ML20151X782
| ML20151X782 | |
| Person / Time | |
|---|---|
| Issue date: | 09/03/1998 |
| From: | Reid D NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Kane W NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 9809170249 | |
| Download: ML20151X782 (150) | |
Text
l I y UNITED STATES
- { NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006 4001
- September 3, 1998 i
MEMORANDUM TO: William F. Kane, Director Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards f FROM:
n /
Dennis G. Reid, Project Manager ,- , i Synt Fuel Licensing Section j g ./ '
Spent Fuel Project Office 4 Office of Nuclear Material Safety and Safeguards
SUBJECT:
SUMMARY
OF JULY 1-2,1998, MEETING WITH NUCLEAR ENERGY INSTITUTE AND INDUSTRY On Ju!y 1- 2,1998, the Nuclear Energy Institute (NEI) sponsored a meeting between representatives of industry and the Nuclear Regulatory Commission's (NRC's) Spent Fuel Project Office (SFPO). The meeting provided a forum for open exchange of technical issues common to the storage and transport of spent fuel and r'5wed attendees to draw consensus from generic problem areas in an effort to expedite car Nrtification. The meeting, which was noticed and open to the public, was held at the Willard Inter-Continer: fat in Washington, DC, with approximately 120 individuals in attendance representing utilitia, contractors, vendors, and fabricators. This report summarizes the meeting.
The meeting was executed in a workshop format, wherein, open and candid discussions were encouraged. Opening remarks were presentsd by Ralph Beedle, Senior Vice President, Nuclear Generation, and SFPO management (William Kane, Charles Haughney, and Susan Shankman). Topics discussed included staff guidance to applicants, ASME Code issues, 1 accident analyses, thermal issues, cask stability, fuel cladding integrity, shielding, and selective loading.
During the meeting, SFPO management took the opportunity to reiterate the following SFPO policies regarding licensing reviews:
Pre-application meetings with the NRC are encouraged and will be billed to the applicants.
Partial or incomplete applications will be retumed to the applicant.
Applicants should state if the standard review plan was followed in developing their application and safety analysis report. 0}_
NRC will dedicate a specific review team for each major application.
NRC's goal is Dg request for additional information (RAI), however, staff will consider one or two RAls acceptable.
- g1 9809170249 980903 [)1 I!
PDR REVGP ERON M C 03u1S5 7 g Ad am, W VY Y
l W. Kane Industry representatives were informed that, because of continued fabrication problems with spent fuel cask vendors, the staff is considering full implementation of ASME Code requirements, complete with the services of an authorized nuclear inspector and code stamping.
, However, this plan was met with strong opposition from industry representatives. citing l increased fabrication costs, longer schedules, duplication in effort, and added restrictions on the l selection of fabricators and material suppliers. Industry feels that the methods and procedures 1- . established by their quality assurance programs adequately address the fabrication problems encountered in the past. However, the industry acknowledges that it needs vigorous l implementation of codes, standards, and quality records and procedures to correct the situation. .;
Specifically, the industry proposed to institute the following improvements:
l .
Thorough' review of fabrication specification by utility and cask vendor !
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Fabrication oversight by utility and cask vendor l .. Review / approval of weld procedures by cask vendor and utility Third party review of nondestructive examination records by certified inspectors l
In general, the presentations were well received. Industry representatives stated that they were pleased with the staff's approaches on various topics and welcomed the opportunity to l communicate directly with the staff to share concems and to obtain a better understanding of the staff's expectations for an expeditious cask certification process. No regulatory decisions I were requested or made. No proprietary information was disseminated. l It should be noted that the NEl issue papers responded to each topic discussed at the workshop i with the exception of the ASME Code issue. NEl plans to discuss the ASME Code issue in l more detail with the industry and submit an issue paper at a later date.
Attachments: 1. Attendance List
- 2. NEl Slides and Issue Papers
- 3. NRC Slides Distribution:
- NRC File Center PUBLIC NMSS R/F SFPO R/F EEaston CHaughney WHodges LKokajko FSturz PEng WReamer, OGC - SGagner, OPA SShankman *see previous concurrence G2GRWRTGSUM.NEl OFC SFPO SFPO / SFPO /
NAME DReid.dd _
VTharpe EJLeeds DATE 0$fM /98 07/21/98 k [/98 C = COVER E = COVER & ENCLOSURE N = NO COPY OFFICIAL RECORD COPY ,
9/3/98dd I l
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DRYSTORAGEROUNDTABLEWORKSHOP July 1 - 2,1998 m Willard Inter Coniinental Washington m Washington, DC FinalParticipants List Edward C. Abbott Treva Ballar,1 Frcaident Administrative Annistant ARZ,Inc. NuclearEnergyInstitute 4451 Brookfield Corporate Drive, Suite 101 Suite 400 Chantilly, VA 20151 1776 i Street, N.W.
Phone: (703)G317401 Washmgton,DC 20006-3708 :
Fax: (703)6315282 - Phone: (202) 739-8106 Email: abz@ patriot. net Jim Becka Kenneth A. Ainger Senior Engineer, Dry Fuel Storage Group Decommissioning, Licenams Manager Wisconsin Electric Power Company Commonwealth Ediarm Company Point Beach Plant lill Opus Place, Suite 111 6610 Nuclear Road Downers Grove,IL 60516 Two Rivers, WI 54241 Phone: (G30) 6G8 5217 Phone: (920) 755-6500 Fax: (630) 663-5400 Fax: (920) 755-6032 Email: aingeka@wmail. ceco.com Email: np4012@wcpco.com
- Joseph F. Andrescavage Ralph E. Beedle S2tc Manager, Spent Fuel Senior Vice President and Chief GPU Nucisar,Inc. Nuclear Officer, Nuclear Generation Oyster Creek Nuclear Generating Station Nuclear Energy Institute P.O. Bnx 388 Suite 400 Forked River, NJ 08731 17761 Street, N.W.
Phone: (609) 9714862 Washington, DC 20006-3708 Fax: (609) 9714449 Phone: (202) 739-8088 Email: jr.ndrucav=g Wupu.mm Fax: (202) 785-4019 Email: rbanei.arg Abdulit. Ba Project Managt Jry Spent Fucl Storage Project Robert M. Bernero GPU Nuclear, Inc. Consultant Oyster Creek Nuclear Gewating Station Robert M. Dernero bute 9 huth, P.O. br 388 201 Summit HallRoad Forked River, NJ 08731 Gaithersburg, MD 208771825 Phone: (609) 9712152 Phone: (301) 926 3844 Fax: (609) 971-1005 l*'ax: (30J) 9261368 Email: abaigegpu.com Email: rmbernerussol.wm Marissa Bailey David Bland Project Manager Project Manager U.S. Nuclear Regulatory Commission Southern Nuclear Operating Company / ARC MS 0-6022 P.O. Box 1295 11555 bekville. Pike Birmingham, AL 35242 0355 Rockville, MD 20852 Phone: (205) 992-6697 Phone: (301) 415 8531 Fax: (205) 992 0362 Fax: (301) 415 8555 Email: david.w.blandbuc.com Email: mgh&nre. gov ATTACHMENT 1 I
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d Eric A. Blocher Julius Bryant i Manager Licenuing Engineer Parsons Power Duke Power Company 2675 Morgantown Road McGuire Nuclear Station i Heading, PA 19607 12700 Hagers Ferry Road Phone: (clo)855 2071 Huntersville, NC 28078 Fax: (610) 856 2509 Phone: (704) 875-4162
, Email: oric.a blocher# parsons.com Fax: (704) 875-4279 Email: jwbryant@ duke-energy.com Jayant Bondre j Manager, Engineering D!-19% James Clark Trannauclear West,Inc. ISTSIProjectManager j 39300 Civic Center Drive, Suite 280 Southern California Edison Company i Fremont, CA 94538 San Onofre Nuclear Generating Station Phone: (510) 744-6043 P.O. Bat 128 i
Fhx: (510) 744-6002 San Clemente, CA 92677 Email: jayant.bundreetranenuclear.com Phone: (949) 868 6240 Fax: (949) 368 1490 i Rita C. Bowser Email: clarkjr@eongs.see.com
! Deputy Director, Spent Nuclear Fuel Programs j Westinghouse Electric Corporation Sidney L. Crawford Suite 1200 6 Spinning Wheel Court 600 New Hampshire Avenue N.W. Germantown, MD 20874 Washington, DC 20037 Phone: (301) 515 6398 i Phone: (202) 945-6458
- - Fox
- (202) 945 G404 William J. DeCaa-a 5
Email: rebowser9ix.netcom.com Project Manager, High level Waste Framatome Technologies. Inc.
