Text

SAR for NFS-4/NAC-1 Spent Fuel Shipping Cask
	ML20012D172
Person / Time
Site:	07109183
Issue date:	03/15/1990
From:	; NAC INTERNATIONAL INC. (FORMERLY NUCLEAR ASSURANCE
To:	;
Shared Package
ML20012C403	List: ML20012C402; ML20012D172;
References
	NUDOCS 9003260582
	Download: ML20012D172 (85)
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{{#Wiki_filter:_. I< -l: Safety Analysis Report lg For The NFS-4/NAC-1 Spent-Fuel Shipping Cask-I:

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.I March 15,1990

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I g Nuclear Assurance Corporation 6251 Crooked Creek P.oad Norcross, Georgia 30092 Telephone: (404) 447-1144 Telex: 6827020,6827114 Facsimil )447 1797 C1990 Nuclear Assurance Corporation Printed in the U.S.A. 'I Docket No. 9183 prva myog; N^C - IL ' ~ ~ } .,,,,,,,f a'

[ RECORD OF REVISIONS-I REVISION DATE EFFECTIVE PAGES QLSCRIPTION OF CHANGE 6/30/1984 N/A Original Issue' 9/1984 7 1, 7 2, 7-3, 7-4 Changes to operations procedures 2/1987 ii, vii, 1-4, 1 6, To reflect copper shielding 1-8, 1-9, 1-10, 3-1, and allow shipment of 3 2, 3 3, 3 4, 3-5,. irradiated material uranium-3 6, 3-7, 3 8, 3-9, fuel rod cool one year. 3 10, 3 11, 3 12, 3-13, o E 3-14, 3 14a, 3-14b, 'E 5-1, 5-2, 5-6, 5 7-9/1987 4-3, 4-5, 4-6, 4-7 Valve handle and chamfer closure lid 5/1989 1, vii, 1-11, 1-12, To allow for shipment of 1-13, 1 14, 1-15 metallic fuel 3/1990 ii, iv, 2 97, 2-98, To allow for shipment of-t ,lE 2-99, 2-100, 2 101, failed metallic fuel- '5 2-102, 2-103, 2 104, 2-105, 2-106, 2-107, R 3-6, 3-6a, 3 6b, 3-7, 3 14c, 3-14d, 3-14e, 3-14f, 3 149, 3-14h, 7-5, 7-6, 7-7, 7-8, j 7-9,'7-10, 7-11, 7-12 f i h I N, I

3 V I t I TAB 12 0F CONTDrtS j I Page- .1.0 General Information., 11 1.1 Introduction................... 1-1 1.2 Package Description.... 11-

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1.2.1 Packaging........... 12 1.2.2 Operational Features. 1-8 1.2.3 Contents of the Package... 18 1.3 Appendices to General Information....................... 1 11 1.3.1 References...................................... 1-12 1.3.2 Record of Submittals........ 1-13

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1.3.3 NAC 1/NFS 4 License Drawing, E10080, Sheets 1 through 4, Revision 22.............. 1 15 l 2.0 Strucxral Evaluation........ 21 2.1 Structural Design............................ -2 1 2.1.1 Discussion................. 21 2.2 We i gh ts a nd Ce n t e rs o f G ravi ty......................... 2 t 2.3 Mechanical Properties of Haterials.... 2-4 2.4 General Standards for All Packages....................., 25 L 2.4.1 . Chemical and Calvanic Reactions................. 26 2.4.2 Positive Closure........................... 26 2,4.3 Lifting Devices..... 2-7 2.4.4 - Tie Down Devices................................ 2-8~ 2.5 Standards for Type B and Large Quantity Packaging....... 2 15 2.5.1 Load Resistance.......... 2 16 2.5.2 External Pressure............................... 2-18 2.6-Normal Conditions of Transport.......................... 2 20 2.6.1 Heat............................................ 2-20 2.6.2 Cold.................... 2 27 2.6.3 Pressure - 0.5 Times Standard Atmospheric L Pressure......................... 2-31 2.6.4 Vibration Vibration Normally Incident to Transport....... 2-31 2.6.5 Water Spray.......... 2 31 2.6.6 Free Drop................... 2 31 2.6.7 Corner Drop.................... 2 40 1 2.6.8 Penetration. 2-40 2.6.9 Compression......... 2-41 L Revised May 1989 i

I. & L I TABl.E OF CONTENTS, Contd. 2.7 Hypothetical Accident Conditions................. .........., 2 41 2.7.1 F r e e D r o p........................................... 2 41 g' 2.7.2 Puncture........................................ 2.7.3 Th e rms.1.............................................. 2 8 0 g 2 85 2.7.4 Water Immersion. 2 90 2.7.5 Summary of Damage................................ 2 90 l; 2-8 S p e c i a l Fo rm................................................ 2 9 0 2.9 Fu e l R o d s.................................................. 2 9 0 1 ' 2.10 Appendices to the Structural Evaluation..................... 2 91 l 2.10.1 References.. 2-91 2.10.2 Balsa Properties..................................... 2-93 2.10.3 Structural Evaluation of Failed Fuel Cans and Liners (Basket)....... 2 97 3.0 Thermal Evaluation............ 3-1 3.1 Discussion.................................................. 3 1 3.2 Thermal Properties of Materials.............................. 32 3.3 Technical Specification of Components........................ 33 1 3.4 Thermal Evaluation for Normal Transport Conditions........... 3-3 "'1 3.4.1 Description of the SCOPE Program..................... 34 3.4.2 Cask Model Description............................... 3 4 3.5 Thermal Evaluation of the Hypothetical Fire Accident......... 35 3.6 Appendices to Thermal Evaluation............ ..............., 3 6a 3.6.1 References........................................... 3 6a l' 3.6.2 SCOPE Input and Results.............................. 3 7 W ! 3.6.2.1 Me tall i c Fue l............................... 3 - 7 3.6.2.2 Failed Metallic Fuel........................ 3 14c i 3.6.3 Original'SAR Design Basis Thermal Analysis Summary... 3 15 1 l 3.6.4 Original SAR Description of the TAP Computer Code and l Thermal Mode 1........................................ 3-L 3.6.5 ' Original SAR Thermal Analysis for Normal Conditions of Ttansport.......................................... 3-48 3.6.6 Original SAR Thermal Analysis for Hypothetical Accident Conditions................................... 3-61 3.6.7 original SAR Thermal Analysis References.............. 3 71 4.0 Containment............ 4-1 l 4.1 Containment Boundary...................................... 4-1 4.1.1 Containment Vessel.............. 41 4.1.2 Containment Penetrations............... 41 n i o M, 11 Fevis ~ .ar,6 19; I

a I TABLE OF CONTENTS, Cont. Page Seals and Welds..................................... 4.1.3 Closure.................. ............................... 4-6 t 4.1.4 4-6 i 4.2 R equi renent s f o r Norma l Condi ti ons of Transport................... _4-6 t 4.2.1 Release of Radi oa ct i ve Mate ri a l........................... 4-6 4.2.2 P res suriza t i on of Cont ai nme nt Vessel....................... 4-6 .I 4.2.3 Coolant Contamination...................'.................. 4-8 4.2.4 Coolant loss.............................................. 4-8 4.3 Containment Requirements for the Hypothetical Accident Conditions. 4-8 4.3.1 F i s s i on G a s P r odu c t s...................................... 4 - 8 4.3.2 R e l e a s e s o f C ont e nt s...................................... 4-9 4.4 Appendi ces. to the Cont ai nment Evalu ation........................ 4-10 r 5.0 S h i e l d i n g E v a l u a t i o n.................................................. 5, 5.1 Discussion and Results........................................... 5-1 s 5.2 NAC-1 Des i gn Bas i s Source Speci fication.......................... 5. 5.2.1 G am m a S o u r c e.............................................. 5 - 2 s 5.2.2 Neu t r on S ou r ce............................................ 5 - 2 5.3 Mode l S p ec i f i ca t i on.............................................. 5 -2 1 5.3.1 Description of the Radial and Axial Shielding Configuration 5-2 _5.3.2 Shi el di n g Regi ona l Dens i ti e s............................... 5-2 ~ 5.4 S h i el di n g E va l u at i on............................................. 5-5 5.4.1 Shiel ding Evaluation for the Metalli c Fuel................ 5-6 1 1 L 5.5 Appendi ces to the Shiel ding E valuati on........................... 5-8 5.5.1 References................................................5-8 5.5.2 ANISN Input Data, Tables 5-4, 5-5 and 5-6................. 5-9 l 5.5.3 Metallic Fuel Source Terms vs. Cool Time (Table 5-7)..... 5-14 6.0 C r i t i c a l i ty E v a l u a t i o n................................................ 6-1 6.1 D i s cu s s i on an d R es ul t s........................................... 6-1 6.2 P a c k a ge F u e l L o a d i n g............................................. 6 - 3 6.3 Appendi ces to the C ri ti ca li ty Eva luation......................... 6-6 b iii

p. 1 l l TABl.E OF CONTENTS, Contd. 7. 0 - Ope r a ting Proc e dur e s............................................. 7 1 7.1-Proc edure s for 1.oading the Package.......................... 7-1 1 j 7.2 Procedures for Unioading the Package..,..... ................ 72 7.3 Preparation of the' Empty Package for Transport.............., 73 7.4 Procedures for Loading Sound and Failed Metallic Fuel....... 7 5 -l 7.4,1 Cask Loading......................................... 7 5 { 7.4.2 Loading "stallic Failed Fuel into Canisters........... 7-8 1 7.5 Appendices to the Operating Procedures..... 75 8.0 - Acceptance Tests and Maintenance Program.. 81 1 8.1 Acceptance Tests............................................. 8 - l' 8.1.1 Visual Inspection.................. 81 8.1.2 Structural and Pressure Tests........................ 8-1 8.1.3 Leak Tests........................................... 8-2 8.1.4 Cask Components Test.................................. 82 8.1.5 Tests for shielding Integrity........................ 8-2 8.1.6 Thermal Acceptance Test............................... 82 1 .i 8.2 Maintenance Program......................................... 8 3 8,3' Appendices to the Acceptance Tests and Maintenance Program... 8-4 8.3.1 References........................................... 8 4 3 8.3.2 Hydrostatic Pressure Test Procedure.................. 8 5 8.3.3 Shielding Test Procedure............................. 8-10 8,3.4-Thermal Acceptance Test Procedure...................... 8-21 8.3.5 Worcester Valve Company Test Report................... 8 26 i l iv Revised March 1990

