ML20211P373

Revised Pages to SAR

ML20211P373
Person / Time
Site:	07109183
Issue date:	02/28/1987
From:	NAC INTERNATIONAL INC. (FORMERLY NUCLEAR ASSURANCE
To:
Shared Package
ML20211P358	List: ML20211P354 ML20211P373
References
27879, NUDOCS 8703020352
Download: ML20211P373 (34)

Text

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l TABLE OF CONTENTS Page 1.0 Ge ne ra l I n f o rma t i on...................................................

1 -1 1.1 I n t r odu ct i o n.....................................................

1 - 1 1.2 P a ck a ge De sc ri p t i on.............................................. 1 -1 1.2.1 P a ck a g i n g.................................................

1 - 2 1.2.2 Operational Features......................................

1-8 1.2.3 Contents of the Package...................................

1-8 1.3 Appendi ces to Gene ral I nfo rmati on............................... 1-11 1.3.1 R e f e re n c e s................................................ ' l - 1 1 1.3.2 Revision, Refornetting and Consolidation Reference.......

1-12 1.3.3 NAC-1/NFS-4 License Drawings, E10080, Sheets 1 through 4, R evi s i o n 20...........................................

1-20 2.0 S t ru ct u ra l E va l u at i o n.................................................

2-1 2.1 S t ru ct u ra l D es i g n................................................ 2-1 2.1.1 D i s cu s s i o n....................,......................... 2 - 1 2.2 Wei ghts an d Cente rs ' of Gravi ty................................... 2-3 2.3 Mechani cal Prope rti es of Mate ri al s............................... 2-4 2.4 General Standards for all Packages...............................

2-5 2.4.1 Ch emi cal and Gal vani c Reacti ons........................... 2-6 2.4.2 P os i t i ve C l os u r e.......................................... 2 - 6 2.4.3 L i f t i n g D e vi ce s...........................................

2-7 2.4.4 Ti e D own D ev i ces.......................................... 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 No rmal Condi ti ons of Transport.................................. 2-20 2.6.1 Heat.....................................................

2-20 2.6.2 Co1de....................................................

2-27 2.6.3 Pressure - 0.5 times Standard Atmospheric Pressure.......

2-31 2.6.4 Vibration - Vibration Narmally Incident to Transport..... 2-31 2.6.5 Wat e r S p r ay.................... ;......................... 2 -31 2.6.6 F r ee D r op................................................ 2 - 31 2.6.7 C o rne r D r o p.............................................. 2 -40 2.6.8 Penetration..............................................

2-40 2.6.9 C omp r es s i o n.............................................. 2 -41 0703020352 670210 DR ADOCK 07109193 j

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TABLE OF CONTENTS, Cont.

Page 2.7 Hypotheti cal Acci dent Condi ti ons................................ 2-41 2.7.1 F r ee D r op............,................................... 2 - 41 2.7.2 Puncture.................................................

2-80 2.7. 3 Th e ma 1.................................................. 2 -8 5 2.7.4 Wat e r Imme rs i on.,........................................ 2-90 2.7.5 Susum ry of D amage........................................ 2-90 2.8 S peci a l F o nn.................................................... 2 -90 2.9 Fu e l R o ds....................................................... 2 - 90 2.10 Appendi ces to the S t ructu ral E val u ati on......................... 2-91 2.10.1 References..............................................

2-91 2.10.2 Bal s a P rope rti es........................................ 2-93 3.0 Thermal Evaluation..............................'......................

3-1 3.1 Discussion.......................................................

3-1 3.2 The rmal Properti es of Materi al s.................................. 3-2

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3.3 Techni cal Speci f1 cati on of Conquonents............................ 3-3 3.4 Thermal Evaluation for Normal Transport Conditions................ 3-3 3.4,1 Descri pti on of th e SCOPE P r6 gram........................... 3-4 3.4.2 Cask Model Desc ri pti on..................................... 3-4 3.5 Thermal Evaluation of the Hypothetical Fire Accident.............

3-5

.1. 6 Ap pendi ces to Thermal Eva l u ati on................................. 3-6 3.6.1 R e f e ren c es...............................................

3 - 6 3.6.2 SCOPE Input and Results for Metallic Fuel................. 3-7 3.6.3 Original SAR Design Basis Thermal Analysis Summary.......

3-15 3.6.4 Original SAR Description of the TAP Computer Code and Thermal Mode 1............................................

3-19 3.6.5 Original SAR Thermal Analysis for Normal Conditions of T ra ns po rt............................................. 3 -48 3.6.6 Original SAR Thermal Analysis for Hypothetical Ac ci dent Co ndi ti ons...................................... 3 -61 3.6.7 Original SAR TherJnal Analysis References................. 3-71 4.0 C ont a i nmen t........................................................... 4 - 1 4.1 Cont ai nment B ounda ry............................................. 4-1 4.1.1 Contai nmen t Ves se1........................................ 4-1 4.1.2 Containment Penetrations..................................

4-1 Revised 99 Feb. 1987 h

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LIST OF TABLES Page Table 1-1 Fuel Assembly Characteri stics..............................

1-8 1-2 Non-Fuel Material Characteri sti cs..........................

1-9 Table 2-1 Major Component Weights of NAC-1 Cask......................

2-3 2-2 Weights of NAC-1 Cask......................................

2-4 2-3 Mechanical Properties of Material s.........................

2-4 2-4 Summary of Evaluation of General Standards..'...............

2-5 2-5 Summary of Analysis Results for Type B and Large Quanti ty Packagi ng Load Condi tions........................

