ML20235D585
| ML20235D585 | |
| Person / Time | |
|---|---|
| Site: | 07105942 |
| Issue date: | 07/06/1987 |
| From: | Bernero R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Audin L AFFILIATION NOT ASSIGNED |
| References | |
| CON-#387-3960 2.206, DD-87-12, NUDOCS 8707100228 | |
| Download: ML20235D585 (133) | |
Text
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1 D D-87-12
- ([0 i UNITED STATES OF AMERIC A
- r; NUCLEAR REGUL ATO R Y COMMISSION o 0FFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS Robert M. Bernero, Acting Director 87 JUL -6 P 3 47 In the Matter of .n Docket No. 71-5942 00Cr! * '
GENER AL ELECTRIC COMPANY "'I
) !
) (10 C . F. R . 5 2.206) j' (Puncture Analysis of Model )
GE-700 Shipping Cask)
DIR E C T O R'S D E CISIO N U N D E R 10 C .F. R. G 2.206 IN T R O D U C TIO N By Petition dated January 10, 1987, Lindsay A udin (Petitioner) I requested the Director of Nuclear Material Safety and Safeguards to review l the Safety Analysis Report (S A R) fcr the GE-700 shipping cask in order to i
reevaluate its puncture test analysis. Mr. Audin further requested that the '
1 cask be used only in its non-extended mode until it can be shown that the extended mode complies with all aspects of 10 C.F.R. Part 71.
Petitioner states three concerns related to the puncture test analysis which was performed on the GE-700 cask, based upon his review of the S A R for the cask as developed in 1980 and the GE-100 cask as developed in 1968.
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Petitioner's first concern is that the puncture test analysis was based on ,
testing of the much smaller GE-100 cask, which does not have an extension.
1 According to Petitioner, the extended version of the G E-700 offers a j potentially v ulnerable point (the juncture between the extension and main cask body) not present in the GE-100. The second concern is that use of j the GE-100 test takes credit for the lead shielding behind the cask's outer I
wall. Petitioner asserts that puncture evaluations should not take credit for j 8707100228 070706 PDR ADOCK 07205942 !
B PDR g3g} l
the lead shieldin g because "recent findings , (i.e. , N U R E G/ C R-0930,1980)" 1/
indicate puncture resistance would be weakened if the lead were softened or I melted by a fire which occurred before the puncture stress. Petitioner's third concern is based on N U R E G / C R-0930, which, according to Petitioner, j found that the empirical methods used to analyze casks for puncture were
" crude and unreliable" and failed to give accurate results when tested against real punctures. Petitioner asserts that since the scaling up of the GE-100 is dependent upon such analyses, a proper puncture analysis should utilize the l 1
NIKE2D or similarly sop histicated computer simulation to be certain of its l accuracy. l l
Receipt of Mr. A udin's Petition was acknowledged by letter dated
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February 17, 1987. A notice that the Petition was under consideration was published in the Federal Register on February 23, 1987. 52 Fed. Reg. 5511.
1 We understand the Petitioner's request for a review of the S A R with regard to evaluation of its puncture test analysis to constitute a request that ]
l the cask be reanalyzed with regard to its ability to withstand puncture. For l the reasons set forth below , I have determined that no sufficient reason exists to justify reanalysis of the puncture resistance of the cask, and that the information in Mr. Audin's Petition does not demonstrate that the GE-700 s hip pin g ca s k fails to co m ply with 10 C . F . R . P a rt 71.
1/ N U R E G/ C R-0930, " P u nctu re of S hielded Radioactive Material S hip ping C o ntain e rs ," by R . A . 1.a rder a n d D . A rth u r, A p ril 1980.
i DIS C U SSIO N The regulations in 10 C.F.R. Part 71 set forth the requirements for packaging and trans portation of radioactive material. In addition to the I general packaging requirements for all packages, sections 71.51(a), 71.71 and 71.73 establish safety and design standards for packages known as Type B l
packaging. These standards require Type B packages to withstand conditions incident to normal transport. P u rs uant to Part 71, the N R C reviews and specifically approves each package design, including Type B, to assure that the design meets applicable requirements. The approvals are issued in the form of a certificate of compliance for each package design.
The standards in 10 C.F.R. i 71.73 specify certain hypothetical accident
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conditions for which a cask must be assessed. These include a thirty-foot drop onto a flat, essentially unyielding s u rface , a forty-inch drop onto a i
six-inch diameter steel pin (" forty-inch puncture test") and an exposure to a thirty minute fire environment of 1475 F. 2_/ Following exposure to these accident con dition s , the cask must provide adequate containment of its contents, experience only limited los:, of s iielding and maintain its contents in a sub-critical condition. The regulations in 10 C.F.R. Part 71 also set forth various procedural, administrative, and technical requirements to be followed J
i
-2/ 10 C.F.R. 5 71.41 provides, however, that as an acceptable alternative !
to use of a sample package, a licensee may evaluate a cask design for j the conditions specified in section 71.73 by subjecting a scale model '
specimen to the test or by usin g other methods acceptable to the C ommission .
l for use of Type 8 packages. T hese standards are designed to assure adequate containment of the radioactive material, adequate control of the radiation emitted by the material, and prevention of nuclear criticality during transport of radioactive materials. In adoition ,10 C. F. P., S 71.101 specifies Q uality Assurance stan dards under w hich packages m u st be desig ned ,
fabricated , s hip pe d , and used and requires that the Q uality Assurance Program be approved by the N R C.
Background on the GE-700 Cask Design With Respect to Puncture Evaluation The GE-700 cask design was originally approved by the Atomic Energy Commission ( AEC) as an amendment to the General Electric Company's license S N M-130 on A ugust 24, 1961. At that time, the cask design did not include 1 either the protective steel jacket or the optional extension discussed in the Petition . The cask design was approved in accordance with the criteria in the then-proposed regulations in Part 72, w hich did not require either a thirty-foot drop test or a forty-inch puncture test.
On June 26, 1964, the General Electric Company requested AEC approval of an optional extension which could be bolted to the top of the GE-700 cask body to lengthen the inner cavity. In support of this application, General Electric subjected a quarter scale model of the GE-700, with extension, to two fifteen-foot drops. The scale model specimen was first drop-tested onto a flat, essentially unyielding surface with the c'ask impacting at an angle of 45 onto the extension at the upper end. In the second test, the same scale model specimen was dropped in a horizontal position onto a rounded steel bar lying on the essentially unyielding surface. The tests were conducted in a
1 manner intended to place maxim u m stress on the joint and its bolted con nection between the cask body and the upper extension. The Safety Evaluation prepared by the AEC staff concluded that the test results demonstrated satisfactory strength of the GE-700 extension and its attachment to the cask body. 3/ The GE-700 extension was subsequently authorized by the AEC on October 28,1964 ( A mendment No. 25 to S N M-130; Docket 70-154).
On August 22, 1966, the AEC revised its packaging regulations (now in 10 C.F.R. Part 71) to require, among other things, a thirty-foot drop test and a forty-inch pu nctu re test. The General Electric Company, by application dated December 23, 1968, as amended February 4,1969, requested I approval of the GE-700 cask under the new regulations. 4/ Because of the more stringent test conditions in the new reg ulations , the design of the GE-700 cask was revised to include a protective jacket constructed of two 5/8-inch thick steel s hells . The applicant's assessment of the protective 1
Jacket was based on extrapolating (scaling up) the results of tests previously
-3/ The scale model tests and results are described in GE Report No.
APED-4522, " Destructive Drop Tests of a Model Fuel Cask," dated April 15, 1964. The report is attached to the Ju ne 26,1964 a p plication sabmitted by the General Electric Company for approval of the optional extension.
4/ A safety analysis report (S A R) was submitted in connection with the December 23, 1968 a p plication.
i conducted on the GE-100, a similar but smaller cask. 5/ The GE-100 cask is equipped with a protective jacket constructed of two 1/4-inch thick steel s hells . When the GE-100 cask was subjected to the forty-inch puncture test, its protective jacket ex perienced only localized yielding and was not 1
penetrate d. 6_/ The applicant used the laws of similitude to scale up the l thickness of the GE-100 protective jacket to the size that would be needed in order to provide comparable p u nctu re protection for the heavier G E-700 cask. 7/ The AEC staff agreed with the applicant's conclusion that the G E-700 protective jacket provided ample steel thickness to protect against p u nctu re , see " Safety A nalysis by the Irradiated Fuels B ra nch " (with attachment), March 11, 1969, and approved the revised design of the GE-700 cask with protective steel jacket ( Amendment No. 71-27 to SNM-960; Docket 70-754).
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~5/ The use of scaling techniques or similitude to evaluate shipping casks is l in accordance with the requirements of 10 C.F.R. 5 71.41. See footnote 8 2, su pra. ,
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6/
The test results are described in the General Electric Company ap plication for the G E-100 cask, dated Septem ber 18, 1968 (Docket !70-754). l I
l 7/
The laws of similitude used to scale-up the results of the GE-100 tests are described in "Some Stu dies of Structural Response of Casks to Impact," by H . G. Clark, Jr., Proceedings of the Second International Sym posiu m on Packa and Trans portation of Radioactive fiaterials (October 14-18, 1968) gin g l
)
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\ l The puncture analysis submitted in the 1968 S A R was incorporated, with essentially no changes, into subsequent ap plications u ntil 1982. On June 22,1982, the General Electric Company requested an administrative l l
amendment to the NRC Certificate of Compliance for the GE-700 cask. On !
l August 10, 1982, followin g a determination by the licensee that the actual l measured weight of the package was greater than that presented in previous SARs, the puncture evaluation was revised to account for the increased I I
weig ht. The revised puncture evaluation used the same methods and laws of l similitude as were used in the February 4,1969 application which requested i 1
initial approval of the protective jacket. The results demonstrated that a !
l thickness of 0.525 inches for the two steel shells in the protective jacket would provide puncture protection camparable to the results observed in the l G E-100 test (i.e. , only localized yielding of the protective jacket and no l penetration) . The design of the GE-700 protective jacket specifies 0.625-inch .
1 thick steel shells, which is more than necessary as indicated by the revised puncture evaluation. Accordingly, the N R C approved the GE-700 cask with corrected weight on August 25,1982 (N R C Certificate of Compliance No. 5942, Revision 5).
Puncture Evaluation of the GE-700 Cask Petitioner's first concern is that the G E-700 S A R uses the results of testing of the G E-100 cask as the basis for demonstrating the ability of the G E-700 to withstan d p u nctu re . Petitioner asserts that this is deficient because the extended version cf the GE-700 offers a potentially vulnerable
J point at the juncture between the extension and the main cask body which is not present in the GE-100. 8/
The GE-100 tests to which Petitioner refers were used in the February 4, 1969 and June 22, 1982 applications to show that the protective jacket of i
the GE-700 would not be penetrated. As discussed above, using the laws of j similitude to account for differences in weight, test results from the smaller l l
but similar GE-100 cask were scaled up to show that the steel shells of the protective jacket around the heavier GE-700 cask were sufficiently thick to provide comparable protection. The evaluation indicated that the steel shells are thicker than necessary for the purpose of preventing puncture.
The. GE-700 cask has an optional extension which the smaller GE-100 cask does not have. Petitioner is concerned that the joint between the cask body 1 l
and the extension could be vulnerable to the puncture test since the joint was not considered in scaling up the GE-100 test results.. However, as discussed above, the strength of the joint was demonstrated in the General Electric l Company application dated June 26, 1964. A scale model of the GE-700 cask, with extension , was subjected to two Meen-foot drop tests without the i
benefit of a protective jacket. The tests were conducted in a manner 1
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-8/ Petitioner states that his concerns are based upon a review of the S AR for the GE-700 cask as developed in 1980. Petitioner does not identify the s pecific 1980 SAR that constitutes the basis for his concerns.
Moreover, as stated above, the puncture analysis submitted in any 1980 SAR would have been essentially the same as that submitted in the December 23,1968 S A R.
_g.
intended to place maximum stress on the joint and its bolted connection. In the judgment of the N R C staff, the stresses and forces to which the joint was subjected in these two fifteen-foot drop tests were greater than those that would be expected in a forty-inch puncture test of the GE-700 cask. Thus, while the joint was not considered in scaling up the GE-100 test results, the stren gth and adequacy of the joint and its bolted con nection was substantiated by another set of tests which were conducted before the tests l
to which Petitioner refers. Petitioner may not have been aware of the earlier scale mooel tests . The earlier tests, con d ucted without benefit of the protective jacket, demonstrate the integrity of the joint. Therefore, there is no basis to grant the Petitioner's request based upon the assertion that the joint could be vulnerable to puncture.
Sequence of Puncture and Fire Tests Petitioner next asserts that puncture evaluations should not take credit for lead shielding because puncture resistance would be weakened if the lead were first softened or melted by a fire . Althou g h 10 C.F.R. 9 71.73 specifies that the puncture test is to be conducted before the fire test, Petitioner asserts that the sequence of events could easily be reversed in an actual accide nt. As an e x a m ple , Petitioner s peculates that a rail car collapsing in a fire could tip over onto aajacent railroad tracks or reinforced p rotrusion s.
Petitioner's argument constitutes an impermissible collateral attack upon an agency rule. Generally, the proper forum for directly challenging a rule is the rulemaking proceeding. See Union Electric Co. (Callaway Plant, U nits
l l
l 1 and 2), AL A B-352, 4 N R C 371, 374 (1976). A petition u n der 10 C . F. R . i 5 2.206 requestin g e nforcement action is not a vehicle for challengin g a Commission rule. Rather, if a member of the public is dissatisfied with an exisin g rule, he may petition the Commission to institute a rulema kin g .
1 proceedin g to change the rule. Since the G E-700 cask was tested and approved in accordance with the Commission's regulations, it cannot now be l l
fou n d u nsatisfactory based merely on Petitioner's dissatisfaction with the Commission's rules themselves. !
With respect to Petitioner's im plication that the Commission's rules governing the approval of shipping casks are inadequate, it should be noted i that d u rin g the past several years, the Commission has reexamined its regulations concerning the transportation of radioactive materials, 9/ and has concluded that existing regulations are adequate to protect the public against unreasonable risk. See 40 Fed. Reg. 23768 (June 2,1975) and 46 Fed. Reg. 21619 ( A pril 13,1981). See also N U REG-0170, " Final Environmental Statement on the T rans portation of Ra dioactive Material By Air and Other Modes,"
December 1977. The Com mission s pecifically reaffirmed the adequacy of 10 C.F.R. Part 71 with respect to the safety of radioactive waste transporta-
-9/ Special note is made of the fact that on August 17, 1979, a total revision of 10 C.F.R. Part 71 to make it more compatible with International Atomic Energy Agency (IAEA) regulations was published for comment as a proposed rule (44 Fed. Reg. 48234). After consideration of comments, the NRC pu blished revised 10 C.F.R. Part 71 in final form on A u g ust 5, 1983 (48 Fed. Reg. 35600).
tion in a rulemaking on advance notification. See 47 Fed. Reg. 596, (January 6,1982). Petitioner has not presented any information which would indicate !
that the test sequence in 10 C.F.R. Part 71 of puncture fellowed by fire is !
not sufficient to provide adequate assurance of public health and safety, or that the sequence of puncture followed by fire could easily be reversed during an accident. As such, in addition to impermissibly challenging the Commission's rules, Petitioner has failed to provide the factual basis for his request with the specificity required by 10 C.F.R. 9 2.206, and action need not be ta ken on his req uest for this reason also . See, e.g., i l
Philadelphia Electric Co. (Limerick Generatin g Statio n , U nits 1 and 2),
9 0-85-11, 22 N R C 149,154 (1985) .
)
Method of Analysis )
Petitioner asserts that N U REG /C R-0930 found that puncture evaluations based upon empirical analysis were " crude and unreliable" and failed to give accurate results.10/ He further asserts that the scaling of the GE-100 test I
10/
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Althou g h N U REG /C R-0930 does state that empirical methods of analysis are " crude and unreliable," the report's conclusions and recommendations l are that empirical methods are too conservative; i.e., empirical equations ;
conservatively underestimate the energy needed to produce puncture by about 60% and use of a more sophisticated method of analysis would permit thinner jacket thicknesses to be used when puncture controls the
, design. See N U REG /C R-0930, at pages 29 and 32. Moreover, the N R C staff is not aware of any instance where the empirical methods used to e valuate pu ncture of lead shieldeo casks have failed to predict a conservative result when the cask design was su bjected to actual physical testing.
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results is dependent voon such analysis, and that a proper puncture analysis should utilize NIKE2D or a similarly sophisticated computer program to be certain of its accuracy.
It is not clear why Petitioner believes that the puncture evaluation of the G E-700 cask is based on em pirical desig n eq uations. As previously j discussed, test results of the uaaller but similar GE-100 cask were based on calculations using the laws of similitude, not on empirical equations. The use of similitude to scale behavioral differences between large and small objects of j similar design is a well established and accepted engineering practice. J1/ As indicated supra, see note 5, the use of similitude to evaluate shipping casks is in accordance with the requirements of 10 C.F.R. 5 71.41. Thus, there is no basis to grant Petitioner's request for a reevaluation of the GE-700 cask based upon this concern.
C onclusion In sum, the Petitioner has failed to set forth sufficient grounds for his request that the SAR for the GE-700 container be reviewed in order to j reevaluate the puncture test analysis for this cask, or for his request that the cask be used only in its non-extended mode until it can be shown that !
the extended mode complies with all aspects of 10 C.F.R. Part 71. As ex plained above, Petitioner has failed to provide sufficient information JJ/ See, e.g., T. Duffey, Scaling Laws for Fuel Capsules Subjected to Blast Impact and Thermal Loading (1971).
8 showing that the puncture evaluation which was. performed on _the GE-700 cask was inadequate or that the cask in its ' extended mode fails to comply with 10 C . F. R . Pa rt 71. Consequently, I decline to take the action requested by I the Petitioner.
A copy of this Decision will be filed with the Secretary for the l Co m mission's review in accordance with 10 C.F.R. 6 2.206(c) of the Commission's regulations.