2.10.3 Structural Evaluation of Failed Fuel Cans and Liners (baskets) This evaluation documents'the thermal and structural-adequacy of six element and three element failed fuel cans and liners (baskets) for the NFS 4/NAC 1 j Spent Fuel Shipping Cask. The maximum normal operating temperature is calculated to be 211'F. The conservstively calculated minimum margin of safety for any component is 1[L4.t9.. The failed fuel cans and liners (baskets) are structurally adequate to satisfy all regulatory requirements. b 2.10.3.1 Discussion Nuclear Assurance Corporation proposes to ship chree or six encapsulated failed metallic fuel rods in the NAC 1 cask. A separ n e liner (basket) and failed fuel can have been designed for a three rod capacity shipment and for a six rod capacity shipment. In each case, the failad fuel can is a sealed, dry aluminum canister. Only one failed metallic fuel rod is placed in each failed fuel can, i The failed fuel can must maintain containment for all loading conditions because it serves as one of the containment barriers. required for the transport of high level radioactive material. Since it is not possible g under any circumstances for the six metallic fuel rods to attain a critical 3 . configuration, the liner (e) (basket) acts as a convenient support and spacer in the cask cavity. No permanent deformation.of the liner-is permitted. Each metallic fuel rod weighs 125 pounds and is approximately 124 inches long. The total heat load for six metallic fuel rods is 30 watts. 2.10.3.2 Method of Analysis A one dimensional thermal analysis of the NAC 1 cask with a total heat load of 30 watts was performed using the SCOPE thermal analysis computer program i to determine the maximum normal operating temperature of the failed fuel can .and the liner (basket). 2 97 Revised March 1990

g I Classical stress analysis methods are used to evaluate the failed fuel cans for buckling during the end impact and for bending during a side impact. The tubes in the liners (baskets) are also analyzed for bending during a side impact. The impact loadings include the g factors determined in the-NAC 1 cask Safety Analysis Report (SAR). The calculated stresses in the failed fuel cans and the liners are conservatively compared to the material yield strength to demonstrate that containment is maintained by the failed fuel cans and that no permanent deformation of the failed fuel cans or the liners (baskets) occurs. 2.10.3.3 Input Geometry & Data 1. Total Heat Load - 30 Watts (For Six Metallic Fuel Rods). l-L 2. Metallic Fuel Rod Weight - 125 Pounds / Rod). l-3. NAC 1 Cask Geometry: (Ref. SAR, Section 1.3.3) I; l-Ir.ner Shell (Cavity) I.D. - 13.50 inches l-Inner Shell Thickness 0.313 inches Load Shell Thickness 6.63 inches outer Shell Thickness 1.25 inches -Neutron Shield Thickness 4.5 inches ~ l Neutron Shield Shell Thickness 0.165 inches 4 Free Drop Impact G Loads: (Ref. SAR) ~ Normal Operation

1-Foot Side Drop 12.3g (Pg. 2-37)

Accident 30-Foot Top End Drop 44.67g (Pg. 2-47) 30 Foot Bottom End Drop 76.6g (Pg. 2-53) 30-Feot Side Drop 96.06 (Pr,. 2-67) 30-Foot Corner Drop 45.5g (Pg. 2-72)

Ref. 10 CFR 71 and Regulatory Guide 7.8.

2-18 Revised March 1990

2.10'.3.4 Mechanical Properties of Materials -l L 1. 6061-T6 Aluminum Alloy (Ref. MIL HDBK 5E) S, - 42 ksi (70'F) S - 35 kai (70*F) (Page 3 222) y At 250*F: S - 0.86(S.") - 36.1 ksi (Page 3-227) 7250 70 lI[ S - 0.88(S ) - 30.8 ksi (Page 3-228) 7250 70 2, 6063-T832 Aluminum Alloy (Ref. ASME B210) S - 40 ksi S - 35 ksi '(Page 194) ug At 250'F: S - 0.86*S - 34.4 ksi S - 0.89*S - 30.8 ksi y y r: l. l 2.10,3,5 Thermal Evaluation The SCOPE thermal analysis computer. program is used to evaluate the NAC 1 cask containing the 2.75 inner diameter (I.D.) failed f.tel can liner and six 2.75 I.D. Failed Fuel Cans loaded with one metallic fuel rod each for a total heat load of 30 watts. !.i Section 3,6.2.2 presents the input data and the resulting calculated temperaturas from the SCOPE analysis. The calculated temperature for the failed fuel can and the liner (basket) is 211*F.

The strength variation with temperature is assumed to be the same as that for 6061-T6 Aluminum Alloy 2 - 99 Revised March 1990

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I 2.10.3.6 Structural F. valuation The failed fuel cans are evaluated to demonstrate that containment of the failed fuel rod is maintained for all loading conditions. The maximum stress occurs in the shells of the failed fuel cans for the 30 foot side drop load case. Buckling.of the shells is evaluated for the-30-foot end drop load case. The liners (baskets) for the failed fuel cans are evaluated to demonstrate that rupture (ultimate failure) does not occur for any loading condition, The maximum stress occurs in the housing (tube) of the liners for the 30-gg foot side drop load case. 2.10.3.6.1 Failed Fuel Can 2.75 Inner Diameter 2.10.3,6.1.1 Shell Bending (Ref. Ow3 340108-D2, Rev.) Loadiny I-30 Foot Side Drop Acceleration - 96 g Support Spacing - 33.66 in (Ref. Dwg. 491-001, Rev. 1) % s: c: Fuel - 125 lb/124 in - 1.008 lo/in Shell-[(3.02,2.75]1x0.10-0.113lb/in 2 Total - 1.121 lb/in Conservatively assume the shell is simply supported at the support disks; then, the moment during impact is: M-fxl'.121x33.66 x 96 - 15241 in lb 2 2 - 100 Revised March 1990

,y...., ~ ~ %A[ I' 'I - ) Shell Properties: l 3 I/C.0.7791 in ( Material Properties: (ASTM B221 Type 6061 T6) l 5;i (S]250 - 30.8 ksi 7 i, Stresses-1 Sb .15241/0.7791 - 19562 psi s M. S. - (S ] /Sb ~ iSLIll y

2. 10. 3. 6. '1. 2. Failed' Fuel Can - 2.75 Inner Diameter l

i: 1 '.. l Shell Buchling !. f -z; Estimated Height-of Can Assembly - 20 lb l-Bottom End Drop Acceleration - 76.6 g n. '~ The compressive stress.in the shell due.co its weight during- .v s impact is: l'. l... 20~(76.6) s*. { (3.0 ' 2.75 ], 7357 p,g 2 2 y. I' li. \\;- ~,, The margin ef safety on yield is: e i t M.S. : y 1- +21.7

p. -

l 2 - 101 Revised March 1990 v

1 =.. ! }'. mi1 ~ The buckling of the cylindrical shell under the auction of uniform axial <s ~ compression may be evaluated using equation 11-1 on page. 458 of Theory of Elastic stability by Timoshenko and Gere. 4 t Et 8 ~ r;3(1 v { 0.5 2 cr Ii g

where, E - 0.97-(10.1) - 9.8 ksi at-250'F t - 0.125 in r - 2.81 in, env. radius-for all tubes in assembly g

v - 0.33 g-P 5 Jc' The margin of safety on buckling is: L1 M.S.-- Scr/8

1 ~ 81b E c

2.10. 3. 6.1.'3 - Conclusion The stress 'in the cylindrical shell caused by a bottom end drop with 76.5 g deceleration is much lower than the yield stress and the critical buckling. 1 stress. j ly 2.10.3.6.2 Liner - Failed Fuel can 2.75 Inner Diameter r,, l' Housing -Bending ( (Ref. Dwg. 491-001, Revision 0) Loading q 30 Foot Side Drop Acceleration - 96 g Sapport Spacing - 33.66 in r I 2 - 102 Revised March.1990 4

I-hight: Fuel - 125 lb/124 in - 1.008 lb/in 'I AluminumShell-{(3 2.75]x1x0.10 - 0.113 lb/in 2 2 Housing-{(3.75 3.5 j x 1 x 0.10 - 0.142 lb/in 2 2 Total - 1.263 lb/in Conservatively assume the housing simply supported at the disk supports; then,'the moment during impact is: M - f (1.263 x 33.66 ] 96 - 17,172 in lb 2 Housing-Properties: -I.. d,' d /32d, I/C - x I (3. 75' - 3.5 )/(32 x 3.75) - 1.2485 inch 0 3 -w l Material Properties: (ASTM B210 Type 6063-Tt32) i (S ) - 30.8 ksi. Y 250 l Stresses: Sb - 17172/1.2485 - 13,754 psi M.S. - (S ) /Sb - +1.24 Y 250 The other accident drop conditions are: 30 Foot Top End Drnp a - 44.67 g 30 Foot Bottom End Drop a - 76.6 g 30-Foot Top Corner Drop a - 45.5 g 2 - 103 Revised March 1990

i Compared t9.a 30 foot side drop accident. condition, with a - 96 g, the-above accident conditions are less critical. Therefore, neither rupture nor .: yielding of the liner housing will occur. 2.10.3.6.3 Failed Fuel Rod Can 4.00 Inner Diameter .Shell' Bending

/

(Ref. Dw6 340-108 D1, Rev. 9) Loading 30-Foot Side Drop-Accel? ration = 96 g

Support Spacing - 50.37 in Weight:

Fuel - 125 lb/124 in - 1.008 lb/in I Shell-}(t25 - 4.0 ) x 1 x 0.10 - 0.162 lb/in 2 2 Total - 1.170 lb/in m conservatively assume the shell is simply supported at the support disks; then, the moment.during impact is: M-f(1.170x50.37]96-35,621inlb 2 hl S e l Properties: (4.90 I.D., t - 0.125 in) 0 0 3 I/C - x (4.25 4 )/(32 x 4) - 1.7243 in Material Properties: ~(ASTM B221 Type 6061 T6) (S ) - 30.8 ksi 7 250 I 2 104 Revised March 1990 ',6 j

,_f, i 6 ' Stresses: Sb - 35621/1.7243 = 20,658 psi M.S. - (S ) /Sb - +0.49 y 2.10.3.6.4 Liver 3 Element NAC 1 Cask Tube Bending. (Ref. Dwg 347-211 F19, Rev. 5) Loading 30 Foot Side Drop Acceleration = 96 g Support Spacing - 50.37 in Weight: Fuel - 125 lb/124 in - 1.008 lb/in I: Shell-{(4.25 4.0]x1x0.10 - 0.162 lb/in 2 2 I Tube-{(5.625 - 5.375 ] x 1 x 0.1 - 0.216 lb/in 2 2 Total - 1.386 lb/in 3 Conservatively assume the tube is simply supported at the support disks; then, the moment during impact is: M-f(1.386x50.37)96-42,198inlb 2 Tube Properties: 5.375')/(32x5.625)-2.9053in I/C - r (5.625 3 0 g 2 - 105 Reviseo March 1990

n-9, 8 - 1 1 Material. Properties: (ASTM B210 Type 6061 T6) (S ) - 30.8 ksi g' Y 250 5- -Stresses: Sb - 42198/2.9053 - 14,524 psi M.S. - (S ) /Sb - +1.12 y 2.10.3.7 Results and Conclusion f The maximum normal operating temperature of the failed fuel cans and the liner (basket) is calculated to be 211'F. The structural evaluation-conservatively uses material properties at 250'F. The calculated xcrgins o! rafety are: 1, Failed: Fuel Catt 2.h Inner Diameter i Shell Bending +0.57 Shell - Buckling +32.1 2. Liner Failed Fuel can - 2.75 Inner Diameter Housing Bending +1.24 3. -Failed Fuel Rod Can - 4.00 Inner Diameter Shell Bending +0.49 4. Liner - 3 F.lement - NAC 1 Cask Tube Bending +1.12 2 - 106 Revised March 1990

l-No permanent deformation occurs in the failed fuel cans or the liners for the critical loading conditions. Containment of the failed metallic fuel rods is maintained and the liner structure remains intact; therefore, structural adequacy is ensured. 2.10.3.8 References I 1. Safety Analysis Report for the NSF 4/NAC 1 Spent Fuel Shipping-Cask, Nuclear Assurance Corporation, June 30, 1984. 2. NAC Drawing: 340-108 D1, Revision 9, Failed Fuel Rod can 4.00 I.D. 3. NAC Drawing: 340 108 D2, Revision 9, Failed Fuel Rod Can 2.75 1.D. 4 NAC Drawing: 247 211 F19. Revision 5, Liner - 3 Element, NAC 1 Cask. 5. NAC Drawing: 491 001, Revision 0, Liner Failed Fuel Can, 2.75 I.D., NAC 1 Cask. g-6. MIL HDBK 5E, " Metallic Materials and Elements for Aerospace _ Vehicle B= Structures," U.S. Department of Defense ~, May 1989. I j I 7. "ASME Boiler and Pressure Vessel Code," Section II, Material ] Specifications, hr': B Nonferrous Materials, The American Society of !!echanical Engineers,1989. 8. Timoshenko and Cere, Theory of Elastic Stability, 2nd Edition, New York,- McGraw Hill, 1961. 9. F. Kreith, Principles of Heat Transfer, 2nd Edition, Scranton, PA, International Textbook Company, 1965.

1-i 2 - 107 Revised i

March 1990 l l

.~...,~ i ? r i l* p.< I, .I 1 l l-l This page intentionally left blank. 3 l t 1. s h l' I l l l 1 l i l 4 2 - 108 Revised March 1990 t e

4 m.

LI l} decay heat load of 750 watts was analyzed. The neutron shield tank was assumed - to contain dry air. The cask component dimensions describing the radial model is presented in Table 3-5. Table 3-5. SCOPE CASK RADIAL MODEL DESCRIPTION Thickness Radius from Cask Center (inches) Component Material (inches) Outside Surface Inside Surface I Cask Surface 321 55. 0.165 19.600 19.435 Neutron Shield dry air 4.500 19.435 14.935 Outer Shell 321 SS 1.250 14.935 13.685 Ganna. Shield Lead 6.623 13.685 7.063 Inner Shell 321 SS 0.313 7.063 6.750 3.5 Thermal Evaluation of the Hypothetical Fire Accident The thermal model for the analysis of the hypothetical accident is basically identical to the model for the analysis of normal transport conditions. The I-ambient conditions are changed to remove the insolation and increase the. temperature of the surroundings to 1475'F to represent the fire. Immediately' before and after the 30 minute fire, the ambient temperature is at 130'F. A 5 summary of the cask and fuel temperatures are presented in Table 3-4 and 3-6 3 for the steady-state. (normal transport condition) and transient analysis (fire accident condition). respectively. The maximum temperature in the neutron shield tank (1354'F) occurs at the end of the fire accident. The ideal gas law is used to determine the maximum pressure of the shield tank. Prior to fuel loading the average temperature of I the neutron shield fluid (air) is assumed to be -40'F (worst case). Using the ideal gas equation P2=V.J. X ,2_ x P1 T E. where: T1. 40*p. 420'R 5 V2 T1 T2 = 1357'F = 1814'R K1=V 2 1 = 14.7 psia Thus: P2 = (14.7) x (1) x (1814/420) = 63.5 psia = 48.8 psig Table 3-6.

SUMMARY

OF TEMPERATURES AND PRESSURES FOR HYP0THETICAL FIRE ACCIDENT CONDITIONS Without Shipping Container With Shippino Container Maximum Pressure Maximum Pressure Component Temperature ('F) (psic) Temperature ('F) (psic) Fuel Rod Surface 540 540 Inner Shell (Cavity) 419 16 419 17 Lead 437 I 437 Outer Shell (Shield Tank) 1354 49 485 18 Surface 1357 488 j Shipping Container 1357 Revised 3-5 Feb. 1987

s .q 'l! ' I lIl This page intentionally left blank. = 'l 1 l I I I I 36 Revised March 1990 . ~ -... ~..

I l 1 3,6 Appendices to the Thermal Evaluation This section presents references and other information that supplements the I information presented in Section 3. .I 3.6.1 References 1. " Safety. Analysis Report for Nuclear Fuel Services, Inc. Spent Fuel Shipping Cask Model No. NFS 4," Nuclear Fuel Services, Inc., U.S. NRC Docket 6698, September 1972. 2. "ASME Boiler and Pressure Vessel Code, Section III, Nuclear Vessels," I-The American Society of Mechanical Engineers, New York, NY, 1971. 3. Nuclear Systems Material Handbook, Volume I, Design Data, Revision 2, August 1, 1977. 4 Courpilation of Thermal Property Data for Computer Heat Conduction Calculations, Edwards, A. L. (University of California, Lawrence Livermore Radiation Laboratory), February 24, 1969. 5. Principles of heat Transfer, Second Edition, Kreith, F. (International

I

Textbook 6, Scranton, Penn.), 1965.

6.

Jakob, M., and hawkins, G.A., elements of heat Transfer, John wiley and Sons Inc., New York, NY, 1951.

3.6.2 SCOPE Input and Results I-3.6.2.1 Metallic Fuel Normal Transport and Hypothetical Fire Accident ~ Conditions 3.6.2.2 Failed Metallic Fuel Hypothetical Drop Accident Condition I 3.6.3 Original SAR Design Basis Thermal Analysis Summary I 3 6a Revised March 1990

- -. _ _ -. -. ~.. -..... s 3.6,4 Original SAR Description of the TAP Computer Code and Thermal-Model i Description s 3.6.5 Original SAR Thermal Analysis for Normal Condition of Transport 3.6.6 Original SAR Thermal Analysis for the Hypothetical Accident Conditions 3.6.7 Original SAR Thermal Analysis References i 5 l I I I I l 3 - 6b Revised March 1990 mm .1

9 3 '. 6. 2. SCOPE Input and Results <g 3.6.2.1' Metallic Fuel' Normal Transport and Hypothetical Fire Accident I E-Condition. /

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I i, - 3.6.2.2 Failed Metallic Fuel Hypothetical Drop Accident Condition The SCOPE thermal analysis computer program is used to evaluate the NAC 1 cask containing the 2.75 inner diameter (I.D.) failed fuel can liner and six 2.75 I.D. Failed Fuel Cans loaded with one metallic fuel rod each for a total heat load of 30 watts. The following pages present the input data and the resulting calculated temperatures from the SCOPE ca.alysis. The calculated temperature for the failed fuel can and the liner (basket) is 211'F. Since the three-element liner and failed fuel cans will have one half of the six element heat load, the temperatures will be significently lower and, thus, less critical. t I l I 3 - 14e Revised March 1990