2-16 2-6 Comparison of the Effects of the Reference PWR Fuel and the Metallic Fuel (130*F) During Normal Transport.....

2-20 2-7 Comparison of Calculated Stress to Allowable Stress for Normal Transport Condi tions...........................

2-21 2-8 Comparison of Calculated Stress to Allowable Stress 1

for Hypothetical Acci dent Condi tions......................

2-42 2-9 Comparison of the Post Fire Accident Effects of the Reference PWR Fuel and the Metallic Fuel..................

2-86 i

Table 3-1 Comparison of Reference PWR and Metallic Fuel Characteristics............................................

3-1 3-2 Comparison of the Maximust Temperature and Pressure Effects of the Reference PWR and Metallic Fuels............

3-2 3-3 Me tal l i c Fuel Descri ption..................................

3-4 3-4 Summary of Temperatures and Pressures for Normal Transport Conditions.......................................

3-5 j

3-5 SCOPE Cask Radial Model Description........................

3-5 3-6 Summary of Temperatures and Pressures for Hypothetical Fire Accident Conditions...................................

3-6 Table 4-1 Fission Gases (C1) vs. Cool Time (Per 1000 kg. Initial Natural.U, o.7115 W/0 U-235................................

4-8 Table 5-1 S umma ry of Radia ti on Dose Rate s............................

5-2 5-2 ANISN Mixing Table and Material Identification for Neutron Shielding Calculations.............................

5-4 5-3 Comparison of the PWR and Metallic Fuels...................

5-6 5-4 ANISN 15$ and 16* Array Data for Neutron Shielding Calculations..............................................

5-10 5-5 ANISN Geometry Data for Neutron Shielding Calculations....

5-11 5-6 ANISN Angular Quadrature Constants for Neutron Shielding Calculations....................................

5-13 5-7 Metallic Fuel Source Terms vs. Cool Time..................

5-14 Table 6-1 Summary of Criticality Evaluation Ft::ile Class I..........

6-2 Table 8-1 Periodic Main tenance Schedul e..............................

8-3 Revised vii Feb. 1987

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STSTUMMERSHEy3(REP) 1.15 BALSA DNtDER PLATE UPPORTGUSSET TTOM luPACT UutTER GASKET R tMPACT uutTER CAVITY UPPER FLANGE CANtT1LDNER FLANGE LE AD GAMMAGHIELD EDTATION TRUNN10M(2) uD spacT uMtTER uPAcT uutTER ATTACH BOLT (4) CAGK UD Figure 1-1.

NAC-1 CASK CONFIGURATION

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prior to shipment.

This penetration does not directly communicate with the cask cavity because the inner 0-ring always provides a seal between the cavity and this penetration.

However, this penetration is considered as part of the contai nment boundary to assure that there is no loss of containment curing shipment.

Valve access is provided by a 5 inch inside diameter thin wall tube extending from the outer shell.of the cask to the outer perimeter of the lower impact limiter.

The recessed position of the valves protects them under accident conditions. Each 5 inch diameter port cover captures an elastomer 0-ring seal, providing a secondary seal.

The cask shielding consists of the cask inner and outer s' hells and the lead that is cast in place between them.

This shielding attenuates gamma radiation from

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the cask cavity contents, and consists of 6.625 inches of lead and the two stainless steel shells that are 0.3125 and 1.25 inches thick.

The ends of the cask are entirely stainless steel, 8.0 and 7.5 inches thick at the bottom and top respectively.

The thickness of the lead is reduced by 1.25 inches in the upper 25 inches of

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the cask wall to reduce the weight of the cask.

This reduced shielding is adequate to attenuate the radiation emitted by the cask contents in this region.

A stainless steel encased void region is provided at the lower end of the lead region to provide a volume that is avatlabrie to accommodate any swelling of the lead due to thermal expansion or slumping'of the lead during an impact.

There is also a small void between the ends of the lead and the end casting which is a result of the shrinkage of the lead that is solidified after pouring.

The NAC-1 cask has been designed and constructed with a neutron shield tank and an expansion tank.

This arrangement would allow neutron radiation from Light Water Reactor (LWR) spent nuclear fuel to be attenuated by borated water in the neutron shield tank.

Water serves to reduce the energy of the neutrons so that the boron is more effective for absorbing the neutrons. The water is mixed with 9thylene-glycol to form a solution that has an extremely low freezing point to avoio the volumetric expansion that accompanies the phase change.

The shield tank and expansion tank are divided into four separate pairs of compartments which are independent to minimize the risk of a single failure causing complete loss of neutron shielding.

Each section of the shield tank is separated from the others by radial gussets that are seal welded in place and supported by adjacent stiffeners.

The stiffeners have lightening holes to reduce their weight and avoid stagnating any shield tank fluid. Each section of the shield tank is connected to a section of the expansion tank to provide volume to accommodate thermal expansion of the fluid.

The openings between the shield tank and expansion tank are covered by baffles to assure that water covers the opening at all times and in all orientations.

Similarly, the expansion tank has baffles in the lower portion of each section to minimize splashing of the fluid which could either entrain air or permit the opening to the shield tank to be uncovered.

Each compartment of the shield tank is protected by a rupture disk that is designed to relieve pressure that exceeds 100 psig.

1-5

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I During the operation as proposed in this application, the neutron shield tanks would be dry. The contents of the cask, as described in Section 1.2.3, below, do not produce significant neutron radiation; therefore no neutron shielding is required.

