ML20199E978
| ML20199E978 | |
| Person / Time | |
|---|---|
| Site: | 07104909, 07104986, 07109009, 07109019 |
| Issue date: | 11/20/1997 |
| From: | Vaughan C GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-97-002, IEB-97-2, NUDOCS 9711240058 | |
| Download: ML20199E978 (5) | |
Text
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p memnv November 20,1997 -
U. S. Nuclear Regulatory Commission ATIN: Document Control Desk
~ Washington, DC 20555-0001-4
Dear Sir / Madam:
Subject:
References:
(1) Doc.kets: 71-4909,*/1 4986,71 9009,7l 9019 (2) Information Notice 97-47 (3) NRC Inspection 97 213 (4) Letter from GE to USNRC dated 10/13/97 on the subject of NRC Hulletin 97-02 (5) Letter from USNRC to Addressee List dated 10/17/97 on the subject of extending the response period fcr NRC Bulletin 97-02
-. On October 13,1997, GE responded as requested to Bulletin 97 02, and indicated that we
-intended to comply with its requirements and schedules.
~ Since receiving Bulletin 97-02, GE has been working to complete the tasks identified by the NRC. Because of the importance of this matter and the need to formally conduct and retain the results of this work, GE identified this as a potential safety concem and as such it was entered into GE's fonnal procedures used to comply with 10 CFR 21. The
~ attached technical responses for the USA /4986/AF and the USA /9019/AF resulted from
- Se findings of this fonnal review process. Also, the details of GE's actions addressing
" Requested Actions"(1),(2) and (3) are included in the attachment to this letter.
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Document Control Desk November 20,1997 Page 2 of 2 In summary, Gli has detennined that the USA /9009/AF ownership will be transferred to Nuclear Fuel Services, Inc., and that the USA /4986/AF and the USA /9019/AF packages meet the conditions discussed in the llulletin, are safe for continued shipment without any special precautions and that funher testing is not required. Based on cunent information, we expect that we will be requesting the NRC to cancel the cenificate for the USA /4909/AF.
Gli requests that the NRC promptly notify us if there are questions or concerns with our response. You should direct your communications to me (910) 675 5656.
Pursuant to 10 CFR 2.204, I declare under penalty of perjury that the foregoing is true and correct. lixecuted on November 20,1997.
Sincerely,
/f(f kf W C M C. hi Vaug ian Acting hianager Facility Licensing cc:
Ch1V 97 028 CJllaughey h1F Weber S Shankman R Chappell PlIng A Gooden SC O'Connor RW lloyle
USNRC Document Control Desk November 20,1997 Pay,e 1 of 3 Response to NRC llulletin 97-02 Hequested Actions (I),(2) & (3)
- 1. Certificate - USA /4909/AF, GE-21 PF-1
'Ihis package is no longer in use by GE. We have notified registered users of GE's intent to request cancellation of the certificate by December 22,1997, unless a registered user requests transfer of the certificate. To date, GE has received no such requests.
- 2. Certificate-USA /1986/AF, RA-2 and RA 3 Series llundle Shipping Package Review puncture test asse snientfor the package and deterinine If the test was perfortnedin accordance with 10 CFR 71.73 (c)(!;.
GE has reviewed the puncture testing to the RA package. The original testing of the package (consisting of an overpack and metal inner container) goes back to 1966.
During this testing, the engineer concluded that the puncture test following the 30-foot drop was not required. please reference the statement made in submittal dated 7/28/67. Docket 70 754/70-1007 Page 3 8, dated 11/18/66, Section 3.1.4.3 of Appendix D relative to the puncture test: "In view of the fact that the inner metal container remained firmly within the outer packaging and, since the inner package was properly designed to remain suberitical with water in leakage such as that v,hich would result from a puncture, an actual puncture test was not conducted."
Additional tests were conducted in 1974. In an i1/18/74 letter from A. N. Tschaeche, GE to C. E. MacDonald, USNRC, a 30 foot drop test on a RA-3 package (overpack and metal inner) was discussed. According to the discussion and the photographs, there was relatively minor drmage to the overpack portion of the package and the puncture tests were not performed since engineeringjudgment indicated that it would have no significance in relation to the performance of the package in the 71.73 test sequence.