Robert M. Bernero, Jr. , Acting Director Office of Nuclear Material Safety and Safeguards Dated at Silver Spring, Maryland this 6th day of L\3 , 1987 l
NUCLEAR REGULATORY COMMISSION
[DOCKETNO. 71-5942]
GENERAL ELECTRIC COMPANY (Puncture Analysis of Model GE-700 Shipping Cask)
ISSUANCE OF DIRECTOR'S DECISION UNDER 10 CFR 2.206 (00-87-12)
Notice is hereby given that the Director, Office of Nuclear Material Safety and Safeguards, has taken action with regard to a Petition for action under 10 C.F.R. 5 2.206 received from Mr. Lindsay Audin, dated January 10, 1987.
l The Petitioner requested that the NRC review the Safety Analysis Report for the GE-700 container in order to reevaluate the puncture test analysis for this cask and that the cask be used only in its non-extended mode until it can be shown that the extended mode complies with all the requirements of 10 C.F.R. I Part 71. The Petition alleged that the puncture analysis was based or. the ;
1 testing of a much smaller cask, the GE-100, and that this resulted in a deficient analysis of the GE-700 cask with its extension. 1 i
The Director of the Office of Nuclear Material Safety and Safeguards has j determined to deny the Petition. The reasons for this denial are explained in the " Director's Decision under 10 C.F.R. 5 2.206," (DD-87-12) which is available for public inspection in the Commission's Public Document Room 1717 H Street, !
l
NW, Washington, DC 20555. A ccpy of this decision will be filed with the Secretary for the Comission's review in accordance with IC C.F.R. 6 2.206(c) of the Comission's regulations. As provided by this regulation, the decision till constitute the final action of the Comission 25 days after the date of issuance of the decision unless the Comission on its own motion institutes a review of the decision within that time.
Dated at Silver Spring, Maryland this 64 day of Tul] , 1987.
FOR THE NUCLEAR REGULATORY COMMISSION
~ , - -
1 Robert M. Bernero, Acting Director Office of Nuclear Material Safety and Safeguards i
JUL 0 61967 Mr. Lindsay Audia One Everett Avenue Ossining, NY 10562
Dear Mr. Audin:
This letter is in response to your petition dated January 10, 1987, concerning the adequacy of the puncture analysis for the GE-700 shipping cask.
In your petition you requested that I review the Safety Analysis Report for the GE.700 cask in order to reevaluate its puncture test analysis and that the cask be used only in its non-extended mode until it can be shown that the extended mode complies with all the requirements of 10 C.F.R. Part 71. For the reasons set forth in the enclosed " Director's Decision under 10 C.F.R. 9 2.206," I fino no adequate basis in your petition for taking the action you have requested.
A copy of this decision will be filed with the Secretary of the Commission for its review in accordance with 10 C.F.R. 5 2.206(c) of the Commission's regulations.
As provided by this regulation, the decision will constitute the fin 61 action of the Commission 25 days after the date of issuance of the decision unless the Commission, on its own motion, institutes a review of the decision within that time.
Also enclosed for your information is a copy of a notice which is being filed with the Office of the Federal Register for publication.
Sincerely, (Signed) Robert AL Bernem Robert M. Bernero, Acting Director Office of Nuclear Material Safety and Safeguards
Enclosures:
- 1. Director's Decision Under 10 CFR 2,206
- 2. Federal Register Notice cc: General Electric Company Distribution: w/encls NR55 Office R/F EPEaston SSchidakel EDD 2#13 RFBurnett CRChappell NRC PDR JAResner EQTenEyck NMSS r/f SECY!(5)
CEMacDonald SGTB r/f OGC OGC SSchidakel by telghont 6/30 /87 ( 4 0FC: SGTB SGTBy 5 SGTR S
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NAME:EPEaston: alm CRU appell CEM n d EQTe yck RFB rnett RMB rnero flLThompson DATE:06/30/87 06/gD/87 06/fp/87 Oy/[/87 (y/l/87 Off//87 Of// /87 0FFICIAL RECORD COPY
DD 12.
Incoming Petition and Supporting Documents 1
Petition J A. Petition from Lindsay Audin Dated January 10, 1987 I
Major Supporting Documents
- 1. NUREG/CR-0930, " Puncture of Shielded Radioactive Material Shipping Containers." l
- 2. General Electric Company Report APED-4522, " Destructive Drop Tests of a ,
Model Fuel Shipping Cask"; April 15, 1964. ~!
- 3. AEC Amendment No. 25 to SNW-130 dated October 28, 1964 containing 4 Safety Analysis Report authorizing GE-700 cxtension.
- 4. Application from General Electric Company dated September 18, 1968 describing GE-100 test results.
- 5. Application from General Electric Company dated December 23, 1968 containing puncture analysis of the GE-700 cask.
- 6. AEC Amendment No. 71-25 to SNM-960 dated January 14, 1969 containing Safety Analysis Report for GE-100 cask.
. 7. AEC Amendment No. 71-27 to SNM-960 dated March 11, 1969 approving revised design of GE-700 cask with protective jacket.
- 8. General Electric Company application dated June 22, 1982 and follow-up correspondence dated July 16, 1982 and August 10, 1982.
- 9. NRC Revision 5 to Certificate of Compliance No.5942 dated August 8, 1983.
%/ fi Al4 %p - M1 10 January 1987 Director of Nuclear Material Safety and Safeguards k'
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Certification of Compliance No 5942 Dear Sir; pursuant to 10CFR2.206, I request a review of the Safety Analysis Report (SAR) for the GE-700 container (Coc #5942) with regard to evaluation of its puncture' test analysis. I believe the analysis of the cask with its extension may be deficient. I therefore request that the cask be used only in its non-extended mode until it can be shown that the extended mode complies with all aspects of 10CFR71.
My concerns are based on a review of the Safety Analysis Report (SAR) for the GE-700 cask as developed in 1980 and the GE-100 cask as developed in 1968. . If there are more recent SAR's that address my concerns, I would appreciate receipt of a copy of the relevant sections with your response to this letter.
I have three concerns related to the puncture test analysis.
- 1. The GE-700 is a scaled up version of the much smaller GE-100. The GE-100 was drop tested on the standard plug and (according to its SAR) "the protective jacket yielded on impact but no ma]or fracture occurred. No damage occurred.to the cask."
The GE-700 SAR utilizes this result as a scale model verification of its ability to withstand the puncture test. The ev* ended verrson of the GE-700, however, offers a potentia 2' 'able nin' li.~.,
it6 innetura hatween the extension ar. 'sk nody) no: present in the GE-100. I therefore belic GE-100 puncture test cannot be used as proof that t r. ;
version of the GE-700 has been tested in the most seve:.
for puncture.
- 2. Use of the GE-100 test automatically takes credit for the lead shielding behind the cask's outer wall. Most casks certified by NRC in the last 12 years have not taken credit for the puncture resistance of the lead. I think this is a wise policy since recent findings (i.e., NUREG/CR-0930, 1980) indicate that casks.with lead shielding are more readily punctured when their lead is softened or melted in a fire. While I realize the testing standard assumes the puncture stress occurs before the-thermal stress, reality could easily reverse the order of these events (e.g., a rail car collapsing in a fire tips over onto adjacent railroad tracks or reinforced protrusions). I therefore-believe the GE-700 puncture analysis should be re-evaluated for both the extended and non-extended versions.
e
o 4
- 3. NUREG/CR-0930 also found that the empirical analyses of past punctures were " crude and unreliable" and failed to give accurate results when tested against real punctures. Since the scaling up of the GE-100 is dependent upon such analyses, I believe a proper puncture analysis should utilize the NIKE2D
'or similarly sophisticated computer simulation to be certain of its accuracy (NIKE2D was utilized in the 0930 study),
please respond to each of these concerns separately and send me a copy of any changed license or other documents resulting from my request.
Thank you for your continued efforts toward transportation safety.
Yours truly,
, /
lm lbvb$
Lindsay Audin One Everett Avenue ;
Ossining, New York 10562 (
)
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- 4,
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4 NUREG/CR-0930 UCRL/52638 Volume i RT Puncture of Shielded Radioactive Material Shipping Containers Analysis and Results M:nuscript Completed: December 1978 D te Published: April 1980 A , D. Arthur Eore,I9b Prepared for Division of Safeguards, Fuel Cycle and Environmental Rose. arch Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20666 NRC FIN No. A0127-8 1
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t .
PUNCTURE OF SHIELDED RADI0 ACTIVE MATERIAL SHIPPING i CONTAINERS ANALYSIS AND RESULTS--PART I INTRODUCTION AND
SUMMARY
i Accidental release of radioactive material is a potential hazard associated with shipment of nuclear material. Each shipping container (cask) must be proven safe to be licensed by the Nuclear Regulatory Commission (NRC) for use, by nongovernment shippers. Cask designers may demonstrate puncture resistance by testing, analysis, or a combination of the two. Analytical verification is highly desirable because of the great expense of full scale testing. Existing test data is inadequate and analytical methods, largely empirical, are crude and unreliable. This document is the final report to the NRC on e project to improve analytical methods for treating cask puncture, and to broaden the experimental data base on puncture phenomena.
Typical shipping casks use either lead or depleted uranium, U238, f r ganna radiation shielding. The shielding is sandwiched between a thick outer jacket of stainless steel and a relatively thin inner liner of stainless steel (SS).
Penetration of the outer stainless shell is usually considered a failure because a subsequent fire could melt some portion of the lead and produce a radiation level above defined standards. The NRC requires that the cask be able to survive a 40-in. drop onto a 6-in.-diam by 8-in, long mild steel punch without releasing radioactive material or losing radiation shielding effectiveness. Because a loaded cask can weigh up to 100 tons or more, such a drop, elthough producing relatively low impact velocity, requires the cask to absorb considerable impact energy. The punch must impact the cask at its mnst vulnerable spot. Current design practice uses empirical curves of scaled incipient puncture energy to establish minimum shell thicknesses. The experimental basis of these curves is not adequate to cover current cask designs; our tests confirm this. In addition, the empirical formulas do not give designers the insight into puncture phenomena they need to produce a rational, safe design. .
4
( .. .. . .- , .n.~..-
t Puncture analysis of laminated cylindrical casks is complicated by many factors. For example, the most vulnerable spot on the cask is not known beforehand. To minimize expense and project duration, we limited the scope of this project while retaining the essential ingredients of the cask puncture problem. The result was an analytical and experimental investigation into the axispmetric puncture behavior of flat, circular, laminated plates. Our simulations included the large deformations that occur due to the ductility of 'I cask materials. Puncture causes gaps and slipping.to occur between the liner, the shielding (backing), outer jacket, and the punch. This behavior was also simulated. We demonstrated that cask puncture is not just a localized phenomenon; elastic and inutta forces throughout the structure also play an important role.
We statically and dynamically tested a series of 12-in.-diam lead- and uranium- backed stainless steel plates. Each plate was clamped in a fixture that allowed an 8-in.-diam area to be unsupported. A backing plate of either lead or uranium was clamped behind the test plate. One-tenth to one-quarter scale punches (0.6 to 1.5-in. diam.) were used on 0.05 to 0.54-in.-thick plates. Punch force, deflection, and punch and plate strains were monitored in the static puncture tests. In the dynamic tests, where the fixture is dropped on a stationary punch, we measured punch force, punch and plate strains, and acceleration.
Initially, we tried analyzing our puncture tests with several existing finite element programs, but fnund them inadequate. No finite element program was available that could economically handle the large deformations and slipping essential to the analysis of cask puncture. As a result of working closely
, with LLL's Engineering Methods Development Group, the necessary improvements
.are now available in a program called NIKE20.1 We found that analytical puncture prediction requires two essential elements:
(1) a finite element code that accurately tracks stress and displacement j throughout the event; and (2) criteria for identifying when and if puncture j
occurs. The puu -:ve event is not self-evident from the finite element -
calculations. )
i 2
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The puncture criteria that worked over the range of parameters we studied were a shear stress criterion for the lead-backed plates and a maximum effective plastic strain criterion for the uranium-backed plates. Figure I shows the agreement between the calculated and measured response of a typical plate.
Figure 2 shows how the calculated and experimental forces at failure for all lead backed plates agree within the scatter of the test data. Based upon the wide range of test parameters for which NIKE20 was validated, it is safe to extend the NIKE2D calculations to more general axisymmetric puncture problems. There is every reason to believe that the finite element methodology in NIKE2D wil,1 work equally well in a three-dimensional code for i
the puncture of shipping casks.
I I We recommend that the NRC and cask designers use NIKE2D or a similar validated
! code to demonstrate cask puncture perfonnance, instead of the current empirical formula which is too simple to account for the structural dynamics and material behavior that vary from design to design. A code such as NIKE2D will account for all the important aspects of cask' puncture. It will result in greater insight into cask puncture behavior and greater confidence in cask puncture performance.
PUNCTURE TESTS Part II describes our puncture tests in detail. Here we present a brief overview of the test program to put the following sections in perspective. We conducted 19 static puncture tests and 40 dynamic (drcp test) puncture tests.
All test specimens were 12-in.-diam flat plates. The cylindrical punches were flat-ended. All tests were axisymmetric; the normal axis of the specimens coincided with the centerline of the punches.
Table 1 sunmarizes the test specimen configurations. Most test specimens consisted of three plates clamped together at their' outer edge: an outer plate (called the test plate) of Type 316 SS that contacted the punch, a At this writing, LLL's Engineering Methods Development Group is implementing .
a 3-D version of NIKE20. Puncture calculations have not yet been made with it, however.
3
,j
's In addition to the calculations that simulated the test conditions, a static parametric study was performed on lead-backed SS plates.
The results of this study are shown by the curves in Figs. 2 and 23. Figure 2 shows how;the punch j
force at failure varies with the ratio of punch diameter to plate thickness.
The points plotted adjacent to the curves represent the experimental validation of the NIKE2D calculations and the shear stress fracture criteria.
Figure 23 shows similar comparisons for the failure energy.
Shear stress is not adequate to predict the failure of the uranium-backed plates. i It appears that the first fracture occurs in the uranium backing.
Based upon the limited test data and calculations, it appears that the uranium plates f racture af ter undergoing 6 to 10% effective plastic strain. The variability of the material tested precludes the establishment of a more precise failure criteria. Figure 24 shows contours of calculated effective plastic strain in the uranium backing at the same force that caused failure Jn a dynamic test.
CONCLUSIONS Our test results do not agree with empirical formulae derived from previous.
puncture tests.
Previous researchers used plots of scaled puncture energy, E/ ut vs d/t, to unify the data and form the basis of an empirical puncture prediction scheme.
Our puncture energy data, when scaled this way is roughly 60% higher than the nearest previous puncture predictor. . In addition, our results do not support the reported 30 to 60% increase in puncture energy of dynamic over static conditions. In fact, the present data show that dynamic puncture can require less energy than static.
We believe this disagreement stems from the inadequacy of the scaling procedure. For example,
} the scaling law does not account for outer jacket ductiitty or the I l
i contribution to puncture resistance made by the backing (radiation shield).
The simple scaling law cannot account for the overall structural dynamic response of a cask during puncture.
Puncture energy and force depend heavily :
on whether bending or membrane forces predominate and on the natural '
freauencies excited by the impact. f 29 4
i
[ ;
Improvements to our computer code, NIKE20, coupled with a shear stress puncture criterion (for lead-backed plates) allowed us to analytically duplicate our static and dynamic puncture test results very closely. The calculation predicted force, displacement, and energy over a wide range of test parameters (i.e., plate thickness, backing thickness, punch diameter). We believe NIKE2D will work equally well for full-scale axisynmetric cask puncture problems.
RECOMMENDATIONS We recormlend that the NRC and cask designers consider changing the way the cask puncture problem is analyzed during design and licensing. Instead of the existing empirical formula, use NIKE20 or a similar validated code to demonstrate ~ cask puncture performance. We believe this approach provides valuable insight and guidance to the designer, improves confidence in a cask's puncture perfortnance, and allows a thinner jacket where puncture controls design.
Prior to analyzing the full-scale puncture problem, subscale tests similar to ours may be necessary.
The tests would validate both the computer code and the analytical puncture criterion where the cask laminate materials and thicknesses are significantly outside the range of our subsca19 tests. Our test specimen materials and thicknesses were selected with cask design in mind, however.
It is our intent that the test data in Part II serve the validation purpose whenever a cask .
laminate design is enough like our specimens.
RECOMMENDATIONS FOR FUTURE WORK s
Finite element puncture calculations on structures more complicated than our test specimens will involve considerable computer expense.
This expense warrants a search for an approach more efficient than
" brute force" finite element analysis. It may be advantageous to represent the global or overall response of a cask to puncture by its linear elastic normal modes, and the local region around the punch by a detailed finite element model. This variation on the component-mode synthesis method reduces the number of degrees of freedom that must be .
considered. This approach was tried early in the program as a means of understanding the dynamics of puncture.
The method and the moderate success it achieved are described in Appendix A.
t 32
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e A very simple puncture analysis scheme is needed for preliminary design and as a check on finite element calculations. The method l
should be far simpler and cheaper than finite element analysis, yet accurate enough and sensitive enough to changes in design variables to
, be useful, i
Based on our test results, we believe a simple bilinear spring can represent the force-displacement behavior of flat, circular, laminated plates during puncture. It seems likely that additional work could uncover a fa0rly simple relationship between the bilinear spring parameters (i.e., two slopes and the intercept) and the geometry and material properties of the laminated plate.
This knowledge, coupled with the simple average shear stress criterion for puncture noted earlier (Fig. 10) could be the basis of a very simple puncture estimator for full scale and subscale plates like ours. The method may even prove useful for rectangular flat plates and curved plates, e We believe that additional finite element analyses should be done on puncture of. laminated cylinders. The objective would be to determine how different the structural response and plate shear stress distribution is from the flat plate axisymetric case. The motivation for this study lies in the side puncture problem for cylindrical casks.
e The effect of strain-rate on material properties was not adequately resolved by this study. Further work should be accomplished to:
- 1. Perform more detailed comparisons between our static and dynamic plate puncture calculations.
, 2. Upgrade NIKE20 to accomodate the effect of strain rate on material characteristics.
l' 3.