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Heat dissipation within the NAC-1 cask is accomplished by passive means. All heat released by the contents is transferred to the cask wall by a combination of conduction, convection and radiation.

The heat is then conducted through the wall of the cask where copper fins at the interfaces between the lead and stainless steel walls provide a path to transfer heat across any gap that develops between the lead and the stainless steel shells due to their different rates of thermal expansion.

The configuration of the copper fins and their attachment to the stainless steel walls is shown in Figure 1-3.

I A combination of conduction, convection and radiation in the shield tank will transfer the heat from the cask wall to the outer surface of the shield tank.

Convection and radiation transfer the heat to the shipping container that is used in the normal transport configuration.

Natural circulation of ambient air and radiation transfer the heat from the shipping container to the environment.

The impact limiters are a part of the outer surface of the cask and transfer i

heat to the environment; however, the presence of the wood (an insulator) in the impact limiter severely limits the amount of heat that follows this path.

The impact limiters are sacrificial

===h-s attached to both ends of the cask to absorb energy by crushing during impact _

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An impact limiter is locate'd at the Tower end of ths cask body.

The impact i

limiter is a stainless steel ring which surround the cask lower casting.

It contains a balsa wood disc placed adjacent to the cask bottom.

A 1/8-inch sheet of asbestos is positioned between the balsa and the cask bottom. Within

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the impact limiter, extending radially from the center of the cask, are eight 3/8-inch thick stainless steel gussets.

The bottom section of the impact limiter also functions as a pedestal for supporting the cask in the vertical j

position.

An impact limiter is located at the upper end of the cask body to absorb the j

energy of the design basis side drop accident. The impact limiter consists of a stainless steel-sheathed, balsa-filled ring which surrounds the cavity flange.

An 1/8 inch section of asbestos is positioned between the balsa and the cask outer shell.

Within the impact limiter, extending radially from the center of the cask.are eight 3/8-inch thick stainless steel gussets.

The upper impact limiter is also a cask support member during transport, resting in the trailer cradle frame crossmember.

i The cask lid is protected during impact by a 12-inch thick balsa filled impact i

limiter that covers and overlaps the cask lid and cavity flange. The balsa is enclosed within a 0.104-inch thick stainless steel container.

A 1/8-inch sheet of asbestos is positioned between the balsa and the steel sheet metal adjacent to the cask.

The impact limiter is attached to the cask lid by four 1-inch diameter bolts. There are elastomer 0-rings in grooves under the heads of the 1-inch bolts and a neoprene gasket on the perimeter of the impact limiter to prevent surface contaminants from escaping from the outer surface of the cask lid.

Revised 1-6 Feb. 1987

1 Weld Lead Copper Stainless Steel Typical Cross Section of Fin t

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Section Through Cask Body Figure 1-3.

COPPER FINS CONNECTING LEAD AND STAINLESS STEEL 1-7

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Two 8 5/8-inch diameter by 3-inch long flanged trunnions are located on the perimeter of the upper impact limiter. The cask is lifted by a special handling yoke attached to the two trunnions.

Located on either side of the lower impact limiters are two 6 5/8-inch diameter by 3-inch long trunnions for rotating the cask to and from the horizontal position on the cask trailer.

The lower trunnions are offset from the cask centerline so that when the cask is lowered onto the container it will rotate to a horizontal position as the crane hook is lowered.

1.2.2 Operational Features The NAC-1 cask is designed to be easily loaded and handled at any nuclear facil-ity. The outer surfaces of the cask have been bead blasted, and the configura-tion of the surfaces selected to aid in decontamination.

Likewise, the loca-tions of the drain and vent valves were selected to permit rapid access and easy operation.

The cask does not permit the release of any of its coolant during either normal operation or hypothetical accident conditions.-

This reduces envi ronmental impact of spent fuel transport and provides economic benefit.

Defective fuel assemblies do not need to be encapsulated prior to shipment, eliminating time, expense and total radiation exposure.

The loaded cask, tractor and trailer weigh less than 73,280 pounds permitting unrestricted travel at night and on weekends This improves overall shipment efficiency and cask utilization. Since:the cask is mounted on its own dedicated trailer, it is independent of the need for-a rail connection to the site.

1.2.3 Contents of the Package The NAC-1 can transport irradiated natural uranium fuel as described in Table 1-1 or irradiated control components, hardware or waste if the maximum fission product radioactivity does not exceed the quantities presented in Table 1-2.

Table 1-1.

FUEL ASSEMBLY CHARACTERISTICS Naturally Enriched NRX-type Metallic Fuel Number of Assemblies 21 Fuel Assembly

Description:

(each)

Envelope (inches) (diameter) 1.36 Fuel Length (inches) 120.5 Enrichment (%)

.712, Fuel Weight (KgU) 54.5 Cool Time 365 days The maximum fission product radioactivity of the material that the NAC-1 may contain is, in accordance with 10 CFR 71 Appendix C, as follows:

Revised 1-8 Feb. 1987

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Table 1-2.

NON-FUEL MATERIAL CHARACTERISTICS Element Mixed Fission Products Transport Group II Radioactivity (Curies) 3.07 x 106 In addition to typical fission product radioactivity the material planned for shipment in the NAC-1 cask will emit neutron radiation which originates from 1) spontaneous fissioning of transuranium elements, 2) Y, n reactions with light nuclei, and 3) subcritical multiplication / fissioning of fissile nuclei.

