length See SAR Chapter 3 Nominal Cross-Sectional Envelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter 3 (w/ fuel channels) No. of Assemblies per DSC s 52 intact channeled assemblies Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches S.==:! Ch==::i:!= ^:=y !!=: " c. = p= Fu^! ^.=:=S!y c 0.37 k?'(:hi: v:!= i:==imur for =y give

=emt!y, =d
y ::: be :v=:ged for :!! 52

== r SM =) Radie!:g!=! Ch==::it:= Bu= p 5 35,^^^ "WD/'rU P=: h:di-:!: T+me 2 5 y:=:.

u.. : _.. _, _ :. : _ i, _.. : _ r.._.._.:

. t. _..._._... _ 4..n,,_, y,,,,,, n e c.., e. n.,,,,,,. t__, a.._. _ _., _ _... " ::: r ' ::!:' U=:i = C=:='

5
=5= p!::=)

s M:::r= S ;re: P= ^.:==b!y c :no E;f== rg f !.0!ES d=: ; "5 :p=:== be :&d by 'h:: - C:::: Sc== P= ^.:=r5!y

c. t.. r_. _ _- cein-G 2.53E!5 ph:::d=: ; '5 :p=:== 5:=&d by th::

5 Ch:,-:= ? :f S.^ R Ntulear Parameters FuelInitiallattice Enrichment $4.0 wt. % U-235 Fuel Burnup and Cooling Time Per Table 10.3-4 Alternate Nuclear Parameters initial Enrichment $4.0 wt. % U-235 Burnup $35,000 mwd /MTU Decay Heat $0.'7 kWper assembly Neutron Source $1.01 x 10' n/sec per assy with spectrum bounded by that in Chapter 7 ofSAR Gamma Source $2.63 x 10" g/sec per assy with spectrum bounded by that in Chapter 7 ofSAR l (G October 1998 72-1004 Amendment Page C.33 Revision 2

Table 10.3 3 PWR Fuel Oualification Tabte for the Standardized NUllOAIS* 24P DSC (Alinimum required years of cooling time after reactor core discharge) Burnup (GWd/ Initial Enrichment (w/o U-235) MTlHM) 2.02.12.22.32.42.52.62.72.82.93.03.13.23.33.43.53.63.73.83.94.0 to 15 ,,,,5 Not Acceptable 20 1 5 per 25 g 5 Figure 10.3-1 28 a in 30 3 W3 32 f u 34 3 e; 36 M O 38 W sr._=_ 40 W 6%ADWh%InimE4MM1%*l56%dll 8 l 7 l 7 l 7 l 6 l 6 l 6 l 6 l 6 l 6 l 41 A4, tfaj$MR662$l Mitt Mlf8 08l8l8l8l8l8l6l6l6l6l 42 . k.-- ~~iir6n- - 9l9l9l9l8l8l8l8l8l 43 e + &D*5W Not Acceptable ,9 rrmi 10 l 10 l 9 l 9 l 9 l 9 l 9 l 9 l 9 l 44 a tml%d5 titel*@5nWiCSS 10 l 10 l 10 l 10 l 10110 l 9 l 9 I ^""

  • 45

?C *!Mt.P"?N 3'f9R IniBNRREBl312 l 11 l 11 l 11 l 11 l 10 l 10 l 10 l 46 ?18' Tr#1RefBW 14 'M 0f* f!WPtRl@t@,M 13 l 13 l 12 l 12 l 12 l 12 l 12 l 5 Notes: Use burnup and enrichment to lookup minimum cooling time in years. Licensee is responsiblefor ensuring that uncertainties infuel enrichment and burnup are correctly accountedfor duringfuel quahfication. Round burnup UP to next higher entry, round enrichments DOWN to next lower entry. Fuel with an initial enrichment less than 2.0 w/o U-235 must be quahfiedfor storage using the alternate nuclearparameters specified in Table 10.3-1. Fuel with an initial enrichment greater than 4.0 w/o U-235 is unacceptablefor storage. Fuel with a burnup greater than 50 GWd/AfTIHM is unacceptablefor storage. Fuel with a burnup less than 15 GWd/MTIHM must be quahfiedfor storage using the alternate nuclear parameters specified in Table 10.31. Example: An assembly with an initial enrichment of 3.65 w/o U-235 and a burnup of 42.5 GWd/MTlHM is acceptablefor storage after a nine-year cooling time as defined at the intersection of 3.6 w/o U-235 (rounding down) and 43 GWd/MTlHM (rounding up) on the quahfication table. October 1998 72-1004 Amendment Page C.34 Revision 2