L. Dave Brevig P.O. Box 10935 Manager, SONOS/Decomminioning Project Lynchburg, VA 21506-0935
} Southern California Edison Company Phone: (801) 882 2640 1 5000 Pacific Coast Highway Fax: (804) 832 2932 i San Clemente, CA 92672 Email: bdecooman@framatech.com J
Phone: (714) 3G8 7820 Fax: (714) 3G8 7844 Max M. DeLong Email: brevigid@eongs.ece.com Executive Engineer Northern States Power Company Gregory Broadbent 414 Nicollet Mall (Res Sq 7)
Senior Engineer Minneapolis, MN 55401 1998 Entergy Operations, Inc. Phone: (G12) 380 5850 ,
P.O. Row 756 Fax: (612)330 5958
- P<rt Gibsoa, MS 39150 Email: max.m.delong@nspco.com Phone: (601) 437-6224 Fax: (601) 487 2146 Jerry Delezenski Email: gbroadh@cntergy.com Superintendent. Quality and Compliance Licenaing/Admimatration John Broschak Sacramento Municipal Utihty District
. Dry Fuel Program Manager 14440 Twin Cities Road Consumers Energy Herald, CA 95688 Palisades Nuclear Plant Phone: (016) 452 3211 X4914 27780 Blue Star Highway Fax: (209) 748 2244 Covert, MI 49043 Email: jdeleze@smud.org Phone: (616) 764 2650 Fax: (G16) 7G4 3200 Email: jpbrosch@rmsenergy.com
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Paul G. Delosier David J. Foss President Project Engineer Debaier Consultina,Inc. PECO Nuclear 3202 Gemstone Court PBAPS 1848 Lay Road Oakton, VA 22124-2708 PS 2 2 Phone: (703) 5911684 . Delta, PA 17814 Fax: G03) 5912292 Phone: G17) 456 4311 l
Email: nuccxprt@ laser. net Fax: 017)456 3433 Email: Afoes@peco-oncrgy.com James W. Doman Manaser, NRC IJennaing Robert Fraser WASTREN,Inc. Engineering Director
! 22 Executive Park Court Maine Yankoc/Entergy Germantown, MD 20874 P.O. Box 408 I
Phone: (801) 540-0022 Wiscasset,ME 04578 Fax: (301) 540 0088 Phone: (207) 882 4553 Email: wantrenemnl.com Fax: (207) 882 5859 Email: fraserr@myape.com John L. Donnell ,
Project Director Ted Gado '
Private FuelStorage Director, Nuclear Material Storage Programs 7677 East Berry Avenue Foster Wheeler Englewood, CO 801112137 8 Peach Tree Hill Road Phone: (303) 7417000 Livingston, NJ 07039
- Fax
- (303) 741 7806 Phone: (973) 597 7122 Email: John.donnell@stoneweb.com Fax: (973) 597 7589 Email: tsade@fwene.mm John Duffy Sales J.Darren Gale Ranor, incorporated Manager, Corpurmte Quality maal P.O. Box 458 Employee Development l Westminster MA 01473 Framatome Technologies,Inc.
Phone: (978) 874 0591 3315 Old Forest Road l
Fax: (978) 874 2748 Lynchburg, VA 24506-0935 Email: du#y@ranor.com Phone: (804) 832 2804 l
Fax: (804) 832 2475 Robert G. Ebl=Jr. PE Email: jgaleaframatech.com Engineering Supervisor i Duke Power Company William D. Gallo 2650 Park Tower Drive Vice Preddent Vienna, VA 22180 Trananuelear,Inc.
Phone: 003) 204 8G57 Four Skyhne Drive Fax: (708) 204 8530 Hewthorne, NY 10532 2120 Email: robert.ehle#rw. doe. gov Phone: (914) 347 2345 Far: (914) 847 2346 Patricia i,. Eng Email: trananclr@aol.com Chief, Transportation and Storage Inspection U.S. Nuclear Regulatory Commiaaion Paul H. Genom Spent FuelProject Office Project Manager, Plant Support
. M/S 0 6 F 1 A Nucicar Energy Inst 2tute
+
Wanhington, DC 20555-0001 Suite 400
- Phone
- (301) 415-8577 1776 ] Street, N.W.
Fax: (301) 415 8555 Washinsten, DC 20006 3708 Email: pleSnre. gov Phone: (202) 739-8034 Fax: (202) 785 1898
John P. Gerety Brian Hallett Group Manager. Nuclear Engineering Serviose . Manager, Geological Repository Dispnaal katon Edison Company Dettia Atomic Powerlaboratory Pilgna Nuclear Power Station P.O. Box 79 600 Rocky HillRoad,MS 88 West Mifnin. PA 15122 0079 Plymouth, MA 02860 5599 Phone: (412)476 5849 y I
Phone: (50H)830 7800 Fax: (412) 578 502G t
! Fax: (508)830-8855 Alan S. Hanson Daniel Oildow President and Chief Executive OfBeer IBFSI Paject Manager Transnuclear,Inc.
Portland GeneralElectric Company Four Skyline Drive
'lYojan Nuclear Plant Hawthorne, hT 10582 71760 Columbia River Highway Phone: (914) 847 2345 ' i Esinier, OR 97048 Fax: (914) 347 2346 Faone: (503) 55G-7527 Email: transactremol.com Fax: (503) 556 7840 g Email: dan _sildowepen.com Charles J. Haughney '
Deputy Director, Spent Fuel Paject Office Alan J. Gould U.S. Nuclear Regulatory Commission Senior St4K Speciahst Mail Stop 06F18 Florida Power & Tight Company Washington. DC 20555 0001 P.O. Box 14000 Phone: (301) 415-8560 Juno Beach, FL 35408 Fax: (301) 415 8555 Phone: (561)694 4199 Email: ejhearc. gov Fax: (561) G04-3072 ,
Email: alsould@email.fpl.com Lynnette Hendricks t Director, Plant Support l Philip J. Grant Nuclear EnergyInstitute Chief Operating OfBeer i.
17761 Street, N.W., Suite 400 WASTREN,Inc. Washington, DC 20006 3708 i
2 Waverly Drive Phone: (202)739-8109 l Hummelatown, PA 17036 l Phone: (717) 566 6780 Linda Hertzog Fax: (717) 466-6780 Special Evente Manager Nuclear Energy Institute
- Don J. Green 1776 I Strest, N.W., Suite 400 Chief, Mechanical / Nuclear / Engineering Mechanics Washington, DC 20006 Tennes ee valley Authority Phone: (202) 73S8026 1101 Mariu t. Street,1E 4J Faa: (202) 739-8171 '.
Chattanoon,TN 37402 Email: lh@nei.org
- Phone: (423) 7518468 Fax: (423) 751 '.* John F. Hilbish Email: digreen$tva. gov Manager Parsons Energy and Chemicals Croup Inc.
Kimberly Ann Gruss 2675 Morgantnwn had Materiale Engineer ' Reading, PA 19607 U.S. Nuclear Regulatory Commission Phone: (610) 855 2547 M/S 06 F16 Fax: (610) 855 2509 Washington, DC 20555 Email: johnJ_hilbish@ parsons.com Phone: (301) 415-8586 Fax: '301) 415 8555 l
Email: kaglanrc. gov
L l
l
- j. Elaine Hirue Roger Hudak j Managing Editar Project Manager
. McGraw Hill Precision Components Corporation .
Suite 1100 500 Linonin Street 1200 G Street, N.W. York, PA 17405 Wachington.DC $606G Phone: (717) 8481126 i Phone: (202) 883-2183 Fax: (717) 845-4533 Fax: (202)883 2125 Email: rhudahepec-yark.com J l Email: ehiruemh.com Donald L. Hutson James S. Hobbs Senior Project Manager, Nuclear Fuel Marketing Manager Tennessee Valley Authority BNFLFuelSolutions Corporation 1101 Market Street. DR 6A C Suite 1060 Chattanonga, TN 37402 2801 90017th Street Phone: (428)151-4759 Washington, DC 20006 Fax: (423) 751 4959 l Phone: (202)785 2635 Emmil: d1hutaonetva. gov 4 Fax: (202) 185-4087 Charles G. interrante Steve Nognett Senior McLallurgieml Engineer Nuclear Engmur U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission USNRC/SFPO/SFTR l MailStop: 06022 MailStep: 06022 i Washington, DC 20555 0001 Wuhington. DC 20555-0001 '
l Phone: (301)415 8537 Phune: (301)415 8967 ,
l Fax: (301) 415-8565 Fax:'(301) 415-8555 i Email: ash 9nrc. gov Email: esiente. gov Bruce W. Holmgren Andrea Jeanetta Engineering Manger PolicyAnalyst Duke Engineering and Services Washington Nuclear Corporation 580 Main Street 4600 North Park Avenue, Suite 111 Bolton, MA 01740 Chevy Chase, MD 20815 Phone: (978) 568 2704 Phone: (301) 652-9500 Fax: (978) 568 8735 Fax: (301) 6541200 l Email: bwholmgr@dukungineering.com F. mail: njanyncra.mm l William C. Hopkins Roger L. Johnson Principal Engmeer ti Licensing Supervisor
- Bechtel Power Corporation Pacine Cas and Electric Company l 9801 Washingtonian Boulevard Mail Code N9B l Gaithersburg, MD 20878 5356 P.O. Box 770000 l Phone
- (801) 417 8341 San Francisco, CA 94177 l Fax: (301) 869 7084 Phone: (415) S73-1784
- whopkina$bechtel.com Fax: (415) 978 0074 l
Email: rlja@ pre.com Chrin6en t. Howard Senior Engineer Colin J. Jones Southern Nuclear Operating Company U.S. Market Research Associate
- P.O. Box 1295 RNFL. inc.
Birmingham, AL 35201 90017th Street, N.W., Suite 1050 .