Determine the effect of increased strain rate on the ultimate shear strength of cask shell materials.
i 33
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A MODEL FUEL SillPPING CASK G.L. SAYWELL
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4 DESTRUCTIVE DROP TESTS OF A MODEL FUEL SHIPPING CASK G. L. Saywell
SUMMARY
l Deactivation of the Vallecitos Boiling Water Reactor brought about i a need for a shipping cask that could accommodate 54-inch-long irradiated fuel elements. The 40-inch cavity of the existing 700-series shipping casks could be lengthened by using a simple spool-piece. The results of a stress analysis of the modified cask showed that the conditions of 10CFR72 (proposed)*
would be met. To verify this conclusion, a one-quarter-size model was made and dropped twice from a height of 15 feet onto an unyielding surface. Although deformation occurred, the cask retained structuralintegrity.
INTRODUCTION l
Because the Vallecitos Boiling Water Reactor (VBWR) w::.s deactivated it became necessary to provide a shipping cask that could accommodate fuel elements up to 54 inches long. The exist-ing General Electric Test Reactor (GETR) 700-series fuel shipping casks (see Figures 1 and A-1) have cavities of only 40 inches; therefore a means for extending the 700-series c isk cavity by approximately 14 inches was required. In_ January 1964, a simple " spool-piece" extension was designed (Figures A-2 and A-3) to add the necessary 14 inches to the 700-series cask cavity length. To obtain an AEC license for this modified cask, it was considered desirable to conduct a drop test of a scale model of this modified cask to show that the spool-piece design was structurally sound.
One of the first steps undertaken during the initial design of the spool-piece modification for the GETR 700-series cask was to determine analytically the stresses for various loading condi-tions to ensure that the tentative design met the requirements of 10CFR72 (proposed). This analysis showed that the tensile stresses in the connecting bolts (Figure A-2, part 18) and the shear stresses in the joint between the spool-piece and the cask body approached the limiting values under the loading specified by 10CFR72 (proposed). While the spool-piece extension de-sign theoretically met the 10CFR72 (proposed) requirements, it was felt that destructive testing should be made to show that the cas,t and its extension could survive the maximum trnpact ,
conditions specified by 10CFR72 (proposed). .
6 J
- Code of Federal Regulations. Title 10-Atomic Energy, Chapter 1 - Atomic Energy Commission, Part 72 - Regulation to Protect Against Radiation in the Shipment of Irradiated Fuel Elements (Proposed).
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I Research and tests conducted at the Franklin Institute' supports "... the conclusion that -
quantitative predictions of the behavior of a prototype cask can be made from tests of exact scale models. .In any tctual situation where a container design is particularly critical..., tests of exact scale models provide a practical economical way to conclusively establish the integrity of the cask"'.
In February 1964, a decision was made to construct a scale model of the 700-series cask and the proposed spool piece extension and subject them to severe impact conditions to verify experi-mentally the structural integrity of the proposed spool-piece design.
THE MODEL t The somewhat simplified, one-quarter-size model was built in the development shop of the Vallecitos Atomic Laboratory. The one-quarter-size was chosen because this was about the I smallest scale in which the design detail could be reproduced accurately at a reasonable cost.
Additionally, the size and weight were convenient. l t
All simplifications in the design detail of the model 'were made to reduce the cost of the model.
All deviations from the full-size cask design were made on the conservative side, that is, the f i
design modifications made the model structurally weaker than the actual cask and spool-piece.
1 Figure 1 is a simplified illustration of the existing cask. The drawings in Appendix A are of the existing cask and proposed spool piece. Figure 2 is a line sketch of the model used in the drop-test::. Figures 3 through b snow the various components of the one-quarter-size model of the existing 700-seriesfuel shipping cask and proposed spool-piece extension before the drop test.
Ii By comparing the drawings of Appendix A with Figure 2, the major simplifications which were made in the model can be seen. Briefly, these simplifications were 4
(1) The inner cavity wall was made a constant diameter through both the spool-piece !
and the cask boc:y. This necessitated a smaller diameter lid plug. The shear strength l1 of the lower portion of the spool-piece (Figure A-2)'vas simulated by placing a cylinder
{
in the center cavity (Figures 2 and 5), the wall thickness of which corresponded to i the thickness of the two steel cylinders in the lower portion of the actual spool-piece.
(2) The bottom of the model cask was a flat steel plate, the thickness of which corresponded !
to the thickness of the convex bottom plate of the actual cask. The square bottom edge i of the model was we,aker than the rounded bottom of the actual cask.
- H. G. Clarke, Jr. and W. E. Onderko, "Model Impact Tests Pertaining to Shipping Containers for Radioactive Materials." Summary Report of AEC Symposium on j
Packaging and Regulatory Standards for shipping Radioactive Material, nu<t tiDI, j Office of Technical 5efvices, Us. Department of Commerce, Washington, D.C.1962. *
Ibid, p 2 41. ]
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SECTION A - A WEIGHT OF WODEL APPR0XIMATELY 500 lb.
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r (3) The entire base support was omitted from the model. This made the bottom of the model cask considerably weaker.
(4) The welds of the model were not radiographed as were the welds of the existing cask.
l (5) A lif ting bracket was added to the model cask (Figure 3) to permit raising and dropping " '
the cask at variouc angles of inclination. I The model was sprayed with several coats of clear lacquer in the hope that the lacquer would crack upon impact, showing stress lines. It should be noted that the lacquer peeled rather than cracked as shown in Figure 21, for example.
1 THE DROP TEST -
I l
A requirement of 10CFR72.32 is that the cask be capable of withstanding an impact force caused by a free fall through a distance of 15 feet upon an unyielding horizontal surface without either I (1) Exceeding the ultimate strength of any structural portion of the cask, or (2 ) Deforming the cask to an extent which would permit the escape of fuel elements j or portions of them, or permit the level of external radiation to exceed 1 Roentgen per hour at any point 1 meter from any accessible surface of the cask. j I
The first step in determining the drop test location was to find an " unyielding" surface.
I The facility chosen for this purpose was the concrete pad in the VBWR parking lot. This concrete pad is 12 inches thick and has two courses of 5 '8 inch steel reinforcing bar in a 12-inch grid in it.On this slab were placed two plates of carbon steel. The bottom plate was 4 feet x 8 feet x 1 '
inch. The top plate was approximately 4 feet 4 f eet 1-1;2 inches.
A mobile crane capable of raising the lower edge of the model 15 feet above the drop surface
- was provided by an outside contractor. The drop site and equipment are shown in Figure 9.
A remote release hook (Figure 10) was used to drop the model from a height of 15 feet in a
,' free f all. A "Releas- A-Matic Hook," Model H44-3 was used. A 35-foot length of manila line was attached to the release hand!e on the hook which permitted an individual standing a safe distance from the point of impact on the ground to open the jaw of the hook. A simple pull on the line opened the hook which dropped the cask in a free f all.
i To mahc an accurate record of the drop test, many photographs were taken of the actual -
drop.s as well as of the model af ter each drop. Two motion picture cameras were used (one as
- a backup) to photograph the actual drops to determine the actual cask deformation upon impact.
Still photographs were taken befut e and ailer each drop to record the permanent deformation of bl the model.
I 7
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After being postponed because of inclement weather, the drop test was conducted on the morning of March 13, 1964. I I
Before dropping the model, test drops were made with one lead brick in a gallon bucket to l practice the timing of the start of the high speed camera. On the second of these preliminary drops, the remote release hook failed to open. The release mechanism of the hook had become worn. This required repair before further testing could be done. Figure 11 shows the pre-liminary test set-up.
The drop test consisted of two drops. The first drop was to impact the model on the upper edge of the lid and cause large shear forces on the joint between the spool-piece and the cask body. The position of the model just before being dropped is shown in Figtire 12.
The second drop was designed to place large tensile forces on the connecti?g bolts and l shear forces on the joint between the spool-piece and the cask body. This was dune by dropping l the model, in a horizontal position, on a length of 2-1/2 inch carbon steel round stock placed normal to the longitudinal axis of the model, i
The position of the model just before the second drop is shown in Figure 13. }
l The results of these two drop tests are summarized in the following sections. [
A. Drop 1 Results .I l
In Drop 1, the model was dropped so that the upper edge of the spool piece extension made the initial contact with the steel plate. Figure 14, a series of frames from a high-speed movie, I shows the deformation of the model at the instant of impact. ' l Figures 15 through 18 are close-up photographs which show the model deformation after the i drop. Figure 15 shows the major deformation. The initial impact caused the 'large flattening of the extension spool piece. Figure 16 shows the slightly bent lid. This deformation of the lid '
elongated bolt 3 (Figure 23C 1,32 inch). !
A small amount of rain ~ water had collected in the cavity of the model cask before the drop f test was performed. Figures 15 and 17 show that this water leaked out upon impact. Figure 17 j l also shows the sme.Il impression the model left on the steel plate during the initial contact. This loss of water is not criticalbecause the cask does not require water for either shielding or heat -
transfer; the cask can be shipped dry. .
After the model initially hit on the upper edge, it rebounded (Figure 14) and landed on the ' '
lower side, which deformed the lower side and bottom of the model. Figure 18 shows this -
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l deformation in detail. This figure also shows the failure of the weld around the lead pour hole in the bottom of the model. This failure would not occur in the actual cask because: (1) these lead pour holes are not present in the existing casks, (2) the bottom of the existing cask is convex,
)
and (3) the bottom supports of the existing cask would absorb most (probably all) of the impact l
energy imparted to the bottom of the cask. It should be noted that all welds in the existing cask il and the proposed extension will be checked for defects. The welds of the model were not checked
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for flaws.
Figure 19 gives the dimensions of the deformation the model received from Drop 1.
It was concluded that no portion of the model failed (except a poor weld which, for reasons stated above, is inconsequential) under the severe impact conditions of this, drop. Some structural members had permanent deformation, but they still satisfied the requirements of 10CFR72.32 ,
(proposed).
B. Drop 2 Results l
Drop 2 was made with the same model that was used in Drop 1 Drop 2 iniparted a force to the model which placed the joint between the spool-piece and the cask body in taision and shear. l' The impact also placed the outer shell in tension. This was accomplished by dropping the model in a horizontal position onto a length of 2-1/2-inch round stock. Figure 20, a series of frames from a high-speed movie, shows the instant of initial impact in Drop 2.
l Figure 21 shows how the side of the model was deformed from the impact on the 2-1/2 inch !
round stock. l The tensile forces imparted in the upper shell of the cask body by the initial impact were great enough to cause the connecting bolts to yield, The maximum elongation was 3/32 inch (Figure 23). In addition to elongating the bolts, the upper plate of the cask body was pulled out-l ward by the bolts causing the permanent deformation shown in Figure 22. The sketch in Figure j 23 shows this deformation in exaggeration. !
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MODEL TEST FULL SIZE MODEL TEST FULL SIZE l RESULTS, RESULTS. RESULTS, RESULTS, DIMENSION INCHES INCHES DIMENSION INCHES INCHES a 1/16 14 j 11'8 41'2 b 3/16 3 '4 k 213'16 111 4 e 1/8 l '2 ( l1332 5 l '8 '~
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APPENDIX This appendix contains the engineering drawings of the existing cask (Figure A-1), the proposed spool piece extension (Figure A-2), and an assembly drawing of th'. cask and spool-t piece (Figure A-3).
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OCT 2 81964
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Anemensat to License e l-130 Gemaml Electric Campesy Ataic Power Equipment Department 175 curtmer Avemme ama Jose, calitsemia 95103 .
Attention: Mr. E. L O'anntes, General Eneager 1rsatistion Proessaing Opstanties Samtlemmat pareumst me ===a.atotoyour license ===a=P=t request dated Ames 36,19fA
( este, Liesame Me.38-130 is hereby amaneed te metharise a omries of shipments of irsediated femal raua year bilmettos Atomie laboratory to the flamiamiam's Savammah River FLast, in the entifica 700 series easks. this lieanse amenement is subjoet to the 2n11estag sentitiems:
- 1. Emoopt as spesitteelly providet by other memettiens et te license, the transfers hereby 11eenset shall be ones la accordamos with the statammets, representations, and passetures sentaimaa la Someral E1ertrie coupesy's 11menos enamenset appliantiam ested June 36,19dA, as supplemented to date.
- 2. The 11eemene shall inspect the ensk prior to ensk shipment to deteet possible viksatiemal er other emmage. h aald any sismitieset emmase be metas, the 11eensee shall supert this emet to the sivisten at asteriale r., , gap,e ,
with the details of the methat to be employet to eerseet the damage and =d=8=taa soeurremos, _. _. ._. _ -
3 Se liesmose en11 met transport er sense to be transportes ,
- irradiatet fant elements la emy ansk thiek the 11eemose knowe er has samman to believe is esfestive la emy seaport havtag rf h a potentially significant etverse stemet en the effietammy at Cc -
the aank. A esmagst aesk mest be seteeted before summe. '
A hipumarte maear thie 14a==== shall be limitet be too ensks '
per shipment. A skipannt shall met be pleest utthin'a ~~~ ~~ -
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earrier with other spesial enslaar anterials, nor p1ssed ute.
_ . , _ in toesty feet er other spe Sal amelaar materials at peines of --
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OCT 2 e 1964 [-
landing, W"7 t, sts; rase, er enliary aumeyt that the presense et other ==elame materials is other vakieles '
in the vietaity need apt be seasidered.
i 5 It a abisment is to be maae with liquid samlaat and a ensk l
land producius less thma one kilouatt et secar heat, een !
embient temperatures may be laser thea -800F, the 11eemose ,
shall submit for approval the means to be suployed to prevent fressias er the seelaat. l l
- 6. The 11eensee shall sempir atth the ageiruussts et seestan 72 51 through 73 53 af 10 ms 12 (proposed . These relate respoetinly te mettriaaties er anc, as w)s, ama Insp o stan one Tests.
m ts: Am aze a m or assansens Original Signed by Lyall Johnson l
i.au se , estias atr=t., 4 alvid ea of Materials Lisemetag Bac1caurus amfety Aealysis Distribution:
Addressee 1 Supp4 - 05$
REG. Read.
~INL Read.
Compliance (2) .
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9 UNITED STATES ATOMIC ENERGY COM(ISSION SAFETY ANALYSIS BY TIIE IRRADIATED FUELS BRANCH r
DIVISION OF MATERIALS LICENSING IN THE MATTER OF GENERAL ELECTRIC COMPANY DOCKET NO.70-15b SLNMARY By letter of June 26, 1964, General Electric Company applied for an ,
amendment to Special Nuclear Material License SNM-130 to authorize I the shipment of approximately 200 fuel element assemblies from the Vallecitos Boiling Water Reactor (VBWR) to the Countission's Savannah RiverPlant(SRP). The application requested authorization to use a modified version of shipping casks previously approved under this license. The casks have been modified by adding an extension which provides fourteen inches of additional cavity length to accommodate longer fuel elements.
We have completed an evaluation of the modified cask design based upon the infozustion and data submitted and have concluded that the design of the equipment is adequate to protect health and minimize' danger to life and property. In making our evaluation, we used as criteria the requirements in 10 CFR 72 (proposed) " Protection Against Radiation in the Shipment of Irradiated Puel Elements" as published in the Federal Register on September 23, 1961. These proposed regu-lations include requirements governing license applications, general {
packaging, cask construction details, shielding, materials and methods of construction, nuclear safety, heat removal, shipping procedures, testing, notification of AEC, records and inspection.
SUBKITTALS I,etter dated June 26,196k, with Application Amendment No. 25 to License SNM-130.
Letter dated July 29, 1964, Modification No. 1 to Application Amendment No. 25 l
Letter dated September 9, 1964, Modification No. 2 to Application Amendment No. 25 . 1 Letter dated October 2,1964, Modification No. 3 to Application -
Amendment No. 25 l
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, DESCjtlPTION OF CASK Gena}-g Shape: An upright circular cylinder with integral base.
Size: 37 inches in diameter by 77 inches high, including a removable extension but exclusive of attachments.
Construction: A lead-filled, steel weldsent with stainless steel piping. A 4-inch thick, wooden fire shield surrounds and is attached to the cask.
Capacity: Ten VBWR irradiated fuel elements.
Loaded weight: 28,500 Pounds, exclusive of the 2,200-pound fire shield.
Body Outer shell:
Original ca,sk, 3/8-inch thick AS1M A-212 steel plate 37 inches in diameter by 56 inches high, with a dished head botton 7 inches deep.
Extension, j-inch thick AS1M A-304 stainless steel plate, 37 inches in diameter and 14 inches high in the exposed portion, 21 inches in diameter, and 10 inches high in the portion which extends into the cask.
Cavity:
l Originalcask,15inchesI.D.by40tincheshigh,l-inch AS1M A-240 stainless steel shell and bottom plate.
Extension,15 inches I.D. by 14 inches high, l-inch AS1M A-304 stainless steel.
Closure: Top of extension recessed for shielding plug; eight, 3/4-inch diameter by 14-3/4 inch long stainless steel bolts are provided for extension hold-down and for lid Closure.
Penetrations: A siphon drain which terminates in a pipe plug on the upper surface of the cask extension and a vent line from.
- the side of the cavity which terminates in a valve.
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Shielding thickness: ApproximatelylOhinchesoflead.
Shielding expansion void: None.
Lifting device: Twovertical,12-inchstructuraltees,3} feet long, velded to opposite sides of the cask exterior shell, and to the base, with reinforced lifting slots located in the web.
Tie-down device: Circumferential, wrap-around wire cables which pass through the cask lifting slots and the fire shield for tie-down to the truck floor.
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j Temperature sensing device: None. !
Pressure rating (with closure): Cask designed for 20 psig and tested at 30 psig.
Valves: Stainless steel valve provided for venting.
Heat dissipation fins: None.
Primary coolants. Water or air.
Means for sampling: By use of vent valve and drain line, both of which are closed by pipe plugs during transit.
Pressure seal: A vide, flat, neoprene rubber gasket between the cask body and the extension.
Lid Shaper Flat plate and cylindrical plug.
Size: Plate is 30} inches diameter by }-inch thick, and plug is 21 inches diameter by 10 inches high.
Construction: Stainless steel veldment, lead-filled.
Penet mtions: Pipe extends from bottom of plug to filtered relief valve on top.
Pressure relief device: Belief valve set to operste at 15 psig..
Filters: A high efficiency, fiber glass filter on the discharge of the relief valve. .
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Shielding expansion void: None.
Lifting device: A 3/4-inch thick vertical plate is welded to the lid with holes to accomunodate lifting equipment.