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7.4. Procedures for Loading Sound and Failed Metallic Fuel 7.4.'l Cask Loading I.. NOTE: The procedure for loading dry metallic. sound (intact)-fuel baskets, small diameter failed fuel canisters (FFCs) and. large diameter.FFCs into the cask is the same; however, only L three large diameter FFCs or three sound fuel baskets can be placed into the cask, while six of the smaller FFCs can be

used, p

7.4.1.1 Perform a receiving survey of the empty cask and closed container and inspect for damage. I. 7,4.1.2' Position the trailer in the designated area. Set the trailer brakes and block the wheels against movement in either direction. 7.4.1.3 Remove the lid from the ISO container. H 7.4.1,4 Perform a Health Physics survey-of the cask and adjacent surfaces of the container. NOTE: A receiving survey of the cask and transporter rust be .I performed as soon as practicable after arrival at the site to ensure compliance with 10 CFR 71.87(i), 10 CFR 71.47, and to ensure timely reporting of any transportation noncompliance. 7.4.1.5 Remove the top impact limiter. i' l 7.4.1.6 Remove the cask tiedown straps. 7.4.1.7 Ucing the cask lifting yoke, engage the lifting trunnions of the - I front end of the cask. Raise the cask to a vertical position on the rear cask support, moving the crane as required to keep the cask engaged in the trailer rear rotation supports and the crane cable vertical. L.' hen the cask is fully vertical, lift the cask g-from the container. 3-75 Revised March 1990 I

i i I Iil .I 5 I o - 7.4.1.8 Place the cask in the dry loading stand. Disengage the lifting yoke. 7.4.1.9 Remove the vent and drain valve port covers. Visually inspect the; valved quick disconnect nipples and replace them, if necessary, g 5 7.4.1.10 Remove the closure lid bolts. Attach the. lid lift slings to the closure lid. Remove the closure lid and set it on a support that = is suitable for radiological control and for maintaining the cleanliness of the closure lid. Inspect the o ring seating g surface on the cask, 3 7.4.1.11 Visually inspect the inner cavity for foreign material or damage. Verify the proper cask cavity liner is in place. 7,4.1.12 Replace the closure lid, but not the lid bolts. 7.4.1.13 Identify the fuel to be loaded. Note: An ORIGEN (or equivalent) analysis must be completed on the g. fuel bearing components prior to loading to ensure that the E-proposed contents conform to the Certificate of Compliance. 7.4.1.14 Place the carriage assembly on top of dry loading stand. 7.4.1.15 Position the carriage assembly for cask lid removal. Connect the cask lid to the removal cylinder and remove the lid from the cask. 7.4.1.16 Position the carriage for fuel loading. 7.4.1.17 Match mark the carriage assembly and turntable to each fuel ) canister location in the cask cavity. For sound fuel, there will be three canisters with five rods each; with small diameter FFCs, there will be six canisters with one rod each; and with, large diameter FFCs, there will be three canisters with one rod each, 7.4.1.18 Move the carriage and turntable to fuel canister position 76 Revised March 1990 B.

q .c ( t j.l 7. 4.1.1'9 After retrieving the fuel (sound or failed) from the pool, place- [ the shielded transfer cask containing a fuel canister (as many as five sound or one failed metallic fuel rods) onto the turntable. i 7.4.1.20 Lower the fuel canister from the transfer cask into the shipping } cask. 'sg. 7.4.1.21 Repeat steps 7.4.1.18, 7.4.1.19 and 7.4.1. 20 to access fuel ]m. canister positions 2 and 3 and complete the cask loading. For small diameter FFC loading, repeat for all six positions. 7.4.1.22 Replace both o rings on the closure lid. I. 7.4.1.23 Replace the closure lid onto the cask by re positioning the carriage assembly and actuating the cask lid cylinder. Visually confirm that the lid is properly seated. 7.4.1.24 Remove the carriage assembly from the loading stand. 7.4.1.25 Tighten all six closure lid bolts to 650 foot pounds. I 7.4.1.26 Attach the pressure test fixture to the cask cavity vent port. I Pressurize the cavity with air to 30 psig and hold for 10 minutes. No pressure drop is-permitted. 7.4.1.27 Using the pressure test fixture, pressurize the annulus between the two lid o ring seals to 30 psig through the o ring port. Observe the air pressure gauge for 10 minutes after closing the isolation valve. If no drop in air pressure is observed, the seal is acceptable. If air pressure drops, remove the valve port cover and replace the seals, then repeat the test. I. 7.4.1.28 Close the vent, o-ring and drain valves and then replace the valve port covers. 7-7 Revised March 1990 LI e

7 4.1.29 Install the tamper indicating seals (3) on the lid closure bolt I heads. 7.4.1.30 Deco ataminate the cask. Survey the cask for surface contamination and radiation dose rates. 7.4.1.31' Using the cask lifting yoke, tronsfer and lower ths cask to the trailer. Engage the lower rotation trunnions with the rotation supports in the container. Lower the cask to rest on the upper trunnion _ saddles,. moving the cask lifting yoke as required to keep the crane cables vertical. Disengage the cask lifting yoke from the cask lifting trunnions and set it aside. 7.4.1.32-Install the cask tiedowns; then install the cask impact limiter. 7.4.1.33 Install the container lid. 7.4.1.34 Complete a Health Physics survey and record vehicle radiological compliance data. Ensure compliance with 10 CFR 71.87(i) and 10 CRt 71.47. Complete the shipping documents and apply placards and labels to the cask and container. 7.4.2 Loading Metallic Failed Fuel into Canisters l l 7.4.2.1 Small Diameter Canisters l-7.4.2.1.1 Examine the small diameter failed fuel canister (FFC) and check it for damage. 7.4.2.1.2 Place the FFC inside the containment barrier portion of the l pool. Position the FFC in the failed rod loading station. 7.4.2.1.3 After verifying the accountability records, place the designated failed fuel rod into the FFC. If the rod is broken into two or more pieces, verify that the lid thread and seal area is not fouled during rod insertion. 7-8 Revised March 1990 I

. I< k I 7.4.2.1.4 When the can is loaded, install the lid using the FFC Lid i Installation Tool. 1 7.4.2.1.5 Using the FFC handling tool, move the loaded FFC through the containment barrier door and place the FFC horizontally into the upender. 7.4.2.1.6 Operate the hand vinch to move the FFC to the vertical position. 7.4.2.1.7 Torque the FFC lid to 100 foot pounds for the small canister. l 7.4.2.1.8 Connect the nitrogen supply line to the vent valve. l 7.4.2.1.9 Open the nitrogen supply valve and pressurize the FFC to force out the water. Blow gas through the FFC for 5 minutes after the first visible traces of bubbles appear. Remove the gas supply line. l 7.4.2.1.10 Invert the FFC in the upender and install the pipe plug, 7.4.2.1.11 Re invert the FFC in the upender. 7.4.2.1.12 Attach the vacuum pump to the FFC' vent valve. Evacuate the FFC I. to a pressure below 1 inch of mercury for 15 minutos. Remove the vacuum pump and backfill with nitrogen. I 7.4.2.1.13 1 Remove the FFC from the upender and place it into temporsry storage. l 7.4.2.2 Large Diameter Canisters 7.4.2.2.1 Examine the large dia.neter FFC and check it for damage. 7.4.2.2.2 Place the FFC inside the containm:nt barrier portion of the pool. Position the FFC in t'.e failed tod loading station. 79 Revised March 1990 m s.- -.m -- - mm.-- m

m I 7.4.2.2.3 This step is to be used when loading a previously canned fuel -rod into the large diameter canister. After verifying the-accountability records, remove the ceramic filter from the top of the original failed rod can. Position the can plug with aluminum screen onto the open can. Install the plug. 7.4.2.2.4 Verify the accountability records for the fuel to be loaded, 7.4.2.2.5 Place the designated fuel into the FFC, If the rod is broken into two or more pieces, verify that the lid thread and seal g, area is not fouled during rod or can insertion. 3 7.4.2.2.6 Wen the canister is loaded, install the lid using the FFC Lid Installation Tool. '7.4.2.2.7 Using the FFC handling tool, move the loaded FFC through the. containment barrier door and place the FFC horizontally into the upender. 7.4.2.2.8 Operate the hand winch to move the FFC to the vertical position. .? 7.4.2.2.9 Torque the FFC lid to 130 foot pounds for the large canister. 1 7.4.2.2.10 Connect the nitrogen supply line to the vent valve. l 7.4.2.2.11 Open the nitrogen supply valve and pressuri::e the FFC to force out the water. Blow gas through the FFC for 5 minutes after the j first visible traces of bubbles appear. Remove the gas supply line. 7.4.2.2.12 Invert the FFC in the upender and install the pipe plug-7.4.2.2.13 Re invert the FFC in the upendor, 7 - 10 Revised March 1990

I 7.4.2.2.14 Attach the vacuum pump to the FFC vent valve. Evacuate the FFC to a pressure below 1 inch of mercury for 15 minutes. Remove I, the vacuum pump and backfill with nitrogen. 7.4,2.2.15 Remove the FFC from the upender and place it into temporary storage. l t (- l r-l 11 l... I l 1. I 7 11 Revised March 1990

.,y .. -. ~ . -. ~. ,, {., v i. Ii - 7. 5 Appendices to the Operating' Procedure 3 Deleted 1 1 i 'f I l 7 - 12 Revised March 1990

~ INSTRUCTIONS FDR IMPIRENTATION OF MARCH 1990 REVISION J PACES TO BE REMOVED PACES TO BE INSERTED Cover Page 05/89 Cover Page 03/90 I Record of Revisions 03/90 Section Pare Number Date Section Pare Number Date = 1 05/89 i 05/89 11-02/87 11 03/90 tii 06/84 111 06/84 iv 06/84 111 03/90 2 2 2 97 03/90 through 2 108 03/90 3 35 02/87 3 35 02/87-36 02/87 36 03/90 3 6a 03/90 3 6b 03/90 37 02/87 37 03/90 38 02/89 3-8 02/87 3 14c 03/90 3 14h 03/90 I 7 7-5 06/84 7 7-5 03/90 7-6 06/84 through 7-12 03/90 (Page 1 of 1)

'gl ' I Safety Analysis Report For The NFS-4/NAC-1 Spent-Fuel Shipping Cask .I W I Lli March 15,1990 1 I i-L Nuclear Assurance Corporation 6251 Cmoked Creek Road Norcross, Georgia 30092 Telephone: (404) 447-1144 Telex: 6827020,6827114 Facsimile: (400 447 1797 l @1990 Nuclear Assurance Corporation ~ Printed in the U.S.A. Docket No. 9183 NAC E-804 e ; - - - -===: .==w-1

