ML19350E974
| ML19350E974 | |
| Person / Time | |
|---|---|
| Site: | 07109071 |
| Issue date: | 06/01/1981 |
| From: | Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Jerome Murphy ANEFCO, INC. |
| References | |
| NUDOCS 8106230707 | |
| Download: ML19350E974 (2) | |
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NUCLEAR REGULATORY COMMISSION o.
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ANEFCO,~ Incorporated m//'
ATTN: Dr. John D. Murphy N
222 Mamaroneck Avenue 6'
White Plains, NY 10605 Gentlemen:
This refers ~ to your application dated January 2,1980, as amended, requesting an approval to transport spent fuel in the Model No. AP-101 packaging.
In connection with cur review, we need the information identified in the enclosure to this letter.
Please advise us within thirty (30) days from the date of this letter when this information will be provided. The additional infomation requested by this letter should be submitted in the fom of revised pages.
If you have any questions regarding this matter, we would be
. pleased to meet with you.
Sincerely, Charles E.' MacDonald, Chief Transportation Certification Branch Division of' Fuel Cycle and Material Safety
Enclosure:
As stated 8106230 M
j, Model No.'AP-101 Packaging Docket No. 71-9071 Encl to'ltr dtd: JUN 01 E81
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~1.
The SAR for the_Model No. AP-101 packaging has not adequately shown that with spent fuel as. contents, 'the cask meets the requirements of 10'CFR 71. The design criteria of the SAR was based on-Section VIII of the ASME Boiler and Pressure Vessel Code.
As described in Regulatory Guide 7.6,.the acceptable design criteria _for structural analysis of the containment vessels.of Type B packages used to transport irradiated nuclear fuel is based on Section III of the ASME Code. With spent fuel as contents, the design temperature
- and pressure are also much higher than those previously used in
'the cask structural evaluation.
Furthermore, Regulatory Guide 7.8 has presented a. set of initial conditions and specific loading combinations of thermal and mechanical-loads to be used in the structural analyses of the casks used to transport irradiated nuclear-fuel. Therefore, appropriate structural evaluation of the cask in accordance with the design criteria and methods of analysis delineated in Regulatory Guides 7.6 and 7.8 should be presented to demonstrate that the cask meets the requirements of 10 CFR 71.
2.
Please clarify whether or not the filled lead shielding is bonded to the inner containment vessel and the outer cask shell.
Provide an analysis to show the stresses induced in the containment vessel
. and 'the outer shell due to lead slump under impact condition.
3.
Inconsistency exists in the materials used to construct the energy absorbing impact limiter.
Some parts of the SAR has stated that the impact limiters are made of high density urethane polyisocyanurate foam totally enclosed within the hermetically sealed aluminum containers. Other parts of the SAR (pages 1-26,1-35,1-42,etc.)
have repeatedly referenced balsa wood and aluminum shells.
4.
. Provide drawings of the packaging (including fuel baskets) which clearly summarize the safety features considered in the analysis, e.g., dimensions, welds, materials of construction, closures, gaskets, etc.
5.
Please re-evaluate the maximum dose rate from the packaoe under normal and accident conditions of transport. ORIGdN-S(NUREG/CR-0200, Section F7) and XSDRN (ORNL-TM-2500) computer calculations indicated both higher gamma ray and neutron dose rates than reported in the application by factors of 4.4 and 1.7, respectively, for four DWR assemblies ~with 25,000 MWD /MTU burnup and 6 year cooling.
The calculations were made with a 22(n)-18(y-ray) coupled cross-section library based on the ENDF/B IV cross-section data base.
Nineteen isotopes (including 14 actinides) were updated during the fuel burnup calculations.
.