The maximum total neutron source strength that the NAC-1 cask may contain resulting from any one or all of the above sources is:

5 neutrons /sec 2.29 x 10 The NAC-1 cask may contain the following maximum quantity of fissile constituents during one shipment of twenty-one (21) natural enrichment metallic fuel assemblies:

Fissile Constituent Quantity (Kg)

U-235 8.15 Kg The NAC-1 cask will contain, in solid metallic form, up to 21 intact metallic fuel assemblies; or up to 10 defective assemblies.

All intact assemblies will be of fixed diumnsion and supported by a fuel basket. Defective assemblies will be separately canned and then inserted 1sia s'pecial fuel basket for shipment.

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The maximum weight of the NAC-1 cask contents is 3300 lbs excluding internal coolant and the fuel basket.

Both the basket and the coolant within the cavity i

are considered to be a part of the contents of the cask because there is the possibility that either can acquire a part of the radioactivity of the fuel assemblies that are being transported.

When non fuel bearing components are being transported the coolant and basket retain their designation as cask contents because they may contain radioactivity from previous shipments.

The original SAR maximum design heat load for the NAC-1 cask is 11.4 kw (42,000 Btu /hr).

Certain calculations described in this report were performed at this maximum value.

The maximum allowable heat load for the cask metallic fuel contents is.75 kw.

This difference in the maximum design heat load and the maximum allowable heat load for the contents results from:

e the high thennal energy of 150 day cooled LWR fuel e the low thermal energy of the natural enrichment metallic fuel e removal of the neutron shield tank liquid Under normal operation the spent fuel decay heat is transferred from the fuel assembly to the inner cask liner by conduction and by natural convection of the contained air; through the cask inner stainless steel shell, lead shielding, and outer shell by conduction; across the neutron shield region by natural conduction, convection and radiation; and from the outer neutron shield containment tank surface to the surrounding air environment by convection and radiation.

Revised 1-9 Feb. 1987

i With the cask enclosed in its shipping container, in an ambient temperature of 130*F, and a heat load of.75 kw, the maximum transport temperatures and pressures are:

Temperature Pressure Cask Cavity 244*F 10 psig Outer Cask Surface 210*F Under the hypothetical accident conditions, the cask is subject to a fire at a temperature of 1475'F for 30 minutes. The empty neutron shield tank acts as a thermal shield thereby retarding the flow of heat from the fire to the cask structure.

During and following the fire, the decay heat of the fuel is transferred across the empty neutron shield tank by radiation; heat transfer through and from other regions and surfaces is the same as under normal conditions.

The maximum operating conditions resulting from the fire accident are:

Temperature Pressure Cask Cavity 419'F 16 psig i

Outer Cask Surface 1357'F The cask is designed to retain its contents under this pressure as well as under all other aspects of the defined; hypothetical accident (i.e., 30 foot drop test -40 inch puncture, etc.).

As a result, the cask provides a zero release system under both normai operating and hypothetical accident conditions.

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1

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Revised Feb. 1987 1 - 10

i 3.0 Thermal Evaluation The safe transport of spent nuclear fuel requires that the heat generated by radioactive decay within the fuel assemblies be transferred to the environment.

The NAC-1 spent fuel shipping cask has been designed to provide assurance that the fuel assemblies are adequately cooled during all phases of' normal operation as well as during a hypothetical accident that involves a fire.

This chapter presents the thermal analyses which demonstrate the adequacy of the cask to withstand normal operating conditions as well as survive the hypothetical accident scenario.

3.1 Discussion The NAC-1 cask design is based on the requirement that the cask reject the heat of a reference 150 day cooled light water reactor PWR assembly having a thermal output of 11.5 kw. The ability of the cask to reject the heat of the reference PWR fuel and to survive the hypothetical fire accident was established by the analysis in the original Safety Analysis Report prepared by Nuclear Fuel Servi ces, Inc. (Reference 3.6.1.1).. and was subsequently verified by thermal testing during the period the casks operated under Certificate of Compliance No. 6698.

The qualifying analysis, using the TAP-N Computer Code, and the Code modelling of the cask are presented in the appendices.

y.

The TAP-N analysis established bounding t;superatures for various components and these components were selected to havel the capability of operating in the thermal and pressure conditions established by the design basis or reference PWR assembly.

c.r The thermal and pressure conditions that result as a consequence to transporting the natural enri chment metallic fuel are based on a SCOPE analysis.

A descriptic, of this analysis is provided in Section 3.4.

For convenience, the characteristics of the design basis PWR fuel and the i

natural enrichment metallic fuel are compared in Table 3-1.

Table 3-1.

COMPARISON OF REFERENCE PWR AND METALLIC FUEL CHARACTERISTICS Reference Metallic PWR fuel Number per shipment 1 assembly 21 rods Average Burnup MWD /MTU 35,000 1600 Cool time (days) 150 365 Decay heat (kw) 11.5

.75 Fission gases (curies) 4950 609 Neutron Source (n/sec) 4.609 x 108 2.289 x 105 The thermal energy of the reference PWR fuel required that the PWR assembly be shipped with the cask cavity partly filled with water to limit.the temperature Revised 3-1 Feb. 1987

n i

of the PWR assembly rods.

The metallic fuel is shipped dry (without water in the cavity).

Cask component and structure temperatures are lower with the metallic fuel than with the reference PWR fuel.

The neutron flux of the reference PWR required that the cask neutron shield tank contain borated water to reduce the neutron dose rate.

This neutron shield was t

assumed to be lost in the accident condition.

The neutron flux rate of the metallic fuel is sufficiently low that no neutron shield is needed; therefore, the neutron shield is also left dry.

