ML20136E183
| ML20136E183 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/05/1981 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| Shared Package | |
| ML20136E188 | List: |
| References | |
| FOIA-85-513 IAL, NUDOCS 8108060139 | |
| Download: ML20136E183 (2) | |
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.lMMEDI ATE ACTlCM LETTER
- June 5, 1981 U
Docket No. 50-409 Dairyland Power Cooperative ATTN:
Mr. F. W. Linder General Manager 2615 East Avenue - South Lacrosse, Wisconsin 54601 Centlemen:
's This' refers to the telephone conversation of June 4,1981 in which Mr. R. E. Shimshak and members of his staff described to Messrs. L. R. Greger and C. C. Peck of my staff the measures you propose to take to assure that external package contamination levels do not exceed DOT limits during trans-port of the NAC-lD cask from 1ACBWR to Morris Operation on June 5,19.81.
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It is our understanding that these measures will include the following Decontamination of the cask to less than the DOT limit.
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1 Packaging of the cask in double plastic bagging which will. cover the entire external surface of the cask except the trunnions.
The trunnions are to be covered with tape. The plastic is to be secured with tape and plastic banding.
A " chase car" is to accompany the ship =ent.
Ahealthhhysicstech-nician, or equivalent will make the trip. Monitoring instruments i
and extra plastic tape will be carried.
The package integrity will be checked periodically.
The first check will be made about twenty miles from the plant.
Subsequent checks will be made at intervals not exceeding eighty miles.
. The entire trip will take place during daylight hours to permit observation of the package condition.
This office finds the above measures acceptable; and sufficient to permit release of the NAC-lD fuel cask from your site.
1MMEDI ATE ACTICN LETTJ' R 9
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e Dairyland Power Cooperative June 4, 1981 If you believe our understanding of this matter to be incorrect, p1 ease contact this office by telephone immediately.
S'incerely,
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James G.;Keppler
.j Director
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Mr. R. E. Shimshah, Plant Superintendent
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(RIDS) lj Resident Inspector, RIII Mr. J. J. Duffy, Chief Boiler Inspector Mr. Stanley York, Chairman Public Service Commission
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DEC 121979 ELD I&E FCTC: RHO RJones, SD 71-6693 State Health Official NRC PDR Docket File DWeiss EBrown l
I&E HQ(6)
Those on attached list CRChappell Gentlemen:
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-The attached order:
(a) Amends Certificate of Compliance No. 6698.
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j (b) Lifes'susper.ston of the general license to use casks.gSerials Nos. NAC-1B, NAC-lD, and NAC-1E.
I This order is effective immediately.
Sincerely, i
Wgned) Wi]Unm J, Dircks William J. Dircks, Director Office of Nuclear Material Safety and Safeguards
Enclosures:
1.
Order Concerning Model No. NFS-4 packaging 2.
Certificate of Compliance No. 6698, Rev. No. 9 cc: w/enci Mr. Richard R. Rawl Depart:nent of Transportation Dr. Donald M. Ross D
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Il depart:aent of Energy 12/3/79 12/ /79 72/-l/79 I entical orders sent to those on attached list, ycf I,9 8'0633a rp FQ{
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MODEL NO. NFS-4 PACKAGING USA /6698/B( )F Addressee's (w/encis)
Ltr dated:
DEC 121979 Nuclear Fuel Services, Inc.
Florida Power and Light Company ATTN: Mr. James R. Clark ATTN: Mr. Robert E. Uhrig 6000 Executive Boulevard, Suite 600 P.O. Box 529100 Rockville, MD 20852 Miami, FL 33152 Commonwealth Edison Baltimore Gas & Electric Company ATTH: Mr. L. D. Butterfield, Jr.
ATTN: Mr. A. E. Lundvall, Jr.
P.O. Box 767 P.O. Box 1475 Chicago, IL 60690 Baltimore, MD 21203 Maine Yankee Atomic Power Co.
Battelle Columbus Laboratories ATTN: Mr. L. H. Heider ATTN: Mr. Harley L. Toy Turnpike Road (RT 9) 505 King Avenue Westboro, MA 01581 Columbus, OH 43201 Nuclear Assurance Corporation Babcock and Wilcox Company ATTN: Mr. John R..Donnell ATTN: Mr. D. W. Zeff 1
24 Executive Park West P.O. Box 1260
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Atlanta, GA 30329 Lynchburg, VA 24505
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Wisconsin El.ectric Power Ccmpany Boston Edison Company ATTN: Mr. Sol Burstein ATTN: Mr. G. Carl Andognini 231 West Michigan 800 Boylston Street Milwaukee, WI 53201 Boston, MA 02199 i -
Rochester Gas & Electric Corporation Dairyland Power Cooperative ATTN: Mr. L. D. White, Jr.
ATTN: Mr. R. E. Shimshak 89 East Avenue P.O. Box 135 Rochester, NY 14649 Genoa, WI 54632 Jersey Central Power & Light Company Westinghouse Electric Corporation ATTN: Mr. J. T. Carroll, Jr.
ATTN: Mr. Ronald P. DiPiazza P.O. Box 388 P.O. Box 355 Forked River, NJ 03731 Pittsburgh, PA 15230 Duke Power Company Florida Power Corporation ATTN: Mr. W. O. Parker, Jr.
ATTN:
Mr. J. T. Rodgers 422 South Church Street P.O. Box 14042 Charlotte, NC 28242 St. Petersburg, FL 33733 Southern California Edison Company
. General Electric Company ATTN: Mr. William H.. Seaman ATTN: Mr. D. M. Dawson, MC 861 P.O. Box 800 175 Curtner Avenue Rosemead, CA 91770 San Jose, CA 95125
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
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NRC CERTIFICATE OF COMPLIANCE
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Docket No. 71-6698 NO. 6698 FOR RADI0 ACTIVE
)
MATERIALS PACKAGES
)
ORDER AMENDING CERTIFICATE AND TERMINATING IN PART ORDER TO SHOW CAUSE I
By Order dated April 6,1979, the Director, Office of Nuclear Material Safety and Safeguards suspended the use of seven casks used for transpor-tation of spent reactor fuel pending a showing that these casks were fabricated in accordance with the de i of Compliance No. 6698 (Revision 8)._j/gn approved by the NRC in Certificate As required by the April 6th Order.as a condition to rescission of the ordered suspension, the Nuclear Assurance Corporation (NAC) and Duke Power Company have reviewed their quality assurance records and have made physical measurements of five of the g/ ected casks (Serial Nos. NAC-1A, NAC-lB, ff NAC-lC, NAC-lD, and NAC-1E).1 The Commission's Office of Inspection and Enforcement has. inspected the quality assurance records of the casks built by Excelec> Developments, Inc. for NAC.
