ML20129E442

Certificate of Compliance 5874,Rev 3,for Model WAPD-40. Approval Record Encl

ML20129E442
Person / Time
Site:	07105874
Issue date:	05/31/1985
From:	Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20129E427	List: ML20129E424 ML20129E442
References
NUDOCS 8506060492
Download: ML20129E442 (6)

Text

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• CERTIFICATE OF COMPLIANCE

l. to crR n Fra RADCACTIVE MATERIALS PACKA!ES E

1 e.CE JiFiCArE NUMBER D REVISION NUMBER c.PACMAGE sOENr:FICArlON NUMBER d PAGE NUMBER e.TOTAt NUMBER PAGES g g

5874 3

USA /5874/B( )F 1

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c. Tn.s certificate is issued to certify tnat ine packaging and contents desenbed in item s beiow. meets tne appiicabie safety standards sei fo,tn in Titie to. Code l

l l of Federal Regulations. Part 71, Packaging of Radio &ctive Materials for Transport and Transportation of Radioactive Matenal Under Certain Conditions."

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' 3. This certificate does not reheve the consignor from compliance with any requirement of the regulations of the U S. Department of Transportat on or other I

apphcable regulatory agencies. including the government of any country through or into which the package will be transported.

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ll U.S. Department of Energy Safety Analysis for Radioactive Material p -

I Division of Naval Reactors Shipping Cask No. WAPD-40 dated N

I Washington, DC 20585 December 1984.

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This certificate is conditional upon fuifalling the requTrernents of 10 CFR Part 71, as appbcable. and the conditions specified below g

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(a)

Packaging s*

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(1) Model No.: WAPD-40 l

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l (2) Description

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End or. top loading cylindrical, 10 inch-lead shielded, 304L stainless p

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steel" clad cask for:the shipment.of irradiated test specimens.

Cask I

I has integral skid welded'to the body.

Cylindrical cavity is 1/4 inch I

I thicki304L stainless steel. tube with 2 ; inches bore by.135.25 inches N

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llg length. ' There'aresstainless steel clad 10 inch thick lead shielded g

plugs bolted to each enda Each lid weighs 100 pounds. Overall size 2

g of the' cask is.24 inches in diameter x.168 inch skid length. Gross

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weight with skid is 27.100 pounds.

Using the four lifting trunnions lN as tiedowns' to a truck is forbidden; hence, a special holddown cradle N

is used during truck shipments..This cradle weighs approximately 5000 l

pounds.

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(3)

Drawings p

n The WAPD-40 cask was originally fabricated in accordance with Westing-N 1

house Assembly Drawing No. 936F577, Rev. 4, and later modified in I

accordance with Battelle Memorial Institute Drawing No.100-E, Rev. O.

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Product Containers p

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1 The contents of the package must be packaged in inner product containers.

W lI The inner product containers are constructed in accordance with the l

ll following Westinghouse Electric Corporation Drawing Nos.:

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gl Product Container Drawing Nos.

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IN-40 979C282, Rev. 1 and 9710362, Rev. 1 f

LRS 979C194, Rev. 2 l

LLR 979C277, Rev. 3 y

E 8506060492 850531 8

PDR ADOCK 07105874 F

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.Page 2 - Certificate No. 5874 - Revision No. 3 - Docket No. 71-5874 E

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(b) Contents I

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Type and~ form of material K

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l Byproduct and special nuclear material contained within product containers. I g

The contents must be dry and unmoderated (H to X atomic ratio (.2).

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!I (2) Maximum quantity of material per package r-1 I

I The fissile content of the cask must be limited to a maximum of 350 E

l equivalent grams of U-235.