Table 3-2 susanarizes the maximum temperatunts and pressures for nomal and hypothetical accident conditions for the reference PWR and-metallic fuel.

Metallic fuel results assume transport in the shipping container. This table shows that the maximum temperature and pressure effects resulting from transport of the metallic fuel are less.than the effects of transport of the reference PWR fuel. Since the. cask valves and seals are adequate to the reference PWR fuel conditions they are also adequate to the metallic fuel conditions.

Table 3-2.

COMPARISON OF THE MAXIMUM TEMPERATURE AND PRESSURE EFFECTS OF THE REFERENCE PWR AND METALLIC FUELS Normal Transport (130*F)

Hypothetical Accident-Reference Metallic Reference Metallic

'PWR fuel-Fuel PWR fuel Fuel Cavity Temperature (*F) 345 M

'.244 530 419 Cavity Pressure (psig) 157 'j 10 984 16 Neutron Shield Tank Temperature (*F) 322 242' 1265 485 Neutron Shield Tank Pressure (psig) 91 10

>100*

18 External Surface Temperature

(*F) 304 210 1300 1357

• Neutron shield fluid is assumed to be lost at the start of the accident condition.

Because there is no liquid coolant in the cask cavity or in the neutron shield tank, there is no danger of internal freezing of liquids.

Because of the lower pressures that will exist in the cavity and in the neutron shield tank, a rupture disk rated at 300 psig will be installed in the cavity relief line; and the neutron shield rupture disk will remain at a rated 100 psig.

The spring operated relief valve is set to relieve at 200 psig.

Calculated pressures for the accident conditions for the cavity and neutron shield tank are well below the ratings of the rupture disks.

3.2 Thermal Properties of' Materials The original themal analysis evaluated the effects of a reference PWR assembly using the TAP-N computer code. The themal properties of materials used in that original analysis are presented in Appendix 3.6.3.

Revised 3-2 Feb. 1987

3.3 Technical Specification of Componen*s The cask components that are significant to the thermal analysis are as follows:

- Vent valves

- Drain valves

- Cavity rupture disks

- Shield tank rupture disks ring seals The temperatures of each component must not exceed the rated temperature of that component to assure that containment is not lost during normal transport or accident conditions.

The vent valves and drain valves are Miser Fire Valves, Part No.1/2FG466TSW or equivalent with modified sealing balls.

These valves are manufactured by Worcester Controls Corporation.

This valve is fire rated which indicates that all components of the valve have been selected to maintain their seals in a high temperature environment.

The modification to the sealing balls consist of drilling a relief hole in the ball so that no water is trapped in the passage when the valve is in the closed position.

This prevents damage to the valve under either freezing or high temperature conditions.

Three typical valves were subjected to leakage tests at 520*F and 985 psig using nitrogen as the working fluid.

Typicabfage rates through the valves during

.these tests were between 2 and 3 x 10F ^ SCFM with no leakage detected at the mountings of the valves.

1 The test report is included in Reference 8.3.5.

The rupture disk that is intended to relieve cavity overpressurization is designed to relieve at 300 psig and 600*F.

The rupture disk is manufactured by In-Val-co or equivalent and is certified by the ASME.

i l

The fixture which supports the rupture disk is modified to accept a relief valve that is intended to seal the cavity when the pressure is 200 psig or less.

There are four rupture disks which protect the shield tanks from overpressuriza-tion. Each disk is designed to rupture at a pressure of 100 psig and 315 F.

All of the rupture disks are designed to operate at the saturation temperature l

that corresponds to the pressure that is to be relieved.

Consequently, there are no temperature limitations on the rupture disks and only the pressures need be examined to assure that the rupture disks are not forced to release cavity contents or shield tank contents.

3.4 Thermal Evaluation for Normal Transport Conditions The steady-state analysis assumed 'a maximum decay heat load of 750 watts.

Revised l

3-3 Feb. 1987 L

8 p

The outside surface temperature was determined by radiative and natural convective heat transfer at an ambient temperature of 130*F.

The total heat flux imposed on the surface is equal to the external heat flux plus the internal decay heat flux.

Conduction is used as the only heat transfer mechanisms through all the steel shells and the lead shield.

Heat is transferred across the neutron shield tank by conduction, radiation and natural convection.

Natural convection is neglected in the analyses for the sake of conservatism.

The fuel consists of 21 natural uranium metallic fuel rods clad in aluminum.

The fuel dimensions are presented in Table 3-3.

Table 3-3.

ETALLIC FUEL DESCRIPTION Length 120.5 inches Diameter 1.36 inches Total Payload 21 rods Total Decay Heat 750 watts The SCOPE computer code was used to calculate the nomal transport (cask in j

shipping container) temperatures and pressures.

These results are shown in Table 3-4.

The SCOPE Program input and results are presented in Appendix 3.6.2.

The Program is described in Section 3.4.1.

Table 3-4.

Sul004RY OF TIMERATURES AND PRESSURES FOR NORMAL TMASPORT CONDITIONS

?

i Component Temperature Pressure 0F)

(psia) i Fuel Rod Surface 372 l

Inner Shell (Cavity) 244 10 Lead 243 l

Outer Shell (Shield Tank) 242 10 Surface 210 3.4.1 Description of the SCOPE Program The temperature calculations performed for this analysis were made by the SCOPE Code (distributions comparable with those produced by the HEATING-5 ORNL/CSD/TM-149, TTC-0316 J. A. Buckholz).