The results of these various reviews and inspections indicate that two casks (Serial Nos. NAC-1A and NAC-lC) are bowed and therefore do not meet the requirements of Certificate of Compliance No. 6698.
It is not known at this time whether the cause of the bowing in these two casks occurred during their manufacture or during their actual use. The NRC staff has found that 1/ Seven casks designated as Model No. NFS-4 have been fabricated to the design approved in the Certificate: Serial Nos. NAC-1A, NAC-lB, NAC-lC, NAC-10, NAC-lE, NFS-4A, and NFS-48. These casks are used to transport spent reactor fuel. The basis for the April 6th Order was the discovery that measurements of the dimensions of the inner shell of one cask (Serial No.
NAC-lA) indicated that the shell was warped or bowed, but the exact cause or extent of the warp or bow was not known at the time.
2/ These casks were originally manufactured by Excelco Developments, Inc.,
Tor NAC. Duke Power Company is the present owner of casks Serial Nos. NAC-1A and NAC-18. NAC owns the remaining casks in the NAC-1 series. Casks Serial Nos. NFS-4A and NFS-48, owned by Nuclear Fuel Services, Inc. (NFS), were also subject to the April 6th Order, but NFS has not submitted any information in response.to the Order. Accordingly, the suspension as provided in the April Order remains in full effect with respect to these two casks.
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casks Serial Nos. NAC-18, NAC-lD, and NAC-lE substantially conform to Certificate of Compliance No. 6698 (Revision 8), except for certain minor differences that are unrelated to bowing or warping of the cask cavity shells._3/ Therefore, the NRC staff believes that these three casks can be returned to service, subject to certain restrictions incorporated in Revision 9 to Certificate of Compliance No. 6698 (attachment to this Order and discussed in Part II infra), without undue risk to the health and safety of the public.
II While the review of casks Serial Nos. NAC-1B, NAC-1D, and NAC-lE show that these casks conform to the design authorized by Certificate of Compliance No. 6698 (as revised to include the minor technical amendments discussed in footnote 3), the staff has determined the three casks should not be returned to unrestricted service without further technical evaluatfon, on the basis that:
3/ As part of the quality assurance review under the April 6th Order, Nuclear X'ssurance Corporation (NAC) reviewed and compared the drawings approved in Certificate No. 6698 with fabrication drawings to which the casks were actually built, reviewed the quality assurance records, and made physical measurements of the casks' cavities. This review indicated that certain tolerances and
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dimensions on the fabricated casks differed frora those shown on Nuclear Fuel Services (NFS) Drawing No. E10080 (Revision 16), which is the drawing approved in Certificate of Compliance No. 6698. By letters dated June 8. July 26, and October 31, 1979, NAC requested an amendment to Certificate of. Compliance No. 6698 to incorporate changes in dimensions, tolerances, and notations, that were unrelated to bowing or warping of the cask cavity shells. Revising the Certificate to include these requested changes would resolve the discrepancies between the drawings approved in Certificate No. 6698 (Revision 8) and the as-built casks, Serial Nos. NAC-18, NAC-lD, NAC-lE..NAC has requested that Certificate of Compliance No. 6698 be amended to incorporate Revision 19 of NFS Drawing No. E10080 as the approved design drawing. Casks Serial Nos.
NAC-18, NAC-lD, and NAC-lE have been constructed in accordance with Revision 19 to Drawing No. E10080. The staff is issuing Revision 9 to Certificate No. 6698 (attached as Appendix A to this Order) to incorporate the requested changes as well as to impose various restrictions further described in this Order. All items identified in the NRC's inspections will be resolved to the NRC's satis-faction with the incorporation of Revision 19 of Drawing No. E-10080 and revised Condition No.13 into Certificate of Compliance No. 6698 (Revision 9).
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the bowing observed in casks Serial Nos. NAC-1A and NAF-lC, both of which are of the same design as the other casks, may be an indication that casks of this design are susceptible to buckling of the inner shell in unrestricted use; 2.
the cask design has not been analyzed to show that the inner shell would not buckle in unrestricted operation; and 3.
it cannot be concluded without such analysis that the cask would meet the requirements in 10 CFR Part 71 under Normal and Hypothetical Accident Conditions.
Accordingly, Certificate of Compliance No. 6698 has been revised (Revision 9) to include additional restrictions which would permit limited use of casks Serial Nos. NAC-18, NAC-10, and NAC-lE, and which would also require certain additional inspections of the casks at periodic intervals. The staff believes these restrictions are sufficient to assure the integrity of the inner shell under ' normal conditions of transport and that the possibility of buckling under accident conditions would be. reduced. Revision 9 to the Certificate reduces the maximum decay heat load of the contents permitted in the cask from 11.5 kilowatts te 2.5 kilowatts. The casks are also required to be shipped dry (i.e., water is no longer permitted in the cask cavity during shipment). These provisions will substantially reduce the pressures within the cask cavity.under the normal and hypothetical accident conditions.
. specified in 10 CFR Part 71.
In addition, the temperatures and thermal gradients througl the cask
- walls and longitudinally along the shells would also be reduced. These restrictions will lower the stresses present in the inner shell during use of the casks under normal and accident conditions.
The revised Certificate also requires the inner shell of each cask to be inspected and measured at intervals not to exceed six months. The casks are to be withdrawn from service if the measurements indicate any changes in the dimensions of the inner cavity. Also, the cask cavity is required to be pressure tested quarterly for possible leakage.
Even if the inner shell were to buckle under the hypothetical accident conditions in 10 CFR Part 71, and if this should lead to a violation of the integrity of the inner shell, the amount of material released would not be expected to exceed the amount pennitted in 10 CFR Part 71 based on the guidance provided in Regulatory Guide 25.
III Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, and the Coninission's regulations in 10 CFR Parts 2 and 71, IT IS HEREBY ORDERED THAT:
(1) Certificate of Compliance No. 6698, Revision 9, supersedes l
in its entirety Certificate of Compliance No. 6698, Revision l
8, dated October.25, 1978; and j
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-4 (2) The suspension of the use of NAC casks with Serial Nos. NAC-1B, NAC-10, and NAC-lE is hereby rescinded.
The Order of April 6,1979, except as stated above, remains in effect in accordance with its terms.
IV Any person who has an interest affected by this Order may request a hearing, within twenty (20) days of the date of this Order, with respect to this Order. Any request for a hearing shall be addressed to the Director, Office of Nuclear Material Safety and Safeguards, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555. Any request for a hearing shall not stay the immediate effectiveness of this Order.