- Phone
- (205) 992-5564 Washington, DC 20006-2501
! Fax: (205) 992 0362 Phone: (202) 785 2635
! Email: christen.l.howardenne.com Fax: (202) 785-4037 l Email: ejones@bnfline.com i
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t Noran hish Stevon P. Erah Vice President Director, Spent Nuclear Fuel Management Nuoon Systema,1nc. Nuclear Energy Institute Two Penn Plaza, Suite 1500 1776 I Street, N.W., Sub 400
~ New York, NY 10121 Washington, DC 20006 Phone: (212) 292-50188 Fhune: (202) 't39-5116 l-Pnx: (212) 292-5089 Fax: (202) 293-3451 Email: nuconeworldnet.att. net Email: sphenei.org I
l Masahiko Kakand Ray W.Lanabert Director Consultant to EPRI Japan Electric PowerInfarmation Center EPRI 1120 Connecticut Avenue, N.W., Suite 1070 3412 Hillview Avenue Washington, DC 2003G Palo Alto, CA 94804 Phone: (202) 955-6610 Phone: (650) 855 2788 Fox: (202) 955-5612 Fax: (650) 855 1026 l l- Email: gendentiepic.com Email: ralambe:Gepri.com v l Williaan F. Esse Eugsac A.Lanning l Director. Spent Fuel Prohet Ofhee Nuclear Fuels Supervisor
! U.S. Nuclear haulatory Commission Nebraska Public Power District u MailStop: 06F18 Cooper Nuclear Station
! Washington, DC 20555 P.O. Box 98 Phone: (301) 415 8500 13rownville, NE 68321 Fax: (801) 415 8555 Phone: (402) 825 5287 l Email: wik8 arc. gov Fax: (402) 825 5102 l Email: ealannenppleon i Jon Kapits 4
! Project Manager . David L. Larkin t Nurdaern States Power Company Nuclear FuelProgram Menaser
- Praire Island Nucicar Power Plant Washington Public Power Supply Syatem
! 1717 Wakonsde Drive, East MailDrop PE 23 l Welch, MN 55089 P.O. h x DGA l Phone: (612) 886 6758 X4619 Richland, WA 99352 0968 i Fax: (oit) 330-0247 Phonc: (509) 377 4201 l Email: jon.kapita$nepoo.com Fax: (509) 377 4099 l Email: dilarkin$wnp2.com j Ray Kollar i
Dry Fuel Storage Project Manager Willington J. Lee .c ;
Entergy Operationa. Inc. % President and Chief hgineer ; i l 1448 SR333 NAC internati<mni i
! Rumellville, AR 72802 655 Engineering Drive l Phone: (501) 858-1688 Norcross, GA 80092 i Fax: (501) 858 7909 Phone: (770)447 1144 i Email: rkellar9entergy.com Fex: (770) 447 1797 Email: bill _leeAnacintl.com John J. Kosiol Manager, Government Programs Henry Lee ABB Combustion Engineering Nuclear Power Senior Structural Engineer 2000 Day Hill Road U.S. Nuclear bgulatory Commission 4
Windeor, CT 06095-0500 M/S 06G22 1* '
, Phone: (860) 285-5598 Washington, DC 20555
- Fax: (A60) 285 9368 Phone: (301)415 A522 l Emmil: JohnJ.kosio)@ussev. mail.abb.com Fax: (801) 415-8555 4 Email: hwl@nte. gov 4 I '
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Eric Leeds David E. McElwee Section Chief, Spent Fuel Licensing Section Limna Enginur U.S. Nuclear Regulatory Comminion Vermont Yankee Nuclear Power Corporation Mail hp 06G22 RD 5, P.O. Box 169 Washington, DC 20555 185 Old Ferry Hood ..
Phone: (301)415-8540 Brattleboro, VT 05301-0128 '
Fax: (301) (15-8555 Phrmo: (802) 25A 4112 Email: ejl8nrc. gov Fax: (802)258-2102 Email: david.mcclwco@vynpc.com Arpad L. Lengyel Advinnry F,ngineer/Scisativt Timothy McGinty lockheed Martin Idaho Technologien Company Project Engineer P.O. Tiox 1625 U.S. Nuclear Regulatory Comminalan MS 3135 Mail Stap: 06022 Idaho Falla.ID 83415-3135 Washington, DC 20555 Phone: (208)526-9683 Phone: (301) 415 8580 Fax: (208) 526 5337 Fax: (301) 415 8555 '
Email: arp@inel. gov Email: tjmi8are. gov
- Albert J. Machiels Hank McGuire Program Manager Vice President, DOE Programs EPRI Foster Wheeler Environmental Corporation P.O. Box 10412 3200 George Washington Way, Suite G Palo Alto, CA 94304 Richland, WA 99352 Phone: (650)855 2054 Ptone: (509) 372-5815 Fax: (650) 855-7945 Fax: (509) 372 5801 Email: amachiclAcpri.com Email: hmesuiresfwenc.com John Mageski Mike McNamara Director Director of Projects ABB Comhustion Enginnring Nuclear Power Holteeinternational 5000 Executive Parkway, Suite 145 555 IAncoln Drive Waat San Ramon, CA 94583 Marlton, NJ 08053 Phone: (510) 830-8311 Phone: (609) 797 0900x005 Fax: (510) 830 5134 Fax: (609) 787-0909 Email: john.mageski@ussov. mail.nhb.com W. Baird McNaught Ramtin T. Mahini Proicct Manager Manager lockheed Martin Idaho Technologies Company .
EPRI P.O. Box 1625 0 3412 Hillview Avenue Idaho Falls,ID 83415 3114 Palo Alto. CA 94304 Phone: (208) 526 3678 Phone: (650) 855 2543 Fax: (208) 526-4002 Fax: (650) 855-1026 Email: rmahinieepri.com Robert L. Moscardini President William J. McConaghy U.S. Tool & Die, Inc. (UST&D, Inc.)
Vice President, Western U.S. Operations Keystone Commons NACInternational, Suite 430 200 Braddock Avenue 226 Airport Parkway Pittsburgh, PA 15145 ,
San Jose, CA 95110 Phone: (112) 823 3773 6 Phone: (408) 453 8900 Fax: (412) 823-6669 Fax: (408) 453 3950 Email: ustd.agi. net Email: bmcconag9nacintl.com
Tara Neider Ron Parkhill Vice Pruident. Engineering Senior MechanicalEngineer Transnuclear. ine. U.S. Nucicar Regulatory Commission Four Skyline Drive OWF 6G22 Hawthorne, NY 10532 Washmgton, DC 20555 i Phune. (914) 847 2345 Phone: (301)4151376 Fax: (914) 347 2346 Fnr: (301)415 8555 Email: transnc1 resol.com Email: rwp8nre. gov Alan Nelson Bruce H. Patton Senior Project Manager, Plant Support Project Manager. High LevelWaato Nuclear Energyinmeitute Pacific Cas and Electric Company 1776 I Street, N.W., Suke 400 Diablo Canyon Power 1%nt Washington.DC 20006 3706 P.O. Box 56 Phone: (202)139-8110 Avila Beach. CA 93424 Fax: (202) 785-8195 Phone: (805) 545 4809 ,. ;
Email: apnenei.nrg Fax: (805) 545 0992 -
Email: bhpl@pge.com Paul E. Netusil T.L. Projects and Reliability Fuela and Analysis Charles W. Pennington Niagara Mohawk Power Corporation Group Senior Vice President, Nine Mile Point Nuclear Station NACInternational P.O. Box 63. Lake Road 655 Engineering Drive. Suite 200 Lycoming, NY 13093 9965 Narcross, GA 30092 Phone: (315) 349 7935 Phone: (770) 447 1144 Fax: (315) 3491579 Fax: (770)447-1797 Email: cpenning@nacintl.com Richard J. Netsel Project Manager Holger Pfelfer
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Surpnt & Lundy Mannpr. Nuclear Analysis 65 East Monroe Strut,23rd Floor NACInternational Chicago,IL 60603 655 Engineering Drive ,
Phone: (312) 209 71.4 3 Norcross, CA 30092 l Fax: (312) 267 2617 Phone: (770) 4471144 i Email. nchard.j.netzc16sichicago.infonet.com Fax: (770) 447 1797 Email: hpfeifer@nacintl.com Raymond N, Ng -
Principal Engineer Kenneth Phy Bechtel Power Corporation Senior Project Manager !
G025 Guntmm D he (nLis 1) Mr. Yorir Pawer Antharity Frederick, MD 21703 8388 James A. FitzPatrick Nuclear Power Plant Phone: (301) 228-6266 P.O. Box 41 Fax: (301)S46 9993 Lycoming. NY 13093 0041 Email: rnng@bechtcl.com Phone: (315) 849-G9G7 Fax: (315) 349-6148 Matt O' Connor Rnneil. play.kGuyp=. gov President Nucen Systems,Inc. Rodney G. Pickard Suite 1500 Engineer, Nuclear Fuel l
2 Penn Plaza Amerienn Electric Power i !
New York, NY 10121 500 Circle Drive I I
Phone: (212) 292-0088 Buchanan, M149107 Fax: (212)292 5089 Phone: (616) 697 5084 ,
Email: nucon@worldnet.att. net Fax: (616) 697-5574 Email: rodney pickard$aep.com
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Paul Plante Dennis Reid PrincipalEngineer Project Manager ,
Wine Venires Atnmic Pneer Ommany ilS NnninnrhenlatoryCommimion l P.O. Box 408 11555 Rockville Pike Wiscosset.ME 04578 Phone: (207) 883 5806 M/S 06F18 Hoelmlle, MD 20852 f-Fax: (207)m2 uss Phone: (801) 415 8556 l
Email: plante8myape.com Fax: (801) 415-8555
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Email: derenrc. gov l Richard A.PlasseJr.
! Senior Nuclear Lleenang Engmeer John A. RichardsonP.E.
New York Power Authority Project Manager Nuclear Programs anel Projecta
, James A. Fiti.petrick Nuclear Plant Raytheon Engineers & Constructors Inc.
P.O.Ihm 41 CN 5287 Lycoming, NY 18098 004i Prinecton, NJ 08548 5287 Phone: (815) 349 6793 Phone: (609) 720 3213 v Fax: (315) 349-6383 Fax: (609) 720 3592 l Email: plasse.nenypa. gov Email: jarrm8aol.com l Robert D. Quinn M.J. Rose-Lee l l Manager of klegulatory Compliance Nuclear Engmeer '
Westinghouse Electric Corporation U.S. Nuclear Regulatory Commision l l 2242 Camden Avenue, Suite 208 MailStop 06F13 l l San . lose, CA 95124 Washington. DC 20555 !
Phone: (408) 369-6741 Phone: (801) 415-3781 l Fax: (408) 369 6747 Fax: (301) 415-8555 l l Email: rdquinneix.netcom.com Email: mir2enrc. gov ,-
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Geoffrey Quinn Morris nichreim Senior Engineer Energy Consultant ,
Sdeatech, Inc. Cummonwealth Edison Company '
010 Clopper Road 1310 Maple Avenue l- Caithersburg, MD 20878 Evanston,IL 60201 Phone: (301) 258 1869 Phone: (H47) M9 6297 Fax: (301)258 2575 Fax: (817) 869-6297 Email: squinn@eeientech.com Steve Schulin Michael G. Raddats Owner .,e l
Senior Pro)cet Manager IBEX Group :
U.S. Nuclear Regulatory Commission P.O. Box 6674
. Mail Stop OGG22 Rockville,MD 20655 l
Rockville,MD 20852-2738 Phone: (301) 762 6714
- Phone
- (801) 415 8544 Email: eteve.achnhn@nuelear.com l Fax: (301) 415 8555 Email: agrenre. gov Susan Frant Shankman Deputy Director Licensing and Inspection
' Joe Y.R. Raehld U.S. Nuclear Regulatory Comminnion ,
Chairman SFPO, U.S. NRC !