m Table 10.3-I \\ BWR Fuel Oualification Tabte for the Standardized NUHOMS%528 DSC t ] (Minimum required years of cooling time after reactor core discharge) Burnup (GWd/ Initial Enrichment (w/o U-235) MTlHM) 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 15 3 3 3l3l3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 20 5 5 5l5l5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 25 5 5 5l5l5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 30 44 ef h*4l 5 l 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 32 "~ ' ~ ~ ~ 6 6 6 5 5 5 5 5 5 5 5 5 5 5 5 5 5 ~" 34 L 8 8 6 6 6 6 6 6 6 6 6 6 6 6 6 5 35 m PNt rR M Ib 9 9 9 9 9 8 8 8 6 6 6 6 6 6 5 36 >4A (M W4_ 10 10 10 10 10 9 9 9 9 9 9 9 8 8 8 37 @ 6atid WP.teSW 4 e% 12 11 11 11 11 10 10 10 10 10 10 9 9 9 38 gj Not cceptable t 14 14 14 13 13 13 12 11 11 11 11 11 10 10 39 @t / M > 16 16 15 15 15 14 14 14 14 13 13 13 13 11 40 h l Not A lyzed 21 20 20 20 19 17 17 17 16 16 16 15 N O I E 19 17 17 17 16 16 16 15 15 15 15 14 14 42 Se t, _ ":" Whi 3RS 11tSt 44 De sW c@lSA $ WW S M6 23 22 22 21 21 21 20 20 19 19 17 17 45 M BM M W 3 C 23 23l 22 22 21 21 21 20 20 20 19 ~ Notes: O) Use burnup and enrichment to lookup required cooling time in years. Licensee is e responsiblefor ensuring that uncertainties in fuel enrichment and burnup are correctly accountedfor duringfuel qualification. Round burnup UP to next higher entry, round enrichments DOWN to next lower entry. Fuel with an initial enrichment less than 2.0 w/o U 235 must be quahfiedfor storage using the alternate nuclearparameters specified in Table 10.3 2. Fuel with an initial enrichment greater than 4.0 w/o U-235 is unacceptablefor storage. Fuel with a burnup greater than 45 GWd/hfTlHhi is unacceptablefor storage. Fuel with a burnup less than 15 GWd/MTlHM is acceptable after three years cooling time provided the physicalparametersfrom Table 10.3-2 have been met. Example: An assembly with an initial enrichment of 3.05 w/o U-235 and a burnup of 34.5 GWd/MTIHM is acceptablefor storage after a nine-year cooling time as defined at the intersection of 3.0 w/o U-235 (rounding down) and 35 GWd/MTlHM (rounding up) on the quahfication table. O i.] October 1998 72-1004 Amendment Page C.35 Revision 2

1] O 60 55 - UNACCEPTABLE i i a 50 i l 45 - 40 - S

  • 35 -

E o @ 30 - QU ALIFIED. [ (Equiv. Enr. < 1.45 w/o) $ 25 - = 3 m 20 - J l 15 - NOT QU ALIFIED 10 - (Equiv. Enr. > 1.45 w/o) 5-0 1.40 1.80 2.20 2.60 3.00 3.40 3.80 INITIAL ENRICHMENT (w/o U 235) Figure 10.31 PWR Fuel Criticality Acceptance Curve October 1998 72-1904 Amendment Page C.36 Revision 2 ' a ,w w-