If a hearing is requested by a person whose interest is affected by this Order, the Commission will issue an order designating the time and place of hearing. The issue to be considered at such hearing will be:
Whether the restrictions added by Revision 9 to Certificate of Compliance No. 6698 are necessary to p,rovide reasonable assurance that the casks will meet the Normal and Hypothetical Accident Conditions in accordance with 10 CFR Part 71.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION
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Williadi J. Dircks, Director Office of Nuclear Material Safety and Safeguards Effective Date:
DEC 121979 Silver Spring, Maryland 9
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- m NRC 618 U.S. NUC1. EAR REGULATORY COMMISSION CERTIFICATE OF COMPLIANCE
.0 CFR 71 For Radioactive Meterials Peckages 1.tal Certificate Number 1.(b) Revision No.
1.(c) Package identification No.
1.(d) Peges No. 1.(e) Totsi No. Pages ti693 9
USA /6698/B( )F 1
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- 2. PREAM8LE 2.lal This certificate is issued to satisfy Sections 173.393a.173.394,173.395, and 173.396 of the Department of Transportation Haaerdous Materiais Regulations (49 CFR 170189 and 14 CPR 103) and Sections 146-19-10e and 146-19-100 of the Department of Transportation Dangerous Cargoes Reeviations (46 CFR 146-149). as amended.
I 2.(b)
The packaging and coments described in item 5 be6ow, meets the safety standeres set forth in Subpert C of Title 10. Code of Federot Regulations. Port 71,"Peckaging of Radioactive Meteriais for Transport and Transportation of Radioactive Material Under Certain Conditions.*
i 2.fcl This certificate does not relieve the consignor from compliance with any reovirernent of the regulations of the U.S. Departrnent of Transportation or other applice:We regulatory agencies, including the government of any country through or into which the package i
will be trensoorted.
- 3. This certificate is issued on the bases of a safety analys.e report of the package design or application-l 3.lal Prepared by (Name and address):
3.(b)
Title and identification of report or ecolication:
Nuclear Fuel Services, Inc.
NFS application dated October 6, 1972, P.O. Box 124
.as supplemented.
Best Valley, NY 14171 j
3.(el Dock. No.71 6698 j
- 4. CONDITIONS This certif cote is conditional upon the fulfilling of the requirements of Subpert O of 10 CFR 71, as appliceele, and the conditions specified in ai.m s o ow.
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- s. Description of reckaging and Avihorised Conte, ts. undes Nu,noer ri.e.i. Crase. O ner conditions and R.e.rences:
(a) Packaging i
(1) Model No.: NFS-4 i
(2) Description A steel, lead and water shielded shipping cask.
The cask is a right circular cylinder with upper and lower steel encased balsa impact limiters. The overall dimensions are 214 inches in length and 50 inches in diameter. The gross weight of the cask is approximately 50,000 pounds. The inner cavity is 178 inches long and 13.5 inches in diameter. The thickness of the inner.
shell is 5/16 inch and 1-1/4 inches for the outer shell. The two stainless steel shells are welded to a 2-inch thick stainless steel shield disc at the bottom. The' annulus between the inner and outer shells is filled with j
lead (max. lead thickness 6-5/8 inches, minimum 5 inches).
The lid is stainless steel frustum of cone 7.5 inches thick. The lid is secured to the cavity flange by six ASTM-A320, Grade L43,1-1/4 inch diameter bolts. The seal is provided by two polytetrafluoroethylene 0-rings.
Four neutron shield tanks, each with surge tank and rupture disc, provide a 4-1/2 inch thick (borated) water-ethylene glycol mixture around the outer i
shell. Four trunnions, two located on either side of the upper or lower 1
impact limiter, are provided. Other cask features include two drain valves located in the bottom shield disc vent valve, head closure gasket leak s
check valve, rupture disc-pressure relief valve system located in the cavity flange, fuel canisters for PWR and BWR shipments, and spacers to accommodate shorter fuel assemblies.
For transport the cask may be enclosed 1
in an expanded metal cage.
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s, Page 2 - Certificate flo. 6698 - Revision No. 9 - Docket No. 71-6698 4
5.
(a) Packaging (continued)
(3) Drawings The NFS-4 shipping cask is constructed in accordance with Nuclear Fuel i
Services, Inc., Drawing No. E 10080, Rev.19 (Sheets 1 through.4).
The i
fuel assemblies are positioned within the fuel canisters shown in Figure 2.1.3 of the application dated October 6,1972. Spacers may be used to accommodate shorter fuel assemblies within the fuel canisters.
(b) Contents 4
(1) Type and form of material The niinimum cooling time of each fuel assembly and rod shall be 120 days, and (i) Irradiated PWR or BWR uranium oxide fuel assemblies with the following maximum active dimensions and maximum compositions prior to irradiation:
Fuel Assembly Data PWR BWR Envelope, inches 8.60x8.6Ox150 5.44x5.44x144 Enrichment, w/o U-235 3.6 3.0 I
Weight of Uranium, kg 480 197 H/U atomic ratio 5.51 (ii) Fuel assembly enriched in the U-235 isotope to not more than 2.5 w/o, with active fuel dimensions not to exceed 4.2" x 4.2" x 110" long.
(iii) Byproduct and special nuclear material in the form of irradiated uranium oxide fuel ro'ds.
(iv) Solid nonfissile irradiated hardware and neutron source components.
(v) Fuel assembly enriched in the U-235 isotope to not more than 4.1 w/o, with active fuel dimensions not to exceed 7.8" x 121" long.
i (vi) Byproduct and special nuclear material in the form of irradiated uranium and plutonium oxide fuel rods.
Prior to irradiation, the j
maximum enrichment in U-235 plus plutonium not to exceed 4.0 w/o.
(vii) Irradiated PWR uranium oxide fuel assemblies including additional irradiated fuel rods inserted and secured in the guide thimbles. The fuel assemblies shall confonn to the maximum active dimensions as described in Item 5(b)(1)(i) and partially disassembled fuel assemblies shall be equipped with an assembly carrier as shown in Battelle Drawing No. 00-001-676, or equivalent.
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Page 3 - Certificate No. 6698 - Revision No. 9 - Docket No. 71-6698 5.
(b) Contents (continued)
(1) Type and form of material (continued)
(viii) Irradiated BWR uranium oxide fuel assemblies.
Prior to irradiation, the maximum enrichment in the U-235 isotope shall not exceed 4.0 w/o with active fuel dimensions not to exceed 5.63" x 5.63" x 83.8" long.