ANATECli onparation MailStop: 06F8 5435 Oberlin Drive . Washington, DC 20555 i
! San Diego, CA 92121 Phone: (301) 415 2287 Phone: (619) 455 6350 Fax: (801) 415 8555 l
- Fax
- (619) 455-1094 Email: afs8 arc. gov !
j Email: joe $anatech.com i
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Lisa Shell Carrick J. Solovey Engineer {
Vice President -
Virginis Power Preemon Components Corporation Nuclear Analysis and Fuel P.O. Box 15101 5000 Dominion Boulevard York, PA 17405 7101 ;
Glen Allen,VA 23060 Phone: (717)8481128x2232 j Phone: (804) 273 8826 Fax: (717) 846 6282 i Fax: (804)273 8543 Email: geolovey@pecyork.com Rmail: lisa shel19espower.com Carl J. Stephenson K. P. Singh Senior Licensing Engmeer President and Cinaf Executive OSicer Rio Technical Servicos a HoltecInternational 190 Red Rock Road 565 Lincoln Drive West Sedona, AZ 86351 Marlton, NJ 08058 Phone: (520)244-5049 Phone: (609) 797 0900 X636 Fax: (520) 284 5049 Fax: (609) 797-0909 Email: riocarl@aol.com Email: kris.Angh6holtee.com David Stoecket Timothy E. SmithEsq. Refueling Engineer President Southern California Edison Company Covernmental Strategies,Inc. San Onofre Nuclear Generating Station 11803 Wayland Street P.O. Box 128 MS DSB v Oakton,VA 22124 San Clemente, CA 92G72 -
Phone: (703) 716-4846 Phuns: (949) 868 6758 Fax: (703) 716-0043 Fax: (949) 368-9007 Email: satrategian@campuserve.com Email: stoeckda@ sone.aoe.com R.Howard Smith R.Jon Stouky h President Quality President NAC International Wga-Tech Services, Inc.
655 Engineering Drive 2804 Woodley Court i Norcross, GA 30092 Jamestown, NC 27282 Phone: (770) 4471144 Phone: (336) 316-0707 ,
Faa: (770) 4471797 Fax: (336) 316-0550 Email: hamithenacintl.com Email: jetnukyseol.onm i Howard L. Sobel Fritz Sturs Vice President, SpecialProjects Section Chiaf, Technical Review Section Edlow international Company U.S.* Nuclear Regulatory Commission 1660 Connecticut Avenue, N.W., Suite 201 fipent Fuel Pmject Ohn Washmgton, DC 20009 0GG22 Phone: (202) 483 4959 Washington, DC 20555 Faa: (202) 488 4840 Plwue. (301) 415-8580 Email: Wobelpenol. cum Fax: (301) 415 8555
- David Sokolsky Email: fcsanre. gov ~
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Senior Licensing Engineer Anthony P. Suda Pacific Gas and Electric Company Principal Engineer 1000 King Saloman Avenue West Valley Nuclear Services,Inc.
Eureka, CA 95503 P.O. Dax 191 Phone: (707) 444 0801 West Valley, NY 141710191 Fax: (707) 444 0786 Phone: (716) 942 2452 Email: ddepp.com Faa: (716) S42 2306 Email: oudat6doo.wv. gov b
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Eileen M. Supko Robert Van Namen Senior Consuhant Manager. Nuclear"uol Management Energy Resources international, Inc. Duke Power Company i 101518th Street N.W.. Suite 650 P.O. Bar 1006, EC08F i Washington, DC 20086 Charlotte, NC 28201 1000 < ;
Phone: (202) 7a5 ansa Phone: (704) 382-4524 l
Fax: (202) 785-8834 Fax: (704) 882 7852 J Email: supko@energyresources.com Email: rvannameduke :nergymm l
- Paul Synnons George E. Vaughan Senior Shielding Ensinoer Director, U.S. Sales l
l ALSTOM Automation Ltd. NACInternational l Cambridge Road 655 Engineering Drive Whetstone,14icester, LE8 6LH Norcrose, CA 30092 England Phone: (770) 4471144 Phane 0111401162750760 Fax: (770) 447 6577 .c Fax: 0114401182n: A237 Email: gvaughardimacintinnm
- l David Tang John A.Vinoent Senior Structural Engineer Senior Engmcer, Nuclear Fuel {
l U.S. Nuclear Regulatory Commuasion CPU Nuclear,Inc. !
Mail Stop: 0-6G22 One Upper Pond Raad l Washington, DC 20555 Parsippany, NJ 070511095 l Phone: (301) 415 8535 Phone: (973) 316 7289 ;
Fax: (301) 415-8555 Fax: (973) 316 7841 - l
- Email: dttenre.com Email: jvinocat$gpu.com i Thomas C. Thompson David Wade ,
Director, Licensing and Competitive Assessment Manager, Shipping Container Analysis NAC International Bettis Atomic Fowerlaboratory 655 Engmcering Drive P.O. Box 79 I Norcross. GA 20092 West MiNlin. PA 15122 0079 l Phone: (770) 4471144x321 Phone: (412) 476 6184 l Fax: (770)1471797 Fax: (412) 476-5026 Email: tthompsu$nucintl.com John Walkin Gary T. Tjerstand PrincipalMechanical Engineer l Director of Licer.aing and Product Development Sacramento Municipal Utility Dutrict Holtecinternational P.O. Box 15880 M/S N801 Y 555 Lincoln Drive West Sacramento. CA 95M521830 Mariton, NJ 08053 Phone: (916) 452-3211x4142 Phone: (609) 797 0900 X611 Fax: (916) 732-5151 Fax: (G09) 797 0909 Email: jwalkinGemud.org Email: gary tjersian68holtec.com Everett M. Washer Len Tremblay Project Manager Senior Enguwer Stone & Webster Enginecrma CorporatJon Duke Engineering and Services 245 Summer Stevet 580 Main Street Boston,IaA 02210 Bolton, MA 01740 Phrma: (978) MiM-23MM Phone: (017) 580 1724 l l Fax: (617) 589-5892 '
Email: everett.wanhesentonewebwm I
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MichaelD. Waters Gary Zimmerman Project Engineer Licensing Engineer U.S. Nuclear Regulatory Commimion Portland GeneralElectric Company OWFN-6 0 22 Trojan Plant 11555Rockvnile1%c 71760 Columbia River Highway Rocknile, MD 20n52-273s Rainier, OR 97048 Phone: (301) 415 3875 Phone: (503) 556 7278 -
Fax: (301) 415-8555 Fax: (503) 556-7002 Email: adwlenre. gov Email: gary zimmerman8pgn.com '
Paul D. Watts Manager. QA and 14 Ranor. Incorporated P.O. Box 458 Wutminster, MA 01478 Phone: (781) 874 0591 i Fax: (7A1) M74 2748 Email: pd wattstranor.com David H.Williamson Senior Engineer Scienoe ApplicationsInternational Caporation -
11251 Roger Bacon Drive Reston, VA 20190 Phone: (703) 818 4659 Fax: (703) 709 1042 Email: david williamson8cpqm.saic.com Carl J.Withee Senior Criticality and Shielding Engineer U.S. Nuclear Regulatory Commission Mail Stop: 0 6 G22 Washington, DC 20555 '
Phone: (301) 415-M584 Fax: (301) 415 8555 Email: ejwenre. gov Brian Wohlers Project Manager Alliant IES Utilities.Inc.
Duane Arnold Energy Center 3277 DAEC Road Palo,IA 52324 Phone: (819) 851 7400
. Fax: (319) 85173G4 Erasil: BrianWohleregalhant. energy.com
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l . DRY STORAGE ROUNDTABLE WORKSHOP j July 1 - 2,1998 m Willardinter ContinentalWashington a Washington,DC 1
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e ATTACHMENT 2
DRY STORAGE ROUNDTABLE WORKSHOP July 1 - 2,1998 a WillordInter ContinentalWashington a Washington,DC Workshop Proceedings
. Final Agenda e Final Participant List
. Industry Conunitment to Quality
. Nuclear Energy Overview-
"NEI Sponsors Landmark Meeting with NRC Staff on Dry Cask Storage Issues"
. Industry Issue Summary
- 1. ASME code issues
- 2. Accident analyses / accident source terms
- 3. Thermalissues
- 4. Cask stability (Further evaluation will be required)
- 5. Cladding integrity (failed fuel and retrievablity)
- 6. GTCC issue was not discussed at this meeting (the issue number assignment will be maintained)
- 7. Selective loadings, e.g., o! der fuel on the outside.
- 8. Shielding (Requires further development) e industry Presentations
- NRC Staff Guidance with Applicants l
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l DRY STORAGEROUNDTABLEWORKSHOP July 1 - 2,1998 m Willard Inter Continental Washington a Washington, DC Final Agenda Wednesday. July 1.1998 8:00 a.m. - 5:00 p.m.