Pressure seal: A flat, neoprene rubber gasket provided between ,
top plate of lid and top of cask extension. {
l Fuel Basket i Shave: "Eggerate", with ten 3-inch by 3-inch compartments inside a cylindrical shen.
Size: 14-5/8 inches diameter by 36-3/16 inches itigh.
Construction: Fabricated from aluminum, and altaminum-clad cadmium neutron poison strips.
Capacity: Ten fuel elements held verticany in a 3-4-3 array.
Lifting device: Aluminum plates welded to basket at opposite sides, with holes for lifting hooks.
Fuel Cans: Ruptured fuel elements or loose rods will be enclosed in cans. Two different size cans will be used. The aman cans win be placed in the fuel basket whereas the basket win be removed when using the large cans. A spacing insert will be used to position 6 large cans in the fuel cavity.
Protective Structure Shape: Box-shaped fire shield enclosing entire cask including lifting devices.
Size: 68 inches by 57 inches, by 97 inches hish.
Construction: Four-inch thick redwood itsaber, supported from within by a steel framework which is attached to the cask.
Fuel Total: Approximately 200 miseenaneous fuel elements used in various development programs are to be shipped. Esch fuel element .
vin contain 9,16, or 25 rods and win weish face 10 to 14 3 kg. Fuel cladding win be stainless steel, Inconel, Incoloy, '
i Bastency "X", or Zirconium. In addition, a quantity of
- disassembled rods win be shipped. Since the disassembled '
rods as well as some of the fuel assemblies win have intentional defects they win be shipped in special containers.
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Maximum cask load: Ten fuel elements containing no more than ;
5,500 grams U-235 Assuming a maximum operating power of k50 kw per element, a cooling time of 180 days, and an operating time at power of 500 days, a cask load of ten such elements would contain a calculated 550,000 curies of fission products and produce 1,850 watts of decay heat.
SAFETY EVALUATION The design specifications for the cask and procedures for its use appear to meet the requirements of 10 CFR 72 (prcposed) provided that the recommended conditions are fulfilled. Discussed below are the results of our review of the cask design and procedures for its use, against the pertinent criteria contained in Parti 72. The Pmposed License Conditions appear at the end of the analysis.
Handling Equipment and Procedures 72.21(a)(6) Handling equipment and facilities must be adequate at loading, transfer, and unloading sites.
The loading facilities and equipment are the same as we have ;
l previously approved. No transshipment will be required and the unloading vill be done at a Commission site.
72.21(c)(2,3.k) Procedures must be adequate for loading, shipping, ,
l unloading, special precautions, and emergencies.
These procedures vill be the same as we have previously approved.
72.21(c)(5) Procedures must be adequate for measuring temperstures, I contamination and external andiation levels; for testing l l
1eaktightness; and for confirming effectiveness of neutzen poisons.
The procedures used for previous shipments will be adequate when supplemented as noted in the application.
Structural Integrity I 72.32(a) Cask must withstand a distributed load of 10 times its w n veight when supported at its ends as a simple beam, without exceedig vlt Nate strength.
The designe.*'s calecations show that under these conditions, .
the animum Mresw on the bolts, which attach the extension -
to the original cask, are 35,500 psi in ahear, and 117,500 psi in tension, wherer.s ultimate strength of these bolts are stated to be 108,000 psi in shear and 180,000 psi in tension. The
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! " 1 r maximum nexural stress in the exterior shell was calculated by the designer to be 9,740 pai, based on the assumption that ;
j the lead-fill vould prevent buckling. Our more conservative analysis which allowed no credit for lead stiffening of the outer snell indicated a maximum stress of 68,000 psi which is below the 70,000 psi ultimate strength of the shell material.
72.32(b) Cask must withstand a specified 15-foot free fall or an impact force of 60 times the weight of the cask.
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A simplified, one-quarter size model was built and subjected to tyd 15 foot drop tests onto a 12-inch thick concrete pad j topped by a 1hinch steel plate. In the first test the model J was dropped on the top edge of the cask extension. Although ;
some structural members had permanent defomation, no portion )
of the model failed, i.e., exceeded the ultimate strength of j the materials of construction. The second test consisted of J a side-drop onto the side of a 2hinch diameter bar in order to subject the cask extension attachment bolts to a severe impact condition. This side impact caused some of the connecting bolts to yield in tension, with a marimum permanent elongation of3/32 inch.
The results of the drop tests shoved satisfactory strength of the extension, and its attachment to the original cask. The mode 3 tested did not have lifting devices, support beams or a wooden fire shield all of which are expected to give additional impact absorption strength to the full-scale cask. It is, there-fore, concluded that the modified cask design has satisfactory strength against the effects of the specified 15-foot free fall.
72.32(e) Cask closure must withstand 60 times its weight plus contents, or 15 times the weight of loaded cask.
Our calculations show aatisfactory strength of the eight, 3/4" hold-down bolts to withstand 60 times the weight of the parts involved, which includes the weight of the extension.
The analysis showed a mar 4=u= stress of 150,000 psi in tension, which is less than the 180,000 psi ultimate tensile strength for the specified high-strength attachment bolts.
Although the designer's calculations show that the attachment bolts alone es.nnot withstand a horizontal force of 15 times the full weight of the ' ash, c a satistsetory analysis was pre- ,
sented shcring compliance with the requirement. This analysis l
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o took advantsge of a slight yield in the bolts, which pemits credit to be taken for the strength of the 3-inch thick extension wall. The assumption that the attachment bolts will 1 yield but not fail, was confirmed in the 15-foot drop test on. {
a2}-inchdiameterbar. The combined shear strength of the l
attachment bolts and the 3-inch thick extension wall was calcula-ted by the designer to safely withstand the required impact force.
72.32(d) Scaled cask must withstand an internal pressure of 20 psig f or twice the operating pressure.
Our calculations indicate that the cask cavity could withstand the design pressure of 20 psig. For the proposed shipments, 3 the cask will not be pressurized.
Internal Structural Components f 1
72.33(a) Fuel holders must be adequate to avoid criticality in the event of specified accidents.
Construction details indicate the strength of the basket is adequate to prevent significant fuel rearrangesent..
72.33(b) Neutron poisons must remain effective.
The cask design indicates that the poison will retain its i effectiveness durin6 the specified accidents and the pro-cedure for experimentally verifying the presence of the poison appears adequate.
72.33(c) Internal fuel container must be leaktight and mechanically adequate.
The internal fuel containers are meelu.nically adequate but f are not leaktight. However, the licensee has shown by test that failed fuel of this type vill not release into water coolant, activity in excess of 1/10 that permitted by 72.45(d) during the time normally required for shipment.
Exterior and Attachments 72.3h(e) Cask exterior must withstand an impact of 30 times loaded weight over 6-inch diameter area or must have a specified .
minimum shell thickness. .
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o 8 Part 72 (proposed) requires a h inch tnick steel weldment for a cask of this weight. The extension has the required thickness; however, the original cask, which was approved prior to the ;
release of Part 72 (proposed), has only a 3/8-inch thickceso. j The addition of the 4-inch thich redwood fire shield which will )
surround the cask during traneport is considered sufficient to i compensate for the 1/8-inch deficiency in steel thickness of !
l the original shell.
72.3h(b) Lead shielding must be completely encased in 1/4-inch thick steel with all joints welded, with provisions to accommodate themal expansion, and with no fusible plugs 1 to lose shielding.
The lead shielding is encased in a steel veldsent having a j minimumthicknessofh-inch,exceptforthethreadedpipecap assemblies enclosing the nuts for the lid and extension attachment bolts. These fittings haver been tested at 15 psig pressure to assure the adequacy of ses.1 provided by the threaded connection. Our evaltation of the cask indicates that the wooden fire shield will prevent the lead from melting in a standard one-hour fire.' Acco dingly, the usual requirements for thermal expansion are not applicable to this des 16n. Fusible j plugs are not used. !
72 3k(c.d;e.f) Cask must have adequate tie-down devices for 2 times )
loaded weight vertically and 10 times horizontally, and lifting devices for 6 times loaded weight.
The cask tie-down devices do not have the strength to satis-factorily resist the required horizontal lateral force of 10 j times the weight of the loaded cask. Calculations indicate j that a horizontal, lateral force of bg would cause damage j
- Tests have recently been performed in both the U.K. and U.S. wherein shipping containers employing wooden protective shells have been exposed to fire.- The tests have shown that the containers thus protected suffered very little damage in proportion to that which occurred when they were similarly tested without the benefit of protection. On the basis of these test results, it is reasonable to conclude that the protection afforded by the wooden fire shield surrounding the cask is sufficient to prevent any significant melting I of the lead shielding as a result of exposure of the protected ,
cask to a standard one-hour fire.
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to the cask lifting devices. The weakness in the tie-down system is justified however, in that: (1) the damsge to the lifting devices would be slight and would in no way affect the operability and safety of the cask; and (2) by procedure the licensee vi n assure that the tie-down method which will be used win not subject the cask to latersi forces greater than kg.
Our calculations indicate that the most severe st n ss produced by the above described lifting forces vin ~ be a shear stress of approximately 12,000 psi in the cask lifting devices, which is wen within the material yield strength in shear of 18,000 psi.
72.%(a) Cask must withstand vibration in normal transit.
Cask structure, lead-fill, and appurtenances appear capable of withstanding nozual vibration in transit. Fuel may rattle in j basket, but damage if any would be principany to bottom end J fittings and insignificant to safety. See Condition No. 2.
72.%(h) Provid'e security seal on closure. j One lid hold-down bolt has been drilled to accommodate a wirs security seal. j l
72.%(i) Provide means for measuring internal cask wall temperature.
Measurements made on previous cask loadings have sheen that the internal cask van temperature can be adequately deduced from the external temperature.
1 72.%(.i .k.1) Attachments must be protected from damage to avoid j doseratesover1r/hratonemeter,leakagemustbe j retained, and valves must be locked.
Attachments are protected. Since the cask is equipped with a siphon drain, leakage during normal condition transport cannot occur. The vent .sive is locked.
72.M(m.n) A pressure zelief device must be provided on a pressurised cask, and vent or relief devices must be filtered and l protected.
A pressure relief valve'has been provided and set for 15 Psig. , !
A high efficiency fiber glass filter has been instaued on i the discharge of the relief valve. .
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r 72.3k(o) Iodine release from cask must be limited to 10 curies I-131 or 1 curie I-129 in the event of specified accidents.
The quantities of I-131 and I-129 for these shipments win be less than the allowable amounts, For this reason, no iodine removal device has been instaned.
72 3k(p) Pmtect against possible explosive mixture in cask.
Experience with water-cooled casks which contain this type and quantity of irradiated fuel has indicated that explosive mixtures vin not be fomed. -
72 34(o) Provide means for sampling.
Removable pipe plugs are provided on the drain line and the vent line valve.
72.3k(r) Provide means for equalizing internal pressure with atmosphere.
The cask is equipped with a vent valve.
72.3k(s) Assure adequate pressure relief capacity.
1 Based on the analysis by the licensee, and experience frca '
previous shiIments of fuel elements of essentially came irradiation history, no significant pressure vill develop.
A procedure has previously been submitted which describes a safe method for reintroducing coolant into a dry cask.
72.3k(t) P m vide pipe connections for removing liquid.
Drain and vent lines are provided.
72 3k(u) Piping through lead must withstand defomation.
The design of the lead cavity and cask piping indicates that no significant deformation shouM result from normal ham 11ag and
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transport.
72.3k(v) Pmvisions to prevent freesing if it would $sM.r cask efficiency or damage the cask or contents. ,
Cask vi n oe shipped either wet or dry. If dry, there abould be ,
no concer;.. If a lfquid coolant is used, see Proposed License Condition No. S.
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c Shielding 72.35(a) External radiation levels must not exceed 200 nr/hr at the accessiblesurfaceor10mr/hratonemeterfromthesurface (or 10 mr/hr at three meters from surface for exclusive use shipment).
Specified preshipment tests will assure that requirements are met.
72 35(b) Radiationlevelmustnotexceed1r/hratonemeterafter specified accidents. )
Our evaluation indicates that neither a standard.one-hour fire lacement norimpactconditionsvillproducesignificantleaddisp/hr or deforination. Accordingly, the radiation level of 1 r vill not be exceeded.
72.35(c) Protect against excessive radiation beaming.
No calculations have been submitted; however, specified pre-shipment tests vill assure that this requirement vill be met.
72.35(d) Shielding must not change position during nozinal transport or if melted.
The cask design indicates that the lead cannot shift in nozise.1 transport, nor is it expected that the lead vill melt in a standard one-hour fire.
72.35(e) Pipes containing radioactivity must be shielded to meet requirements of 72 35(a).
Shielding not required as only insignificant quantities of solid radioactive material could reach the valve or pipe plug, and release of solids or gases from the fuel is not anticipated.
Materials and Methods of Construction i
72 36(a) There must be no chemical, galvanic or other reactions between cask materials or with fuel.
Based on:past experience with the cas[, no adverse reactions will occur during the transit time required for these shipments.
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o 72.36(b c) There must be no inaccessible pockets or crevices in
, external surfaces of cask, and exposed surfaces must be free of pits, cracks and porocity.
Previous experience with original cask indicates external cask surfaces are reasonably accessible to decontamination.
The modified cask with the extension should also be relatively easy to decontaminate.
72.36(d) Welding must be workmanlike, free of defect, and the -
mechanical efficiency and melting points not less than 85% and 1000 0 F, respectively.
Welding on the cask extension will be adequate, 'sceording to the specified tests. Drawing specifications assure adequate welding on the original cask.
72.36(e) P mvent holes in shell through which there could be a loss of shielding material.
See7234(b).
Standards for Control of Criticality 72.37(a) Fuel in one cask must not exceed 75% of the nusber of elements for criticality, or the k,ff must not exceed 0 9 GE calculations of k show values ranging up to 0.82, for the casks when loaded wikthe most reactive fuel and with allowance for extension of the fuel beyond the poisoned basket. Our independent calculations confirm that k for the curzently proposed loadings will not exceed 0 9. ,ff 72.37(b) For7237(a),assimmecaskisfloodedandimmersed, fuel is at marimun reactivity, structursi materials and neutron poisons are at minimum effectiveness, and include considers-tions of the effects of specified accidents.
- - These conditions were assused. For the case of exposure of the cask to a s+mndard one-hour fire, the protection provided by the fire shield would prevent melt-down of the fuel or the cadmiust poison.
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- # 72.37(c) Each shipment must be constituted to avoid criticality from interactions among casks, or between the casks and other special nuclear material in transit.
See Proposed License Condition No. 4.
Heat Removal 0
72.38(a)(1) Accessible surface temperature must not be over 180 F, surface in contact with dunnage or vehicle not over 350 F (or 1800F in water transport).
Specified preshipment test will assure meeting requirements for the proposed shipments.
72 38(a)(2) Primary liquid coolant must not be circulated outside of the cask, must remain 200 F below boiling, must not cause significant corrosion of the fuel and must not reach a pressure in excess of 50 psig or 50% of design pressure, l whichever is lower. In addition, any liquid coolant must not freeze nor undergo any detrimental chemical changes.
The temperature of the water coolant will be measured for each shipment to assure that it will be 20*F below its normal boiling point. There will be no significant corrosion and the cask pressure will be negligible. The applicant also states that, based on previous experience with the cask, the liquid coolant will not undergo any detrimental chemical changes. Regarding freezing, see Proposed License Condition No. 5 72 38(a)(3,k) Normal maximum fuel temperature must not exceed 3000 F, or temperature in reactor, or 300c7 below fuel failure temperature, whichever is highest.
Tests were conducted with the cask, using electrical besters of fuel element configuration. The tests indicated that the par 4=um decay heat load anticipated in the proposed shipments would produce a fuel cladding temperature of 255 F in the dry cash. This is lower than the fuel temperature in the reactor, and more than 300 F Lelow failure temperature.
1 72.38(b) Loss-of-coolant fuel temperature must be at least 2000F below failure temperature.
, i See 72 38(a)(3,4).
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- 14 -
- 72.38(c) Calculations of temperatures must assume still ambient air at 1000 F, marim'a solar load, and loss of any sunshade if ,
coolant is lost.
The heat transfer tests were conducted in an ambient temperature of 870F. Extrapolation of the test results to correct for an ambient tempersture of 1000F would produce a negligible change.
Other Considert.tions 72.k1 A lenktight internal container must be provided for defective fuel.
Elemente containing rods with exposed fuel, or disassembled rods with exposed fuel, will be shipped in vented cans. See paragraph T2 33(c) above.
72.k2 A defective cask must not be used.
See Proposed License Condition No. 3 72.k3(a) Requires tests specified in 72.44 and 72.45 The submitted cask specifications and procedures indicate the t
performance of such tests.
72.k3(b) A cask involved in an accident must be ratested before reuse.
See Proposed License Condition No. 3 72.k4 Preliminrry tests for heat transfer and shielding.
The tests have been completed and satisfactory results were obtained.
72.k5 Boutine and periodic testa for radiation level, contamination, pressure, coolant activity, and neutron poisons.
Tests are specified. ,
72.46 Preshipment temperature and pressure measurements must be made.
The applicant has stated that these tests wi21 be perfezued. ,
I. .. . -
r .
9 PROPOSED LICENSE CONDITIONS Inordertoassurecompliancewith10CFR72(proposed)thefollowing I are recommended as conditions of a license issued to authorize transfer ,
I of the specified fuel elements in -the cask described:
- 1. Except as specifically provided by other conditions of the license, the transfers hereby licensed shall be made in accordance with the ;
statements, representations, and procedures contained in General 1 Electric Company's license amendment application dated June 26, 1964, as supplemented to date.
- 2. The licensee shall inspect the cask prior to each shipment to j detect possible vibrational or other damage. Should any signi-ficent damage be noted, the licensee shall report this fact to the Division of Materials Licensing, together with the details of the method to be employed to correct the damage and minimize recurance.
3 The licensee shall not transport or cause to be transported irradiated fuel elements in any cask which the licensee knows or has reason to believe is defective in any respect having a potentially significant adverse effect~ on the efficiency of the cask. A damaged cask must be retested before reuse.
- 4. Shipments under this license shall be limited to two casks per shi; ment. A shipment shall not be placed within a carrier with other special nuclear materials, nor placed within twenty feet of other special nuclear unterials at points of loading, tranashipment, storage, or delivery except that the presence of other nuclear materials in other vehicles in the vicinity need not be considered.
