ML20236U369

Forwards Request for Addl Info Identified in Encl to Ltr, Requesting Certificate of Compliance for Model 3750A Package

ML20236U369
Person / Time
Site:	07109275
Issue date:	07/23/1998
From:	Chappell C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:	Roughan C AMERSHAM CORP.
References
NUDOCS 9807300081
Download: ML20236U369 (7)

Text

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e July 23, 1998 )

Ms. Cathleen Roughan '

Amersharn Corporation 40 North Avenue Burlington, MA 01803

Dear Ms. Rot,

ghan:

This refers to your application dated October 31.1996, requesth., a Certificate of Compliance for the Model No. 3750A package.

ln connection with our review, we need the information identified in the enclosure to this letter.

Please advise us within 30 days from the date of this letter when this informaticci will be provided. Additionalinformation requested by this letter should be submitted in the form of revised pages. If you have any questions regarding this matter, we would be pleased to meet with you and your staff. Bernard White is the project manager for our review of your application. Mr. White may be contacted at (301) 415-8515.

Sincerely, Original /s/ by:

Cass R. Chappell, Chief Package Certification Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket No. 71-9275

Enclosure:

As stated j

"C" = Copy without attachment / enclosure "E" = copy with attachment / enclosure "N' = No copy OFC SFPO E SFPO E SFPO b SFPO E SFPO. F N 9

I NAME White CLBro n BBarto hg Dhiktinsky l DATE t /W/98 7 /#3/98 ~T / 2Z/98 7 /22/98 i/N98 OFC SFPO E SFPO f- SFPO E SFPO 6 i NAME NL ood ang hZiegler CRChappell DATE i /11/98 ) /u /98 7 /22/98 r/ /B3/98 Distribution: PUBLIC NRC File Centar NMSS r/f SFPO r/f 9807300081 980723 < .. m

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71-9275 Encl. to ltr. dtd. July 23. 1998 Drawings

1. Revise the drawings to include a section view of the cask in elevation which clearly shows the principal features of construction, the overall dimensions of tb cask, the l thicknesses of the shells and plates, the location and size of the welds, the

! materials of construction and the penetrations that pass through the inner and outer shells. Provide a drawing which clearly shows how the package is closed and sealed. Provide a drawing that clearly shows how the cask body is attached to the j l frame. Enclosed for your information is a copy of NUREG/CR 5502, " Engineering i Drawings for 10 CFR Part 71 Package Approvals." j 1

i l 2. Revise Drawing No. A25237 to include:

a. Tolerancas for all dimensions,

b. A detail with dimensions and tolerances showing the stepped joint on depleted uranium shields,

c. A detail showing all penetrations into the package and their method of closure,

d. A detail showing the design of the maintenance plugs and seals, for both the I lid and body,

c. The diameter, length and materials of construction of the closure bolts, and

f. The size, materials of construction and a detail of the closure gasket.

3. Revise Drawing No. A25237, sheet 1 to: I

a. Clarify whether the combined weight of the basket and sourca capsules is  ;

27 kg.

b. Verify that the drawing specifies the correct length and height of the package. Note that these dimensions are inconsistent with the values given on page 1-2 of the application. Revise Sheet 2 of 6 to include the diameter of the cask.

c. Specify the torque, and its tolerance, for the lid bolts.

4. Revise Drawing No. A25237, sheet 2 to:

a. Clarify what is being illustrated on detail V V and show its location on the package. Detail V-V refers to note #9, which does not appear on the drawing. ,

b. Revise note #3 to clarify whether all steel surfaces are to be sprayed with j the copper plasma spray. I

c. Revise the drawing and appropriate sections in the application text to correct inconsistencies in the depleted uranium shield dimensions. Drawing No.