- General Session Ballroom Welcome introductions NEIIndustry opening remarks Ralph Beedle, Senior Vice President, Nuclear Generation NRC openi.. remarks William Kane Director, Spent Fuel Project Office Charles Haughney, Associate Director, Spent Fuel Project Office l Susan Shankman, Deputy Director, Licensing and Inspection ,
i Workshop conduct and format '
Lynnette Hendricks, Director, Plant Support Charles Haughney, Associate Director, Spent Fuel Project Office Alan Nelson, Sr. Project Manager, Plant Support Round Table Toolcs for discussion:
- 1. ASME code issues NRC: Ron Parkhill Industry: Tara Neider, Transnuclear
- 2. Accident analyses / accident source terms l NRC: Charles Haughney i Industry: Bill Lee, NAC International
- 3. Thermalissues ' - -
Industry: Bob Quinn, Westinghouse l a. high burnup fuel cladding issue 4 NRC: ChuckInterrante
- b. internalconvection i c. helium conduction
- d. Ession gas release NRC: Steve Hogsett
Wednesday. July 1.1998 s Round Table Topics (cont.)
- 4. Cask stability NRC: David Tang Industry: K.P. Singh, Holtec International
- a. low. velocity impact test
- b. high seismic Gs
- c. pad analysis
- 5. Cladding integrity (failed fuel and retrievability)
NRC: Marissa Bailey NRC: Mike Raddatz Industry: Ray Lambert, EPRI
- 6. GTCC issue will not be discussed at this meeting (the issue number assignment will be maintained)
- 7. Selective loadings, e.g., older fuel on the outside.
NRC: Tim McGinty Industry: John Broschak, Consumers Energy
- 8. Shielding NRC: Mike Waters NRC: CarlWithee Industry: Ted Gado, Foster Wheeler
- a. streamlining
- b. skyshine
- c. array calculations
- d. neutron measurement Time for bree.ks and lunch will be made at appropriate times.
At the conclusion of the day sufficient time will be provided for public comment.
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'o Thursday. July 2.1998 l l
8:00 a.m. - 10:30 a.m.
Break Out Sessions 2.d Floor Meeting Rooms If additional discussions are needed due to insufficient time on day one,
- breakout sessions may be delayed.
4 Vendor Owners Groups will review application status and impact based on round )
table discussions. -
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. BNFL l
. TN/TNW l . Holtec
- . Foster Wheeler i
j 10:30 a.m. - 12:00 p.m.
General Session Ballroom l Breakout session feedback 4 Workshop summary- where do we go from here
- j. Meeting adjourned -
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Dry Cask Storage Industry Commitment to Quality July 14,1998 The nuclear energy industry is committed to dry cask storage quality as a mechanism for ensuring public health and safety. Implementation of design and fabrication standards will be ensured through, vigorous quality oversight and procedures. This commitment assures that quality will be incorporated into work practices on a daily basis. Illustrative of this commitment will be a coordinated industry oversight effort including:
licensee, owners groups, designers, fabricators and the newly formed Dry Cask Quality Working Group under the NUPIC* charter. The review will include dry cask storage design, materials, and inspection. This
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commitment includes the communications and sharing ofindustry experiences pertaining to dry cask storage management, licensing, l procurement, fabrication, implementation, quality assurance, and regulatory compliance. !
' Nuclear Utility Procurement Issues Committee
v Industry Drv Storare Round Table Sn===rv Willard Hotel, Washington DC July 1 & 2,1998
- 1. ASME code issues (Further evaluation will be required)
- 2. Accident analyses / accident source terms
- 3. Thermalissues
- 4. Cask stability (Further evaluation will be required)
- 5. Cladding integrity (failed fuel and retrievablity)
- 6. GTCC issue was not discussed at this meeting (The issue number assignment will be maintained)
- 7. Selective loadings, e.g., older fuel on the outside.
- 8. Shielding (Requires further development)
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l NEI Dry Storage Roundtable Workahop Willard Inter ContinentalHotel Washington, DC July 12,1998 Accident Analysis Source Terms and Release Rates 4
Issue Description Accident evaluations shall be performed for both 10CFR71 transportation packages and 10CFR72 storage systems in accordance with the applicabic regulations, and Standard Review Plans. This Industry Position Paper presents industry recommendations for the accident analysis source tenns and release rates to j demonstrate with compliance of storage components with 10CFR72.
1 Accident evaluations for storage components should include all credible events
- which could occur during storage. (Accident evaluations for transport packages are prescribed in 10CFR71 and NUREG-1617 (draft). Industry comments on NUREG-f 1617 should be incorporated into the final version of the document.)
Source terms and release fractions of radioactive isotopes to the interior of the
! containment 1 boundary should be consistent between the storage and transport regulations and the associated Standard Review Plans. Standard Review Plans
- should be sufficiently flexible to maintain consistency between storage and
. transport requirements.
Industry Presentation Should the evaluation of the credible events not result in a breach of the storage containment boundary, a non mechanistic 100% release evaluation need not be performed.
The storage canister / cask containment boundary shall be tested subsequent to fuel loading and demonstrated to be sufficiently leak tight and in compliance with the leaktightness limits specified in the cask system 10CFR72 C of C or Site License.
The accident source terms and release fractions (to the interior of the containment boundary) should be based upon macroscopic data and measurements from representative fuel assemblies. Specific data should be provided to support the source terms and release fractions assumed for the evaluations.
' 10CFRF72.122(h)(1) and (4) refers to " confinement systems" while 10CFR71.4 refen to " containment" system.
For the purpose of this paper " containment"is used to distinguish the boundary from a requirement for intemal
" canning" of damaged fuel assemblies in "confmement" container.
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For accident evaluations resulting in the release of radioactive material from the r containment boundary to the environment, the behavior of the material from the release point to the destination may be considered in the evaluation, when ;
supported by analytical methods or experimental measurements. This is especially i applicable to the behavior of volatiles and fines.
Conclusion It is recommended that current SRP (NUREG-1536) be updated to include the above provisions .
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. I NEI Dry Storage Roundtable Workshop i Willard Inter Continental Hotel Washington, DC July 1-2,1998 1
ThermalIssues i l
License Annlications for Storare of Hiah Burnun Fuel Existing studies which are based on testing of moderate burnup commercial !
zircaloy clad fuel are widely accepted for use in the licensing of dry storage systems under 10CFR72 for low and moderate burnup spent nuclear fuel (SNF). The lack of comprehensive test reports which adequately address the behavior of higher burnup fuel under dry storage conditions has resulted in uncertainty in !
determining an approach which is acceptable to NRC for licensing of high burnup l fuel. ;
NRC appears to be heading toward determination of cladding temperature limits l based on a creep based methodology. Industry encourages such an approach.
However, until such time as this methodology is in place, the conservative application of existing DCCG methodology for moderate burnup fuel can be used to i provide acceptable temperature limits for high burnup fuels. This includes (1) the conservative determination of fuel internal pressure due to the increased radionuclides present in high burnup fuel; and (2) the use of reduced fuel cladding thickness based on the conservative assumption of the 10CFR50 limit on fuel cladding thinning (17%).
Relative to fracture toughness of spent fuel cladding for high burnup fuels, there is evidence that zircenium based alloys experience increases in strength and decreases in ductility as a result of radiation exposure. However, there is also data that shows that exposure to temperatures of greater than 300'C provides annealing of the cladding, resulting in reductions in strength and recovery of ductility. Since storage temperature limits are generally greater than 300'C, and since fuel temperatures experienced during vacuum drying are significantly higher, on the order of 400-500'C, the higher burnup fuel cladding will be annealed and, as such, there is no significant difference in the storage of moderate burnup and high burnup fuels as related to fracture toughness.
Internal Convection Within The Snent Fuel Cantater Appropriate determination of the effects ofinternal convection will provide temperature distribution in the basket which more accurately reflect that observed in tests of storage casks. As such, conservative application ofinternal convection needs to be given serious consideration in the design and licensing of dry fuel
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storage and transportation systems.
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- Many areas of heat transfer engineering rely on the presence of natural convection j for a signi6 cant, if not a majority, of the heat removal capacity for the design. The significance that natural convection within an enclosure can play in the overall heat transfer can best be appreciated by noting important application areas. A primary non-nuclear example is the electronics industry, including consumer, commercial, and military electronics, where convection heat transfer is viewed as
- . the principal method of heat rejection.
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- Within the SNF industry, buoyancy driven convection heat transfer within ventilated storage casks has long been used to provide a significant portion of the overall heat removal from the external surface of canisters / casks in storage via airflow in vertical or horizontal anmdi reound the canister. Over the past 10 to 15
, years, the presence and strength of ba pacy induced flow within non-ventilated i dry storage canisters / metal casks has been evident in many full scale thermal testa j and evaluated and correlated in analytical studies. Examples of physical testing include the Caster-V/21, the TN-24, the REA 2023 (MSF-IV), the MC-10, the
! NUHOMS, and the VSC-17 storage casks.
The results from these tests document the presence and effects cf convection heat i
transfer within the fuel basket. The conclusion to be drawn fr m the available test data is that convection heat transfer does occur within all of the fuel basket designs
! examined. This documentation is evident in the comparison of the basket and fuel i temperatures under vacuum, nitrogen backfill, and helium backfill conditions and
, with the cask in the vertical and horizontal orientations.
l Based on these results, credit for convection heat transfer within the guide tubes for l any of the basket designs under vertical orientation will shift the predicted j maximum temperature location by one to two meters over that predicted with no
! credit for convection, which may be an important effect depending on the design.
j Although little or no reduction in the maximum temperature is seen between j vertical and horizontal orientations when helium is used as the backfill gas, the benefits of convection will still be signi6 cant since it is seen as being equal to that l- obtained with credit for direct contact between the fuel assemblies and the fuel basket. Therefore, the analysis of the baskets with a helium, air, or nitrogen gas backfill should include the effects of the buoyancy driven convection in order to accurately predict both the shift in the peak temperature location and the reduction in the maximum temperature. Analytical codes for evaluation of convective heat transfer analyses should be appropriately benchmarked against the available test data.
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l Mirina Of Cover Gases With Fission Games And Its Affect On Ther==1 Conductivity.