This Code generates temperature cosamnly used for cask safety analysis, but temperature distributions are in the radial dimension only.

The maximum temperatures occur at the fuel midplane, and this is the location at which the radial temperature profile is calculated.

The code version used by Nuclear Assurance Corporation has been t

benchmarked against the HEATING-5 Code. The calculations of fuel temperatures are made in SCOPE via the Wooten-Epstein correlation method, which treats i

square arrays of rods.

It is to be expected that this method will cause the central rod temperature to be overpredicted for metallic fuel, which has a l

triangular lattice, since the SCOPE analysis is based on a square lattice,-

which will give a higher temperature for the central rod.

l 3.4.2 Cask Model Description Both the steady-state and transient analysis were performed using a one dimensional radial mode through the center of the cask.

A maximum spent fuel 3-4 Revised Feb. 1987

i 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 i

Thickness Radius from Cask Center (inches)

Component Material (inches)

Outside Surface Inside Surface Casc 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 Gamma Shield Lead 6.623 13.685 7.063 Inner Shell 321 SS 0.313 7.063 6.750 3.5 Thennal Evaluation of the Hypothetical Fim Accident l

The thermal model for the analysis of the hypothetical accident is basically identical to the model for the analysis of normal transport conditions.

The 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 j

summary of the cask and fuel temperatures are presented in Table 3-4 and 3-6 for the steady-state (normal transport condition) and transient analysis (fire accidentcondition),respectively.

7

)

i The maximum temperature in the neutrose shield tank (1354*F) occurs at the end of the fire accident.

The ideal gas few is used to determine the maximum J

pressure of the shield tank.

Prior to fuel loading the average temperature of l

the neutron shield fluid (air) is assumed.to be -40*F (worst case).

Using the ideal gas equation 2=V.,,1. X 1.2., X P1 where:

T1 = -40*F = 420*R 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.

SUMERY OF TEMPERATURES AND PRESSURES FOR HYPOTHETICAL FIRE ACCIDENT CONDITIONS

.Without Shippine Container With Shippina Container Maximum Pressure Maximum Pressure Component Temperature (*F)

(psia)

Temperature (*F)

(psia)

Fuel Rod Surface 540 540 Inner Shell (Cavity) 419 16 419 17 Lead 437 437 Outer Shell (Shield Tank) 1354 49 485 18 Surface 1357 488 Shipping Container 1357 Revised 3-5 Feb. 1987

I 3.6 Appendices to the Thermal Evaluation This section presents references and other information that supplements the 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.

Docket 6698, September 1972.

2.

ASE Boiler and Pressure Vessel Code,Section III, Nuclear Vessels, 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.

5.

Principles of Heat Transfer, Second Edition, Kreith, F.

(International 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 Resu.1ts for MetANkfueF7 3.6.3 Original SAR Design Basis Thermal Analysis Summary 3.6.4 Original SAR Description of the TAP Computer Code and Thermal Model Description 3.6.5 Original SAR Thennal Analysis for Normal Conditions of Transport 3.6.6 Original SAR Thermal Analysis for the Hypothetical Accident Conditions 3.6.7 Original SAR Thermal Analysis References Revised 3-6 Feb. 1987

Appendix 3.6.2 ScoDe Input and Results for Metallic Fuel Normal Transport and Hypothetical Fire Accident Conditions.

}

. ~.,

f fr l

r Revised Feb. 1987 3-7

i SC0Pt INPUT (ID cat > 1*'ASE FetIAT) FOLLOWSr 21 RODS C.21N WES, 7504, 219008, SASEET/ WALL SAP, NG=144.5, MAC1, SEA, INS) 21 1400 2 0 250 2.21t+05 3.937t+14 1 7 5 f.50 0.125 12.000 0 12.000 5 0.22 0.20 0.00 0.44 14 0.55 to 0.125 3 13.5 4*4 0.3125 1.25 0 145 1 1 0 0.500. 999 50 130.0 1 1 1 3 4.425 34.0 0

PeoPERTIES OF MATERIALS CURRENT'LT IN TME DATA LIBRART i

RATERIAL DENSITT CONSUCTIVITY NEAT CAPACITT TERPERATURM LIRif CAPITAL COST (LS/CUFT) (BTU /NR/FT/F)'

(BTU /LE/F)

(DESRSES F)

(SILB) 1 Ps 708.54 18.0000

.0320 418 3.000 2

FE 483.24 24.0000

.1200 1950 2.000 1

U 1189.25 15.0000

.0280 1450 9.000 4

CU 559.35 210.0000

.0950 1730

.540 5

AL 14R.49 140.0000

.2280 1050

.280 4

SS 494.43 11.0000

.1200 1800 4.000

?

MA 4T.00 38.0000

.3000 1400

.-.200 5

LI 30.00 20.0000 1.0000 1400

• Pe-L 144.40 4400

.1540 1200 19.000 4.000 10 CONC 707.40 18.0000

.0320 410

.140 11 ALSI 170.00 80.0000

.2000 1945 4'..

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i 12 towA 42.00

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W.98 342 43 381.54 388.38 109.93 189.84 189.82 2.97 2.53. S 384.88 384.88 386.00 386.89 384.72 344, '.M4.42 382.8F 380.98 3F9.75 189.72 189.68 189 61

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e 3.$G 3.30. S 377.94 3FF.94 37F.94 377.94 377.84 77.