)
5 If a shipment is to be made with liquid coolant and a cask load !
producing less than one kilowatt of decay heat, where ambient temperatures may be lower than -200 F, the licensee shall submit for approval the means to be employed to prevent freezing of the coolant, j
- 6. The licensee shall comply with the requirements of Section 72 51 through ;
72 53 of 10 CFR 72 (proposed). These relate respectively to '
Notification of AEC, Records, and Inspection and sts. l
)
A. E. Aikens, Jr., Chief -
Irradiated Fuels Branch '
Division of Materitis Licensing 4
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GENER AL h ELECTRIC "uc"".y"saar D l vI s i o NReg.. -.
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." -; Ille C. y, vAttscitos Nucts AR CENTER. vAttacrTOs Ro AD. PLEAsANTON. ca tiro R NI A 94566 1R R ADIAflON PROCf sslNG OPER ATloN (415) 862 2211 l W
F .. l Se ptembe r 18, 1968 e . :' q l F l
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R. B. Chitwood, Chie f l i O Irradiated Fuels Branch '
s - 2i Division of Materials Licensing :
U. S. Atomic Energy Commission > [>
Washington, D. C. 20545 b l2 d
Dear Mr. Chitwood:
General Electric is authorized to use special nuclear material at its Vallecitos Nuclear Center under License SNM-960, Docket 70-754.
Appendix D to the application was submitted on November _.18 _1966-containing information required by 10 CFR 71. Appendix D includes Section 5.0, " Irradiated Fuel Container Systems." Action on Section 5.0 is currently pending.
j General Electric has been engaged in a program to develop improvements to its irradiated container systems and hereby submits information which shall {'
comple tely supersede Section 5.1 of Appendix D, concerning the GE Shielded Container - Model 100. In sub3Equent submittals, it is anticipated that much, if not all, of the information contained in Section 5.0 will be replaced or substantially modified.
The protective jacketing concept utilized for the Model 100 is planned for a number of other (including larger) models for subsequent submittal.
Wood jacketing will be employed for the remaining GE shielded containers.
Sincerely, !
, Qh,. -
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Walter H. King Administrator-Licensing WHK;pc 'y [.!
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1R R ADI ATjoN s(RyjC($ AND R ADiotSO70PE 5 FOR RE5t ARCH AND INDUSTRY
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" $ h 5.0 IRRADIATr D FUEL CONTAINER SYSTEMS 5.1 General Electric Shielded Container . Model 100, J
5.1.1 Package Descr:pt;on . Fachaging tal Geaeral All containers of this model, for purposes of constructing additional contateers of this model, will have dimensions of plus er minus five percent of the container dimensions s pecified in this application, and all lifting and/or tiedown devices for additic.nel containers of this model, if different from the lifting and/c,r tiedown devices described in' this ]
l application. nill s atis'y the requirements of 10CFR71. 31 ;
(c' and 'd).
Shape: An upright circular cylinder with attached base.
l l
Size: 201/4 inchee diameter by 271/8 inches hich.
i l
Construction: A lead Elled steel weldment with sta'.nleR F F reel piping. A double j we?. led protective jacket of 1/4 inch steel plate surrounds the cask during transport, j Vf e ight: 3250 pcunds exclusive of liners, j e.d e': and pallet. .3370 pounds with lead-f'lled l'.ner, 3520 pounds with uranium-filled liner, and 3510 pounds with tungaren-f811ed liner. The pro-Liconse %. wu _ o (, n Dod e No. ~'O ~54 Sect. No. 5.1. I p. ,
Appendtx D Amend. No. 2 p, gmems,y s n;68 NEV/
A.,ng, $.e,,(,) l'
External Shell of Cask, v.
Top, 16 5 ' F ta External Shell of Cask, Bottom, 156
- F Protective Jacket, out-side surface, Top 99'F Protective Jacket, out-side surf ace, Side, 95'F Ambient Temperature, 78'F
- Based on 64.2 curies of Cobalt 60 per thermal watt. Watt equival-ents for various isotopes are standardized on a curie basis and deter-mined from the total energy expended per atomic disintegration.
5.1. 3 Package Evaluation There are no components of the pack-(a) Gener al i
aging or its contents which are subject i!
to chemical or galvanic reaction; no coolant is used during transport. The J protective jacket is bolted closed during transport. If that portion of the protective jacket which is used in the tiedown system or that portion l which constitutes the principal lifting l device failed in such a manner to allow the protective jacket to separate i from the tiedown and/or lifting devices <
the basic protective features of the protective jacket and the enclosed cask would be retained. The package, (con-70 754 Sect. No. 5.1. 3 Page License No. s N u _o/>n Deeket No.
Appendix D 2 Det. Septembe r 16, 1968 Amend. Sect.( ) NEW -
,8 Amend. No.
i
F tents, c ast anu pt -wo m r....,
. ] l regarded as a simple beam supported
'. at its ends along its major axis, is i capable of withstanding a static load, normal to and distributed along its entire length, equal to five times its fully loaded weight, without generating }
stress in any material of the packaging 1
in excess of its yield strength. The 4 f
packaging is adequate to retain all contents when subjected to an external ;
l pressure of.25 pounds per square inch i
I The calculative methods employed in the design of the protective jacket subjectec to the tests described in this section ar based on strain rate studies and calcu-Intions and on a literature search'of the effects on materials under impact con. ;
)
ditior s . The intent was to design a pre i
tective jacket that would not only satisi .
the requirements of the U.S. Atomic Energy Commission and the Departmer of Transportation prescribing the pro-cedures and standards of packaging an-shipping and the requirements govern-ing such packaging and shipping but would protect the shleided cask from 70 754 Sect. No. 5.1. 3 - page SNM o60 - Docket No.
License No. Appendix D p ,, September 16, 1968 Amends Sect.(s) _ NEW 9 2
Amend.No.
h any deforn ation in the event of an
' accident. In the event that the package was involved in an accident, a new pro-tective jacket could be readily supplied e
and the shipment continued with minimal time delay. To ' demonstrate the effect-iveness of the strain rate calculations and engineering intuitiveness employed in the protective jacket design, the hypothetical accident condition tests were conducted without benefit of the energy absorbing pallet.
'(b) Normal Transport Conditions Thermal: Packaging components, i. e. , steel shells and lead, uranium / tungsten shielding, are unaffected by temperatur<
extremes of 40*F and 130*F. Package contents, at least singly encausulated or contained in specification 2R con-tniners but not limited to special form, will not be affected by these temper- i ature extremes.
- 1. TID 7651 Summary Report of AEC Symposium on Packaging and Regu-s latory Standards for Shipping Radioactive Material, December 3.5,1962 l
\
- 2. SC RR 65-98 Proceedings, International Symposium for Packaging and '
Transportation of Radioactive Materials, January 12-15, 1965.
SNM 960 Docket No. 70 754 Sect. No. 5 1 1 Page License No. 4 Appendix D 2 p.,. Septembe r 16, 1968 Amends Sect.(s)
NEW 10 Amend. No.
k (c) Hypotnetical Accident Conditions M
Gene ral: Both the drop test and puncture test were conducted on a concrete drop pad 20 feet by 15 feet by 6 feet thick A carbon steel plate,10 feet by 8 feet by 3 inches thick was located on the pad.
. To determine maintenance of container 2.ntegr!ty, the cask cavity was filled with water. A mobile crane capable of raising the cask and protective jacket assembly to a 30 foot height was used.
Drop Test: The 30 foot drop was designed to impac-the package (contents, cask / protective jacket assembly) on the upper edge of the protective jacket, where the lifting and tiedown devices are located. The.
package struck the horizontal steel plate and flipped onto its base. The drain plug on the cask was fractured j l
slightly. This was caused by the angle ;
1 iron energy absorption cushions, in the l protective jacket base, partially collapsing, pinching the drain plug against the base collar. A few droplets of water were released through the damaged drain plug on impact with the drop pad. This problem no longer-License No. SNM 960 Decket No. 70.754 Sect. No. 5.1. 3 p ., .
Appendix D h and.No. 2 -
0,,,so,+,mh - iA io68 hends Sect.(s) NEW 12
exists witn the model 100 shielded I container as the drain plug is recer,6ed i flush with the outer surface of the f
cask shell.
Figure 3 shows a close-up view of the extent of deformation and damage re ceived by the lifting and tiedown devices of the protective jacket at impact. Figure 4 shows an internal view of the fracture of the top portion l
of the protective jacket. The circular dents in the top of the protective jacket were caused at the moment of impact when the cask lid and protective l
jacket were crushed together. The 1
cask lid remained secure during the test; there was no evidence of water leakage through the lid seal. No ild bolts were stripped, and the lid was easily removed, as was the protective !
jacket, after the tests were completed l
. Puncture Test: The puncture test consisted of droppin i the assembly in a horizontal position from 40 inches onto a 6 inch diameter l 8 inch long steel puncture bar. The assembly struck the puncture bar and came to rest as shown in figure 5. T'4 l
70 754 Sect. No. 5 1 3 Pop ':
License No. SNM _06 o Decket No.
Appendix D Septembe r 16, 1968hends Sect.(s) Nrw .13 hand. No. _ 2 Dete _
protective %cket yielded on impac* ba- 2:
no major fractures occurred. No damage occurred to the cas. . Af te r *he tests were complcted, the protecMve jacket was removed without difficult, .
Since the model 100 cask sustained ,
The rmal Test: 1 negligible damage and only minor dam.
age occurred to the protective jacket in the drop and puncture tests, it is reasonable to consider the resultant package, for purposes of thermal resistance, as essentially undamaged.
Accordingly, the package was assessed using the General Electric Transient Heat Transfer Computer Program, version D (THTD), which allows the analysis of the general transient pro.
blems involving conduction, convection and radiation. The program allows the thermal properties of the materials to be entered as a function of temperature and the boundary conditions to be entered as a function of time.
The significant assumptions, approx.
imations, and boundary conditions used for the analysis are listed below:
1 Sect. Noi 5.1. 3 Pop SNM - 960 Docket No; - 70.754 License No. Appendix D NEW 14 A,nond. No. 2- Date S e a' e '" h * " 16 1%8 Amends Sect.(s)
to sect 5.1 is presented as a further explanation of THTD.
Water Immersion: Since optimum moderation of product material is assumed in evaluations of criticality safety under accident con-i ditions the water immersion test veas not necess ary.
Summary and Con-clusions: The accident tests or assessments described above demonstrated that the package is adequate to retain the pro-duct contents and that there is no change in spacing. The re fo re , it is concluded that the General Electric shielded container niodel 100 is ade-quate as packaging for the contents specified in Section 5.1.2 of this section.
5.1. 4 Procedural Controls >
Vallecitos Site Instructions have been established and implemented I to assure that shipments leaving the Vallecitos Nuclear Center (VNC) comply with all specific licenses and special permits required for the j i
various shipping container models utilized by the VNC in the normal i conduct of its business. Routine audits are performed to assure com-pliance with these licenses and permits.
Each cask is inspected and radiographed prior to first use to ascertain that there are no cracks, pinholes, uncontrolled voids or other defects License No. SNM. O Docket No. 70 754 see,. No. 5.1. 4 Appendix D p,
Amend. No. 2 NEW Dete .Sgntembe r 16, 1968 Amends Sect.(s) 16
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FIGURE 2. CASK AND PROTECTIVE J ACKET ASSEMBLY WITHOUT P ALLET Page Sect. No. Illu s t r a t io n Docket No. 70 754 Licens. No. SNM 060 Appendix D NEW _
1 Amends Sect.(s)
Amend. No. 2 Date Seetember 16. 1968
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FIGURE 3. DAM AGE TO PROTECTIVE J ACKET FOLLOWING THE 30-FOOT DROP TEST (CtOSE.UP) r.
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Appendix D l License No.SNM 060 4 16, 1968 Amends Sect.(s) NEW _
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Tungsten Uner Uranium Liner Leod Liner GE IPO Casks No.101 through No.112 Cash Wt. 3250 lb Lead Liner We.120 lb , ,
Uronium Liner Wt. 270 lb Sex Section Outer Jocket 32in. l
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Tungsten Liner Wo. 260 lb c-l l 0 @
Protective Jock et Wt. 570 lb I Cosk Pollet Wo. 365 It- Soo Section inner Jocket 3 0in.
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Cosk Lif ting f ors 28 3/4in. -- '
f _ Inner Jocket 8.d. 22-1/4in. Top View k " "g , s , , .
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FIGURE 6 GENER AL ELECTRIC - MODEL 100 5HIELDED CONT AINER Wesse saw r/*4 ,
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l NUCLE AR ENERGY GEN ER AL $ ELECTRIC /
l o: VISION C0MPANY Re M VALLEClio$ NUCLE AR CENTER. VALLECITo$ Ro AD, PLEAS ANYoN. CALIFoRNI A 94566 tR R ADIAtloN PRoCE$$ LNG oPER Af ton (415) 862 2211 Dec embe r 23. 1968 1 9 4 4
e D W e,E!';p !
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C V:' _'v Q Mr. R. B. Chitwood, Chief Irradiated Fuels Branch
[ h,A .Q. Q~ ,
- s. l Division of Materials Licensing - k W.'i - (' ? )
U.S. Atomic Energy Commission % 4'N l Washington, D. C. 20545 \
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G. I l
l
Dear Mr. Chitwood:
General Electric is authorized to use Special Nuclear Material at its Vallecitos Nuclear Center under License SNM-960, Docket 70 754 Appendix D to the Application was submitted on November 18, 1966, containing information required by 10CFR71. Appendix D includes Section 5. 0, " Irradiated Fuel Container Systems . " Action on Section
- 5. 0 is currently pending.
General Electric has been engaged in a program to develop improve-ments to its irradiated container systems and hereby submits infor-mation on GE Shielded Container Model 700. This container was included in the November 18, 1966, Appendix D submittal. However, General Electric proposes to modify its Model 700 application with this completely superseding submittal. Subsequent submittals will modify and in some cases completely supersede other parts of Section 5. 0. j l
This Application, as well as future submittals, will reference data presented with the Model 100 Application, Section 5.1 of Appendix D to SNM-960. Docket 70-754, dated September 18, 1968, as amended October 25, and December 19, 1968.
Sincerely, S
e ' occnno 2
usuc -
1969 p., -3 eau, c_ N.
Walter H. King 6~ M gastmtf 1 Administrator-Licensing VNC sut 151"5 8 WHK:m sg
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IR R ADI Af ton SE RVICE S AND R ADiol$oToPE S FoR RESEARCH AND INDUSTRY
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,. 5. 7 GENERAL ELECTRIC SHIELDED CONTAINER MODEL 700 (w/w/o Extension)
- 5. 7.1 Package Description Packaging (a) General All containers of this model, for purp.oses of constructing additional containers of this model, will have dimensions of plus or minus 5% of the container dimensions speci-fied in this application, and all lifting and/or tiedown devices for additional containers of this model if different from the lifting and/or tiedown devices described in this application will satisfy the. requirements of 10CFR
- 71. 31 (c) (d). This container will be used with and without the extension. The same protective jacket is used for both situations. I This container is detailed in G.E. Drawings dl7B1170, V13 5C 5573, 197E538,/237E325, 477E400 and 77E401, attached.-
Shape: An upright circular cylinder shielded cask and an upright circular cylinder protective I
jacket with attached square base plate. j l
Siz e: The shielded cask is 36 13/16 inches dia--
meter by 64 3/4 inches high. With the extension, the cask is 78-7/8 inches high.
The protective jacket is 81 inches high by 66-1/2 inches across the box section.
Construction: The cask is a lead filled carbon and stain-less steel weldment. The protective jacket ;
'Is a double walled structure of 5/8 inch j carbon steel plate and surrounds the cask l during transport. . >
l I
License No. SNM %0 Docket No. 70 754 Sect. No. 5.7.I p.g.
Appendix D hen 4 No. 3 Date Decembe r 23,1968 g,g, g ,,,(,) New
Wetgnt: A ne v on ou,g., ..... c _..... _ . . . . _ ,
with extension weighs 28,900 pounds. The protective jacket and base weigh 5,000 pounds.
(b) Cask Body Outer Shell: 3 /8 inch steel plate, 63-1/2 inches high by 36 13/16 inches diameter with 3 /8 inch top and bottom plates. The extension is 1/2 inch thick steel plate 36.13/16 inches diameter and 14 1/8 inches high in the exposed portion, 21 inches diameter- and 10 1/4 inches high for the portion which extends into the cask. The top plate is 1/2 inch thick and the bottom plate is
~
1/4 inch thick.
C avity: 1/4 inch stainless steel wall and bottom-plate, 15 inches diameter by 40.1/4 inches deep. With the extension, the cavity is 15 inches diameter by 54.1/4 inches deep.
Shielding Thickness: 10-9/32 inches of lead on sides, 9 29/32 inches of lead beneath cavit- and 9 7/8 inches of lead above cavity, both with and without the extension.
Penetrations: (1) A 1/2 inch, schedule.40 stainless steel siphon drain from the cask cavity bottom terminating in a valve and fusable pipe plug ,
I on the upper surface of the cask. The valve is guarded by steel channel attached to the adjacent lifting structure.
(2) A 1/2 inch, schedule 40 stainless steel 4 liquid fill line from the side of the cavity to the side of the outer cask shell terminating J
70.754 5, 7.1 p,,
License No. SNM-960 Docket No. Sect. No.
Appendix D New 2 Amend. No. DateDec e mbe r 23,1968 Amends Sect.(s)
l
. , in a valve anc : sable pipe plug, guarded by a surrounding pipe sleeve and covered by the protective jacket during transport.
Filte rs: None.
Lifting Devices: Two diametrically opposed, vertical 12 inch structural tees, 3.1/2 feet long welded to the cask shell and to the base with reinforced lifting slots located in the web. Covered by protective jacket during transport.
Pressure Rating: Design . 20 psig; tested - 50 psig; normally unpressurized; maximum operating - 15 psig, i with and without the extension.
i Water or air, i Primary Coolant: i Means for Sampling: Vent and drain lines; both closed by pipe plugs and covered by protective jacket during t r ans po rt.
1 Closure Seal:
I A minimum 1/4 inch thick flat gasket between extension and cask body when extension is used.