Postulated rod failures create a mixed gas environment within the canister which has historically not been considered for thermal analysis. The thermal conductivity of fission gases (krypton / xenon)is significantly lower than helium. The presence of fission gas will also alter other thermal properties, such as viscosity and density, which affect convection. Depending on the specific composition of the mixed gases, these effects could be either beneScial or detrimental to the overall thermal performanet ~'
l It appears that for normal and off normal conditions of storage (assumed 1% and l 10% rod failure respectively), the effect of the introduction af fission gases on i thermal performance would be negligible. In the accident case where all rods are assumed to fail, increased fuel cladding temperatures are not a concern since the rod pressures are zero and cladding failure mechanisms are mitigated.
While it is acknowledged that the mixture of rod fission gas and helium results m a decrease in thermal conductivity of the final gas mixture when compared to straight l helium, the mixture of fuel rod fission gases and helium may significantly enhance l the free convection within sealed canister, depending upon the basket geometry and orientation. In many cases, the combined conductive and free convective heat transfer will result in a net enhancement in the overall thermal performance of the i
dry storage system with the postulated failure offuel rods. Therefore for canister i designs that do not account convection for the thermal analysis the conservatisms of l
ignoring convection and the enhanced heat convective removal capacity of the heavier fission gases more than offset the reduced conductivity. For designs which do account for convection, the analysis should explicitly consider the conductive and ,
convective effects due to the introduction of the fission gases.
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1 NEI Dry Storage Roundtable Workshop Willard Inter Continental Hotel Washington, DC l l July 12,1998 4
Cladding Integrity Issue Description Following t'he NRC presentations by Marissa Bailey and Mike Raddatz and the l industry presentation by Ray Lambert (EPRI), it was apparent that the NRC and industry position were very similar. The one area that needed greater definition is the mechanics ofimplementing the agreed upon philosophy relative to the language to be used in NRC licensing documents and in reactor plant procedures and i
protocols.
i The following key points represent those areas where a consensus appears to have been reached:
- 1. " Damaged fuel"is the term that best describes the fuel ofinterest, as opposed to
" failed fuel" which has a great historical connection to in reactor fuel 4 performance and does not best describe the conditions of fuel clad damage or bundle hardware damage that is significant to spent fuel storage or transportation. Also, the term " intact fuel" should be avoided in that it causes confusion because it has been normally used to define fuel which has not been consolidated rather than define the state of fuel clad integrity.
- 2. All fuel assemblies, as long as they remain in the nominal arrangement and geometry of a fuel assembly, are specifically exempted from the double containment requirements of 10CFR71.63 even though the cladding may have varying degrees of through wall penetrations. Spent fuelin the form ofloose pellets, debris, rod fragments, etc. do not qualify for this exemption. Fuel which is not damaged but may require special handling, such as consolidated fuel, should be treated on a case by case basis.
- 3. Damaged fuel defines that fuel which requires special handling because of either cladding defects or because ofmechanical damage to the assembly. To be classi6ed as damaged fuel, the cladding defect definition is agreed to be " fuel containing a breech of cladding greater than pinholes or hairline cracks".
Industry has suggested that in order to better quantify this definition that it be added "--- such that there is the potential for the loss of cignificant fuel particulate". To be classified as mechanically damaged, the fuel assembly's structural integrity would have been compromised such that special handling is j required or that the fuel assembly geometry, because of excessive bow or twist, l
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I no longer permits normal insertion or removal from standard sized spaces in storage racks or canister baskets. For some unique situations, such as in the NSP case where there had been mechanical damage to a few assembly's end fittings that were addressed by the use of a special pickup grapple, the NRC indicated that they saw no reason why these could not be standard configuration assemblies as long as they met all other criteria, but noted that these situations would have to be handled on a case by case basis.
- 4. Once fuel has been deemed to be damaged, the special handling requirements are that the fuel must be canned. The separate can will not need to be a ,
pressure containing vessel, like in the case of double containment, but may use l fine mesh screen to contain any fuel particulates and must maintain appropriate j margins of criticality control. The new can must also assure that the package 1 complies with all normal handling and retrievability requirements. j
- 5. The primary basis for utilities to determine the fuel condition in order to make ,
the judgment of whether an assembly is to be classified as damaged, and therefore require special handling, win be to use reactor operating records to l determine if significant clad failures occurred in the fuel to be loaded and what any subsequent special discharged fuel inspection or sipping tests revealed.
Additionally, the utility must perform a visual fuel inspection prior to the fuel being loaded to certify that the fuel assembly geometry allows for normal withdrawal and insertion into rack or basket spaces and that there is not ,
evidence of any gross cladding failure. Should either of these two qualification steps identify fuel that is suspected of being " damaged", additional testing (such as UT) or inspections (such as careful four-face TV recorded visuals) must be performed and a solid technical case made before the fuel may be classified as standard configuration.
Epund Table Discussion During the course of the round table discussions it was made clear that there remained certain items where some additional work was needed. In nearly all cases, these are action items for the industry and it was agreed that the burden to develop appropriate protocols was the responsibility of the utilities and vendors.
The following items summarize these action items:
- 1. The NRC is open to the concept of some relaxation in the criteria that requires canning of fuel (in the NRC's chart, this means moving the break line between the first and second classes to the right) but the burden will be squarely on'the industry to make such a case.
- 2. The protocols to be used to give evidence that the fuel has been correctly classified by the utilities must be developed and a system offormal paperwork e.nd sign offprocedures put in place to show that the reactor records have been 2
adequately reviewed and that the visual inspections completed.
- 3. In the case of very old fuel where reactor records are not as complete and where less fuel assembly specific information is available, it will be up to the utility to detail the additional testing and/or inspection steps that will be done in order to provide sufHeient evidence to enable the fuel to be classified as standard configuration.
- 4. While it is clearly the responsibility of the NRC to develop the language in licensing documents, such as C of C's, the NRC extended an invitation to the industry to make suggestions or recommendations as to the appropriate wording that might be used.
Conclusion In summary, the exchange ofideas on this topic was extremely beneficial and it is believed that a solid framework of mutual understanding has been established.
The work that remains must now focus on developing the mechanics to implement the agreed upon approach.
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NEI Dry Storage Roundtable Workshop {
Willard Inter Continental Hotel :
Washington, DC !
July 12,1998 Selective Spent Fuel Loading Issue Descrintion Historically, the Nuclear Regulatory Commission (NRC) has approved fuel qualification for spent fuel storage systems been based on the storage systems design basis fuel assembly parameters. Such parameters include size, weight, enrichment, burnup, cooling time, heat generation, gamma source strength, gamma spectrum, neutron source strength, neutron spectrum and a burnup-enrichment
, curve. Increasingly, applicants are relying on additional administrative controls in the selection process of candidate spent fuel assemblies for storage (e.g., targeting specific basket locations for certain types of fuels). Some administrative controls l
can serve to optimize the analyzed fuel assemblies acceptable for storage and i reduce the degree of utility calculations necessary to select candidate fuel
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assemblies. However, applicants will need to focus on the adequacy of the design basis fuel qualifications and the spent fuel selection operating procedures to ensure safety.
Currently, NUREG-1536, Standard Review Plan (SRP) for Dry Cask Storage
- Systems, considers fuel qualification utilizing the traditional approach. This aspect of optimized fuel selection is in the formative stages, and the SRP will need to be
- updated as the staff's position evolves. Furthermore, the NRC prefers engineemd safeguards in lieu of administrative controls to ensure that licensing requirements will be met during operational activities. In addition, if possible, NRC prefers a streamlined approach to generic issues, such as would be defined and approved through a topical report, to delineate an approved methodology. This type of approach reduces staff review time by not necessitating the review of similar issues for each applicant.
J Industry Response
! All applicants currently adhere to the historical design basis fuel assembly l qualification methodology. This methodology of fuel qualification for storage is extremely conservative in that an entire canister is assumed to be loaded with the
, design basis fuel assemblies. In reality, no assemblies ever meet the limit for all
- required parameters. Therefore, a very large amount of design and thus safety i margin exists between the hypothesized design loading and the actual fuel thatis
! 1.oaded into the canister. This excessive margin can and does limit the capability of l storage systems. This limitation of capability per package will most likely result in
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an increased program life-cycle dose due to the need to load additional systems.
Through both analytical and operational experience, cask designers and utility users hnve realized that certain canister basket locations contribute more than others to cask operation dose Targeting these locations for older, cooler fuel assemblies can significantly nduce the dose required to load a canister. This approach to target speci6c bant locations for certain types offuel is simply a '
refinement of the existing desigt basis approach and not a reduction in control to the administrativelevel.
Currently, several utilities speci6cally load older, cooler bundles on the periphery of the basket. The fuel loading plan used by each utility specifies the specific basket location for each fuel assembly, similar to core shuffles and reloads during reactor i refueling. Independent verification (2 separate and independent checks)is used to ,
place the assemblies in their specific basket locations, again, similar to core shufHes and reloads during reactor refueling. The measurement verification of proper system operation is always the final dose rate on the exterior of the package surface regardless of how the internal basket is loaded.
l The proposed approach would require the designer tojustify the loading pattern within the batket and prove analytically that the external dose rate would be within regulatory limits. The loading of assemblies would still require the same i independent verification in accordance with the loading plan developed and approved by the utility. Therefore, the proposed approach to target specific basket locations for certain types of fuel is not a departure from existing methodology, but rather an advancement of the same basic fuel assembly qualification approach. This advancement would require a rigorous and thorough evaluation to justify the fuel qualification. For example, this evaluation may propose that two separate regions in the basket have specific limits rather than a uniform fuel qualification basis.
Each cask designer would need to determine their approach based on the performance improvement desired. Regardless, the evaluation would most likely be more complex than the existing approach to analyze for a single design basis fuel assembly. Irregardless, the industry hopes that NRC will be receptive to advanced forms of an existing methodology and work with designers to approve this approach and utilize more realistic safety margins.
Industry needs to point out that the existing fuel qualification methodology contains excessive safety margin and may not support a full program ALARA approach. Specideally, if capacities cannot be increased by refining the existing fuel qualification methodology, then additional cask loadings will be necessary over the life of the program. The additional cask loadings will,in themselves, cause additional dose to workers. This situation may not be in the best interest ofALARA l and total system safety performawe and should be reason enough to consider
- advanced approaches to design methodology.