(2) Maxiinum quantity of material per package, f
Not to exceed a decay heat generation of 2.5 kw and l
(i) Item 5(b)(i) above:
One (1) PWR fuel assembly, or Two (2) BWR fuel assemblies; or l
(ii) Item 5(b)(1)(ii) above:
Four (4) fuel assemblie's contained within the fuel basket shown in NFS Dwg. No. lA-T-1107 Rev. 0; or (iii) Item 5(b)(1)(,iii) above:
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Maximum Fissile Maximum Enrichment Mass Limit (w/o U-235)
(ko of U-235) 3 2.0 4
1.6 5
1.5; or j
(2) Maximum quantity of material per package (continued)
(iv) Items (b)(1)(iv)above:
As needed, appropriate component spacers shall be used in the cask cavity to limit movement of contents during shipment; or I
.(v) Item 5(b)(1)(v) above:
One (1) fuel assembly; or l
(vi) Item 5(b)(1)(vi)above:
Fuel rods within the fuel canisters described in 5(a)(3). The maximum mass of U-235 plus plutonium shall not exceed 4.0 kg. A suitable i
fixture may be used to secure the fuel rods within the canister; or
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Page 4 - Certificate No. 6698 - Revision No. 9 - Docket No. 71-6698 5.
(b) Contents (continued)
(2) Maximum quantity of material per package (continued)
(vii) Item 5(b)(1)(vii) above:
The maximum compositions of one PWR fuel assembly including additional rods shall conform to Item 5(b)(1); or (viii) Item 5(b)(1)(viii) above:
Two (2) BWR fuel assemblies.
Prior to irradiation, the maximum uranium content per assembly shall not exceed 122 kg.
(c) Fissile Class III Maximum number of packages per shipment One (1) 6.
The cask shall be shipped dry (no water coolant in cask cavity).
7.
The water-ethylene glycol mixture in the neutron shield tanks may contain up to 1.0 weight percent boron. This mixture shall not freeze or precipitate in a temperature range from -40'F to 330*F.
The cask contents shall be so limited under normal conditions of transport that 27 s.
times the neutron dose rate plus 1.4 times the gamma dose rate will not exceed 1,000 millirems per hour at three (3) feet from the external surface of the package.
9.
The vent'and drain valves shall be 1/2" FG466TSW Hiser ball valves (Worchester Valve Company,Inc~.). The ball of the valve may have a bleed hole to equalize the pressure between the cask cavity and the ball passage in a closed position. Alternatively, the vent and drain lines may be sealed with pipe plugs.
10.
In addition to the requirements of Subpart D of 10 CFR Part 71, each package prior to first use shall meet the acceptance ' tests and criteria specified on pages A-21 thru A-34 of the Nuclear Fuel Services, Inc. application dated October 6,1972, as amended, March 1,1973 and Nuclear Assurance Corporation letter dated November 1,1974. The results of these tests shall be documented and retained for the life of the cask.
- 11. At periodic intervals not to exceed (3) years, the thennal performance of the cask i
shall be analyzed to verify that the cask operation has not degraded,below that which is licensed. Following the initial acceptance tests, the heat source may be that provided by the decay heat from the contents of the package provided that the heat source is equal to at least 25% of the design heat load for the package.
Each cask that fails to meet the thennal acceptance criteria given on pages A-21(a) and A-21(b) using the TAP computer program, or equivalent, shall be withdrawn from service until corrective action can be completed.
The rupture discs for the neutron shie.ld tanks shall be type "B" or "DV" (BS&B Safety Systems,Inc.)orequivalent.
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'Page 5 - Certificate No. 6698 - Revision No. 9 - Docket No. 71-6698 13.
In lieu of the requirements of 10 CFR 571.54(h), the licensee shall perform periodic maintenance and testing of 0-rings, drain and vent ball valves, relief valves, and rupture discs of the cask as indicated in the table given below. During inactive periods, the maintenance and testing frequency may be disregarded provided that the package is brought into full compliance prior to the next use of the package.
Cask Component Period Test / Action
)
)
Ball Valve Each shipment Hydro test to 80 psig*
Ball Valve Annually Replace seats and seals 0-rings Each shipment Test to 80 psig*
0-rings Quarterly Test to 167 psig*
0-rings Annually
. Test to 1006 psig*
l Inner Containment Vessel Quarterly Test to 250 psig*
Cavity Relief Valve Annually Test at set point l
Cavity Rupture Disc Annually Replace
.i Neutron Shield Tank i
Rupture Disc
. Annually Replace
.t Impact Limiters Annually Test for leakage
- There shall be no visual (pressure gauge) indications of pressure drop i
for the component under test during a 10-minute test period. Otherwise, corrective action shall be taken and the test repeated until such time as the component meets the specified test.
(Test to pressures equal to or greater than those indicated.)
- 14. At least every six (6) months, corsne'ncing six months following the first shipment of irradiated fuel under Revision No.- 9 to Certificate of Compliance No. 6698, the licensee shall perform physical measurements of the cask inner shell. Any cask whose inner container dimensions are measured to deviate by more than +0.015 inch i
at, comparable points from the dimensions documented in Appendix C to NTC letter dated June 8,1979 shall be removed from service.
- 15. The package authorized by this certificate is hereby approved for use under the j
general license provisions of 10 CFR 571.12(b).
16.
Expiration date: December 31, 1980.
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Page 6 - Certificate No. 6698 - Revision No. 9 - Docket No. 71-6698 REFERENCES Nuclear Fuel Services, Inc. application dated October 6,1972.
Supplements dated: November 9,1972; January 10 and 22 February 1 and 28, March 1,14, and 21, May 4, June 4, and July 26,1973; July 17,1974; May 4,1976; and November 9, 1977.
Nuclear Assurance Corporation supplements dated: November 1,1974; August 13 and December 24, 1975; September 13, 1976; October 20,1977; May 25, July 18, and September 25,1978; and June 8. July 26, and October 31, 1979.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION
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Charles E. MacDonald, Chief Transportation Certification Branch Division of Fuel Cycle and Material Safety DEC 121973 te:
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CEMacDonald RH0degaarden CRChappell JMurray, ELD AGrella, I&E Jul.22ISS)
RJones, S 71-6698 State Health Official NRC PDR Docket File GJackson ZMcDonald IE HQ (6)
W NMSS R/F FCTC R/F
/
Gentlesaen:
The attached order:
(a) Prohibits the use of Model No. NFS-4, Serial No. NAC-lD packaging by NRC licensees until reasonable determination is made by NRC that DOT surface contamination licitts will not be exceeded on subsequent shipraents of this packaging.
(b) Requires further order of the Comission to return the packagings to service.
This order is effective imediately.
Sincerely, (signed) John G. Davis John G. Davis, Director Office of Nuclear Material Safety and Safeguards
Enclosure:
As stated cc w/ enc 1:
See next page Identical orders sent to those on attached list pf$ b f\\,
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U. S. Department of Transportation Materials Transportation Bureau ATTN: Mr. Richard R. Rawl DMT 221 Washington, D. C.