(c) Lid I Shape: Flat plates and a cylindrical plug.
Size: Top plate is 30.1/2 inches diameter by 3/4 inch thick. The bottom plate is 21 inches diameter by 1/4 inch thick. The right cylinder.
is 10-1/4 inches .high.
l Construction: Steel weldment, lead filled.
\
License No. SNM 960 Docket No. 70.754 Sect. No. 5.7.I p. ,
Appendix D <
3 New 3 Amend. No. Date December 23.1968 , g,,,g,) l l
Closure: Eight 3/4 inch .10.UNC.2A by 2.1/4 inches long stainless steel bolts equally spaced 45'
~
apart on a 28 inch bolt circle. The bolts are 14.3/4 inches long when the extension is used. j i
Closure Seal: A minimum 1/4 inch thick flat gasket between body and lid.
Penetrations: A 1/2 inch schedule.40 stainless steel pressure vent line from the side of the cask cavity, through the lid, terminating in a relief valve / filter assembly guarded by a surrounding pipe sleeve and covered by the protective jacket du' ring transport.
Pressure Relief Device: 0 to 15 psi rating.
Filte r: A high efficiency fiberglass or equivalent filter on the discharge of the relief valve.
Covered by' protective jacket during transport.
l Shielding Expansion Void: None, f)
Lifting Device: 3/4 inch thick by 8 inch high by 13 inch long vertical plate welded to the lid with a 1.1/2 inch diameter hole centered 2.1/2 inches from the top edge to accommodate lifting hook or cable. Covered by protective jacket during transport.
(d) Protective Jacket Body Shape: Basically a right circular cylinder with open bottom and with a protruding box section dia.
i 1
License No. SNM 960 ' Docket No. 70.754 Sect. No. 5.7.I p ,,
Appenatx u Amend. No.
3 Dee De c embe r 23,1968 Ands Sect.(s) New 4 l l
I
metrica11y across top and vertically down
~
sides with a smaller box section extending from only one portion of the cylinder.
Size: 81 inches high by 66.1/2 inches across the main box section. Outer cylindrical diameter is 51.1/4 inches. Inner diameter is 46 inches.
Construction: Carbon steel throughout. Double walled con.
struction. The walls are 5/8 inches thick with a 4.1/2 inch air gap between outer cask wall and inner jacket wall and 1.1/2 inch air gap between inner and outer jacket walls.
Attachment s: Eleven 1. inch hex head bolts connecting the jacket to the cask base at the bottom edge of the jacket.
Lifting and Tiedown devices: Two 10 by 7 by 6 inch steel blocks located on the sides of the main box section. A >
4 inch diameter hole is cut through each block to accept cables or clevis.
Penetrations: Air passage from spaces between cask base and protective jacket base plate through a 4 inch diameter hole in the top of the inner protective jacket and out through two 1 inch by 6 inch horizontal slots on top of box section of outer protective jacket.
(e) Jacket Base Plate
~
Shape: Square plate welded to bottom of cask.
Size: Plate is 59 inches by 48 inches by 5/8 inch thick. .
License No. SNM 960 Docket No. _ 70 754 Sect. No. 571 Page Appendix D Am end. No. 3 0,,, De c embe r 23,1968 A.,,g , $,,,,(,) New 5
s Construction: 5/8 inch carbon steel plate welded to bottom of cask.
5.7.2 Package Description Contents (a) General By product material, source material and special nuclear material, with or without cask extension.
(b) _Fo r m Clad, encapsulated or contained in a specifi.
cation 2.R container or equivalent, but not limited to special form.
(c) Fissile Content i. Fiselle fuel plates, rods, rod sections, pellets and specimens subject to the following limitations:
(1) Fuel form: Uranium. Dioxide (2) Maximum uranium enrichment in U-235 prior to irradiation: 6 percent.
(3) Special nuclear material per cask load not to exceed 1. 2 kilograms of U 235 on a pre-irradiation basis.
(4) Ref. SNM.960, amendment 71.20.
ii. Fissile fuel plates, rods, rod sections, pellets and specimens subject to the following:
(1) Fissile rnaterial content not to exceed 2.0 kilograms.
(2) Fissile contents are confined to a cylin.
f.
drical geometry with an inner diameter
/ not exceeding 5.0 inches (nominal).
iii. Fissile fuel plates, rods, rod sections, pellets and specimens subject to the License &. SNM.960 . Docket No.70-754 Sect. No. 5.7.2 p ,,
Appendix U j hand. %. 3 Dete De c embe r 23.1068 hends Sect.(s) New 6 l
._' i following:
, (1) Quantities-of fissile material in excess of 1.2 kilograms.
(2) Quantities of fissile material in excess of 2. 0 kilograms.
(3) Shipments without the 5. 0 inch liner.
(4) Provided a neutron multiplicating study indicates that each container load is sub.
c rit ic al.
A method used in indicating the reactivity of the load is described in Exhibit A to this application. I (d) Radioactivity That quantity of any. radioactive material which does not spontaneously generate more than 6500 thermal watts by radioactive decay and which meets the requirements of 49CFR173. 393 and 10CFR71. 35.
(e) He at Total maximum internally generated heat load not to exceed 6500 watts. An analytical determination, described in Exhibit B to this application, of the container temperature j profile and heat load resulted in the following: l Cask Surface 300* F j Inner Shield 101' F Outer Shield 87* F Ambient 80' F l Heat Load 6566 watts
^l General Electric will analyze by test or other assessment each container heat loading prior to shipment to verify that the requirements of 10CFR71. 35 will be satisfied. Reference is e- -
License No. SNM - 960 Docket No. 70 754 Sect. No. 5.7.2 pop Appendix D Amend. No. 3 p., De c e mb e r 2 4 toA8 Amends Sect.(s) New 7
made to GE Model 100 Application, SS ction l'... 5.1, Exhibit B, for a method of internal heat load analysis and heat dissipation.
- 5. 7. 3 Package Eva?uaticn (a) General There are no components of the packaging or its contents which are subject to chemi.
cal or galvanic reaction; no coolant is used during transport. The protective jacket is bolted closed during transport. A lock wiri and seal of a type that must be broken if the package is opened is affixed to the cask closure device. If that portion of the protective jacket which is used in the tie.
down system or that portion which consti-tutes the principal lifting device failed in such a manner to allow the protective jacke to separate from the tiedown and/or lifting devices, the basic protective features of the protective jacket and the enclosed cask would be retained. The package (contents, ,
cask and protective jacket) regarded as a simple beam supported at its ends along its major axis, is capable of withstanding a static load, normal to and distributed along ,
its entire length equal to five times its fully loaded weight, without gene rating stress in any material of the packaging in excess of its yield strength. The packaging is ade-quate to retain all contents when subjected to an external pressure of 25 pounds per square inch gauge. Reference is made to the GE Mo' del 100 Application,Section 5.1 Exhibit C, for a method of determining static loads.
License No. SNM - 960 Docket No. 70 754 Sect. No. 5. 7. 3. Poge Appendix D bd. No. 3 DeeDecember 23.1968 4,,,g, g,,,,(,) New 8.
r
! s l
. The calculative methods employed in the l design of the protective jacket are based )
on strain rate studies and calculations and on a literature search
- of the effects on 1
materials under impact conditions. The intent was to design a protective jacket that ,
1 would not only satisfy the requirements of the U.S. Atomic Energy Commission and the Department of Transportation pre- 1 scribing the procedures and standards of packaging and shipping and the requirements j governing such packaging and shipping but would protect the shielded cask from signifi-cant deformation in the event of an accident. ]
In the event that the package was involved in an accident, a new protective jacket could f be readily supplied and the shipment con-tinued with minimal time delay.
1 The effectiveness of the strain rate calcu.
lations and engineering intuitiveness in the design and construction of protective jackets was demonstrated with the General Electric Shielded Container Model 100 (Ref: Section 5.1.3 of Appendix D to SNM-960). The pro.
tective jachot design for the General Electric Shielded C atainer Model 700 will be scaled from the design of the Model 100 in accordance with the cask weight and dimen-sions, maintaining static load safety factors ,!
greater than or equal to unity, and in accor- I dance with the intent to protect the shielded !
cask from any deformation in the event of
j l
i License No. SNM 060 Docket No. 70 754 Sect. No. 5.7.3 p 9, Appendix D Amend. No. 3 p ,, December 23,1968 Amends Sect.(s) New 9 1
O en accident.
(b) Normal Transport Conditions l
Thermal: Packaging components, i. e. , steel shells !
and lead, uranium and/or tungsten shield-ing, are unaffected by temperature extremes l I
of 40' F and 130' F. Package contents, at least singly-encapsulated 'or contained in specification 2R containers, but not limited to special form, will not be affected by 1
these temperature extremes.
i
< Pressure: The package will withstand an external pressure of 0. 5 times standard atmospheric pressure.
Vibration: Inspection of the Model 700 casks used since 1958 reveals no evidence of damage of signi-ficance to transport safety.
Water Spray and Free Drop: Since the container is constructed of metal, there is no damage to containment resulting from dropping the container through the standard drop heights after being subjected to water spray.
Penetra tion: There is no effect on containment or over-all spacing from dropping a thirteen pound by 1-1/4 inch diameter bar from four feet onto the mo st vulnerable exposed surface of q the packaging. j l
Compression: The loaded container is capable of with. i l
l L.icense No. SNM 960 Docket %. 70 754 s,c,, g, 5.7.3 p. , )
Appendix D ]
Amend. %. 3 Dee December 23.1968 New to i Ands Sect.(s)
\
- I i
, standing 'a' compressive load' equal to five times its weight with no change in spacing.
Summary and
Conclusions:
The tests or assessments set forth above provide assurance that the product contents are contained in the Shielaed Container -
Model 700 during transport and there is no reduction in effectiveness of the package.
(c) Hypothetical Accident Conditions
, General: ' The effectiveness of the strain rate calcu-lations and engineering intuitiveness in the i design and construction of protective jack'ets was demonstrated with the GE Shielded Con.
tainer Model 100 (Ref. : Section 5.1. 3 of i I
Appendix D to SNM-960). Extrapolations <
of the Model 100 data were used in the design ;
and construction of the GE Model 700 pro-tective jacket. The increased weight and dimensions of the Model 700 container over the Model 100 container-necessitated a pro-tective jacket wall of 5/8 inch steel compared to a 1/4 inch wall for the Model 100.
Drop Test: The design and construction of the GE Model 700 protective jacket was based on an. ;
extrapolation of the proven data generated j during the design and construction of the GE Model 100 and on the results of cask drop License No. SNM- 960 Docket No.- 70 754 Sect. No. 5.7.3 p.g. I Appendix D Amend. No. 3 0,,, Decembe r 2 3. I 968 Amends Sect.(s) New- 11 f
a
. i
, experiments by -
C. B. Clifford IIII } and H.G. Clarke, Jr. N The laws of similitude were used in an analytical evaluation (3)(4) to determine the protective jacket wall thickness that woulc withstand the test conditions of 49CFR173. 398 (c) and 10CFR71. 36 with.
out breaching the in+egrity of the Model 7C cask. The evaluatiw, described in GE .
Model 1000 Applicanon, Section 5. 9, Exhibit A, indicated a protective jacket wall thickness of 5/8 it.ch. The intent of the design for the GE Model 700 is, durin accident conditionc, to s'1 stain damage to the packaging not greater than the damage i sustained by the GE Model 100 during its accident condition tests (Ref. : Section 5.1. 3 (c) of Appendix D to SNM-960). It is expected that damage not exceeding that suffered by the GE Model 100 will result if the GE Model 700 is subjected i to the 30 foot drop test.
I Puncture Test: The intent of the design for the GE Model 700 is to sustain less or equal damage to the packaging during accident conditions (1) C.B. Clifford, The Des ign, Fabrication and Testing of a Quarter Scale of the Demonstration Uranium Fuel Element Shipping Cask, KY-546(June 10,1968)
(2) C.B. Clifford, Demonstration Fuel Element Shipping Cask from Laminated Uranium Metal-Testing Program, Proceedings of the Second International Symposium on Packaging and Transportation of Radioactive Materials, Oct.14-18,1968, pp. 521-556. ,
(3) H.G. Clarke, Jr. , Some Studies of Structural Response of Casks to Impact, Proceedings of the Second International Symposium of Packaging and Trans.
portation of Radioactive Materials, Oct. 14-18,1968, pp. 3 73 398.
(4) J.K. Vennard, Elementary Fluid Mechanics, Wiley and Sons, New York,1962, pp. 256 259.
License No. SNM-960 Docket No. 70 754 Sect. No. 5. 7. 3 p .g.
Appendtx D hd. No. 3 Dete Decembe r 23,1968 Amends Sect.(s)
New 12 l
b
. than the deformation suffered by the GE Model 100. It is expected that deformation not greater than that sustained by the GE Model 100 will be received by the GE Model 700 in the event that the package is subjected to the puncture test.
Thermal Test: Since it is expected that the GE Model 700 cask will sustain negligible damage and only minor damage will occur to the pro tective jacket in the drop and puncture tests, it is reasonable to consider the resultant package, for purposes of thermal resistance, as essentially undamaged. Accordingly, the package was assessed using the General Electric Transient Heat Transfer Computer ;
Prograrn, Version D (THTD), which allows the analysis of the general transient problems "
involving conduction, convection and l radiation. The program allows the thermal properties of the materials to be entered as a function of ternperature and the boundary conditions to be entered as a function of time.
The significant assumptions, approximations, and boundary conditions used for the analysis are listed below:
- 1. Fire temperature 1472* F
- 2. Effective fire Emiss ivity 0.9
- 3. Fire shield surface Emis sivity 0.8 and constant with temperature
- 4. Emissivity of other Surfaces 0.8 .
and constant with temperature.
1 L,1 cense No, SNM 960 Docket No. ._ 70.754 Sect. No. 5.7.3 p. ,
Appendix D Amend. No. 3 Det, neeembe r 2 W8 N o m. 13 Amends Sect.(s)
4 "
- 5. Th' 3 is intimate contact betw2en the lead shielding and the stainless steel shell of the cask.
- 6. There is negligible heat transfer by -
conduction through the pipes used as spacers between the cask and the first shield and between the two shields of the protective jacket.
- 7. There is negligible heat transfer by convection between the two shields of the protective jacket and between the cask and first shield of the protective jacket.
- 8. There is an internal heat load of 6500 watts with assessed tempera-tures as outlined in Section 5. 7.2(e) of this application.
The computer program calculations were run for a 30 minute fire. The calcu-lations indicate a maximum temperature rise to less than 473* F for the lead after 30 minutes and no lead melting could be expected. A coast up analysis (Ref. the Model 100 Application) indicated that a l temperature of 464* F could be expected I at the inne rmost lead node after 34 minute Exhibit A to Section 5.1 (Shielded Con-tainer Model 100 dated September.18,196E as amended October 25,1968) of Appendix D to SNM.960 further describes !
the computer code THTD.
Water immersion: Since optimum moderation of product material is assumed in evaluations of ' l l
criticality .
I License No. SNM-960 Docket No. 70 754 het. No. 5.7.3 p, appenauc u An.end. No. 3 p.,, pecember 2 3. I 968 kg $,,,,(,) New 14 i
6 ,
safety under accident conditions, the water
_ immersion test was not necessary.
Summary and
Conclusions:
The accident tests or assessments described l j
above demonstrated that the package is ade. '
quate to retain the product contents and that there is no change in spacing. The refore, it is concluded that the General Electric Shielded -
Container Model 700 is adequate as packaging for the contents specified in 5. 7.2 of this I
section. 1 i
i i
l l
1 q
I l
1
- I 1
License No. SNM 460 Docket No. 70 754 Sect. No. 5.7.3 p. , l Appendix D 3 Date December 23,1968 Amend. No. gg, g,,,(,) New 15 ,
p - ..
i s a Dis tribution: '
Supple.4 '
IFB reading file Docket No.70-754 g } 4 jggg DML reading file -
State Health ND'.,ulos, DML Ceneral Electric Company Drig: BPBrown Vallecitos Atomic Laboratory P. O. Sox 846 Pisassaton, Califersia 94566 Attention: Mr. Walter E. King Administrater - Licenslag
- Vallecites Nuclear Center l I
Gentlement i i
Enclosed is Amendment No. 71-25 to AEC License 92 issued SIM-960, '
in response to your application of September 18, 1968, as amended This license amendment authorians delivery of byproduct material and i
special nuclear material to a carrier for transport in the CE Shielded Container - Model 100.
sincerely.
Origind pand E[
R. B. Chit *WA '
'R. 5. Chitwood, Chief i
Irradiated Fuels Branch Divisten of Materials Licensing Enclosure
{ As s tated ,
I g cct Mr. William A. Brobst DOT, w/cy eac1. and cy Staff Safety Evaluation I
omcr > ._..DML: I , , , _ , ,
- An>, ,
l '
BPB
$URNAMES ....................,..
kBChitwood 1/9/69 . .1 . . . . /69 carr > . . . . . . .. . . . . . -
Form AEC-318 (Rev. 9-5h) AECM 0240 _.
m a aevanamsm, nannes enses asse e-se.-.o _ .
s e _
9 i
a .
. - a . . . . .
i . -.
UNITED STATES
[s*\* f#+,k 'l ATOMIC ENERGY COMMISSION
.. ;_* g
- WASHINGTON. o.c. 20545
% pY min e Docket No.70-754 License No. SNM-960 Amendment No. 71-25 LICENSE AMENDMENT l
Pursuant to the Atomic Energy Act of 1954, as amended, and Title 10, Code of Federal Regulations, Chapter 1, Part 30 " Rules of Applicability to ;
Licensing of Byproduct Material", Part 70 "Special Nuclear Material", and '
Part 71 " Packaging of Radioactive Material for Transport", and in reliance on statements and representations heretofore made by the licensee, License No. SNM-960 is hereby amended authorizing the licensee to receive, possess, -
use and trans fer the byproduct and special nuclear material described below for the purpose designated below, in accordance with the regulations in said Parts and the conditions set forth below. l
- 1. Licensee:
Ceneral Electric Company Nuclear Energy Division 175 Curtner Avenue San Jose, California 95125
- 2. Pu pose: /
Delivery of the material specified in Item 3 of this amendment in the packaging specified in Item 5 of this amendment to a carrier for transport as a Fissile Class II shipment.