I RECORD 0F REVISIONS I REVISION DATE EFFECTIVE PAGES DESCRIPTION OF CHANGE 6/30/1984 N/A Original Issue 9/1984 7-1, 7-2, 7-3, 7-4 Changes to onerations procedures 2/1987 ii, vii, 14, 1-6, To reflect copper shielding 1-8, 1-9, 1 10, 3-1, and allow shipment of I 3-2, 3-3, 3-4, 3-5, irradiated material uranium 3-6, 3-7, 3-8, 3-9, fuel rod cool one year 3-10, 3-11, 3-12, 3-13, I 3-14, 3-14a, 3-14b, 5-1, 5-2, 5-6, 5-7 i 9/1987 4-3, 4-5, 4-6, 4-7 Valve handle and chamfer closure lid 5/1989 1, vii, 1-11, 1-12, To allow fer shipment of 1-13, 1-14, 1-15 metallic fuel 3/1990 ii, iv, 2-97, 2-98, To allow for shipment of 2-99, 2-100, 2-101, failed metallic fuel I 2-102, 2-103, 2-104, 2-105, 2-106, 2-107, 3-6, 3-6a, 3-6b, 3-7, 3-14c, 3-14d, 3-14e, 3-14f, 3-14g, 3-14h, b,hil 12 7 I I I I I I I

Io ' i I TAB 12 OF CONTENTS I Page 1.0 General Information........................................ 11 1.1 Introduction............................................ 11 I 1.2 Package Description................... 11 I 1.2.1 Packaging............... 12 1.2.2 Operational Features............ 18 1.2.3 Contents of the Package......................... 18 1.3 Appendices to General Information........ 1 11 1.3.1 References...................................... 1 12 1.3.2 Record of Submitta1s........................... 1 13 1.3.3 NAC 1/NFS 4 License Drawing, E10080, Sheets 1 through 4, Revision 22....... 1 15 2.0 Structural Evaluation... 21 ~ 2.1 Structural Design................ 21 . I. Discussion...................................... 2-1 2.1.1 2.2 Weights and Centers of Cravity............ 23 2.3 Mechanical Properties of Materials.. 2-4 2.4 General Standards for All Packages...................... 25 2.4.1-Chemical and Galvanic Reactions....... 26 2.4.2 Positive Closure................................ 2-6 2.4.3 Lifting Devices................................. 27 2.4.4 Tie Down Devices............. 28 I.- 2.5 Standards for Type B and Large Quantity Packaging....... 2 15 2.5.1 Load Resistance. 2 16 2.5.2 External Pressure............ 2-18 2.6 Normal Conditions of Transport.......................... 2-20 . I 2.6.1 Heat......................... 2-20 2.6.2 Co1d....................... 2 27 I-. 2.6.3 Pressure 0.5 Times Standard Atmospheric Pressure........ 2-31 2.6.4 Vibration Vibration Normally Incident to Transport..... 2 31 - I; 2.6.5 Water Spray.......... 2 31 2.6.6 Free Drop................ 2-31 2.6.7 Corner Drop....... 2-40 2.6.8 Penetration. 2-40 2.6.9 Compression.. 2-41 Revised May 1989 i II

lo I TABLE OF CONTENTS, Contd. 2'. 7 Hypothetical Accident Conditions............................ 2 41 2.7.1 Free Drop.............. ,,, 2 41 2.7.2 Puncture.... Th e rm a 1.............................................. 2 8 0 2.7.3 2 85 2.7.4 Water Immersion......................................, 2-90 2.7.5 Summary of Damage........ 2 90 2.8 S p e c i al Fo rm................................................. 2 9 0 i 2.9 Fuel' Rods.................................................... 2-90 .t 2.10 Appendices to the Structural Evaluation..................... 2 91 2.10.1 Leferences....... 2 91 2.10.2 Balsa Properties................... 2 93 2.10.3 Structural Evaluation of Failed Fuel Cans and Liners (Basket)......... 2-97 3.0= Thermal Evaluation............................................... 3 1 3.1 Discussion..................................................31 3.2 Thermal Propertiet o f Mate ri als............................ 3-2 3.3 Technical Specification of Components....................... 3 3 3.4 Thermal Evaluation for Normal Transport Conditions........... 33 -3.4.1 Description of the SCOPE Program..................... 3 4' 3.4.2 Cask Model Descripcion............................... 3-4 3.5 Thermal Evaluation of the Hypothetical Fire Accident......... 35 3.6 Appendices to Thermal Evaluation........................ 3-6a i l' 3.6.1 References........................................... 3 6a l 3.6.2 SCOPE Input and Results.............................. 3 7 i 3.6.2.1 Metallic Fue1......................... 3 7 l 3.6.2.2 Failed Metallic Fue1........................ 3 14c l 3.6.3 Original SAR Design Basis Thermal Analysis Summary... 3 15 l 3.6.4 Original SAR Description of the TAP Ccmputer Code and Thermal Model........................................ 3 19 3.6.5 Original SAR Thermal Analysis for Normal Conditions of Transport..... 3 48 3.6.6 Original SAR Thermal Analysis for Hypothetical Accident Conditions.................................. 3 61 3.6.7 Original SAR Thermal Analysis References............. 3-71 4.0 Containment................... 4-1 4.1 Containment Boundary.................... 41 4.1.1 Containment Vessel.............. 4 1 4.1.2 Containment Penetrations......... 4-1 11 Revised March 1990 au

- - m m a in imii ss -uu siu milis'unBum I TABLE OF CONTENTS, Cont. I Page-L Seals and Welds........................................ 4.1.3 Closure..................................................4-6 4.1. 4 4-6 4.2 Requi renents for Normal Conditi ons of Transport.................. 4-6 4.2.1 Rel eas e of R ad i oa ct i ve Mat e ri a l........................... 4-6 4.2.2 P ressuriza ti on of Cont ai nme nt Vessel....................... 4-6 4.2.3 Coolant Contamination..................................... 4-8 4.2.4 C o o l a nt L o s s.............................................. 4 - 8 '4.3 Containment Requirements for the Hypothetical Accident Conditions. 4-8 4.3.1 F i s s i on G a s P r odu c t s...................................... 4 -8 4.3.2 Releases of Contents...................................... 4-9 4.4 Appendices to the Contai nment E valuation........................ 4-10 5.0 Sh i el di n g E va 1 u a t i on.................................................. 5-1 5.1 Di s cu s s i on an d R es u l t s........................................... 5-1 5.2 NAC-1 Desi gn Basi s Source Speci fi cati on.......................... 5-2 5.2.1 G amm a S o u r c e.............................................. 5 - 2 5.2.2 N eu t r o n S ou r c e............................................ 5 5.3 Model Specification............................................ . 5-2 5.3.1 Description of the Radial and Axial Shielding Configuration 5-2 5.3.2 Shi el di ng Regi ona l Dens i ti es.............................. 5-2 5.4 S h i el di n g E va l u at i on........................................ -.... 5 - 5 5.4.1 Shielding Evaluation for the Metallic Fuel................ 5-6 5.5 Appendi ces to the Shiel di n g Evaluati on........................... 5-8 5.5.1 References................................................5-8 5.5.2 ANISN I nput D ata, Tables 5-4, 5-5 and 5-6................. 5-9 5.5.3 Metallic Fue' Source Terms vs. Cool Tine (Table 5-7)..... 5-14. s 6.0 C r i t i c a l i ty E v a l u a t i o n............................................... 6 - 1 6. '1 D i s cu s s i on an d R es u l t s........................................... 6-1 6.2 P a c k a ge F uel L o a di n g............................................. 6-3 6.3 Appendi ces to the C ri ti cali ty Evaluati on............ -............. 6-6 I i i i

Il ' J TABLE OF CONTENTS, Contd' 7.0- Operating Procedures............................................. 71 7.1 Procedures for Loading the Package........................... 71 4 '7.2 Procedures for' Unloading the Package......................... 72 7,3 Preparation of the Empty Package for Transport............... 73 7.4 Procedures for Loading Sound and Failed Metallic Fuel........ 7 5 7.4.1 Cask Loading......................................... 7 5 7.4.2 Loading Metallic Failed Fuel into Canisters........... 78 7.5 Appendices to the Operating Procedures.. 7-5 8.0 Acceptance Tests and Maintenance Program.................... 8-1 8.1 Acceptance Tests............................................. 8-1 8.1.1 Visual Inspection..................................... 8 1 8.1.2 Structural and Pressure Tests......................... 81 8.1. 3 Le ak T e s t s..........................................., 8 - 2 8.1.4 Cask Components Test.................................. 82 i 8.1. 5. Tests for Shielding Integrity......................... ' 8 2 8.1.6 Thermal Acceptance Test....... 8-2 '8.2 Maintenance Program................ 8-3 8.3 Appendices to the Acceptance Tests and Maintenance Program... 84 8.3.1 References........................................... 8 4 8.3.2 Hydrostatic Pressure Test Procedure.................. 8 5 l 8.3.3 Shielding Test Procedure.............................. 8 10 5 8.3.4 Thermal Acceptance Test Procedure..................... 8 21 8.3.5 Worcester Valve Company Test Report................... 8-26 iv Revised March 1990