The nature and presence of the outer portion or the RA-2/3 package plays a significant part in protecting the inner metal component of the package and its contents particularly in the case of the puncture test of 71.73 (c)(3). The combination of wood, honeycomb and ethafoam pads dissipate and absorb the majority of the energy resulting from the container impacting the cylindrical steel bar. The inner steel container sees little of this energy.1 urthermore, a puncture of the inner steel container would not be significant since the package assumes water in leakage and is safe under those conditions.
USNRC f
Document Control Desk November 20,1997 Page20f3 The NRC has questioned pictures in the consolidated RA 2/3 application which show tests on the bare steel inner container. These tests show a puncture test in which the steel cylindrical bar was clearly not aflixed to the unyielding test pad as evidenced by the fact that the bar fell over at some point during the test. With still photographs, it is not possible to determine the point at which this happened. This information was first added in the 3/1/82 submittal that contained a consolidated application dated 2/24/82. At this point in time, GE was storing fuel at an off site location. Tests were being conducted on the inner steel pucion of the package to detennine its ability to meet the requirements of the 71.73 test sequence in support of various options being considered by GE. The package has ntver been licensed by the NRC in a manner authorizing the bare metal inner portion of the package to ba shipped without the overpack.
The tests on the bare metr.1 inner contairer are noteworthy, however, because they attest to the tot ustness of the inner porton of the package. While the puncur;; test was not demonstrably fully la accord with 71.73 (c)(3), it is clear from the photograph, given the impact mark and the slight symmetrical bowing of the container, that the container absorbed most, if not all, tl.e energy associated with the impact with the nylindrica' bar.
GB has concluded that the engineering determinations made for the container in regard to the puncture test requirement of 71.73 (c)(3) were appropriate and the package is safe and requires no additional evaluation in this regard. The NRC has concurred with this position throughout the cenification process from 1966 to present.
For puncture tests not in accordance with 10 CFR 71. 3 (c)(3), identify any special precautions or operational controls required to asst. m safe use ofthe package pending re-testing.
Puncture tests [71.73 (c)(3)] were not perfonned on the RA-2/3 package. Engineering evaluations found them unnecessary as a part of the testing sequence defined in 71.73.
The package is safe to use as authorized and no additional precautions or operational controls are required.
Plan andschedule to demonstrate compliance with 10 CFR 71.73.
No additional testing is required to meet the requirements of 10 CFR 71.73 (c)(3).
USNRC Document Control Desk November 20,1997 Page 3 cf 3
- 3. Certificate - USA /9009/AF, FL 10-1 GE has been notified by Certified mail that Nuclear Fuel Services, Envin, TN, is requesting the NRC to transfer ownership of that certificate to them. They indicated that they would take full responsibility for the package including responsibilities under NRC llulletin 97-02. GE does not object to this transfer and is transferring all package related information to Nw lear Fuel Services.
- 4. Certillente - USA /9019/AF,11U - 7 Review puncture test assessmentfor the package and determine if the test was performed in accordance with 10 CFR 71.73 (c)(3).
GE has performed the requested review. The BU-7 package was most recently puncture tested b; Nuclear Containers, Inc. on 8/1/95. The specification bar was welded to a 1" thick steel plate, however, the plate was not attached to the essentially unyleiding test pad. The puncture test was video taped.
The video tape was reviewed and GE has detennined that the intent of 71.73 (c)(3) was complied with since there was no discemible movement of the bar during the test. The video was also reviewed by an NRC inspection team on site during the week of September 15,1997, and their conclusion was similar to that reached by GE.
The video tape is on file at the GE facility in Wilmington, NC.
For puncture tests not in accordance with 10 CFR 71. 73 (c)(3), identify any special precautions or operational controls required to assure safe use of the package ending re-testing.
The tests confonned to 71.73 (c)(3), and therefore no additional precautions or operational controls are necessary.
Plan and schedule to demonstrate compliance with 10 CFR 71.73.
The package has been tested to and meets the requirements of 71.73. No additional testing is required for this package.
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