A25237, sheet 2, shows the depleted uranium shielding radial thickness as 146.5 mm, while section 1.2.1 and Figure 5.3 shows the minimum depleted ,

uranium thickness as 150.5 mm and 151 mm, respectively. I

d. Include a detail showing the inner shell and the liner. l 1

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5. Revise Drawing No. A25237, sheet 3 to:

a. Use a common system of measures (i.e. all metric or all English units). Note that the system of units used for the parts list is inconsistent with the system of units used for the engineering drawings. j

b. Revise section Y-Y and detail P to explicitly show whether the drain tube passes through the depleted uranium. (See item No. 3 under Structural, below.) f

6. Revise the application to include an engineering drawing of the source basket that is suitable for reference in the Certificate of Compliance. The drawing should show the safety features of the basket, including dimensions, tolerances, materials of construction and welds. l

7. It appears that wording on the drawings prohibits release of the material to third parties. Provide a letter stating that the drawings may be released for public f inspection. Note that information submitted to the NRC may only be withheld from i public inspection under the, frovisiorss of 10 CFR 2.790.

Structural

1. Revise the application to specifically state whether there were any openings or cracked welds in the inner and outer shells of the scale model packaging following the hypothetical accident condition tests.

2. Verify that the discussion in the drop test reports is consistent with the test drawings. Verify that the discussion of the tests and results is accurate and complete. Note that the following examples of inconsistencies were found:

(a) Test report 1901 states that the package was suspended at an angle of 24 ,

degrees from the vertical, placing the center of gravity over the point of l

impact. However, the corresponding drawing (1901) appears to show that the package is suspended at an angle of 0 degrees from the vertical (i.e.,

f'at-bottom drop). 1 (b) Test report 1913 states that the package was suspended at an angle of 25.5

, degrees from the vertical, placing the center of gravity over the point of impact. However, the corresponding drawing (1913) appears to show that the package is suspended at an angle of 0 degrees from the vertical.

(c) Test report 1909 discusses the procedures and results of the penetration test for normal conditions of transport. However, the corresponding drawing (1909) shows the package suspended over a puncture bar.

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3. ' It sppears from Drawing No. A25237, Sheet 2, that the drain line passes through the depleted uranium. Revise the application to show that the drain line and its welds will not crack or fatigue under normal conditions of transport due to differential thermal expansion and contraction of the stainless steel and depleted uranium.

I 4. Evaluate the package for immersion under 50 feet of water. The evaluation should consider internal decay heat, inleakage of water, (e.g., steam flashing) and the effect of quenching on the cobalt capsules.

5. Show that the stresses in the closure bolts are within acceptable limits under hypothetical accident conditions considering the temperature of the bolts and the maximum combined stress due to bolt pre-load, internal pressure, and impact loads.

Show that the bolts are adequate for the temperatures and pressures that c:. cur during the fire test. Revise the thermal analysis to report the maximuin temperature of the bolts under normal conditions of transport and hypothetical accident l conditions.

Chemical / Galvanic Reactions I

1. Show that there will be no significant oxidation of the depleted uranium under normal conditions of transport or hypothetical accident conditions, (see item No.1 under Thermal, below). The evaluation should consider possible inleakage of water into the shielding cavity.

2. Revise the application to include an assessment of the potential for corrosion or l sensitization of the stainless steel special form capsules under normal conditions of transport. (See NRC Information Notice 96-54, " Vulnerability of Stainless Steel to Corrosion when Sensitized," dated October 17,1996.)

! 3. Show that a 0.1 mm thickness of copper coating is sufficient to preclude formation i of an eutectic reaction between the stainless steel and uranium over the service life l of the cask. Note the discussion of this subject on page 230 of ORNL-NSIC-68,

"The Cask Designer's Guide."

Thermal

1. Revise the thermal analysis to consider air, rather than helium, as the gas present in the uranium shielding cavity.

2. Justify that use of an allowable temperature of 1032'C for uranium (pg. 3 4) is acceptable, considering possible oxidation (see item No.1, under Chemical / Galvanic Reactions, above). Note that the melting temperature of uranium given in paragraph 3.1.4(c)(i) is incorrect.