20590 Department of Energy ATTN Dr. Donald M. Ross MS E-201 Washington, D. C.
20545 Reynolds Electric and Engineering Company, Inc.
ATTN: Mr. Arden E. Bicker P.O. Box 14400 Las Vegas, NV 89114 Oak Ridge National Laboratory ATTN: Mr. William E. Terry P.O. Box X Oak Ridge, TN 37830 Nuclear Assurance Corporation ATTN: Mr. Charles R. Johnson 24 Executive Park West Atlanta, GA 30529 Department of Energy ATTN: Mr. A. T. Newmann P.O. Box 14100 Las Vegas, NV 89114 Department of Energy ATTN: Mr. James M. Peterson P.O. Box 550 Richland, WA 99352
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U Identical orders sent to:
Babcock and Hilcox Company General Electric Company ATTN: Mr. A. F. Olsen ATTN: Mr. D. M. Dawson P.O. Box 1260 175 Curtner Avenue Lynchburg, VA 24505 San Jose, CA 95125 Baltimore Gas & Electric Company Jersey Central Power & Light Company ATTN: Mr. A. E. Lundvall, Jr.
ATTN: Mr. John Sullivan, Jr.
P.O. Box 1475 P.O. Box 388 Baltimore, MD 21203 Forked River, NJ 08731 Battelle Columbus Laboratories Maine Yankee Atomic Power Company ATTN: Mr. Harley L. Toy ATTN: Mr. L. H. Heider 505 King Avenue Turnpike Road (RT 9)
Columbus, OH 43201 Westboro, MA 01581 Mr. L. H. Heider Boston Edison Company Turnpike Road (RT 9) Westboro, MA 01581 ATTN: Mr. G. Carl Andognini 800 Boylston Street Nuclear Fuel Services, Inc.
Boston, MA 02199 ATTN: Mr. Larry Wiedemann P.O. Box 124 Commonwealth Edison West Valley, NY 14171 ATTN:
Director of Nuclear Licensing P.O. Box 767 Rochester Gas & Electric Corporation Chicago, IL 60690 ATTN: Mr. John E. Maiser 89 East Avenue Dairyland Power Cooperative Rochester NY 14649 ATTN: Mr. R. E. Shimshak P.O. Box 135 Southern California Edison Company Genoa, WI 54632 ATTN: Mr. William H. Seaman P.O. Box 800 Duke Power Company Rosemead, CA 91770 ATTN: Mr. W. O. Parker, Jr.
422 South Church Street Westinghouse Electric Corporation Charlotte, NC 28242 ATTN: Mr. A. J. Nardi P.O. Box 355 Florida Power and Light Company Pittsburgh, PA 15230 ATTN: Mr. Robert E. Uhrig P.O. Box 529100 Wisconsin Electric Power Company 1 --
Miami, FL 33152 ATTN: Mr. Sol Burstein 231 West Michigan F!orida Power Corporation Milwaukee, WI 53201 ATTN:
Dr. Patsy Y. Baynard P.O. Box 14042 St. Petersburg, FL 33733
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o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Docket No. 71-6698 NRC CERTIFICATE OF COMPLIANCE NO. 6698 FOR RADI0 ACTIVE MATERIALS PACKAGES ORDER TO SHOW CAUSE.(IMMEDIATELY EFFECTIVE)
I On November 14, 1972, a Certificate of Compliance under 10 CFR Part 71 was issued to Nuclear Fuel Services, Inc., for Model No. NFS-4 cask design. The latest license expired on December 31, 1980, and is currently under timely renewal.
The packaging (" cask") identified as Serial No. NAC-1D is one of seven casks manufactured to the Model No. NFS-4 design.
All seven were suspended from service by the Comission's April 6,1979 Order concerning structural integrity.
On December 12, 1979, after further evaluation of the structural integrity, the Commission permitted three casks, including Cask Serial No. NAC-1D, to return to service with certain restrictions on their use.
II On at least seven occasions between August 1980 and July 1981, following offsite transportation, the cask displayed impermissably high levels of surface contamination under the Department of Transportation's regulations, 49 CFR 5173.397.
Following the discovery of the excessive contaminations, the cask, before reshipment, was required to be decontaminated to the levels permitted by 49 CFR 5173.397.
After transportation following the decontaminations, the cask repeatedly arrived with surface contamination exceeding the permissable limits of 22,000 dpm/100 cm2 by as much as 2
2,000,000 dpm/100 cm.
The increase in surface contamination exhibited following transport suggests that contamination which originally was fixed, was released in transit.
The reason for this excessive contamination, which may be related to the surface finish of the cask, is not fully understood. There appears to be no reasonable assurance that future shipments of the cask would be within the surface contamination limits set forth in 49 CFR 5173.397.
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2 III In view of the repeated instances of excessive surface contamination, in violation of 49 CFR 5173.397, reasonable assurance does not now exist that the public health and safety will not be jeopardized by the continued use of this cask. Therefore, I find that the public health, safety and interest require imediate suspension of use of cask Model No. NFS-4, Serial No. NAC-lD.
IV In view of the foregoing and pursuant to Sections 57,62,81,and161(b) of the Atomic Energy Act of 1954, as amended, and the Comission's regulation in 10 CFR Parts 2 and 71, IT IS HEREBY ORDERED THAT:
(A) Use of the cask designated as Model No. NFS-4, Serial No. NAC-lD, outside the confines of a licensed facility or plant is suspended, effective imediately; provided that, for the sole purpose of attempting to requalify the cask for use outside the confines of a licensed facility or plant, it may be transported (empty) once to an appropriate testing / rehabilitation site, subject to the following procedures:
(1)
Prior to shipment, surface contamination of the cask shall not exceed the levels pennitted by 49 CFR 5173.397.
(2)
The cask shall be packaged in plastic bagging covering the entire external surface of the fcask except the trunnions, which shall be covered with tape.
The bagging shall be secured with tape and banding.
(3) A health physics technician carrying monitoring instruments and extra tape, shall accompany the shipment.
(4) The integrity of the bagging shall be verified at transport intervals of not more than 80 miles, 3w (B) The owner / user show cause, as specified in Section V of this Order, why the suspension of the general license should not be continued until the Director, Office of Nuclear Material Safety and Safeguards, finds there is reasonable assurance that surface contamination levels will not exceed the require-ments of 49 CFR 5173.397 at any point during future shipments of the cask.
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- In determining whether there is reasonable assurance that the cask will not experience excessive contamination levels in transport, the Director will consider among other things:
(1) The extent of the understanding of the cause of the excessive surface contamination (e.g., improper decontam-ination of cask surfaces and condition of cask surfaces).