The number of equivalent grams of U-235 is

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1.6 x grams plutoniumB ' ' il?.0Txfgrams U-235 + 1.4 x grams: U-233 +

j determined by the eguatioit:

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(c) Fissile Class V

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K Minimum transport index to be shown on label

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Maximum decay heaE."per package must not exceed 8,400, BTU /hr.A g

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The acceptance tests and maintenance, program pustilie:f' I

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in accordance with Chapter 8.0 I

I to WAPD-RE0(C)-270, Rev. 3 5 $ S I

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~~2 Expiration date;, May &Q 31 8.

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REFERENCE 1,33

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WWAq 2 91 t d 4 4 5 W

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I Safety Analysis for RaiBoactive/Materiah Shippin'gfCask No. NRBK-40' dated December 1984 E

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R THE E S.~ NUCLEAR, REGULATORY COMMISSION g

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'Chanles E'.*MacDonald, Chief a

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Transportation Certification Branch

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I Division of Fuel Cycle and p

1 Material Safety, NMSS I<

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UNITED STATES NUCLEAR REGULATORY COMMISSION o

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WASHINGTON, D. C. 20565 g

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Transportation Certification Branch j

Approval Record.

l Model No. WAPD-40 Package Docket No. 71-5874 By application dated December 26, 1984, Department of Ener Reactors submitted a revised Safety Analysis Report (SAR) gy, Naval' for the NRC q

Certificate of Compliance No. 5874. The revised Safety Analysis Report 4

has taken into account the cumulative effects of the hypothetical accident conditions as stipulated in Part 71.. In addition, a 1/2-inch thick, 9-inch diameter cover plate has been added to each of the cask end plugs.

All other aspects of the cask are consistent with the original design.

The package has been renewed for a 5-year period.

l STRUCTURAL EVALUATION i

Two lifting trunnions are provided at each end of the cask. They are shown by analys1s to be. capable of lifting three times the maximum package weight. Excessive forces on the trunnions would tear the 1/2-inch fillet weld between the trunnion base and the outer shell, the i

package containment and shielding properties would not be degraded in i-any way.

The WAPD-40 cask and skid (which are welded to each other) are held down

'in the transport vehicle by a tie-down cradle. The cask sits in the tie-down cradle. Two hold-down frames fit over each end of the cask.

Tie-down of the cask was provided through overlapping fit. This tie-down arrangement was shown by analysis to be capable of withstanding the resultant. force from the three specified component forces of 10 CFR 571.45(b).

The lifting trunnions which could~ be used as tie-down devices will be covered during shipment.

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The SAR used existing analytical techniques and reasoned argument to

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evaluate the package for the nonnal and accident tests specified in 10 CFR Part 71. The SAR adequately demonstrates that the package will r,

remain intact and perform its intended safety function under the Nonnal l

Conditions of Transport.

The package was evaluated for a 30-foot drop test in the end, side, stable corner (COG), and oblique corner (20', 40', and 60* from the vertical) orientations. The results of the analyses indicate the following:

(1) None of the end. plug fasteners (at either end) are

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stressed above yield during any of 30-foot drop tests. Thus, the end plugs remain attached to the cask which will ensure that the inner j

container and cargo cannot excape from the cask; (2) the maximum load calculated for the fasteners is less than the preload applied to them.

Thus, the sealing capability associated with the end plug will be maintained.

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' The package was also evaluated for the 40-inch puncture test. The results indicate the 6-inch diameter pin could puncture through the outer shell and into the lead a distance of 0.66 inches. The significance of puncture with respect to potential loss of containment integrity is evaluated in the application. The results of the analyses indicate that a the package defonnation assceiated with the puncture load and package inertia loads would not result in excessive stresses in either the cask inner cylinder or the inner container.

The cumulative effects of the 30-foot drops, the 40-inch puncture drop, and the 1475'F fire accident on the package are reported.

Information presented in'the structural section indicate that shielding remains adequate for the Accident Conditions of, Transport (Chapter 5.0, Shielding Evaluation).

Due to the higher failure temperature of the metallic outer shell' gaskets than that of the 1475'F fire, the outer shell gaskets would remain intact during the fire test. Thus, the package will.not become flooded during the water imersion test.