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( 373.40 3FS.79 349.77 364.62 84.22 184.14 184.12 i

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$47.M 345.34 343.24 342 26 341.14

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i 3 - 14b Page Added j

Feb. 1987

T 5.0 Shielding Evaluation The shielding analysis presented in this section demonstrates that 1) under j

normal conditions of transport the shielding effectiveness of the NAC-1 Cask will not be reduced, and 2) the post accident radiation dose rates external to 1

the NAC-1 Cask will be below the limiting value specified in the NRC l

regulations.

Under normal conditions of tran' sport the cask will meet the following i

requirement from 571.35 (a) (2) of 10 CFR 71:

l The effectiveness of the packaging will not be substantially reduced.

Also, the cask, upon subjection to the specified hypothetical conditions of j

free drop, puncture, and fire (themal), will meet the following requirements from 571.36 (a) (1) of 10 CFR 71:

i The reduction of shielding would not be sufficient to increase i

the external radiation dose rate to more than 1,000 millires per hour 3 l

feet from the external surface of the cask.

5.1 Discussion and Results The NAC-1 cask was originally designed to provide adequate gamme and neutron j

radiation shielding for a PWR assembly.

The gamma and neutron source terms i

for the (typical) assembly are presented below.

The original analysis assumed that the total radiation dose at a point 6 ft.

j from the surface of the transport vehicle (10 ft. from the cask centerline)

L will not exceed 10 mres/hr under nomal shipping conditions when the cask r

contained one PWR assembly.

This assumption is based on the requirement of a j

radiation survey prior to departure of the loaded shipment.

This requirement j

is an administrative control to assure compliance with federal regulations.

The maximum post accident dose rate at a point 3 feet from the cask surface, t

assuming a design basis PWR assembly and the loss of the neutron shield fluid in the accident, is 876 arem/hr. 816erem/hr is due to neutron radiation.

i j

i l

For the proposed metallic fuel, with the neutron shield tanks drained, the I

maximum normal transport dose rate is 1.12 aren/hr gamma and 0.135 ares /hr i

neutron at 6 feet.

The maximum calculated post accident dose rate (gamma plus neutrons) is 3.74 i

j.

ares /hr at a point 3 feet from the cask for metallic fuel.

The calculated dose rates for the reference PWR fuel and the metallic fuel are 4

l susenrized in Table 5-1.

In this table, the normal transport dose rate is j

calculated at 6 feet from the edge of the transport trailer with no credit i

taken for the steel shipping container, and the hypothetical accident dose I

rate is calculated at 3 feet from the cask surface. The normal operation dose rate for the reference PWR fuel assumes that the neutron shield tanks are

{'

full.

The neutron shield tanks are assumed to be empty for the other j

conditions.

~

Revised 5-1 Feb. 1987

)

l Table 5-1.

SLMMARY OF RADIATION DOSE RATES Normal Transport Hypothetical Accident (6 feet) ares /hr (3 feet) ares /hr i

h Reference PWR fuel

< 10 876 Metallic fuel T.26 3.74 5.2 NAC-1 Design Basis Source Specification

]

[

This section describes the original design basis source specifications and provides the analysis model.

These descriptions are necessary because this analysis provides scaling factors required to determine the consequences of 4

transport of the metallic fuel.

5.2.1 Gamma Source a

)

The gamma source term used in the original shielding analysis is comprised of the mixed fission product gamma rays characteristic of PWR spent fuel.

A typical Westinghouse 15x15 fuel assgably, burned to 35,000 MWD /MTU, cooled 150 i

days, is a gesuna source of 2.25 10l* MeV/sec.

This representative gamma source i

ters is used in the metallic fuel analysis.

5.2.2 Neutron Source The total neutron source is p(rimarily thei result of subcritical avtiplicat{gn 1

l of neutrons originating from 1) s fission decays of cme

• and Cm3 ',

j and (2) ( a, n) reactions from fW the fuel.

Approximately 80% of the fixed source neutrons in the design basis fuel are the result of spontaneous l

fission with the remaining 205 resulting from ( a, n) reactions.

Since the l

fission source due to subcritical asitip'ication is the most significant of the neutron sources the fission source spectrum was used in all calculations.

A l

maximum source strength (fixed plus fission source) of 9.2x109 neutrons /sec was t

used. This maximum total neutron source strength would result in a dose rate of 10 mrom/hr at a position 6 ft from tne transport vehicle surface (10 ft. from i

the cask centerline).

l A typical Westinghouse 15x15 fuel assently., burned to 35,000 MWO/MTU, cooled 150

]

days, is a neutron source of 4.60g x 105 n/sec.

This representative neutron source term is used in the metallic fuel analysis.

i l

5.3 Model Specification i

l This section illustrates the cylindrical approximation of the cask and lists the number densities for constituent nuclides used in the ANISN calculations.

5.3.1 Description of the Radial and Axial Shielding Configuration Figure 5-1 illustrates the geometric configuration of the calculational model used in the shielding analysis of the NAC-1 cask.

]

5.3.2 Shielding Regional Densities t

Table 5-2 lists the shielding regional densities for the NAC-1 cask in the ANISN

)

Calculations.

I 5-2 Revised j

Feb. 1987

[

i

5.4 Shielding Evaluation Gamma shielding methods are not discussed since the gamma dose rates external to the NAC-1 Cask are not significantly affected by any of the specified accidents in the hypothetical accident sequence (Appendix B of 10 CFR 71).

A 27-20 neutron-gamma coupled cross section set (Reference 5.5.1.2) including coefficients for a P3 Legendre polynomial approximation to the scattering cross section was used in conjunction with the ANISN (Reference 5.5.1.3) computer program to perform calculations for neutron transport through the NAC-1 Cask shield. All calculations were performed using an S8 angular quadrature.