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2.10.3 Structural Evaluation of Failed Fuel Cans and Liners (baskets) This evaluation documents the thermal and structural adequacy of six element and'three element failed fuel cans and liners (baskets) for the NFS 4/NAC 1 Spent Fuel Shipping Cask. The maximum normal operating temperature is calculated to be 211'F. The conservatively calculated minimum margin of safety for any component is +0.49 The failed fuel cans and liners (baskets) are structurally adequate to satisfy all regulatory requirements. l 2.10.3.1 Discussion Nuclear Assurance Corporation proposes to ship three or six encapsulated l failed metallic fuel rods in the NAC-1 cask. A separate liner (basket) and failed fuel can have been designed for a three rod capacity shipment and for a six rod capacity shipment. In each case, the failed fuel can is a sealed, dry aluminum canister. Only one failed metallic fuel rod is placed in each failed fuel can. The failed fuel can must maintain containment for all loading conditions because it serves as one of the containment barriers required for the transport of high level radioactive material. Since it is not possible under any circumstances for the six metallic fuel rods to attain a critical configuration, the liner (s) (basket) acts as a convenient support and spacer in the cask cavity. No permanent deformation of the liner.is permitted. I Each metallic fuel rod weighs 125 pounds and is approximately 124 inches long. The total heat load for six metallic fuel rods is 30 watts. 2.10.3.2 Method of Analysis A one dimensional thermal analysis of the NAC-1 cask with a total heat load of 30 watts was performed using the SCOPE thermal analysis computer program to determine the maximum normal operating temperature of the failed fuel can and the liner (basket). 2 97 Revised March 1990

r.- \\ Classical stress analysis methods are used to evaluate the failed fuel cans for bucklin6 during the end irapact and for bending during a side impact. The tubes in the liners (baskets) are also analyzed for bending during a side impact. The impact loadings include the g factors determined in the NAC 1 cask Safety Analysis Report (SAR). The calculated stresses in the failed fuel cans and the liners are conservatively compared to the material yield strength to demonstrate that containment is maintained by the failed j fuel cans and that no permanent deformation of the failed fuel cans or the liners (baskets) occurs. HI 2.10,3.3 Input Geometry & Data 1. Total Heat Load - 30 Watts (For Six Metallic Fuel Rods). 2. Metallic Fuel Rod Veight - 125 Pounds / Rod). l 3. NAC 1 Cask Geometry: (Ref. SAR, Section 1.3.3) Inner Shell (Cavity) I,D. - 13.50 inches Inner Shell Thickness 0.313 inches Lead Shell Thickness 6.63 inches Outer Shell Thickness 1.25 inches Neutron Shield Thickness 4.5 inches Neutron Shield Shell Thickness 0.165 inches 1 4 Free Drop Impact G Loads: (Ref. SAR) Normal Operation

1 Foot Side Drop 12.3g (Pg. 2-37)

Accident 30 Foot Top End Drop 44.67g (Pg. 2-47) 30 Foot Bottom End Drop 76.6g (Pg. 2-53) 30 Foot Side Drop 96.0g (Pg. 2-67) 30-Foot Corner Drop 45.5g (Pg. 2-72)

Ref. 10 CFR 71 and Regulatory cuide 7.8.

2 - 98 Revised March 1990 em r

I 2.10.3.4 Mechanical Properties of Materials 1. 6061 T6 Aluminum Alloy (Ref. MIL HDBK 5E) ] 'g; S - 42 ksi (70'F) -S - 35 ksi (70'F) (Page 3 222) u 770 g 70 At 250'F: S - 0.86(S") - 36.1 ksi (Page 3 227) 7250 70 S - 0,88(S ) - 30.8 kai (Page 3 228) 7250 70 2, 6063 T832 Aluminum Alloy (Ref. ASME B210) S - 40 ksi S - 35 ksi (Page 194) ug y S"250 - 0.86*S"70 At 250*F: - 34.4 ksi S - 0.89*S - 30,8 ksi l Y250 Y70 lI 2.10.3.5 Thermal Evaluation The SCOPE thermal analysis computer program is used to evaluate the NAC-1 cask containing the 2.75 inner diameter (I.D.) failed fuel can liner and six l '2.75 I.D. Failed Fuel Cans loaded with one metallic fuel rod each for a l ~ total heat load of 30 watts. l Section 3.6.2.2 preserts the input data 'and the resulting calculated temperatures from the SCOPE analysis. The calculated temperature for the failed fuel can and the liner (basket) is 211'F.

The strength variation with temperature is assumed to be the same as that for 6061-T6 Aluminum Alloy 2 - 99 Revised March 1990

m R 2.10.3.6 Structural Evaluation The failed fuel cans are evaluated to demonstrate that containment of the failed fuel rod is maintained for all loading conditiont,. The maximum stress occurs in the shells of the failed fuel cans for the 30 foot side. ' drop load case. Buckling of the shells is evaluated for the 30 foot end drop load case. The liners (baskets) for the failed fuel cans are evaluated to demonstrate that rupture (ultimate failure) does not occur for any loading condition, The maximum stress occurs in the' housing (tube) of the liners for the 30-g, 3 foot side drop load case. 2.10.3.6.1 Failed Fuel can 2.75 Inner Diameter 2.10.3.6.1.1 Shell - Bending (Ref. Dwg. 340 108 D2, Rev.) Londine l l 30 Foot-Side Drop Acceleration - 96 g Support Spacing - 33.66 in (Ref. DN. 491 001, Rev. 1) Weight: Fuel - 125 lb/124 in - 1.008 lb/in Shell-{(3.0 2.75]1x0.10-0.113lb/in 2 2 Total - 1.121 lb/in Conservatively assume the shell is simply sur'orted at the support disks; then, the moment during impact is: 1' M1' x 1.121 x 33.66 x 96 - 15241 in lb 2 I 2 - 100 Revise 1 March 1990 I' l

l I She11 Properties: [ I/C - 0.7791 in Material Properties: (ASTM B221 Type 6061 T6) I (S ]250 - 30.8 ksi 7 Stresses: Sb - 15241/0.7791 - 19562 psi M.S.-(S] /Sb ~ 15L11 y 2.10.3.6.1.2 Failed Fuel Can - 2.75 Inner Diameter [ Shell - Buckling Estimated Weight of Can Assembly - 20 lb I Bottom End Drop Acceleration - 76.6 g The compressive stress in the shell due to its weight during. impact is: 20 (76.6) S* - 2.75]-1357 psi [(3.0 2 2 The margin of safety on yield is: f-M.S. 1 j - 1 12Ll 0 i I. 2 - 101 Revised March 1990 ?

[ i i ( !.[ gi [ The buckling of the cylindrical shell under the auction of uniform axial u I compression may be evaluated using equation 11 1 on page 458 or Lheory of Elastic Stability by Timoshenko and Gere, t 4 Et r,3(1 v { 0.5 2

f 3

where, E - 0.97 (10.1) - 9.8 kai at 250'F t - 0.125 in l

r - 2.81 in, env. radius for all tubes in assembly v - 0.33 g 5. The margin of safety on buckling is: M.S. - Scr/8

1" E c

g; 2,10.3.6.1.3 Conclusion E,' The stress in the cylindrical shell caused by a bottom end drop with 76.5 g deceleration is much lower than the yield stress and the critical buckling

stress, 2.10.3.6.2 Liner Failed Fuel Can 2.75 Inner Diameter

. Housing Bending (Ref. Dwg. 491 001, Revision 0) Loading 30 Foot Side Drop Acceleration - 96 g Support Spacing - 33.66 in I. 2 - 102 Revised March 1990

e I Weight: Fuel - 125 lb/124 in - 1.008 lb/in I AluminumShell-{(3 2.75]x1x0.10 - 0.113 lb/in f 2 2 Housing-[(3.75 3.5 ] x 1 x 0.10 - 0.142 lb/in 2 2 Total - 1.263 lb/in I conservatively assume the housing simply supported at the disk supports: then, the moment during impact is: M=f(1.263x33.66]96-17,172inlb 2 Housing Properties: I/C - a d,' d' /32d, g (3.75' 3.5')/(32 x 3.75) - 1.2485 inch3 -w Material Properties: (ASTM B210 Type 6063 T832) (S ) - 30.8 ksi 7 250 I Stresses: Sb - 17172/1.2485 - 13,754 psi M.S. - (S ) /Sb - +1.24 y The other accident drop conditions are: 30 Foot Top End Drop a - 44.67 3 30 Foot Bottom End Drop a - 76.6 g 30 Foot Top Corner Drop a - 45.5 g 2 103 Revised March 1990

I! ,a l Compared to a 30 foot side drop a:cident condition, with a - 96 g, the above I accident conditions are less critical. Therefore, neither rupture nor yielding of the liner housing vill occur. 2.10.3.6.3 Failed Fuel Rod Can 4.00 Inner Diameter Shell Bending (Ref. Dwg 340 108 D1, Rev 9) Loading 30 Foot Side Drop Acceleration - 96 g Support Spacing - 50.37 in i L'eight : Fuel - 125 lb/124 in - 1.008 lb/in I; Shell-{(4.25 4.0)x1x0.10-0.162lb/in 2 2 Total - 1.170 lb/in Conservatively assume the shell is simply supported at the support disks, then, the moment during impact is: M-f(1.170x50.37]96-35,621inlb 2 Shell Properties:-(4.00 I.D., c - 0.125 in) 0 0 3 I/C - x (4.25 4 )/(32 x 4) - 1.7243 in Material Properties: (ASTM B221 Type 6061 T6) (S ) - 30.8 ksi Y 250 2 104 Revised March 1990

I 1 I Stresses: Sb - 35621/1.7243 - 20,658 psi M.S. - (S ) /Sb - +0.49 j y l 2.10.3.6.4 Liner 3 Element NAC 1 Cask i Tube Bending I (Re t'. Dwg 347 211 F19, Rev. 5) Loading 30 Foot Side Drop Acceleration = 96 g Support Spacing - 50.37 in Veight: Fuel - 125 lb/124 in - 1.008 lb/in Shell-{(4.25 4.0)x1x0.10 - 0.162 lb/in 2 2 Tube ={(5.625 5.375 ] x 1 x 0.1 - 0.216 lb/in 2 2 Total - 1.386 lb/in Conservatively assume the tube is simply supported at the support disks; then, the moment during impact is: ~ M-f(1.386x50.37]96-42,198inlb 2 Tube Properties: I/C - x (5.625' 5.375')/(32 x 5.625) - 2.9053 in 3 au 2 105 Revised March 1990

i 1 Material Properties: (ASTM S210 Type 6061 T6) i (S ) - 30.8 ksi Ill 7 250 Stresses: Sb - 42198/2.9053 - 14,524 psi M.S. - (S ) /Sb - +1.12 250 I 2.10.3.7 Results and Conclusion The maximum normal operating tettperature of the failed fuel cans and the liner (basket) is calculated to be 211'F. The structural evaluation conservatively uses material properties at 250'F. The calculaced margins of safety are: I 1. Failed Fuel can 2.75 Inner Diameter Shell Bending +0.57 Shell Buckling +32.1 2. Liner Failed Fuel Can 2.75 Inner Diameter l Housing Bending +1.24 3. Failed Fuel Rod Can - 4.00 Inner Diameter I Shell - Bending +0.49 4 Liner 3 Element FAC 1 Cask Tube - Bending +1.12 1 2 - 106 Revised March 1990 I.