CONTAIMENT The containment provisions and associated tests for the Model No.

WAPD-40 cask discussed in the SAR have been reviewed. The staff has concluded that the containment requirements of 10 CFR 971.51 are satisfied.

The containment criteria.5 leaktightness.

by demonstrating no leakage for test sensitivity of lx10-7Leaktightness1ssatj/s.

sfied atm-cm Leaktightness is to be demonstrated before first use (8.1.3, Chapter 8.0 of SAR) and annually (8.'.2, Chapter 8.0 of SAR).

In add assembly verification leax test with sensitivity of 1x10 jtion, gn atm-cm /s is specified (8.2.2, Chapter 8.0 of SAR).

To assure adequate sealing of the containment system, both the metallic and silicon rubber 0-rings of the containment system are replaced before each use (8.1.4.2, Chapter 8.0 of SAR).

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THERMAL

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An independent thermal analysis was completed for the Model No. WAPD-40 ~

shipping cask using the HTASl/ HEATING 6 computer code (NUREG/CR-0200).

The initial conditions, such as ambient temperature, solar insolation,

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and the values used to define the fire accident condition, were taken directly from NRC regulations (Appendices A and B,10 CFR Part 71).

In evaluating the Model No. WAPD-40 cask, it was assumed that the molten lead shielding was completely retained by the inner and outer shells during the fire accident condition, i.e., it was not pennitted to leak out forming insulating air gaps between the remaining shielding and the outer shell. The internal heat load was 2,000 BTU /HR.

. The analysis showed that shield melting occurred under the prescribed fire accident conditions. Lead melting was most pronounced at a point approximately 45 minutes from the onset of the fire or 15 minutes after i

the fire was " extinguished." At that time, approximately 2.5 inches of lead were melted along the sides and one-half inch along the top and bottom. This represented approximately 40-45% of the original lead volume. Temperatures within the shielding ranged from 612*F at the inner wall interface to 621.5 at the outer wall interface; i.e., the whole shield was within 10 degrees of the lead melting point. However.

for the whole shield to melt approximately twice as much heat would have to be provided than was experienced under the hypothetical fire conditions.

i Under the worst case scenarios, the cask would rupture at the 45 minute mark. After the molten lead escaped, the remaining shielding would be about 7-7.5 inches thick. The applicant calculated that the lead thickness j

remaining would be about 5.4 inches thick. The difference between the i

calculations is due to more conservative assumptions by the applicant.

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SHIELDING The application demonstrates that after the thermal fire test, about 4.5 i

inches of lead (Pb) (originally 9.875" of Pb) melts along the side walls i

of the cask. Using the SPAN 4 shielding computer program, the gamma dose rate was calculated at three feet from the package surface for irradiated steel contents to be 1,000 mrem /hr for the above lead (Pb) melt. Applicant's gansna dose rates for the two contents and both normal and accident conditions are given in the following table.

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Gamma Dose Rate (mrem /hr) at 3' from surface of package Contents Normal Condition Accident Condition l

Irradiated Steel' O.17 1.000 Uranium Fuel Spent 0.81 963 J

The staff has calculated its own gamma dose rate for the irradiated steel cas94usingalinesourceandgamm92 build-up-in lead. Assuming S =5.0x10 MEV/sec giving an S 1 MEV/sec-cm, a gamma dose rate a?3feetfromthepackagesurfk=e.5x10 c was calculated to be 0.70 mrem /hr (versus applicant's value of 0.17) for normal conditions.

For accident conditions (5.4"Pb) the result was 965 mrem /hr (versus applicant value of 1,000 mrem /hr). Since the applicant's dose rates include a ~ safety factor of 1.3, the accident dose rates differ by about 25% in a conservative manner.

RENEWAL The package has been renewed for a 5-year period.

P Char es

. MacDona hief Transportation Certification Branch Division of Fuel Cycle and Material Safety, NMSS EIIE Date:

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