ATTACHMENT D Certificate of Compliance Suenested Changed Panes (revision bars represer.t changes relative to Revision 1 of this amendment application) Section 1.2.1 (Entire Section) C \\ October 1998 72-1004 Amendment Page D.1 - Revision 2

.m l l- ( l.2.1 Fuel Specifications i l Limit / Specification: The characteristics of the spent fuel which is allowed to be stored in the I standardized NUHOMS' system are limited by those included in Tables 1-la and 1-lb. i Applicability: The specification is applicable to all fuel to be stored in the standardized NUHOMS' system. Objective: The specification is prepared to ensure that the peak fuel rod cladding temperatures, maximum surface doses, and nuclear criticality effective neutron multiplication factor are below the design limits. Furthermore, the fuel weight and type ensures that structural conditions in the SAR bound those of the actual fuel being stored. Action: Each spent fuel assembly to be loaded into a DSC shall have the parameters listed in Tables 1-la and 1-lb verified and documented. Fuel not meeting this specification shall not be stored in the standardized NUHOMS system. Surveillance: Immediately, before insertion of a spent fuel assembly into a DSC, the identity of each fuel assembly shall be independently verified and documented. ( Bases: The specification is based on consideration of the design basis parameters ~b included in the SAR and limitations imposed as a result of the staff review. Such parameters stem from the type of fuel analyzed, structural limitations, criteria for criticality safety, criteria for heat removal, and criteria for radiolo' ical protection. The standardized NUHOMS* system g is designed for dry, horizontal storage of irradiated light water reactor (LWR) fuel. The principal design parameters of the fuel to be stored can accommodate standard PWR fuel designs manufactured by Babcock and Wilcox, Combustion Engineering, and Westinghouse, and standard BWR fuel manufactured by General Electric. The system is limited for use to these standard designs and to equivalent designs by other manufacturers as listed in Chapter 3 of the SAR. The analyses pre:,ented in the SAR are based on non-consolidated, zircaloy-clad fuel with no known or suspected gross breaches. The physical parameters that define the mechanical and structural design of the HSM and DSC are the fuel assembly dimensions and weight. The calculated stresses gi_ven in the S AR are based on the physical parameters given in Tables 1-la and 1-lb and represent the upper bound. The design basis fuel asenblies for nuclear criticality safety are Babcock and Wilcox 15x15/208 pin fuel assemblies and General Electric 7x7 fuel L. assemblies for the standardized NUHOMS*-24P and NUHOMS*-52B October 1998 72-1004 Amendment Page D.2 Revision 2

1 ( designs, respectively. The NUHOMS*-24P is designed for unirradiated ~\\ fuel with an initial fuel enrichment of up to 4.0 wt. % U-235, taking credit for soluble boron in the DSC cavity water. Section 10.3.15 defines the requirements for boron concentration in the DSC cavity water for the NUHOMS*-24P design only. The NUROMS*-24P DSC has the additional requirement that the assemblies have an equivalent unirradiated enrichment ofless than 1.45 wt. % U-235. Figure 1.1 defines the required burnup as a function ofinitial enrichment to produce an equivalent enrichment less than 1.45 wt. % U-235. The NUHOMS*-52B is designed for unirradiated fuel with an initial enrichment of up to 4.0 wt. % U-235. i The thermal design criterion of the fuel to be stored is that the maximum heat generation rate per assembly be such that the fuel cladding temperature is maintained within established limits during normal and off-normal conditions. Fuel cladding temperature limits were established based on methodology in PNL-6189 and PNL-4835. The radiological design criterion is that fuel stored in the NUHOMS* system must not increase the average calculated HSM or transfer cask surface dose rates beyond those calculated for a canister full of design basis fuel assemblies. The design value average HSM and cask surface p dose rates were calculated to be 48.6 mrem /hr and 591.8 mrem /hr Q respectively for twenty four (24) Babcock and Wilcox 15x15 PWR assemblies with 4.0 wt. % U-235 initial enrichment, irradiated to 40,000 mwd /MTU, and having a post irradiation time of five years. l l l l !O I' h ' October 1998 72-1004 Amendment Page D.3 . Revision 2