(2) The action taken to refurbish the cask surfaces and/or decontamination procedures to be used compatible with user waste treatment facilities.
(3) Tests performed which simulate transport conditions to demonstrate the response to Items (1) and (2) above are correct and that excessive contamination levels will not be experienced.
V An owner / user to whom this order applies may show cause within 25 days of the date of this Order by filing a written answer under oath or affirmation which sets forth the matters of fact and law on which the licensee relics. The owner / user may answer, as provided in 10 CFR 52.202(d),
by consenting to the entry of an order in substantially the form proposed in this Order to Show Cause.
Upon failure of the owner / user to file an answer within the specified time, the Director, Office of Nuclear Material Safety and Safeguards, may issue without further notice an order' continuing the suspension as described in Section IV above.
I VI f
The owner / user or any other person who has an interest affected by this order may request a hearing within 25 days of the date of this Order.
Any answer to this Order or any request for hearing shall be filed with Mr. John G. Davis, Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Comission, Washington, DC 20555.
Copies shall also be sent to the Secretary of the Comission and the 3-Executive Legal Director at the same address.
If a person other than
[
the owner / user requests a hearing, that person shall describe specifically.
in accordance with 10 CFR 52.714(a)(2), the nature of the person's interest and the manner in which that interest is affected by this Order. ANY REQUEST FOR A HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OFSECTIONIV(A)0FTHISORDER.
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4 VII If a hearing is requested, the Comission will issue an Order designating the time and place of any hearing.
If a hearing is held, the issue to be considered at any such hearing shall be whether this Order should be sustained.
F0P THE NUCLEAR REGULATORY COMMISSION W'
Jo n G. Davis, Director Office of Nuclear Material Safety and Safeguards Dated at Bethesda, Maryland this J.2. day of July 1981.
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Attanta Georgra 3o329 4o4)325-420o Teter $49567. 542703 i
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'l C Woemstrasse 9 9001 Zaren Setzenana (01)470844 7ees $7275 March 11, 1983 Mr. John G. Davis, Director Office of Nuclear Material Safety and Safeguards i
U.S. Nuclear Regulatory Commission 7915 Eastern Ave.
Washington, D.C.
20910
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Dear John:
As suggested by you, we are sending to you a background document discussing the events surrounding the NAC-1 licensing efforts. Obviously, we wish to come to some resolution of the NAC-1 Certificate of Compliance and are immensely pleased you are taking a personal interest in the spent fuel transportation area. Please let Chuck Johnson or me know if additional information is
- desired, i
l We are looking forward to interaction with you within the next couple of weeks.
Cordially, NUCLEAR ASSURANCE CORPORATION (Mrs.) Carol S. Thorup Group Vice President Resource Analysis CST:pwh Enclosure bD 'l '0 0
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N NAC-1/NFS-4 SPENT FUEL SHIPPING CASKS Nuclear Assurance Corporation Prepared by Charles R. Johnson, March 1983
Background
The NAC-1/NFS-4 spent fuel shipping cask is designed to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies.
The cask is a stainless steel and lead structure consisting of an outer stainless steel snell welded to an upper flange and a lower end casting with an inner stainless steel shell welded to the same two end pieces. Lead was cast between the two shells to prov1de the principal gamma snielding..
A stainless steel neutron shield tank filled with an ethylene glycol-water mixture witn potassium tetraboride as tne neutron poison is located outside of the outer shell.
Handling trunnions are attacned to both ends of the cask for handling and rotating the cask on its trailer.
Impact limiting devices are incorporated on both ends of the cask to cushion it during the drop accident sequence.
The design weight of the cask is 52,000 lbs including contents.
The actual weight of tne cask with a full complement of spent fuel, is on the order of 49,500 lbs.
Casks of this design have logged over a million miles transporting spent fuel in the United States and have been handled in more than 20 facilities.
They have been tne primary means of spent fuel transport in tne United States.
The NAC-1/NFS-4 cask was designed in tne early 1970's.
Its original certificate of compliance was issued in 1972.
The design basis was Title 10, Chapter 1, Code of Federal Regulations Part 71 (10 CFR 71), which essentially corresponds to tne International Atomic Energy Agency (IAEA) Safety 55 ries 6 Regulations for.
.tne Safe Transport of Ra,cioactive Materials.
i The intent of IAEA Safety Series 6 and the 10 CFR 71 regulations is to es'tablish standards of safety whicn provide an acceptable level of control of the l
radiation hazards to persons, property, and the environment that are associated witn the transport of radioactive material. Normal operating conditions for the packagings are specified, as well as a hypotnetical accident series.
Under normal operating conditions the cask is expected to perform satisfactorily in temperatures ranging from -40' to 130* F, at low atmospheric pressure and under the vibrational conditions of transport.
The hypothetical accident series requires the cask to be dropped from tnirty feet onto an essentially unyielding surface in an orientation that would inflict the greatest damage.
The 30 ft.
drop accident is followed by a forty-inch drop onto a six-incn diameter mild steel bar in an orientation whicn would inflict the greatest damage.
The drop accidents are followed by a fire test and a water immersion test.
Regulatory criteria for acceptability of a cask design is that there must be no significant loss of radiation snielding or release of radioactive material throughout and following the accident series.
Licensing Philosophy In the United States, the philosophy has been to design large packages by analysis rather tnan by tests.
Otner countries rely heavily on tests of actual packagings or scale models to verify their capability of withstanding the hypothetical accident series.
To provide some guidance to cask designers in the United States, Regulatory Guides 7.6 and 7.8 were issued in 1977.
These Guides
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suggest acceptable design criteria and load combinations to be considered during analyses for normal operating conditions and for the series of hypothetical accidents.
In Regulatory Guide 7.6, it is stated that "the vessel should meet the speci fication for linear elistic analysis, given in tnis guide."
This means tnat containment of the cask must not be plastica 11y deformed in normal conditions of service and there must be no instability (buckling) under accident conditions.
These are quantitative criteria expressable in stress limits.
Literally interpreted, (2nd as currently used by regulatory personnel) even a strain limited instability, which could.be so slight as to have no chance of breaching containment, under accident conditions is considered failure and such a design feature is unacceptable.
During the initial accident of the series, the 30 foot dr,op, impact limiters are normally provided to soften the impact and limit forces transmitted to the cask structure such that the suggested design criteria (allowable stresses) are not exceeded.
The 40 inch drop on the 6 inch diameter rod (the second accident in the series) will impact an area of the cask that is not protected by impact limiters and it will cause plastic deformation of the cask shell.