I No permanent deformation occurs in the failed fuel cans or the liners for the critical loading conditions. Containment of the failed metallic fuel I rods is maintained and the liner structure remains intact; therefore, structural adequacy is ensured. I 2.10.3.8 References I Safety Analysis Report for the NSF 4/NAC 1 Spent Fuel Shipping 1. Cask, Nuclear Assurance Corporation, June 30, 1984 2. NAC Drawing: 340 108 Dl, Revision 9, Failed Fuel Rod Can 4.00 I.D. I 3. NAC Drawins: 340 108 D2, Revision 9. Failed Fuel Rod can 2.75 I.D. I 4. NAC Drawing: 247 211 F19, Revision 5, Liner 3 Element, NAC 1 I. Cask. 5. NAC Drawing: 491 001, Revision 0, Liner Failed Fuel Can, 2.75 1.D., NAC 1 Cask. 6. LI MIL HDBK 5E, " Metallic Materials and Elements for Aerospace Vehicle Structures," U.S. Department of Defense, May 1989. 7. "ASME Boiler and Pressure Vessel Code," Section II, Material Specifications, Part B. Nonferrous Materials, The American Society of I Mechanical Engineers, 1989. 8. Timoshenko and Cere, Theory of Elastic Stability, 2nd Edition New York. McCraw Hill, 1961. 'I F. Kreich, frincieles of Heat Transfer, 2nd Edition, Scranton, PA., 9. International Textbook Company 1965. I 2 107 Revised March 1990 I

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l I

I I: I; I-I g. l' i 2 108 Revised March 1990 I:

I I -decay heat load of 750 watts was analyzed. The neutro6 shield tank I was assumed to contain dry air. The cask component dimensions describing the radial model is presented in Table 3-5. Table 3-5. SCOPE CASK RADIAL MODEL DESCRIPTION Thickness Radius from Cask Center (inenes) Component Material (inches) O_utside Surface Inside Surface . I Cast Sur7 ace 721 SS 0.165 19.600 19.435 Neutron Shield dry air 4.500 19.435 14.935 l Outer Shell 321 SS .1.250 14.935 13.685 Gama Shield Lead 6.523 13.6A5 7.063 I Inner Shell 321 SS 0.313 7.063 6.750 3.5 Thermal Evaluation of the Hypothetical Fire Accident The thermal model for the analysis of the hypothetical accident is basically identical to the model for the analysis of normal transport conditions. The I kmbient conditions are changed to remove the insolation and increcse the j temperature of the surroundings to 1475'F to represent the fire. Innediately j before and after the 30 minute fire, the ambient temperature is at 130'F. A 1 I summary of the cask and fuel temperatures are presented in Table 3-4 and 3-6 for the steady sttte (normal transport condition) and transient analysis (fire accident condition), respectively. The maximum temp-ature in the neutrc.n shielo tank (1354'F) occurs at the end of the fire acetdent. The 1 deal gas ins is used to detersics the maiximum pressure of the shield tank. Frior to fuel loading the average terperature of I the neutron shield fluid (air) is assun.ec to be -40*F (worst case), Using the ideal gas equation '2 " $1 X l. X El I I where: 11 = 40*r' = 420'R V2 T1 T2 = 1357'F = 1814'R i K1 = Y li -1 = 14.7 psia Thus: P2 = (14.7) x (1) x (1814/420) = 63.5 psia = 48.8 psig Table 3-6 SU" MARY OF TENFERATURES AND PRE!.URES FOR HYPOTHETICAL FIRE ACCIDENT CONDIT nh5 Without Shipping Container WAShipping Contairer Maximum Pressure Maximum Pressure Component lerperature ('F) (psig) Temperature ('F) (pzii Fuel Rod Surface 540 540 Inner Shell (Cavity) 419 16 419 17 I Leac 437 437 Outer She'l (Shield Tank) 1354 49 485 18 Su-face 1357 488 Shipping Container 1357 Revised 3-5 Feb. 1987 m m

. _.. _ _ _. _... ~. _. _ _ P Ii, I L Il I' Thit DS94 intentionally left blank. I' I iI I: L I I l I I: I I 36 Revised l March 1990

3.6 Appendices to the Thermal Evaluation This section presents references and othet information that supplements the I information presented in Section 3. 3.6.1 References 1. " Safety Analysis Report for Nuclear Fuel Services, Inc. Spent Fuel Shipping Cask Model No. NFS 4," Nuclear Fuel Services. Inc., U.S. NRC Docket 6698, September 1972. 2. I "ASME Boiler and Pressure Vessel Code, Section III Nuclear Vessels," i The American Society of Mechanical Engineers New York, NY, 1971. 3. Nuclear Systems Material Handbook, Volume I, Design Data, Revision 2 August 1, 1977. 4. Compilation of Thermal Property Data for Computer Heat Conduction Calculations, Edwards, A. L. (University of California, Lawrence Livermore Radiation Laboratory), February 24, 1969. F 5. Principles of heat Transfer, Second Edition, Kreich, F. (International I Textbook 6, Scranton, Penn.), 1965. 6. Jakob, M., and hawkins, C. A., elements of heat Transfer, John wiley and Sons Inc., New York, NY, 1951. I 3.6.2 SCOPE Input and Results I 3.6.2.1 Metallic Fuel Normal Transport and Hypothetical Fire Accident Conditions 3.6.2.2 Failed Metallic Fuel Hypothetical Drop Accident Condition I 3.6.3 Original SAR Design Basis Thermal Analysis Summary 3 6a Revised March 1990 I

Il 3.6.4 Original SAR Description of the TAP Computer Code and Thermal Model ) Description Original SAR Thermal Analysis for Normal Condition of Transport I 3.6.5 3,6,6 Original SAR Thermal Analysis for the Hypothetical Accident Conditions 3.6,7 Original SAR Thermal Analysis References I I l I l I l I 1 I I I I I 3 6b Revised March 1990 .~.

4, O I 3.6.2 SCOPE Input and Results 3,6,2.1 I Metallic Fuel Normal Transport and Hypothetical Fira Accident Condition l. 1 ,3 i I I I l l I I I l 3-7 Revised March 1990

~.- - e t$n.. INPUT (IN CA*D*1 MAGE F0" 1TI ISLLOWI8 l 1 ep g 4.21N vie, ?iDw. 2190D 5, g AS Att/W ALL S AP, ut e166.i e Watt 6 IEA* INI) l !.?!!;d' !i,2!,E iYs5!;iY3 Ib' "! ii"oEaI" ' 0 3 6.62 34.0 O ) P90PtR*tti Of PAttetALS tutttNfLT j4 twt [ATA (!$RART r PattelaL p!N13tt CONDU0ftVtti atAt CAPAtttY ftmpffAtutt timtt (APltAL tolf (LS/turt) (9tu/Mettt/st (pty/Lgift (f tntgg 11 (t/LS) i P4

08.56 1?.0000 5320 6?B 3.000 2

't &B* 26 2 6. *f4 0 .1200 19fb 2.000 ti 1189.*$ 15.0000 5230 1450 9 000 4 'u t19.15 210.0000 5990 1730 .360 AL 16*.49 140.*000 .2290 1050 .2t0 6 !? 494.43 11.0000 1200 1000 4.000 NA 4*.00 39.0000 1000 1400 .200 i Lt 39.00 20.0000 1.0000 1400 11.000 9 Ffi-t 146.60 4400 1560 1200 1.800 li t94C

3*.60 18.0000

.0320 610 .160 11 ALst 1**.00 10.0000 .2000 1065 .354 1* 09WA 6*.00 .0?60 1260 600 1.000 1* NE .01 1700 1.2400 1600 .000 14 Att .04 .0360 .2600 1A00 .000 1* M*0 6*.43 .2920 1.0000 250 .000 16 N'iL1 81.20 1000 .0660 1000 .000 1* 1NOT M**.00 .3500 .St10 1600 .0C0 + 19 NUL1 '94.50 1.2000 .0660 1960 .000 19 NLwG 212.00 .?000 .1600 1290 .000 20 NUL2 20?.00 .6000 .0640 1000 .000 21 Pun 2*9.90 1.0000 .1000 1650 .000 22 ewa 1**.20

0000 1000 1650

.000 2T NLwt 11*.00 .2500 .220h 932 .000 No't: 'N!$ tot' WILL st *ettttt onLT 04tte EVEN TNOV4M tut U$tt 4AT HAVE MUL17PLE Sits Of INPUT DATA. [ r (mitt . 00 tatst-*-suuratt ERIS$tvit? 0F THE tatt r ein tas 10atts s) CASE DESIGN Llatts I I c f TfNmat 99*.0 etG.7 ffMM t. 4At!4U4 ALLOW ADL E $UR F Att itR>t R A T' Rt J wnut9t 99.0 t!LO.L83 WGutas.-q AsimUm eLLOWAOLE WE!6ht OF LO ADED C ASE TAmn 130.9 DES.T Tamp... 0Vil!St Angligt ffhPIRATUtt l Nt0LAt 1 450Lat**INCLVliOu 0F $0LAR IN$$L ANtt At 122.92 ttfu/NJi t' T* *2 (1st t S, 2* l Revised Feb. 1967 3-J El

==w+r-r w -* m A- - - - i

I I 3.6.2.2 Failed Metallic Fuel Hypothetical Drop Accident Condition The SCOPE thermal analysis cc.sputer program is used to evaluate the NAC 1 cask containing the 2.75 inner diameter (I.D.) failed fuel can liner c.nd 5.ix . I 2.75 1.D. Failed Fuel Cen loaded with one metallic fuel tod tach for a total heat load of 30 watts. The following pages present the input data and the resultitig calculated temperatures from the SCOPE analysis. The calculated temperature for the failed fuel can and the liner (baskat) is 211'F. Since the three-element liner and failed fuel cans will have one half of the six element heat lead, the temperatures will be significantly lower and, thus, less critical. I I I I I I I 3 - 14e Revised i March 1990 I

g e f i !*)8! :l!:' -:I 'll!.*ll!! Eltl?:.t?iXII: Itallat:: :: IIM ::::1;1:st i h::41 hel 1:d D it! ?S?ll !!!f X1H '.innenisiniin..uninuniinnuinienn,isiinisiin..... unio 'tillt I'll:P ':l! filit lillt !!:I !!G ltt![ ltil! 2 1500 2 0.0 5.0 6.)l.4 !.: N ( 1 ' till !*ll '!*ll I?ll lit:0 I?nt; ) ':!

