ML20217K539
| ML20217K539 | |
| Person / Time | |
|---|---|
| Site: | Paducah Gaseous Diffusion Plant |
| Issue date: | 10/24/1997 |
| From: | John Miller UNITED STATES ENRICHMENT CORP. (USEC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GDP-97-0182, GDP-97-182, IEB-97-002, IEB-97-2, NUDOCS 9710280299 | |
| Download: ML20217K539 (4) | |
Text
United States Enrichment Corporation 2 Democracy Centes 6003 Rockhxige Drive Bethe6da. MD 20817 Tel: (3011564 3200 rar(301)564 3201 JAMES H. MILLER Dr: (301) 664-3309 VICE PRESIDENT, PRoDtlCTioN Fax: (301) 571-8279 October 24,1997 U.S. Nuclear Regulatory Commission SERIAL: GDP 97 0182 Attention: Document Control Desk Washington, D.C. 20555 0001 Paducah Gaseous Difhision Plants (PGDP) l l
Docket No. 70-7001 Response to NRC Bulletin 97-02 On September 23,1997, the Nuclear Regulatory Commission (NRC) issued Bulletin 97 02, Puncture Testing ofShipping Packages Under 10 CFR Part 71. This Bulletin applies to USEC as the holder of the Certificate of Compliance (CoC) No. 6553 for the Model No. Paducah Tiger package. Accordingly, enclosed is USEC's response to the subject NRC Bulletin. As indicated in the encic3ure, based on engineering analyses, the Paducah Tiger package satisfies the requirements of 10 CFR 71.73(c)(3). Therefore, no further actions are required by USEC in response to NRC Bulletin 97-02.
Any questions related to this response should be directed to either Beth Darrough at (301) 564-3422 or Russ Wells at (301) 564-3245. There are no new commitments contained in this submittal.
Sincerely, mes II. Miller I
\\
Vice President, Production
,\\y c4 Enclosure CC:
NRC Director Spent Fuel Project Oflice (C. Haughney)
NRC Resident Inspector - PGDP NRC Special Projects Branch (R. Pierson) 9710280299 971024 PDR ADOCK 07007001
.9-.
PDR Omces in Overmore, Cohfomia Paducah. Kentucky Portsmouth. Ohc Washington. DC
OATil AND AFFIRh1ATION 1, James 11. hiiller, swear and aflirm that I am Vice President, Production, of the United States Enrichment Corporation (USEC), that I am autl.., ; zed by USEC to sign and file with the Nuclear Regulatory Commission this response to llulletin 97 02, that I am frmiliar with the contents thereof, and that the statements made and matters set forth therein are true and correct to the best of my knowledge, information, and belief.
/
A mes 11. hiiller 1
On this 24th day of October,1997, the oflicer signing above personally appeared before me, is known by me to be the person whose name is subscribed to within the instrument, and acknowledged that he executed the same for the purposes therein contained.
In witness hereofI hereunto set my hand and ollicial seal.
hila L LLQfL Ladrie hl. Knisley, Notary Public State of htaryland, hiontgomery County hiy commission expires hiarch 17,1998
,o l
Enclosure GDp97 0182 Page1 l
United States Enrichment Corporation (USEC)
Response to NRC Hulletin 97 02 l
Hackground NRC Bulletin 97-02 requires overpack certificate holders to assess the ability of their overpacks to meet the criteria of 10 CFR 71.73(c)(3) with regards to puncture testing and the use of a puncturing bar which is firmly attached to the test slab such that it will not move during testing.
10 CFR 71.73 (c)(3) states: "A free drop of the specimen through a distance of im (40 in) in a positior, for which maximum damage is expected, onto the upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizont I surface. The bar must I
be 15 cm (6 in)in diameter, with the top horizontal and its edge rounded to a radius of not more that 6 mm (0.25 in), and of a length as to cause maximum damage to the package, but not lem than 20
)
em (8 in)long. The long axis of the bar must be vertical." The NRC has concluded that in order for the puncture test to be expected to cause maximum damage, the bar must be " fastened or attached such that it will not move during the test."
NRC Requested Actions NRC Bulletin 97-02 requested holders of NRC CoCs take the following actions:
1)
Review the puncture test assessment of their certified package design (s). The method used to assess the effect of the puncture test should be determined. For package designs whose assessment was based on physical tests, the certificate holder should determine whether the puncture test was perfoaned in accordance with 10 CFR 71.73(c)(3).
2)
For pac?: age designs whose puncture test assessment was based on physical tests not in accordance with 10 CFR 71.73(c)(3), the certificate holder is requested to identify any special precautions or operational controls that are needed to assure safe use of the package, pending retesting and reassessment of the package design. In addition, the certificate holder for any such package design should prepare a justification showing why there are no health and safety concerns that would require immediate removal of packages of that design from service.
3)
Further, for any package design whose puncture test assessment was based on physical tests that were not in accordance with 10 CFR 71.73(c)(3), the certificate holder is requested to prepare-a plan and schedule for demonstrating the adequacy of the design for the hypothetical accident conditions specified in 10 CFR 71.73. Certificate holders are requested tojustify the timeliness of their schedule.
Enclosure GDP97-0182 Page 2 USEC's Response The ability of the Paducah Tiger to withstand a puncture has been evaluated by analysis. The requirements of10 CFR 71.73 require the puncture test to be conducted in the location expected to cause maximum damage. For the Paducah Tiger overpack, this location is the valve end of the overpack. The cylinder valve is the most vulnerable portion of the cylinder. If the valve is stmck or damaged, the continued leaktightness of the cylinder valve threads cannot be ensured. In order to meet the requirement to assess the location at which maximum damage is expected, analysis was performed of a puncture occurring directly in line with the cylinder valve on the overpack. This analysis (References 1 and 2) assumed that the entire kinetic energy of the overpack gained during the drop was absorbed by the penetration of a 6 inch diameter bar. This assumption is conservative since it assumed the center of gravity of the overpack is directly above the impact location. The center of gravity is actually located about 1 % feet away from directly above the valve, which would tend to cause the overpack to spin during the puncture test, absorbing some of the impact energy during the spin. Since the analysis assumes the head of the bar absorbs all the impact force with nu recoil, the " analytical bar" is essentially modeled as being ofindeterminate length (any length longer than the total distance of puncture) and being mounted to an immovable object (i.e., ofinfmite mass).
The analysis concluded that the Paducah Tiger would absorb the impact suflicient to prevent the cylinder valve from being struck, therefore, ensuring the cylinder valve's continued physical integrity. Although physical tests were performed on a predecessor design of the Paducah Tiger, the data resulting from those tests did not form the basis for the analysis. Rather, material property data (i.e., crush strength of the foam, and the stress and elongation values of the steel) were used in the analysis. Therefore, the actual Paducah Tiger design, upon which the CoC No. 6553 is based, was de.nonstrated by analysis to satisfy the requirements of 10 CFR 71.73(c)(3). Thus, there are no further actions required by USEC in respouse to NRC Bulletin 97-02.
References 1.
KY 6SS, Safety Analysis Report on the "Paducah Tiger" Protectiw Owrpackfor 10-Ton Cylinders of Uranium Hexafluoride.
KY-665 Supplement 1, Safety Analysis Report on the "Paducah Tiger"Protectiw Overpack for 10-Ton Cylinders of Uranium Hexafluoride.
2.
KY-665 Supplement 1 Revision 1, Safety Analysis Report on the "Paducah Tiger" Protective Overpackfor 10-Ton Cylinders of Uranium Hexafluoride.