To meet the criteria suggested by Regulatory Guide 7.6, the plastically deformed shell cannot be part of the cask's primary containment even though the 6 inch rod does not breach tne shell.
As a direct result, all spent fuel shipping casks currently licensed under 10 CFR 71 have at least two shells with shielding material between them.
The Inner shell is considered containment and is not de, formed during this accident because the outer shell and shielding protect it.
-The Regulatory Guides, while not law, are-treated by regulators as the only acceptante design rules. The cask oesigner must " prove" to the regulatory staf f that his design meets the criteria established by the Regulatory Guides.
Proving that a cask design meets the letter and intent of the Regulations (10 CFR 71),
i.e., no significant degradation of shielding or loss of radioactive material today requires total adherence to the Regulatory Guides.
Without total adherence, the staff claims the design is on new ground and the time required and methods for review become indeterminate.
This extremely conservative philosophy results in cask designs th'at greatly exceed the requirements of the Regulations.
This also results in prohibitive design and licensing costs without any commensurate increase in protection of public health and safety.
Chronology of the NAC-1/NFS-4 Licensing Activities The original Certificate of Compliance (C of C) was issued for the NAC-1/NFS-4 cask design in 1972.
Since the original issuance, there have been fifteen amendments to the C of C.
The casks were designed to meet the requirements of 10 CFR 71, whicn requires the retention of shielding and essentially no leakage of radioactive material under normal conditions of transport and only slight release following the hypothetical accident series.
In early 1979, during a routine check of tne quality assurance records for the manufacture of tne NAC-1/NFS-4 casks, NAC determined that the NAC-1 "A" cask was not in complete conformance with the approved design in that a copper patch had been placed on tne outside of the outer shell of the cask to provide some -
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auxiliary shielding.
This patch was apparently attached to the cask witnout any recorded design review or approval.
The discrepancy was described to the NRC by NAC in a meeting, dated March 29, 1979, and confirmed in a letter to the NRC dated April 2, 1979.
On April 6,
1979, the NRC issued a "show-cause" order stating in part tnat "it must.be demonstrated to the Commission that each cask was fabricated in accordance with the design approved by the C of C; this demonstration snould include a QA review and actual measurements of existing packagings."
The rationale behind this snow-cause order was a concern on the part of NRC that during lead pouring operations, the cask cavity or the inner snell was distorted and possibly buckled.
If the inner shell had " buckled" during tne lead pour, however slignt or strain limited, in the opinion of the NRC staff, the casks automatically would not be in conformance witn Regulatory Guide 7.6.
This Regulatory Guide was first published in 1977--some five years after the NAC-1/NFS-4 casks were designed and licensed.
Between April 6 and December 12 1979, a QA review was completed for the NAC-1.
Cavity ovality and straightness in five of the casks were measured, and two casks, the NAC-1 "A"
and "C"
casks had cavity bows outside the drawing tolerances.
On December 12, the NRC partially rescinded the order of April 6, 1979 and three of the casks that substantially conformed to the C of C were allowed to return to service witn restrictions. Furtner evaluation was required on the basis that:
1.
"the bowing observed in Casks Serial Nos. NAC-1A and NAC-1C both of whicn are of the same design as tne other casks, may be an indication that the casks of this design are susceptible to buckling of the inner shell in unrestricted use;"
2.
"the cask design has not been analyzed to show that tne inner shell
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would not buckle in unrestricted operation;" and 3.
"it cannot be concluded, without such an analysis, that the cask would meet 10 CFR 71 under Normal and Hypotnetical Accident Conditions."
In essence, the original snow-cause order and partial rescision of that order triggered a requirement for a re-analysis of the structural capability of the casks' design, using Regulatory Guides which were promulgated five years after the initial cask design was completed and licensed.
Structural analysis following Regulatory Guides 7.6 and 7.8, leads to a very
'l complicated series of analyses that require heavy reliance on computer programs and data processing.
NRC staff made it clear in informal meetings that a sophisticated and thorougn analysis following the Regulatory Guides was required.
On May 21,1980, a Structural Analysis Report was submitted for tne NAC-1 design, based on a sophisticated dynamic analysis ccde.
At a meeting on July 17, 1980 with the NRC, some twenty-seven comments on the initial structural report were presented.
At that point in time the NRC was trying to contract with Southwest Research Institute to review the report.
On November 24, 1980, the NRC forwarded twenty-two requests for additional information on the May 21, 1980 report.
Apparently, the NRC was unable to t
contract witn SRI for review of the report and did not have the capability internally to review the metnods and results.
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In response to the twenty-two requests for additional information and to facilitate review of the structural analyses, NAC submitted Revision 2.to the Structural Report on June 5, 1981.
Specific replies to the twenty-two requests for information were noted.
Revision 2 to the Structural Report was based on static analyses and in essence, was a complete revision of tne structural analysis report, using many manual methods for analysis that could be checked with hand-held calculators.
The NRC was successful in contracting with Lawrence Livermore National Laboratory (LLNL) to review part of the NAC analysis.
In October 1981, LLNL issued a report entitled " Residual Stress Assessment and Impact Analysis for the NAC-1 Spent Fuel Shipping Casks".
LLNL's repcrt concluded that "the current i
design of NAC-1 spent fuel shipping casks meets tne criteria in Regulatory Guide 7.6 for both accident conditions and normal conditions of transport."
On Decemoer 29, 1981, the NPC forwarded twenty-two requests for added i
information on Revision 2 of the Structural Analysis Report.
On July 30, 1982, NAC responded to the twenty-two requests for added information in the form of changed pages and summarized specific responses.
On NovemDer 20, 1982, a meeting was held with the NRC to discuss a draft of twenty-one comments and requests for additional information on the July 30, 1982 submittal. During this meeting, the staff stated that they were concerned that the casks could nave buckled during the lead pouring operation and/or subsequently during cooling following the lead pour.
They were concerned that the inner shell of tne cask might nave been damaged by plastic deformation and that -they tnerefore could not be confident that the cask would meet the requirements of 10 CFR 71.
The inner shell is a 13-1/2 inch diameter pipe that must be plastically deformed to both fabricate it and to straighten it j
sufficiently (or basket insertion.
Fabrication deformations are normal and of significantly greater magnitude than any possible deformation resulting from the lead pouring operations.
NAC's calculations using some very pessiniistic and conservative assumptions about the lead pouring and cool-down operations show that the inner cavity of the casks cooled slower than the outer shell and that the temperature difference between the inner and outer shell will cause a compressive stress in the inner shell that is, at worst, at the yield strength of the material.
The NRC's position is that calculations performed for them by Oak Ridge National Laboratory, using even more pessimistic conditions than used by NAC, show that the inner shell may have yielded in compression during cooling after the lead pour and they believe that there is insufficient analytical justi fication to accept a small margin of safety against the buckling criteria established by Regulatory Guide 7.6.