t I l /' Table 1 la ( PWR Fuel Specifications for Fuel to be Stored in the Standardized NUHOMS*-24P DSC Title or Parameter Specifications. Fuel Only intact, unconsolidated PWR fuel j assemblies, with or without control components, with the following requirements 4 Physical Parameters Fuel Design See SAR Chapter 3 Assembly Length See SAR Chapter 3 Nominal Cross-Sectional Envelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter 3 No. of Assemblies per DSC s 24 intact assemblies L Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches l Nuclear Parameters FuelInitial Enrichment s 4.0 wt. % U-235 Fuel Bumup and Cooling Time Per Table 1-2a Alternate Nuclear Parameters Initial Enrichment s 4.0 wt. % U-235 Burnup s 40,000 mwd /MTU and Per Figure 1.1 Decay Heat s1.0 kW per assembly l Neutron Source s 2.23 x 10 n/sec per assy with spectrum 8 bounded by that in Chapter 7 of SAR Gamma Source s 7.45 x 10 g/sec per assy with spectmm 15 bounded by that in Chapter 7 of SAR 1 l 1 1 1 ,iQ October 1998 ' 72-1004 Amendment Page D.4 . Revision 2

i ,O Table 1-lb BWR Fuel Specifications of Fuel to be Stored in the Standardized NUHOMS*-52B DSC l Title or Parameter Specifications Fuel Only intact, unconsolidated BWR fuel assemblies with the following requirements Physical Parameters Fuel Design See SAR Chapter 3 Assembly Length See SAR Chapter 3 Nominal Cross-Sectional Envelope See SAR Chapter 3 Maximum Assembly Weight See SAR Chapter 3 No. of Assemblies per DSC s 52 intact channeled assemblies Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding breaches Nuclear Parameters FuelInitial Lattice Endchment s 4.0 wt. % U-235 Fuel Burnup and Cooling Time Per Table 1-2b Alternate Nuclear Parameters G Initial Enrichment 5 4.0 wt. % U-235 ). Burnup s 35,000 mwd /MTU Decay Heat 50.37 kW per assembly l Neutron Source s 1.01 x 10 n/sec per assy with spectrum 8 bounded by that in Chapter 7 of SAR Gamma Source s 2.63 x 10" g/sec per assy with spectrum bounded by that in Chapter 7 of SAR 1 i 1 [ } l C/ October 1998 72-1004 Amendment Page D.5 Revision 2 i

I (" Table 12a PWR Fuel Oualification Table for the Standardized NUHOMS*.24P DSC (Minimum required years of cooling time after reactor core discharge) Bumup (GWd/ Initial Enrichment (w/o U 235) MTlHM) 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 10 15 5 5 Not Acceptable 20 5 5 5 per 25 fQ 5 5 Figure 1.1 28 3 y !? 30 M f^ W 32 E F 34 3 3 36 SM L:~ 2 38 3 Y e,, _ - 40 44 m WAFae*lvsTwJ4*dW11Tl*>l 8 I 7 I 7 l 7 l 6 l 6 l 6 l 6 l 6 l 6 l 41 @'N cc# T @ W R8EF&f4 FN fmlWGlul 8l8l8l8l8l8l6l6l6l6l