H

!!.30 i '4:I!!! !I!!!!! '!:l!!! ::.l 't:Ir '[ git t+1gt i 2. 1 '. ).) 40: 1.*5 0.6 II;It

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c i 7.4 Procedures for Loading Sound and failed Metallic Fuel 7.4.1 Cask Leading I NOTE: The procedure for loading dry metallic sound (intact) fuel baskets, small diameter failed fuel canisters (FFCs) and large diameter FFCs into the cask is the same; however, only I thtee large diameter FFCs or three sound fuel baskets can be placed into the cask, while six of the smaller FFCs can be used. I 7.4,1.1 Perform a receiving survey of the empty cask and closed container and inspect for damage. 7.4.1.2 Position the trailer in the designated area. Set the trailer brakes and block the wheels against movement in either direction. 7.4.1.3 Remove the lid from the ISO container. I 7.4.1.4 Perform a Health Physics survey of the cask and adjacent surfaces l of the container. NOTE: ,I A receiving survey of the cask and transporter must be performed as soon as practicable after arrival at the site to ensure compliance with 10 CFR 71.87(1), 10 CFR 71.47, and to ensure timely reporting of any transportation noncompliance. 7.4.1.5 Remove the top impact limiter. 7.4.1.6 Remove the cask tiedown straps. 7.4.1.7 I Using the cask lifting yoke, engage the lifting trunnions of the front end of the cask. Raise the cask to a vertical position on the rear cask support, moving the crane as required to keep the cask engaged in the trailer rear rotation supports and the crane cable vertical. W en the cask is fully vertical, lift the cask -g from the container. l 3 75 Revised March 1990 LI l

. ~.. I: 7.4.1.8 Place the cask in the dry loading stand. Disengage the lifting yoke. ? 7.4.1.9 Remove the vent and drain valve port covers. Visually inspect the valved quick disconnect nipples and replace them, if necessary. 7.4.1.10 Remove the closure lid bolts. Attach the lid lift slings to the closure lid. Remove the closure lid and set it on a support that is suitable for radiological control and for maintaining the cleanliness of the closure lid. Inspect the o. ring seating g' surface on the cask, g 7.4.1.11 Visually inspect the inner cavity for foreign material or damage. Verify the proper cask cavity liner is in place. 7.4.1.12 Replace the closure lid, but not the lid bolts. 7.4.1.13-Identify the fuel to be loaded. Note: An ORICEN (or equivalent) analysis must be completed on the fuel bearing components prior to loading to ensure that the proposed contents conform to the Certificate of Compliance. E' 7.4.1.14 Place the carriage assembly on top of dry loading stand. i-7.4.1.15 Position the carriage assembly for cask lid removal. Connect the cask lid to the removal cylinder and te:nove the lid from the cask. 7.4.1.16 Position the carriage for fuel loading. 7.4.1.17 Match mark the carriage assembly and turntable to each fuel canister location in the cask cavity. For sound fuel, there will be three canisters with five rods each; with small diameter FFCs, there will be six canisters with one rod each; and with, large diameter FFCs, there will be three canisters with one rod each. 7.4.1.18 Move the carriage and turntable to fuel canister position 76 Revised March 1990 5.

i. s6 '. I i 7.4.1.19 After retrieving the fuel (sound or failed) from the pool, place the shielded transfer cask containing a fuel canister (as many as I five sound or one failed metallic fuel rods) onto the turntable. 7.4.1.20 Lower the fuel canister from the transfer cask into the shipping cask. E 12 R*p**t 8t'ps 7.4.1.18, 7.4.1.19 and 7.4.1.20 to access fuel e a canister positions 2 and 3 and complete the cask loading. For small diameter FFC loading, repeat for all six positions. 7.4.1.22 Replace both o rings on the closure lid. 7.4.1.23 Replace the closure lid onto the cask by re positionin5 the carriage assembly and actuating the cask lid cylinder. Visually confirm that the lid is properly seated. 7.4.1.24 Remove the carriage assembly from the loading stand. lI 7.4.1.25 Tighten all six closure lid bolts to 650 foot pounds. 7.4.1.26 Attach the pressure test fixture to the cask cavity vent port, Pressurize the cavity with air to 30 psig and hold for 10 minutes. i No pressure drop is permitted. 7.4.1.27 Using the pressure test fixture, pressurize the annulus between the two lid o ring seals to 30 psig through the o ring port. I observe the air pressure gauge for 10 minutes after closing the isolation valve. If no drop in air pressure is observed, the seal is acceptable. If air pressure drops, remove the valve port cover and replace the seals, then repeat the test. l 7.4.1.28 Close the vent, o ring and drain valves and then replace the valve port covers. 77 Revised March 1990 I

t i s ' <'. a 7.4.1.29 Install the tamper indicating seals (3) on the lid closure bolt

heads, i

7,4.1.30 Decontaminate the cask. Survey the cask for surface contamination and radiation dose rates. 7.4.1.31 Using the cask lifting yoke, transfer and lower the cask to the trailer. Engage the lower rotation trunnions with the rotation supports in the container. Lower the cask to rest on the upper I trunnion saddles, moving the cask lifting yoke as required to keep the crane cables vertical. Disengage the cask lifting yoke from the cask lifting trunnions and set it aside. 7.4.1.32 Install the cask tiedowns; then install the cask impact limiter. 7.4,1.33 Install the container lid. 7.4.1.34 Complete a Health Physics survey and record vehicle radiological compliance data. Ensure compliance with 10 CFR 71.87(i) and 10 CFR 71.47. Complete the shipping documents and apply placards and labels to the cask and container. 7.4.2 Loading Metallic Failed Fuel into Canisters 7.4.2.1 Small Diameter Canisters 7.4.2.1.1 Examine the small diameter failed fuel canister (FFC) and check it for damage. 7.4.2.1.2 Place the FFC inside the containment barrier portion of the pool. Position the FFC in the failed rod loading station. 7.4.2.1.3 After verifying the accountability records, place the designated failed fuel rod into the FFC. If the rod is broken into two or more pieces, verify that the lid thread and seal area is not fouled during rod insertion. 78 Revised March 1990

f I I 7.4.2.1.4 When the can is loaded, install the lid using the FFC Lid Installation Tool. 7.4.2.1.5 Using the FFC handling tool, move the loaded FFC through the containment barrier door and place the FFC horizontally into the upender. 7.4.2.1.6 Operate the hand winch to move the FFC to the vertical position. 7.4.2.1.7 Torque the FFC lid to 100 foot pounds for the small canister. I I 7,4.2.1.8 Connect the nitrogen supply line to the vent valve. 7.4.2.1.9 Open the nitrogen supply valve and pressurize the FFC to force out the water. Blow gas through the FFC for 5 minutes after the I first visible traces of bubbles appear. Remove the gas supply line. 7.4.2.1.10 Invert the FFC in the upender and install the pipe pluS-7.4.2.1.11 Re invert the FFC in the upender. 7.4.2.1.12 Attach the vacuum pump to the FFC vent valve. Evacuate the FFC to a pressure below 1 inch of mercury for 15 minutes. Remove the vacuum pump and backfill with nitrogen. LI 7.4.2.1.13 Remove the FFC from the upender and place it into temporary l storage. 1 7.4.2.2 Large Diameter Canisters 7.4.2.2.1 I Examine the large diameter FFC and check it for damage. 7.4.2.2.2 place the FFC inside the containment barrier portion of the pool. Position the FFC in the failed rod loading station. 79 Revised March 1990

Il ~., t 7.4.2.2.3 This step is to be used when loading a previously canned fuel rod into the large diameter canister. After verifying the accountability records, remove the ceramic filter from the top of the original faileo rod can. Position the can plug with aluminum screen onto the open can. Install the plug. 1 l 7.4.2.2.4 Verify the accountability records for the fuel to be loaded. .j 7.4.2.2.5 Place the designated fuel into the FFC. If the rod is broken into two or more pieces, verify that the lid thread and seal area is not fouled during rod or can insertion. 7.4.2.2.6 k* hen the canister is loaded, install the lid using the FFC Lid Installation Tool. 7.4.2.2.7 Using the FFC handling tool, move the loaded FFC through the i containment barrier door and place the FFC horizontally into the upender. 7.4.2.2.8 Operate the hand winch to move the FFC to the vertical position. 7.4.2.2.9 Torque the FFC lid to 130 foot pounds for the large canister. 7.4.2.2.10 Connect the nitrogen supply line to the vent valve. 7.4.2.2.11 Open the nitrogen supply valve and pressurize the FFC to force out the water. Blow gas through the FFC for 5 minutes after the first visible traces of bubbles appear. Remove the gas supply line. 7.4.2.2.12 Invert the FFC in the upender and install the pipe plug 7.4.2.2.13 Re invert the FFC in the upender. I 7 10 Revised March 1990

I .1 % I I 7.4,2.2.14 Attach the vacuum pump to the FFC vent valve. Evacuate the TTC to a pressure below 1 inch of mercury for 15 minutes. Remove the vacuum pump and backfill with nitrogen. j i 7.4.2.2.15 Remove the FTC from the upender and place it into temporary storage. I I I I l I ' I l 7 11 Revised March 1990

~_...._. i e' A,, 7.5 Appendices to the Operating Procedure Deleted L i t 4 ) Ii i i k I I-I I R 7 12 Revised March 1990 W* g w- ,-w-3--e-e w y, ,,,,,,,,}}

ML20012D172

Text

SUMMARY

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