NRC's calculations actually result in a compressive strain in the inner snell of less than 0.150 inch.
This is a strain of only 0.00086 inches /incn of the 175 inch long inner shell.
Only 0.020 inch of tne 0.150 incn total strain is " inelastic", which, when uniformly distributed over the length of tn'e shell, represents a unit strain of 0.00011 inches / inch.
This strain is limited since tne two shells remain the same length because of their mutual attachment to the upper flange and lower casting.
This amount of strain is nardly a " buckling" of the inner shell except under tne strictest literal interpretation of Regulatory Guide 7.6.
As a result of the uncertainties and discrepancies between the analyses performed by NAC, the NRC and their consultants, particularly Oak Ridge National Laboratory and Lawrence Livermore National Laboratories, the NRC suggested that NAC provide a cask for destructive tests.
The intent was to drop the cask from tnirty-feet onto an essentially unyielding surface at a temperature of -20*F at an angle to be determined by NRC. Criteria for whether or not the cask would be acceptable was to be leakage after the test.
At a meeting January 12, 1983, the NRC declined NdC's agreement to provide a cask for testing on the basis that they had no funds for sucn tests.
In response to Duke Power Company and NAC's suggestions that we might wish to test the cask, the NRC personnel expressed mixed emotions about tne acceptance criteria to be applied subsequent to the testing.
Structurally, tne NRC felt that a leak test (interpreted as a hydrostatic test) could be used to show acceptability subsequent to the drop test series.
However, another member of the staff expressed concern that a nydrostatic test was insuf ficient'ly definitive and that some other leak test would probably be requi red.
The suggestion was made that NAC formulate such a leak test and submit it to NRC for
, their evaluation.
If the NAC suggested test was satisfactory, then an overall test of the cask might be condoned.
Again, the absence of realistic criteria leaves tne designer with open ended risk.
At this time, March 4,1983, the Certificate of Compliance for the NAC-1/NFS-4 cask design is under Timely Renewal and we have had only informal discussions with the NRC subsequent to January 12, 1983.
These informal discussions indicate that the NRC is considering maintaining Timely Renewal of the cask design for the remainder of 1983, issuing a geries of questions and requests for additional information, and if satisfactory responses are not received by the end of 1983, the cask Certificate of Compliance will expire.
Discussion All of tne comercially operating light water reactor (LWR) spent fuel shipping casks licensed in the U.S. were licensed before the publication of Regulatory Guides 7.6 and 7.8.
A cask for transport of Fast Flux Text. Facility (FFTF) spent fuel has been licensed using the Regulatory Guides.
The only existing cask hardware that has been re-analyzed using the Regulatory Guides is the NAC-1/NFS-4 cask design.
As noted in previous paragraphs, the re-analysis of the NAC-1/NFS-4 design has been time consuming,
- tedious, expensive and unsuccessful.
The lack of success, we believe, is principally because of an attempt to 1mpose new, and in our opinion, impractical, criteria on an existing cask originally designed to meet 10 CFR 71 requirements.
l It is our opinion that the Regulatory Guides place an excessively conservative interpretation on the requirements of Safety Series 6 and 10 CFR 71.
Safety Series 6 and 10 CFR 71 require retention of shielding and leak tightness throughout normal and accident conditions.
The Regulatory Guides interpret this as a requirement that the containment of the cask not be even slightly deformed in compression at any time before or during the accident series.
This is a near impossible condition to prove by analysis for the myriad of conditions imposed by Regulatory Guide 7.8.
Using the current literal interpretation and application of Regulatory Guide 7.6, a monolitnic cask with even a 12 inch thick shell would not be licensed _ _
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4 because the 40 inch drop test on tne 6 inch bar would assuredly cause deformation of the single containment of the cask even though tne containment would not be breached.
A finite number of cask and model tests have been performed and correlated with structural analyses.
The correlations have invariably shown that analytical methods conservatively predict tne test results.
The test results have shown tnat spent fuel shipping casks can tolerate significantly more abuse tnan is conceivable under the worst accident conditions before they would lose' their integrity and present a significant nazard to the health and safety of the public.
Some tests have snown that even large deformations of containment shells do not result in breaching containm,ent.
From the experience we have had with the NAC-1/NFS-4 re-analysis, we can only conclude that Regulatory Guide 7.6 places an unrealistic requirement on the containment of a spent fuel shipping cask in that it is interpreted to require analytical proof that no compressive deformation occurs througnout,the accident series. The requirement is unrealistic because compressive deformations must be extremely severe to breach containment.
Maintaining these criteria places a severe burden on the cask designer without commensurate benefit to the health and safety of tne public.
The test as initially proposed by the NRC would, we believe, have shown that the NAC-1 casks meet in letter and intent, the regulatory requirements.
- However, more desirable and of generic value would be a comprehensive test series with an NAC-1 cask that would provide the basis for modifying the Regulatory Guides such that more realistic criteria are use in the evaluation of acceptability of a cask design.
We believe tnat a modification to the Guides that would provide for strain limited deformations under accident conditions witn consideration given to the ductility of the shell materials would bring the Guides in line with the intent of the Regulations without compromising the safety required for shipments of spent nuclear fuel.
Other practical modifications to the Guides could also result from a well planned and instrumented test series such that the amount of time and effort required to design and license casks would be reduced.
Sucn modifications, justified by analytical as well as test results, would not detract from the safety of the package but would significantly improve tne cost effectiveness of the design and licensing processes.
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7 iMMED? ATE ACT10>! LETwm June 3, 1981 Docket No. 50-409 Dairyland Power Ccoperative ATTN: Mr. F. W. Linder General Manager i
2615 East Avenue - South Lacrosse, Wisconsin 54601 Gentlemen:
This refers to the telephone conversation on June 3, 1981, in which Mr. R. E. Shimshak of your staff and Mr. L. R. Greger of this office discussed the high levels of removable contamination measured on the surface of the NAC-1D cask on its arrival at LACBWR on June 2, 1981.
It is our understanding that you will not release the cask feos your facility until measures have been taken, acceptable to representatives of this office, that will assure contamination levels during transit do not exceed DOT limits.
If you believe our understanding of this matter to be incorrect, please contact this office by telephone in=adiately.
Sincerely, James G. Kappler Director cc Mr. R. E. Shimshak, Plant Superintendent DMB/ Document Control Desk (RIDS)
Resident Inspector, RIII Mr. J. J. Duffy, Chief Boiler Inspector Mr. Stanley York, Chairman 4
j Public Service Comnission 35?
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ff-IMMEDIATE ACTION LE7..
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