As demonstrated in the structural analysis (Section 2.0) and the thermal analy-sis (Section 3.0) the heavy metal shield is not subjected to geometric distor-tion under any of the regulatory accident conditions. Therefore, the ganna dose rates during or after any of the prescribed accidents will not change signifi-cantly.

However, sections of the borated water-ethylene glycol neutron shield may be lost during several of the prescribed accidents.

Complete loss of the neutron shield would result in a neutron dose rate increase by a factor of 27.

Since (1) the limiting radiation dose rate under normal transport conditions is 10 mrem /hr for a dose point 6 ft. from the transport vehicle surface (10 ft.

from the cask centerline), and (2) the ganna shield is not signifioantly affected by the hypothetical accidents, the maimum post-accident gamma dose rate 3 ft. from the NAC-1 Cask surf ace may be computed as follows:

Da 10n X xF g 3; g

where:

D

= post-accident gamma ray dose rate measured at a position a

3 ft. from the cask surface.

Dn = normal conditions gamma dose rate measured at a radius rn measured from the cask cylindrical axis (10 mrem /hr).*

n = radial position of nonnal conditions dose point measured r

from the cask cylindrical axis (10 ft. for sole use of vehicle shipments).

rs a radius of cask surface (1.633 ft. for the NAC-1 Cask).

F

= geometric correction factgr for a finite length radiating g

cylinder (assumingcosine surface flux, Fg = 1.4).

then:

10 Da 1 10 x x 1.4 = 30 mrem /hr.

4.633

• Note:

In this analysis the assumed maximum gamma and neutron dose rates of 10 mrem /hr are mutua11y exclusive. The dose rate limits represent extremes of gansna and neutron sources resulting from different, irradiation conditions.

5-5

The maximum post-accident dose rates were also computed using the approach outlined above.

f Da 1 Dn X xFg x Fneutrons (r +3) where:

Fneutrons = neutron dose rate attenaation factor (27) for the 4-/2 inch thick water-ethylene glycol shield.

then:

10 Da 110 x x 1.4 x 27 = 815.8 mrem /hr 4.633 Both values of post-accident dose rates are safely within the prescribed limit of 1.0 rem /hr for the PWR fuel.

5.4.1 Shielding Evaluation for the Metallic Fuel For reference, Table 5-3 provides a comparison of the significant parameters of the original design PWR fuel and the metallic fuel.

Table 5-3.

COMPARISON OF THE PWR AND METALLIC FUELS PWR Metallic

.3 15 15 Fuel Cask Contents L assedly 21 rods Cool time 150 days 365, days Burn-up 35,000 MWD /MTU 1600 MWD /MTU 453 K 1,145.5 Kg Total Uranium 14.0 Kg (g3.09%)

< 8.15 Kg Total U-235 Thermal Output 10 Kw 0.75KYS Total Gamma Source 2.25 x 1016 MeV/sec 1.253 x 10 Mey Total Neutron Source 4.609 x 108 n/sec 2.289 x 105 n/sec Total Fission Gases 4950 curies 609 curies With the PWR fuel the radiation dose rates are:

Normal Transport Hypothetical (6 feet)

Accident (3 feet)

Gamma (mrem /hr)

< 10

< 60 Neutron (mrem /hr)

T 10 7 816 And the neutron and gamma source strength for the reference PWR are:

Sy = 2.25 x 1016 MeV/sec and Sn = 4.609 x 108 n/sec For the metallic fuel:

Revised 5-6 Feb. 1987

15 Sy = 1.253 x 10 MeV/sec = 5.6% of the PWR reference fuel Sn = 2.289 x 105 n/sec = 0.05% of the PWR reference fuel

\\

The original SAR for the cask uses a neutron flux of 9.2 x 109 n/sec.

The neutron flux of the reference PWR fuel is used because it results in a more conservative metallic fuel dose rate.

The metallic fuel dose rates are obtained by multiplying the reference PWR dose rate by the ratio of source strengths:

Normal Transport Hypothetical (6 feet)

Accident (3 feet)

Gasman (mres/hr) 0.56 3.34 Neutron (mres/hr) 0.005 0.4 With the shield tank fluid removed, the gamma dose rate increases by a factor of 2 and the neutron dose rate increases by a factor of 27.

Under this condition the calculated dose rates are:

Normal Transport.

Hypothetical (6 feet)

Accident (3 feet)

Gamma (mres/hr) 1.12 3.34 Neutron (mres/hr) 0.135 0.4 Total N*

ET4 These values are well under the regulatory limits of 10 mrom/hr for normal transport and 1 res/hr for post accident conditions.

t Revised 57 Feb. 1987

5.5 Appendices to the Shielding Evaluation This section presents references and other material that supplements the information presented in Section 5.

5.5.1 References 1.

" Safety Analysis Report for Nuclear Fuel Services, Inc. Spent Fuel Shipping Cask Model No.

NFS-4," Nuclear Fuel Services, Inc.

U.S.

Docket 6698, September 1972. Section 3.3.

2.

Arnold, E.D., Private Communications, ORNL.

3.

Engle, W.W. J r., A Users Manual for ANISN, A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering, K-1693, March 30, 1967.

5.5.2 ANISN Input Data, Tables 5-4, 5-5, and 5-6.

5.5.3 Metallic Fuel Source Terms vs. Cool Time (Table 5-7).

i l

l i

5-8 L

-