3. Revise the application to show that there is adequate axial and radial clearance between the stainless steel and depleted uranium to accommodate differential thermal expansion and contraction. Show that the maximum axial clearance will not allow radiation streaming through the stepped joints in the depleted uranium shielding.

4. Explain and justify how the maximum capsule temperature was determined for normal conditions of transport and for hypothetical accident conditions. The adequacy and relevance of the tests conducted on the Model No. 3300A package is not clear. Revise the acceptance procedures to include a test of the Model No.

3750A package which will confirm the validity of the calculated capsule temperatures.

5. Show that no accessible surface of the personnel barrier will have a temperature exceeding 50*C under normal conditions of transport, as required by -

10 CFR 71.43(g).

6. Justify the assumption used in the thermal evaluation (e.g., pg. 3-20) that individual I effects on package temperatures are additive.

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7. Justify using 38 *C as the ambient air temperature in the fin region to calculate the heat transfer coefficient. Note that as air rises through the fin region the air temperature willincrease, This will decrease the heat transferred from the outer )

shell. Show that the Nusselt number and heat transfer coefficient correlations in j Appendix A are appropriate for the cask geometry and fluid conditions being considered, for both normal conditions of transport and hypothetical accident conditions. Indicate the surfaces te which the resulting heat transfer coefficients will be applied.

8. Show that the calculations for normal and hypothetical accident conditions heat transfer coefficients (Appendices 3A.3 and 3A 6) have sufficiently converged. Note that the temperatures assumed for the calculation in Appendix 3A.3 differ from the I temperatures given in section 3.4.1.1(f).

9. For the axi-symmetric model, show how the vertical surface heat transfer coefficient was determined for normal conditions of transport and how the vertical surface heat fluxes were determined for hypothetical accident conditions. I

10. Justify that the correct temperatures were used to calculated the pressure within the shielding cavity under normal and accident conditions. l

11. Justify using 5 m/s as the bulk fluid velocity in the hypothetical accident condition l heat transfer coefficient calculation. Show that this velocity is typical of a hydrocarbon fuel / air fire with a temperature of at least 800*C. It appears that a value of 10 m/s may be more representative.

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, e Shielding Revise the application to delete the discussion of dose rates for the Model No. 3300A package. Provide a shielding analysis of the Model No. 3750A package which shows that the design meets the external dose rate requirements of 10 CFR Part 71.

Operating Procedures

1. Revise the operating procedures to specify a positive method for drying the cavity (e.g., vacuum drying or purging).

2. Revise the operating procedures to;

a. show how a hot package will be cooled prior to being lowering into a pool,

b. include a determination that the package is in unimpaired physical condition,

c. include a determination that the package has been properly closed, and

d. revise step 17 on page 7-11 to be consistent with the dose rates for an empty package in DOT regulations.

Maintenance Procedures and Acceptance Tests

1. Revise the acceptance tests (Section 8.1) to:

a. Verify that each packaging has been fabricated in accordance with the drawings referenced in the Certificate of Compliance.

l . b. Revise step 8.1.5 to include measuring the dose rate at intervals around the package to ensure there are no streaming paths through the depleted uranium.

c. Specify a thermal acceptance test that will be used for each packaging before first use. Specify the details of the thermal acceptance test. The test specification should identify the heat load, test duration, test instrumentation, and the specific points for the temperature measurement.

Specify a numerical criteria for accepting or rejecting a packaging. Show I' that these criteria are adequate to verify the thermal performance of the package and show how the criteria correlate to the thermal analysis in the application. Note that the criteria in paragraph 3.2.4 of NUREG/CR-3854 may be helpful,

d. Include a pressure acceptance test as required by 10 CFR 71.85(b).

e. Verify the thickness of the copper coating applied to the uranium shielding.

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2. " Revise the Maintenance Procedures (Section 8.2) to include a periodic test of the heat removal capability of the packaging. This test should be able to detect a significant deterioration of the heat transfer properties of the packaging. Specify the numerical criteria for accepting or rejecting a packaging, and show that these criteria are adequate to verify that the thermal performance of the packaging has not degraded. .

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