- 3. Syproduct Material and Irradiated Special Nuclear Material in solid metal or oxide form.
- 4. Maximum Content per Package:
The material specified in Item 3 of this license shall be limited per package such that:
A. The mass of U-235 plus 1.66 times the mass of U-233 plus 1.66 times the mass of Pu is less than 500 grams, and B. The decay heat is less than 400 thermal watts.
l l
. . ~ .
g ..
h M Ceneral Electric Company Page 2 Docket No.70-754
, License No. ShM-960 Amendment No. 71-25
- 5. Minimum Transport Index to be shown on package label for Fissile Class II:
5.6
- 6. Packaging: -
A. Cask:
General Electric Company (GE) Shielded Container - Model 100,. c as described in CE Drawing Nos,. 612D139, Rev. 3 and 693C292, Rev. 3. E The cask may be equipped with one of the cavity lines described in CE Drawings Nos. 985C575, Rev. 1, 693C203, Rev. 2 or 135C5371, Rev. 1.
B. Cask primary coolant: Air C. Protective Jacket: The loaded cask shall be enclosed in the Protective Jacket described in CE Drawing No. 706GE578, Rev. 1, during transport. ," )
I 7.
The transportation of AEC licensed material shall be subject to all applicable regulations of the Department of Transportation and other agencies of the United States having jurisdiction.
When Department of Transportation regulations or regulations of other agencies of the United States having jurisdiction are not applicable to ,
the transportation of AEC licensed material, the licensee shall comply with all applicable requirements of the rules and regulations, as amended, appropriate to the mode of transport, of the Department of Transportation (49 CFR Parts 171-178), Federal Aviation Agency (14 CFR Part 103), and Coast Guard (48 CFR Part 146) insofar as such rules and regulations relate )
to the packaging of the licensed material and to the marking and labeling of the package and placarding of the transporting vehicle and accident reporting to the same extent as if the transportation were in interstate or foreign commerce. Any requests for modifications or exceptions to those requirements, and any notifications referred to in those require-ments shall be filed with, or made to, the Atomic Energy Cocanission. 1 FOR THE ATCMIC ENERGY COMMISSION st.< e fk
. wood, Chief
(' R.
Irr di Fuels Branch ision of Materials Licensing Date of issuance: M 141969 yl0 /()i;. i ,.
L g M'M ,
mI; & mi _ L .- J
UNITED STATES ATOMIC ENERGY CCf3 FISSION SAFETY EVALUATICN BY THE IFRADIATED FUELS BFANCH DIVISICN CF MATERIALS LICHISING
,IN THE MATTER OF GE:EFAL ELEC13IC COMPANY DOCKET No.70-75h GE SHIELLED CONTARIER - MODEL 100 FOR FISSILE CLASS II SHIFFENTS
, j
}
SUMMARY
By an applicatien dated September 18, 1968, General Flectric Cc=pany (GE),
Vallecitos Atomic Laboratory, pleasanten, California, submitted a safety analysis and applied for a license amendment to authorize delivery cf irradiated carrier fissile material for transport in theClass as a Fissile GE Shielded Centainer - Model 100 to a I shipment.
October 25, November 13, and Decemb By submittals dated The submittal dated December 19,19 9,1968, GE amended the application.
to Fissile Class II. requested a change from Fissile Class I pg The fissile content per package will be limiteil such that the mass of U-235 plus 1,66grams.
than 500 times the mass of U-233 plus 1.66 tf.mes the mass of Pu is less Our evaluation of the prepcsed package indicates that the egtipment fulfills the requirements of 10 CFR 71.
CONTENTS OF SUBMIITALS GE submitted the following items for our review:
Application as Exhibit Adated September to the application. 18, 1968, and a related safety analysis Letter dated October 9 and including the following Exhibits:, revising part of the safety analysis B - Data on internal heat, load evaluations.
C - Method of assessing static loads.
D - The following assembly and detailed drawings of the y ickage:
g A
. Drawing No. Title 612D139 Rev. 3 100 Series Shipping Cask 153F9C2 Eev. 1 Base 117B1011 Fev. 1 Drain Tube 115A8101 Rev. 1 Drain Tube End Fitting 693C292 Rev. 3 Lid l 693C293 Fev. 2 Liner (Lead ) !
985C575 Rev. 1 Cesk Liner for GE."NC 100 Series Cask (Depleted U) 135C5371 Rev. 1 Cask Liner for GE-VNC 100 Series Cask I 7C6E578 Rev. 1 Protective Jacket Assembly l 706E59k Rev. 1 Jacket 135C5529 Rev. O Pallet Letter dated November 13, 1968, submitting =sterial to substantiate the heat transfer analysis.
Letter dated December 19, 1968, requesting change from Fissile Class I to Fissile Class II, revising the criticality section of the safety analysis, and including a statement on a lock vire for the cask closure.
DESCRIPTION OF PACKAGE The package consists of a lead filled steel veldment cask enclosed in a '
steel protective jacket with an integrally velded stet 1 pallet. The maximum g weight of the package (with a uranium-filled liner in the cask cavity) is 3520 pounds. Figure 1 shows details of the cask, liners and protective jacket.
A. Cask
- 1. Body Shape: An upright circular cylinder with attached base.
Size: 20-1/h inches diameter by 27-1/8 inches high.
Construction: A lead filled steel veldment with stainless steel piping. Adoublevalledprotectivejacketof1/4 inch steel plate surrounds the cask during transport.
Outer Shell: 1/hinchthicksteelplate,20-1/hinchesdiameter by 26-1/8 inches high with a 3/8 inch bottom plate and 1-1/8 inch top flange.
Cavity: 1/8 inch thick steinless steel vall and bottom plate, 7-5/8 inch diameter by 10 inches deep.
6
~ * .
l d
Shielding Thickness : 5-7/8 inches of lead.
Penetration : One 1/2 inch diameter stainless steel tube, gravity l drain frem center of cavity bcttom to side of outer shell near bottom. Clcsed with fusable core hex headed, lead filled, brass plug. The drain may be permanently closed en come models.
Lifting Devices: T90 diametrically opposed ears welded to sides of cask; covered by protective jacket during transport, i
Pritary Coolant: Air, j
- 2. Cask Lid I
Shape: Truncated cone plug attached to flat plate. Plate is 20-1/h inches diameter by 3/h inch thick, Truncated ccnn ,
is 7-7/16 inches high, major diameter is lk-15/16 inches, !
minor diameter is 10-1/2 inches.
Ccnstruction: Lead filled stainless steel clad plug velded to ,
circular steel plate.
l Closure: Six 1 inch 8UNC-2A stainless steel bolts equally spaced in circular pattern.
Closure Seal: A minimum 1/h ir.ch thick flat necprene rubber casket between body and lid, j Penetrations: None Shield Expansion Void: Iiene Lifting Device: Single steel loop, 1-1/h inches diameter steel rod located in center of lid top. Covered by protective jacket during transport.
3 Cask Liners: (There are three liners, lead, uranium and tungsten).
Shape: The three liners are right circular cylinders with a central bore section.
Size: All three liners are 10 inches high with a T-1/2 inch outside diameter. The lead liner has a 3-3/8 inch diameter central bore and the lead and tungsten liners have a 1-3/h inch diameter central bore.
4
~
. M h Censt ructien : Lead liner - a lead-filled stainless steel veldment. 1 Uranium liner - a uranium (natural cr depleted) filled stainless steel weldment. The tungsten line is of solid tungsten metal.
I The uranium liner has the legend " CAUTION - RADICACTIVE SHIELLING - l URANIUM" impressed on its surface and meets the constructicn require- !
ments of ICC Spec 55. Accordingly, pursuant to ELO 13(c)(6) cf .I Part LO, the uranium is exempt frem licensing.
B. Protective Jacket j
I
- 1. Body
)
i Shape: Easically a right circular cylinder with cpen bottom and with a protruding tcx section diametrically across the top and vertically down the sides.
l Size: 36-3/8 inches high by 32 inches vide across the bcx section.
Outer cylindrical diameter is 25-1/h inches, inner diameter is 22-3/4 inches.
ll Construction: Carben steel throughout. Double-valled construction.
All valls are 1/k inch thick with 1 inch air gap between inner and outer valls, j
i
Attachment:
Two 1-1/2 inch 6 UNC - 2A by 10 inch long bolte diametrically cpposed cennecting the jacket body to its base at the bottom edge of the body.
Lifting Devices: Two rectangular 5/8 inch thick steel locps l located en top cf the bcx section at the cerners. Steel i is 7 inches long by 3 inches hi Sh by 3 inches vide.
Tie Ecun revices: TVo diametrically cpposed 1-1/2 inches thich steel ears velded to sides of bcx secticn, each with 1-1/2 inch hole to accept elevis or cable.
Penetrations: Slots alcng periphery of the protective jacket at the bottom and in box section under lifting loeps. Allows natural air circulation for cooling.
- 2. Protective Jacket Base Shape: Hollev cylindrical veldment with plate.
Size: Bottom plate diameter is 33 inches, cylinder is 21 inches diameter by 3 inches high, !
Construction: 1/4 inch carbon steel plate with internal and external steel angles for energy absorption.
i 1
/-
Attachment:
Two diametrically cpposed tie blocks to accept Jacket attachment bolts.
3 Fallet i
Shape: Square base supported by four I-Beams.
Size: hh inches square. I-Eeams are 3 inches by 5-6/10 inches by kh inches long.
1 Construction: Carben steel thrcughout. I-Eeams are velded to base plate. The pallet is permanently attached to the protective jacket base by a continuous 1/h inch fillet veld around the base. [
m C. Package Contents
{
l Fissile Content: Limited such that the mass of U-235 plus 1.66 times .
I the mass of U-233 plus 1.66 times the = ass of Pu is less than 500 grams. j Form: Solid, either metal or oxide.
Maximum Ikcay Heat: kOO thermal vatts.
Containment: Clad, encapsulated or contained in a specification 2R container.
SAFETY EVALUATION I
I A cask with the deceribed design features and enclosed in the protective !
jacket meets the package standards of 10 CFR 71. The results of our evaluation are discussed below.
General Standards 71.31(a) There shall be no significant chemical, galvanic or other reactions between cask materials or with fuel.
There are no co=ponents of the packaging or its contento which are subject to chemical or galvanic reaction. Only air is used as a coolant during transport.
- 71. 31(b ) The cask shall be equipped with a positive closure which vill l 1 prevent inadvertent opening. '
Both the protective jacket and the cask are bolted closed during transport.
A lock vive and seal that must be broken if the cask is opened is affixed to the cask closure.
\
l x .
1 Pressure
)
The protective jacket is not a sealed vessel and is therefore unaffected by pressure changes. We agree with GE that the cask is capable cf with-standing the pressure differential without deformation.
Vibration ;
There are no shock sensitive areas in the cask or centents. I i
Water Spray i
This would have no effect on this metal package.
Free Drop A four foot free drop would not affect the effectiveness of this package.
The 1,/h inch thick steel walla cf the protective jacket provide this ,i protection.
Corner Drop Not applicable to this metal structure.
Fenetration The specified 13-pound cylinders would not penetrate or effect the 1/4 inch thick steel valls of the protective jacket. ,
Compression We agree with GE's analysis thut the required compressive load for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> would not constitute any stress hazard for the cash and protective jacket and that there would be no change in spacing.
71.35(c) A package . .. shall be designed and constructed and its contents so limited that under the normal conditions of transport specified in Appendix A, the containment vessel would not be vented directly to the atmosphere.
GE considers the cask as the containment vessel for this package (i.e., no credit is taken for claddin6 encepsulation or the 2R container). We agree with GE that the cask will remaia sealed during the normal conditions of transport.
Standards for Hypothetical Accident Conditions 71.36(a) A package . .. shall be designed and constructed and its contents so limited that if subjected to the hypothetical accident con- '
ditions specified in Appendix B it will meet the following con-ditiens: '
l
[ (1) The reduction of shielding would not be sufficient to increase the external radiation dose rate to =cre than
{
l 1,000 millireentgens per hour er equivalent at 3 feet i from the external surface of the package; and j (2) No radicactive material vould be released from the package except for specified limits.
The specified sequential accident conditions of a 30-foot free drcp cnto an unyielding surface, a LO-inch drop onto the flat end of a 6-inch diameter ateel bar, exposure within 1k75 F hest source for thirty minutes, and sub-mersion in water for twenty-four hours have been evaluated and we have determined that the shielding would not be reduced and that no radicactive raterial vould be released from the package. The bases on which we have concluded that the package vill safely meet the hypcthetical accident con-ditions are as follows:
Free Drop and Penetration
,k L
To determine the effectiveness of protection afforded by the protective -
Jacket, GE made actual deep and puncture tests. Both tests were conducted on a concrete drop pad 20 feet by 15 feet by 6 feet thick. A carben steel plate,10 feet by 8 feet by 3 inches thick was located on the pad. To determine maintenance of container integrity, the cask cavity was filled with water. A mobile crane capable of raising the cask and protective jacket assembly to a 30 foot height was used.
We have reviewed the results ef these tests and agree with GE that the design of the protective jacket provides adequate strength to protect the cask l during the hypothetical accidents. Since GE made the drop test without the l pellet, we have made calculations and determined that the presence cf the pellet would not negate the effectiveness of the protective jacket.
Thermal Since the packege sustained only minor damage in the drop and puncture tests, GE considered the resultant package, for purpose of the:=al resistance, as essentially undamaged. We agree that this is acceptable. The package was -
assessed for the specified thermal exposure using the General Electric Transient Computer Program, Version D (THTD), which allows the analysis of the general transient problems involving conduction, convection and radiation. {
The program allows the thermal properties of the materials to be entered as a function of temperature and the boundary conditions to be entered as a function of time. We have reviewed the assumptions, approximations and l boundary conditions used for the analysis and agree that they are reascnable. '
The results of the analgsis showed that the maximum temperature the cavity I will reach is about 556 F, some h0 minutes after the thermal exposure. The maximum temperature of the cask surface is about h55 F, again occurring about h0 minutes after the fire. All temperatures are belev these that would cause lead melting or' cask venting. The two steel enclosures of the pro-tective jacket serve as thermal radiation shields. l I
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Fissile Centent Sofe Nutter of Transport
_ Containers _
Index 300 grams Pu (vet) 9 300 grams U-233 (vet) 5.6 500 grams U-235 30 2h 1,7 2.1 GE stated that in all cases, regardless of fi leading vill be assumed to be exclusively Pu ssile mixtures involved, the Class be safe.II shipment with a minimum transport ind We agree that a Fissile ex of 5.6 per package vill t I
Procedural Centrels Procedures and centrols have been established assure that shipments comply with license and a speci land used many times will permit ecnditions.
GE stated that each new use to ascertain that there are no crackscack vill be incpected and radiographed pI -
other defects which would reduce the effectivepe, pinholes, uncentrolled voidl ss of the package.
After AEC and DOT apprevals, each package vill t as required by the DOT special permiten . quirementssteel cf 10 CFRplate 71 andin accordanc
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CCNCLUSION We cenclude that the GE shielded containe r Model 100 leeded as preposed, meets all pertinent packaging .
criteria of 10 CFR 71 co as described with 10 C E 71. in the suttittal drawings and prints n ex, and the package I will assure ccepliance ,
1 5
Signed: ) . h f 8 9,,,
- b. P. brcun, Chemical i.ngineer Approved:
E.' B. Chitvood, Chief
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Irradiated Fuels Branch Date:
$]4 g Division of Materials Licensing
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MAR 11 1969 r
Docket No.70-754 General Electric Company Vallecitos Nuclear Center F. O. Box 846 Fleasanton, California 94566 Attention: Mr. Walter H. King Administrator - Licensing Vallecitos Nuclear Center Gentlemen:
D2 closed is Amendment No. 73-27 to AEC License No. SNM-960, issued in response to your application of February 4,1969, as amended. This license amendment authorizes delivery of byproduct material to a carrier for transport in the GE Model 700 package.
Sincerely, Originalsigned by
- 11. B. Chitwood R. B. Chitwood, Chief irradiated Ebels Branch Division of Materials Licensing
Enclosure:
As stated I
cc: Mr. William A. Brobse, DOT, I w/cy encl. '
+
i OmCE > -
SURNAME >
em ,. 31"As.....4h'0/69 Form AEC-816 (Rev. E-63)
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.. s. novossant eservise .mes 3 erian-e i
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r Dochet No.70-754 License I;o. S!04-960 Amendment No. 71-27 LlcLUEE hGLGDuLE1 Pursuant to the Atomic Energy Act of 1954, as amended, and Title 10, Chapter 1 Code of Federal Regulations, Part 30
" Rules of Applicability to Licensing of Byproduct Material",
and Part 71 " Packaging of Radioactive Material for Transport",
and in reliance on statements and representations heretofore made by the licensee, License No. SNd-960 is hereby amended authorizing the licensee to receive, possess, use and transfer the byproduct material described below for the purpose designated below, in accordance with the regulations in said Parts and the conditions set forth below.
- 1. Licensee:
i General Electric Company Nuclear Energy Division 175 Curtner Avenue San Jose, California 95125 2.
Purpose:
Delivery of the byproduct material specified in Item 3 of this amendment in the packaging specified in Item 6 of this amendment to a carrier for transport.
- 3. Byproduct material:
Up to 6500 thermal watts per package of byproduct material in solid metal or metal oxide form.
- 4. The byproduct material specified in Item 3 of this amendment shall be clad, encapsulated or contained in a metal encastment in accordance with the statements and representations contained i.n 196 ghe General Electric Company's application dated February 4, f
i General Electric Company License No. SNA-960 l Amendaent No. 71-27 e
- 5. The purpose specified in item 3 of this amendaent is authorized at any locacion in the United States, as defined in Sectioa 30.4(u),
Title 10, Code of Federal Regulations, Fart 30, where the U. S.
Atomic Ene'!gy Comnission maintains furisdiction for regulating the use of byproduct material.
- 6. Packaging:
A. Shipping Cask:
General Electric Company (GE) Cask Model No. 700 with or without the extension as described in GE Drawings No.
197E538 Rev. 5, No. 23fE325 Rev.1, and No. 117B1170 Rev. O.