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Conclusion The use of advanced analytical methods to increase spent fuel storage system ;
capacity while reducing overall dose is a positive development in the evolution of ;
dry fuel system technology. - As with any method, cask designers must demonstrate I adequate protection ofpublic health and safety. Cask designers acknowledge their responsibility to provide adequate justification for proposed methodologies while understanding that additional complexity in analysis may lead to increased NRC review time. The increased review and approval time is worth the potential gain in l system performance.
Regarding a generic or topical approach to this issue, the industry believes that such an approach is not possible at this time due to the variations in system designs and specific performance improvements being pursued by individual cask designers. The industry acknowledges the advantages to NRC to generic approaches and would pursue such an effort in this case if determined feasible in ;
the future. l l
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Dry Storage Round Table Scope and Format Willard Inter-Continental Hotel July 1-2,1998
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Purpose of the Meeting To discuss generic issues associated with the dry cask storage process and to identify and clarify generic problem areas in an effort to expedite cask certification.
. Provide a message to participants and attendees the urgency of coming to clear understanding of the issues in order to obtain dry cask C of Cs.
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Round Table Process Presentation format a NEI and NRC will be round table moderators assuring that discussion flows, equal time and representation is provided.
m NRC will introduce the topic (5-10 minute overview).
m Industry will present approaches or understanding (5-10 minute discussion).
m Round table participants will provide additional points of view. Sufficient time will be allotted in order to permit sufficient airing of the topic.
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Round Table Process (cont'd)
= Positions and clarification will be 1
discussed
= Time will be provided for attendees to provide input 1
g(c 2 2
I Round Table Follow-up
= NEI will coordinate workshop proceedings, including topic understanding, exceptions, and suggestions for resolution. A schedule of resolution coordination will be provided.
. Draft proceedings will be forwarded to NRC for review and comment.
= Final proceedings will be forwarded to NRC (for placement in the Public Document Room) and all attendees.
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l l Dry Storage Round Table i
i l July 1-2,1998
.i Washington, DC j i .
! l
! ASME Code issues l i
,i Industry: Tara Neider l Transnuclear, Inc. -
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- _...... .1 -
l~ NRC Position i
- a FullImplementation of ASME Code
! Requirements a Authorized NuclearInspector a N Stamp (or NTP Stamp)
/
QEI
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i Basis for NRC Position 1
m Continued Fabrication Problems
= Establish a proven process used for l fabrication of nuclear power plant components l
1 9
.. c Industry Response
= FullImplementation of ASME Code is an over reaction.
= Fabrication problems a result of insufficient implementation of existing QA programs.
m Vendor QA programs are sound.
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Industry Response a Utility, vendor and fabricator must work together to ensure quality and compliance.
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- Impact on Industry
= Increased Costs a Longer Schedules i = Duplication of Effort a Restrictions on selection of l
- fabricators and material suppliers I Y'
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Design Requirements
= ASME Code =NRC
. Design specification and . SAR to describe all design Design Repon Required conditions and analyses.
- Design Ommer's Cemficae of
. NRC audits Design Accreditmion prior toissuance of a design speci6 cation. Organization QA Program.
. Weinen agreement with an e Utihty must audit design AuthoruedInspection Agency organimion QA Proparn.
required prior to an . NRCIssuance of C of C.
Applicmion.
. Packaging ownerto prepare, cemfy and issue construenon specificsion.
qicn
,: a. . n , .s Materials Selection
= ASME Code =NRC
. Procure from ASME . Materials to be certificate holders specified in SAR.
only. Audit not . Meetindustry required. standards.
. Only ASME . Materials suppliers Materials can be must be audited by used which have fabricator.
been selected based on pressure vessel , . .
requirements.
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3 Fabrication
= ASME Code =NRC
. N or NTP Stamp . Approved Vendor's l
holder required list.
- (these fabricato25 .10CFR71/72 or
, are in limited 10CFR50 QA l supply). Program.
l . NCA-3800 QA l Program.
. Industry notlarge
- enough to support multiple vendors. I l
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.;y.7. x;- %
Inspection m ASME Code =NRC
. ANIrequired to be . Vendor and utility present during customermust establish fabrication. witness and hold points
. ANIrequired to ensure throughout fabrication compliance with ASME Process.
Code only. . Vendor and utility
. Noinspectionsonnon customer must ensure ASME Code compliance with SAR,C components (poison, of C and SER.
' ~ -
resin, etc.)
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Auditing
= ASME Code =NRC
. Auditing not . All fabricators and required if N suppliers must be stamp holder or audited.
ASME Cenificate . Benefit of using holders are used. cenificate holders negated.
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Specific Fabrication Problems Encountered
. Inadequate fabrication specifications.
. Mistakes during fabrication (e.g., excess grinding of welds)
= Inadequate fabrication procedures (welding)
. Improper review of records ,.
(radiographs)
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Solutions to Specific Problems
= Thorough review of fabrication specifications by Utility and Cask Vendor.
. Fabrication oversight by utility and cask vendor.
e Review / approval of weld procedures by cask vendor and utility.
. Third party review of NDE records. Use of l
SNT-TC-1 A inspectors. l
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Industry Solution
= Commitment to quality through vigorous implementation of the Codes, Standards, and QA/QC plans and procedures that already exist.
= Establishment of a corporate " safety culture" which emphasizes quality.
NII w
Industry Progress a Utility Working Group formed in December,1997 to address quality and perforrnance related issues related to Dry Cask Storage Activities.
= All cask vendors will be audited by end of year to Dry Storage NUPIC checklist. -
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- 37.;;R ;
Industry Progress
= Increased surveillance in fabrication facilities by both cask vendors and utilities.
= Fonnation of cask owners groups to benefit from previous experience.
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1 Summary
= ASME Code implementation will not eliminate fabrication problems. ,
= The methods and procedures established by industry QA programs adequately address the fabrication issues when rigorously implemented.
= It has taken many years to establish the QA programs currently in place at cask vendors and utilities. They should not be replaced without considerable review and thought. g@.g ,
.y: . .
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Summary
= Fullimplementation of ASME Code will have a huge impact on cost and schedule.
= Utilities, cask vendors and fabricators must work together to eliminate fabrication problems.
N'E &
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- Dry Storage Round Table ;
} July 1-2,1998 j Washington, DC
- Confinement Source Terms and Release Rate l Industry: Bill Lee
} NAC International NEkesm21 i
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. _ ;.;y.&rz .._' . :; s.;, . spy ~ . y ; ~ -
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Source Terms / Release Rates
] Background i
l = Regulatory guidance chronology I
i . SRP/NUREG-1536 (draft)(10CFR72)
! I
- . SRP/NUREG-1536 (final)(10CFR72)
I I
. SRP/NUREG-1617 (draft) (10CFR71)
NEkbam2-2 c
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Source Terms / Release Rates Current Regulatory Position a Source terms include:
. Gases
. Volatiles
. Fines (including crud) e Release rate within containment boundary based upon published release rates (e.g. SAND 90-2406
& SAND 88-1358)
. Release to environment based upon leak testing performed (normal condition) e Instantaneous release (accident condition)
.... V8 3 p. .Ae; . ay : . . ;.~.. :pg.,. p yyy v x Source Terms / Release Rates Comparison of 10CFR71 and 10CFR72 Boundaries Containment j Boundary Controlled Area Boundary (72.106) 10CFR72 '
indwidual I Containment (72.104)
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Source Terms / Release Rates industry Recommended Approach a Utilize common source terms for both 10CFR71 and 3 10CFR72 evaluations i
a Refine release fractions and activity levels (NUREG-1617 Table 4-1) based upon integrated source term
, (not peak rod, worst assembly) e Calculate release to environment upon transport scenarios (SAND 80-2124)
= Calculate dose at controlled arez 'youndary based upon characteristics and behavior of material released from containment boundary (settlement of particulates and settlement of fines, including crud) www.s N,E I 3
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! Dry Storage Round Table i
- July 1-2,1998
. Thermal Reviews
(
{ industry: Robert Quinn j Westinghouse Electric Corporation
! N' i ..u u; vc .~;. x w i
l Thermal Issues 1
i
= Storage of High Burnup Fuel I
= Internal Convection
= Fission Gas Effects 1
High Burnup Fuel a Cladding Allowable Temperature Basis e Rod Internal Pressure Determination a Cladding Thinning
.a:- ., =
.~. . n ..
Cladding Allowable Temperature Basis a Current Methodologies Were Developed for Lower Bumup Values
. Based on expedmental data to 33 GWd/MTU -
. Extrapolated to 46 GWd/MTU
= Need Approach for Higher Burnups
~
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l Cladding Allowable Temperature Basis (cont)
= Conservatively Extend Existing Methodology to High Burnup Fuel
= Damage Mechanisms to Consider Include
. SCC
. Hydriding
- . Stress rupture due to creep ,
.DCCG QEI
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Cladding Allowable Temperature Basis (cont)
= Basis
. Provide confinement for fuel particles
. Maintain sufficient structuralintegrity ,
for handling after storage l l
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Cladding Allowable
- Temperature Basis (cont)
= UCID-21181 Failure Mechanism Assessment
. SCC and hyddding would relieve pressure
. Creep damage mechanisms are based on testing at T & P much higher than those for dry storage
. uncertainty in correlation
. DCCG is only mechanism for gross cladding rupture N$I w
- .? . fUS _
' Ah - Nh 1T V jiNk ?fhkf~ T ?
Cladding Allowable Temperature Basis (cont)
= EPRI TR-103949 Observations
. Test conditions much more severe than dry storage conditions
. Cladding strength increases and ductility decreases with irradiation
. High storage temperatures (>350 C) provide annealing , .