  1. Mh ' Not Acceptable ~ "'9hidlT6let h Wl 9 l 9 l 9 l 9 l 8 l 8 l 8 l 8 l 8 l 42

^!D dd 43 9.ii a4H ggWlWlWfl 10 l 10 l 9 l 9 l 9 l 9 l 9 l 9 l 9 l 44 we en M @lMltalml HI 10 l 10 l 10 l 10 l 10 l 10 l 9 l 9 l 45 ab W Not Analyzed 1 ' 46 Y Y t4 ( 47 3 M ~ ( 48 ' { ^ 49 1 3 4 50 W m2 n. g Notes: Use burnup and enrichment to lookup minimum cooling time in years. Licensee is responsible for ensuring that uncertainties in fuel enrichment and burnup are correctly accounted for during fuel qualification. Round burnup UP to next higher entry, round enrichments DOWN to next lower entry. Fuel with an initial enrichment less than 2.0 w/o U-235 must be qualified for storage using the alternate nuclear parameters specified in Table 1-la. Fuel with an initial enrichment greater than 4.0 w/o U-235 is unacceptable for storage. Fuel with a burnup greater than 50 GWd/MTIHM is unacceptable for storage. Fuel with a l burnup less than 15 GWd/MTIHM must be qualified for storage using the alternate nuclear parameters specified in Table 1-la. Example: An assembly with an initial enrichment of 3.65 w/o U-235 and a burnup of 42.5 GWd/MTIHM is acceptable for storage after a nine-year cooling time as defined at the l intersection of 3.6 w/o U-235 (rounding down) and 43 GWd/MTIHM (rounding up) on the qualification table. 1 October 1998 72-1004 Amendment Page D.6 Revision 2

i i Table 12b 4 { BWR Fuel Oualification Table for the Standardized NUHOMS*-52B DSC l (Minimum required years of cooling time after reactor core discharge) i. f Burnup j (GWd/ Initial Enrichment (w/o U 235) MTlHM) 2.0 2.1 2.2 2.3 2.4 2.5.2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 15 3l3l3l3l3l3l3l3l3l3 3 3 3 3 3 3 3 3 3 3 3 20-5l5l5l5l5l5l5l5l5l5 5 5 5 5 5 5 5 5 5 5 5 ~ 25 5l5l5l5l5l5l5l5l5l5 5 5 5 5 5 5 5 5 5 5 5 30 5 5 5 5 5 5 5 5 5 5 5 5 1 32 5 5 5 5 5 5 5 5 5 5 5 i 34' S 6 TTTTTTTTTTT 35 TTTTTTTTTTT l 36 10 TTTTTTTITTT l 37 1 11 11 10 10 10 10 10 10 9 9 9 ^ i 38 14 13 13 13 12 11 11 11 11 11 10 10 4 t Acceptable i 39 5 15 15 14 14 14 14 13 13 13 13 11 l 40 17 17 17 16 16 16 15 15 15 15 14 14 42 1 20 20 20 19 17 17 17 16 16 16 15 44 23 22 22 21 21 21 20 20 19 19 17 17 45 YYYYYYYYYYY Notes: \\ Use burnup and enrichment to lookup required cooling time in years. Licensee is responsible for ensuring that uncertainties in fuel enrichment and burnup are correctly accounted for during fuel qualification. Round burnup UP to next higher entry, round enrichments DOWN to next lower entry. Fuel with an initial enrichment less than 2.0 w/o U-235 must be qualified for storage using the alternate nuclear parameters specified in Table 1-lb. Fuel with an initial enrichment greater than 4.0 w/o U-235 is unacceptable for storage. Fuel with a burnup greater than 45 GWd/MTIHM is unacceptable for storage. Fuel with a l burnup less than 15 GWd/MTIHM is acceptable after three years cooling time provided the physical parameters from Table 1-lb have been met. Example: An assembly with an initial enrichment of 3.05 w/o U-235 and a burnup of 34.5 GWd/MTIHM is acceptable for storage after a nine-year cooling time as defined at the intersection of 3.0 w/o U-235 (rounding down) and 35 GWd/MTlHM (rounding up) on the qualification table. O October 1998 72-1004 Amendment Page D.7 Revision 2

.___.._m......._._________..____ !O l I 60 55 -- UNACCEPTABLE 50 45 - 40 - S H 35 - E O @ 30 - QUA FIED [ (Equiv. Ent: < 1.45 w/o) @ 25 - x o" 20 - 15 - NOT QUAllFIED 10 - (Equiv. Enr. > 1.45 w/o) 5-0 1.40 1.80 2.20 2.60 3.00 3,40 3.80 INITIAL ENRICHMENT (w/o U 235) Figure 1.1 PWR Fuel Criticality Acceptance Curve l 1

G October 1998 72-1004 Amendment Page D.8 Revision 2}}
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