B. Cast Frimsry Coolant: Air.
C. Protec tive Jacket:
The loaded cask shall be enclosed in the protective jacket described in GE Drawings No. 277E400 Rev. O, and No.
277E401 Rev. O, during transport.
- 7. A. The transportation of AEC-licensed material shall be subject to all applicable regulations of the Department of Transportation and other agencies of the United States Laving jurisdiction.
When Department of Transportation regulations in Title 49 Chapter 1, Code of Federal Regulations, Parts 173 - 179 ar,e !
not applicable to shipments by land of AEC-licensed material I by reason of the f act that the transportation does not occur in inters tate or foreign commerce. (1) the transportation shall be in accordance with the requirements relating to packaging of radioactive material, marking and labeling of the package, placarding of the transportation vehicle, and accident reporting set forth in the regulations of the Department of Transportation in il 173.389 - 173.399 173.402 173.427 173.414, 177.842 177.843, 177.861, 49 CFR Part 173,Part 49 CFR " Shippers," and II 177.823, 177, # Regulations Applying to Shipm,ents Made, by Way of Common Contract, or Private Carriers By Public Highways," and (2) any re, quests for modifications or ese[ptions torequirements those requirements and any notifications referred to n those shall be filed with, or made to, the
'tanic Energy Commission. -
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General Electric Company License No. KNM-960 Amendment No. 71-27 i
- 7. B. When Chapterthe1,Department of Transportation Code of Federal Regulations,regulations in not Part 103 are Titleappl 14[cabic to shipments by air of AEC= licensed material by reason of the fact that transportation does not occur in civil aircraft, (1) the transportation shall be in accordance with the requirements relatin:,
to the packaging of radioactive material, markins and labeling of l the package notification of the pilot in command, and accident l reporting se, t forth in the regulations of the Department of l Transportation in 14 CFa tart 103, " Transportation of Dangerous J Articles and Magnetized Materials," and (2) any requests for l modifications or exceptions to those requirements, any requests '
for special approvals referred to in those requirements, and any notifications referred to in those requirements shall be filed with, or made to, the Atomic Energy Commission. l l
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FOR THE ATOKIC ENERGY COMMISSIOh !
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- 89 '
R. B. Chitwood, Chief j leradiated Fuels Branch Division of Materials Licensing Date of Amendment: MAR 11 1969 l i
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UNITED STATES ATa:IC E' ERGY C01NISSION SAFETY ANALYSIS r
DY T.in IRP.ADI ATID FUELS BRWC:!
DIVISION OF HiTERIALS LICENSING I'! T lE M\TTER OF GENT: PAL ELECTRIC C0"DANY MODEL 700 PACKAGE FOR BYPRODUCT :MTERI AL
~
SUM'!ARY By an application dated February 4,1969, General Electric Company (GE), )
Vallecitos Nacicar Centor, Pleasanton, California, applied for a license amendment to authori:e packaging of byproduct material in the CE Model 700 Cask with only air as a primary coolant. In the application, GE incor-porated by reference a safety analysis submitted with an application dated December 23, 196S for packaging of both special nacicar material k and byproduct material in the Model 700 Cask with either air or water as a prinary coolant. Sinco additional analyses are required before we 1 can approve packaging of fissile material and the use of water coolant in the Model 700 Cask, GE requested this approval pending development of '
additional analysis. l We have previously evaluated and approved the Model 700 Cask with a wood shield against the standards of old 10 CFR 72 published in the Federal Ronister on September 23, 1961. To meet the standards of 10 Our CFR 71, Gd has designed a steel jacket to protect the cask.
evaluation of the cask in the steci jacket indicates that the design l of equipment is adequate to protect health and mininize danger to life and property. We conclude that the Model 700 Cask, loaded as proposed and with air as a coolant, meets the 10 CFR 71 packaging standards.
SUB"ITTALS Application dated February 4,1969 requesting package approval and a license for 6500 watts of clad or contained byproduct material as a dry load and which incorporated.by reference a December 23,196S license
- application for the Model 700 package.
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e The referenced application dated Deccaber 23, 1963 contained the following: l
- 1. Package description, i 1
- 2. Package safety evaluation, and
- 3. The following drawings:
Title
_GE _Orawi n- ~.To.
Fuel Shipping Cask l 197E53S Rev. 5 Fuel Shipping Cask Extension !
237E325 Rev. 1 700 Series Cask and Extension 117B1170 Rev. 0 Protective jacket 277E401 Rev. O Ease 135C5573 Rev. 0 Protective jacket assembly l 277E400 Rev. 0 Fuel Shipping Cask (Parts List) ]
197E53S (3 sheets)
Letter dated February 20, 1969 with an analysis of the effects of tie-down l forces on the package.
Letter dated February 27, 1969 with a reanalysis of the effects of tic-down i forces on the package, l I
DESCRIPTION OF PACKAGE A. General Shave: An upright circular cylinder shicided cask and an upright circular cylinder protective jacket with attached square base plate.
. ,'. k.
With an Size: The cask is 36-13/16 inches diancter by 64-3/4 inches high.
extension, the cask is 78-7/S inches high. The protective jacket is 81 inches high by 66-1/2 inches across the box section.
Construction: The cask is a lead-filled carbon and stainicss steel weldment. The protective jacket is a double walled structure of 5/S-inch carbon steel plate and surrounds the cask during transport.
The cask with extension weighs * ;
Weicht: The cask weighs 24,000 pounds.
28,900 pounds. The protcetive jacket and base weigh 5,000 pounds. .
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- 3. SodV Dater Shell:
Orininci Cask: 3/S-inch thick ASTM A-6 steel plate 36-13/16 incacs in diameter by 56 inches high, with a dished head bottom 7-1/2 inches deep.
Extension: 1/2-inch thick AST:-! A-504 stainicss steel plate, 50-13.'16 inches in diameter and 14-1/S inches high in the exposed portion, 21 inches in diancter, and 10-1/4 inches high in the portion which cxtends into the cask.
Cavity:
Original Cask: 15 inches I.D. by 40-1/4 inches high,1/4-inch ASTM A-240 stainicss steel shcil and bottom plate.
1 Extension: 15 inches I.D. by 14 inches high, 1/4-inch AST11 A-5u4 stainicss steel. With the extension, the cavity is 54-1/4 inches deep.
Closure: Top of extension recessed for shicided plug; cight l 3/4-inch diameter stainless stcc1 bolts are provided for extension hold-down and for lid closure.
Penetrations: A 1/2-inch stainicss steci siphon drain which terminates in a pipe plug on the upper surface of the cask l extension and a 1/2-inch vent line from the side of the cavity which terminates in a valvo. The siphon valve is protected by steel channel attached to the adjacent lifting structure and the vent valve is protected by a pipe sleeve, both are covered by the protective jacket during transport.
Shiciding thickness: Approximately 10-1/4 inches of lead on sides, 9-2u/32 2nenes of Icad beneath cavity and 9-7/8 inches of 1 pad above cavity.
Liftine Devices: Two dianetrically opposed, vertical 12-inch structural tees, 3-1/2 feet long, welded to the cask shc11 and to the base with reinforced lifting slots located in the web. Covered by protective jacket during transport. ,
a r
Pressure Ratinc: Design - 20 psic; tested - 50 psig; normally '
j unpressurl:cd; maxinum operating - 15 psig, with and without the extension.
Prinary Coolant: Water or air. Ilowever, for this proposed package only air will be uscJ.
Closure Sen1: A minimun 1/4-inch thick flat gasket between cask and extension is used.
Lid:
Shanc: Flat plate and cylindrical plug.
Si:c: Top plate is 30-1/2 inches dine.cter by 1/2-inch thick, bottom plate is 21 inches diameter by 1/4-inch thick. The plug is 10-1/4 inches high.
Construction: Steel weldment, lead-filled.
Closure: Eight 3/4-inch UNC-2A by 2-1/4 inches long stainless steel bolts equally spaced 45' apart on a 2S-inch bolt circle.
The bolts are 14-3/4 inches long when the extension is used.
Closure Scal: A minimum 1/4-inch thick flat gasket between body I and 11d.
Penetrations: A 1/2-inch schedule-40 stainicss steel pressure-vent line f rom the side of the cask cavity, through the lid, terminating in a relief valve / filter assembly, guarded by a surrounding pipe sleeve and covered by the prot'cctive jacket during transport.
Pressure Relief Devicc: 0 to 15 psi rating, set at 15 psi.
Filter: A high efficiency fiberglass filter on the discharge of the relief valve. Covered by protective jacket during transport.
Lifting Device: 3/4-inch thick -by S-inch high by 13-inch long vertic; plate welded to the lid with a 1-1/2 inch diameter hole centered 2-1/2 inches from the top edge to accommodate lifting hook or cable. Covered by protective jacket during transport, ,
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r C. Protectivo .Tacket Shano: Sasically a right circular cylinder with open bottom and with a protrudin; box section diametrically across top and vertically down sides with a smaller box section extending fron only one portion of the cylinder.
Si:c: S1 inches high by 66-1/2 inches across the main box section, j Outer cylindrical diameter is 51-1/4 inches. Inner diameter is j 46 inches. l Construction: Carbon steel throughout. Doubic walled construction.
Inc walls are 5/S inches thick with a 4-1/2-inch air gap between outer cask wall and inner jacket wall and 1-1/2-inch air gap ,
between inner and oater jacket walls. l Attachnents: Eleven 1-inch hex head bolts connecting the jacket to j tac cask base at the bottom cdge of the jacket. ,
I Liftint and Tiedown Devices: Two 10-by 7-by 6-inch steel biccks located on Inc slees or the main box section. A 4-inch dianctor hole is cut through each block to accept cables or clevis. ,
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Penetrations: Air passage from spaces between cask base and protective jacket base plate through a 4-inch diameter hole in the top of tho inner protective jacket and out through two 1-inch by 6-inch horizontal slots on top of box section of outer protective jacket.
Shock Absorber: The spacc between the top of the cask and the jacket is filled with aluminum "Hexec1" shock absorber (honeycomb), 3-inch thick when the extension is used and 17-inch thick when the extension is not used.
D. Packace Contents Solid metal or metal oxide byproduct material; clad, encapsulated or contained; limited to 6,500 thermal watts.
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r SAFETY EVALUAT!T:
A cash with the described design featurcs and enclosed in the protective jacket ncets the package standards of 10 CFF,71. The results of our evaluation are discussed below.
71.31 General Standards for all Packaacs 71.31(a) The steci, Icad and c,asket natorial in the cask and jacket, arc not subject to galvanic action in the design configuration. The solid netal or netal oxide contents will be clad or contained and no reactions are expected. Only air will be authorized as a primary coolant.
71.31(b) Positive closure of the cash is provided by cight 3/4-inch bolts for the lid. A lock wire and seal that must be broken is affixed to a closure bolt. The protective jacket, bolted to the base, covers the cask closura.
71.31(:) The lifting devices of the lid and cask uill satisfactorily resist the specified force of three times the weight of the lid and cask, respectively. Both are covered by the protective jacket during transport. Devices on the jacket, used for both lifting and tic-doun, will ncet the specified force. No other part of the jacket can be used for lif ting.
71.31(d) Our analysis shows that the tie-down systen is capable of with-standing, without generating stress in any natorial of the package in excess of its yield strength, forces of 2 tines loaded weight vertically, 10 times fore and aft, and 5 times laterally.
Failure of the tic-dova/ lifting devices will still leave the basic protective feature of one 5/S-inch thick steel shell to protect the cask.
71.32 Structural Standards 71.32(a) Considering the conplete nackage (cask and protective jacket) as a bean supported at the two axial ends, an eaually distributed load equal to 5 times the loaded weight of the packago would prodace a maxinun stress in the outer shc11 of the protective jacket of 1100 psi which is insignificant conpared to the yield strength ,
of 59,000 psi for carbon steel.
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In our previous evaluation of this cask with the extension, we showed that the cask with the extension could support 10 times its weight when supported at its ends without execeding.the ultimate strength of the bolts holding the extension to the cask.-
I 71.32(b) The package is adcquate to retain all contents when subjected to a pressure of 25 psig. Stresses caused by this pressure are insignificant and the positive cask closure is not affected.-
71.35 Standards for Normal Conditions of Transnort 71.35 (a and Our evaluation of the package indicates that under the c) conditions of transport specified in Appendix A of 10 CFR 71, normal there will be no release of radioactive material from the package; the effectiveness of the package will not be subt.tantially reduced; there vill be no ' mixture of gases or vapors in the package which could reduce its effectiveness; no contamination of the primary coolant will occur; there will be no loss of coolant, and the package will not vent to the atmosphere. The bases on which we conclude that the package will safely meet the normal conditions are as follows:
Thermal !
l Exposure of the cask to direct sunlight at an ambient Water temperature of 130*F will not be used and -40*F will not affcct the package components.
as coolant. Therefore, freeting'at -40*F is not a probica.
At 130*F' ambient with air as the primary coolant, we estimate from GE's analysis, that the internal temperature of the cask will be about 390*F and '
the pressure about 9 psig which is within the cask relief pressure of 15 psig.
If water would be used as primary coolant, the cask would vent with a.
normal ambient temperature of 130*F; to assure meeting the requirement of no venting, a license condition will specify only air as a primary coolant, Prcssure The protective jacket is not a sealed vessel and is therefore unaffected by pressure changes.
The cask has been satisfactorily tested at 30 psig. l Thereforc, a pressure change of 0.5 times standard atmospheric would not affect the cask.
Vibration ,
GE notes that inspection of Model 700 casks used since 1958 reveals'no - l 1
cvidence of vibration damage significant to safety.
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r Wa:cr Snra) and Free Dron j tlater spray and the specified one-foot drop would not affect this metal ,
packa;c. l 1
Penetration )
1 Dropping a thirteen-pound, 1-1/4-inch diancter bar from four feet would ;
not damage this package protected by the jacket with 5/8-inch thick stcci l walls.
Comorcssion i
This package (jacket) can withstand a compressive load of five times its weight.
71.36 Standards for llynothetical Accident Conditions I I
The specified sequential accident conditions of a thirty-foot free drop l onto an unyielding surface, a forty-inch drop o nto the flat end of a six- )
inch diancter bar, and exposure within a 1475'F heat source for thirty i minutes have been evaluated and we have determined that shielding would 1 not be lost nor would radioactive material be released from the package.
The bases on which we have concluded that the package will safely meet the hypothetical accident conditions are as follows:
1 Free Dron and Penetration j CE based the design and construction of the Model 700 jacket on an extra-polation of the actual dron t9st data of the Model 100 Cask and data 1 reported by C. B. Clifford{1)(2) and 11. G. Clarke, Jr.(3),
)
(1) C. B. Clifford, The Desien, Fabrication and Testin of a Duarter Scalc !
of the Demonstration Uranium Fuel Elencat nalp*ung Casx, KY-540 (June 10, l i
1966). ,
(2) C. B. Clifford, Demonstration Fuel T.1cment Shinnint Cask from Laninated Uranium Metal - Testina Program, Proceedings of the 6econd international Symposium on Packaging and Transportation of Radioactive Materials, October 14-18, 196S, pp. 321-556. j (3) li. G. Clarke, Jr., Some Studies of Structural Resnonse of Casks to ,
Innact, Proceedings of Inc decond international Symposium on Packaging j and Transportation of Radioactive Materials, Cetober 14-18, 1968, pp. j 373-398 !
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/t ilc have reviewed the design and construction of the jacket and its connections to the base. ile agree that the jacket will protect the cask from the puncture and the free fall. Our analysis indicated =that a bottom drop d
with one edge of the base striking first would re ult in the maximum stresses to the bolts connecting the jacket to the base. Our analysis indicated that the design of the bolts will provide adequate strength to insure that the jacket remains connected to the base. The shock absorbing material in the top of the jacket will protect the cask closure from a top-end drop. We therefore agree with GE that the jacket provides adequate protection. .
Thermal Since the analysis indicated that the jacket would sustain only minor danar,c from the puncture and drop tests, GE considered the package, for the purpose
. of thermal resistance, as essentially undamaged. The package was assessed for the _ thermal exposure by the GE TilID heat transfer program. The calculations indicate that the cask cavity will reach a maximum temperature'of 464*F, 34 minutes after the thermal exposure. At this temperature, with air as the primary coolunt, the pressure in the cavity is 10.7 psig--well below the 15 psig relief pressure. A condition of the license will authorize only air as a primary coolant; hence possibic venting and loss of radioactive material because of water-stcam pressure need not be considered. Accordingly, we agree that there vill be no loss of radioactivo material.
The temperatures reached by the cask are below the melting point of the Icad shiciding.
Water Immersion Not applicable to a byproduct package.
Conclusion 1
We conclude that the GE iodel 700 package loaded as proposed, meets all portinent transport criteria' of 10 CFR 71, provided (1) only air is used as a primary coolant, and (2) the cask is enclosed in the protective jacket during transport; these requirements will be specified as conditions of the license along with the conditions specifying the load as proposed and the package description. With those conditions to assure compliance with ,
10 CPR 71, we recommend issuance of the requested license. ,
l Signed: /S Y bu/)
J is. P.' drown Approved:
L D. Chitwood, Cnief l
Irradiated Fucis Branch Division of Materials Licensin DATE MAR 11 1969 q
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GENER AL h ELECTRIC ENGINEERING D GENERAL ELECTRIC COMPANY, P.O. BOX 460, PLEASANTON, CALIFORNIA 94566 DIVIS1ON
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June 22, 1982 1 i T H.T. , - l 4 Da.N
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C. E. MacDonald, Chief Transportation Certification Branch jc' RECENED y.7 j 3 Divis!on of Fuel Cycle and Material Safety I- JUN 25 GE2 h 2j l U.S. Nuclear Regulatory Commission 'l U 3 getga ntcutMGM ~ '
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Washington D.C., 20555 'g "'
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REF: Certificate of Compliance No. 5942, Docket 71-5942 to ,
Dear Mr. MacDonald:
Enclosed are copies of the new certification drawings for the General Electric Model 700 shipping container. These drawings (12904768,129D4769, and 12904770) replace those currently referenced in Section 5(a) 3 of Certificate of Compliance No. 5942.
A check for $150.00 for this administrative amendment is enclosed.
Sincerely, Y $* , l G. E. Cunningham '
Sr. Licensing Engineer GEC:sl r- %
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