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Cladding Allowable Temperature Basis (cont) ;
a EPRI TR-103949 Observations (cont)
. Zirconium based alloys appear to be resistant to radiation induced void and cavity formation
. DCCG methodology provides j conse. tive allowable cladding i temperatures '
4 NEI l!
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l Cladding Allowable j Temperature Basis (cont) i i = DCCG Methodology (UCID-21181) i Not Lirnited to Specific Burnup Values l = Methodology Dependent on
- . Temperatures
+ are higher and decay more slowly j . Rod cladding stress - dependent on l + rod internal pressure (higher)
+ cladding thickness (less) N iE'I w
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- Cladding Allowable 4
Temperature Basis (cont) -
5 m Rod Internal Pressure Must Account l for
. Higher temperatures ,
j: . pressure varies by ideal gas law i . Additional quantity of fission gases l i
. Additional pellet swelling l . partially offset by increase in rod volume l due to irradiation growth j
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! Cladding Allowable
- Temperature Basis (cont) l = Cladding Thickness Reduced Due to
,l Waterside Cladding. Oxidation l l . PNL-4835 provides appropriate oxide l l layer thicknesses for moderate burnup
- fuel l . Conservatively assume 10 CFR 50 limit j (17%) for high burnup fuel , ..
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Cladding Allowable Temperature Basis (cont'd)
= Summary
. Use DCCG approach with conservative inputs for rod internal pressure and cladding wall thinning
. Validation of HBU fuel characteristics may be possible
+ piggyback off of EPRI Robust Fuel Program hot cellinvestigations NE I I
- . __et..,y' *} '@ 'E' _ <e 7 ' . ' % g,
',',_'j _ ' '
i Internal Convection 1
l
= Internal Convection is a Real Phenomenon
= Accounting forIntemal Convection
. Produces more realistic peak l temperature results l . Produces more realistic temperature distribution in the basket and spent fuel - - -
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, I Internal Convection (cont'd) l
= Buoyancy Induced Flow Within Non-Ventilated Packages Has Been l Demonstrated in Full Scale Thermal Tests
. Castor-V/21 MC-10 I
. TN-24 NUHOMS
. REA 2023 (MSF-IV) VSC-17 NEI l w ,
wewnW:: 9 %+ 4xw :. w:y:n .
Internal Convection (cont'd)
= Analytical Codes Are Available to Accurately Predict Effects
. Use industry-accepted analytical codes ,
. Use existing test data to validate analytical modeling approach
. Conservatively neglect convective flow .
through the fuel assembly . - -
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Fission Gas Effects on Heat Removal
= NRC States that Study has Been Performed that Identifies This As An Issue a Industry Needs to Understand Issues Raised by NRC Study Y'
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. ,. a n Fission Gas Effects on Heat Removal (cont'd) l
= Industry Position
)
. For normal (1%) and off-normal (10%),
the effects of fission gases would seem to be 2nd order effects
. For accident (100%), cladding integrity l is assumed to be gone and therefore there is no need to limit cladding , , ,
temperatures, obviating the concern ,
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Fission Gas Effects on Heat -
Removal (cont'd)
. Fission gas has lower thermal conductivity than helium, decreasing thermal con'ductivity of mixed gas l . Fission gas alters mixed gas viscosity I and density, which affect convection !
. Net affect could be detrimental or beneficial to overall thermal performance 4
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i Dry Storage Round Table l l
l July 1-2,1998 )
Washingtun, DC l Cask Stability and Cask Drop issues Industry: Dr. K.P. Singh -
Holtec International ,
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GENERAL ISSUES l
\
l A. Seismic Excitation of Casks l
. . Cask Kinematic Stability I
. Inter-caskImpact? l
) . IntemalImpact Loads
, B. Low Velocity Impact (Drop /Tipover)
. Industry Concurrence with LLNL Methodology
. Practicability of Nomograph Method
- C. Storage Pad Design
. Acceptable Design Methodology NIE I
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A A. HIGH SEISMIC LOAD ISSUE DESCRIPTION
= CASK RESPONSE UNDER HIGH SEISMIC g's BY DYNAMIC METHOD?
s a CONSIDERATION OF SOIL-STRUCTURE INTERACTION 7
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HIGH SEISMIC LOAD SRP APPROACH (NUREG-1536)
= APPENDIX A OF 10CFR100 m 1.1 SAFETY FACTOR AGAINST SLIDING AND OVERTUPSING
= INTER-CASK IMPACT PERMITTED E
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HIGH SEISMIC LOAD INDUSTRY POSITION
= Static Method Currently Used O.K. for Low Seismic Sites
= Time-History Method Is Appropriate for High ZPA Sites
. Seismic Earthquake of Proven Pedigree Should Be Used
= Acceptable Earthquake for a Cask Can Be Represented by a Response Spectrum
. A Family of Response Spectra Can Be Developed
. Input Time-History to the Pad Should include Soil-Stnicture Interaction, as Appropriate s
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. m e: p g g . y yy a y p k;m4 w e x HIGH SEISMIC LOAD ACCEPTANCE CRITERIA UNDER DBE
= MAXIMUM CASK DISPLACEMENT
= MAXIMUM CASK CENTERLINE ROTATION
= COMPLETE LIFT-OFF NOT PERMITTED
= INTERNAL IMPACT FORCES -
ACCEPTABLE FOR THE SNF NEI s W
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HIGH SEISMIC LOAD REQUIREMENTS OF i
DYNAMIC ANALYSIS
= Validation Against Experimental
( Data (Available from Japan)
= Time-Histories Satisfy SRP 3.7.1
- a Numerical Sensitivity Analysis j Required to Establish Confidence
{ = Parametric Analysis Required to Establish Confidence NEi
(
_ ff s),.R$,.i.. W; $1;_ki uff 4YO?)*:75$f tf s9f fi - l
- l B. PAD DESIGN:
4 ISSUE DESCRIPTION l = INDEPENDENT SPENT FUEL STORAGE INSTALLATION PAD IS SUBJECT TO A l WIDE VARIETY OF LOADS.
j = LATITUDE IN INDEPENDENT SPENT FUEL STORAGE INSTALLATION PAD i DESIGN IS SEVERELY LIMITED DUE TO l
OPPOSING DESIGN CONSIDERATIONS.
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- BECAUSE PAD IS DECOUPLED l l FROM THE CASK. l
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PAD DESIGN INDUSTRY POSITION
= LOAD ON THE PAD DEFINED IN THE CASK TSAR AND SITE DATA
= LOAD COMBINATION PER ACI-349
= STRUCTURAL STRENGTH COMPLIANCE WITH ACI-3249 j
= PAD DESIGN SHOULD MAKE MAXIMUM USE OF AVAILABLE DESIGN VARIABLES (E.G.,
ENGINEERED FILL, SUB-SURFACE SOIL) TO MINIMIZE DECELERATIONS UNDER DROP I
- ~
EVENTS to
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C. LOW VELOCITY I IMPACT: ;
ISSUE DESCRIPTION 1
m PREDICT RESPONSE OF THE 1 CASK UNDER POSTULATED l DROP AND NON-MECHANISTIC TIP-OVER SCENARIOS
,, d'
- :~ m eg: .. z .Wr > +>: t .n s . : -'s;+,;9% n &:
LOW VELOCITY IMPACT SRP APPROACH e IN AN ANALYTICAL APPROACH TO CALCULATING CASK DECELERATIONS, THE HARD RECEIVING SURFACE FOR A DROP OR TIPOVER ACCIDENT NEED NOT BE AN UNYIELDINO SURFACE, AND THE FLEXIUILITY OF THE STORAGE PAD AND SOIL FOUNDATION MAY BE INCLUDED IN THE MODELING. HOWEVER. THE ANALYTICAL MODEL USED SHALL BE VALIDATED. NUREG/CR-6608 REPORTS A 4
SERIES OF LOW VELOCITY IMPACT TESTS OF STEEL BILLETS.IT ALSO DEMONSTRATES AN ACCEPTABLE APPROACH TO USE THE BILLET TEST DATA TO VALIDATE A BILLET PAD. SOIL INTERACTION MODEL AND ITS ADAPTION FOR A CASK-PAD-SOIL SYSTEM TO CALCULATE CASK DECELERATION LOADS. $ -
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LOW VELOCITY IMPACT l INDUSTRY POSITION l l = NUREG/CR-6608 -I.LNL METHODOLOGY IS VALID AND l SOUND.
. LLNL DATA HAS BEEN BENCHMARKED BY EPRI.
i e SUPPLIERS' PREDICT 10N MODEL SHOULD BOUND LLNL PREDICTED SOLUTION: NOT TEST DATA WHICH HAS WIDE SCATTER
, I s I NOMOGRAPHS WILL LIhBT INDUSTRY'S EFFORTS TO ACHIEVE SUITABLEPAD DESIGN.
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i Dry Storage Round Table j July 1-2,1998 Washington, DC l
l Fuel Cladding integrity i
l Industry: Ray Lambert
! EPRI 1
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j 5.1 Applicability of 71.63(b) to ;
j Damaged Fuel 5.2 Disconnects in " Failed Fuel" Terminologies Industry. Approach / Comments 2
1
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, 1. Industry agrees consistent terminology needs to be applied for spent fuel storage and transportation
. " Failed fuel" has a common reactor driven meaning that is not appropriate for storage and transponation
. " Damaged fuel"is a better description
. Two types of damage may exist (as described in EPRI repon TR-108237):
. Mechanical damage - refers to bundle structure where the i damage is such that special handling is required
. Cladding damage - cladding perforations greater than pin holes and hair line cracks that have the potential ,
for loss of significant fuel particulate. Special EI handling is required 3
agiw+tWa+Ww 7 cmsA4 %+wW #
. Intact fuel" tenninology should be dropped as historically it has been used to differentiate between normal fuel and consolidated fuel assemblies
. Misapplication of terminology could result in requiring special handling for more than 10,000 assemblies rather than the few hundred for which it is appropriate i
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- 2. Industry strongly believes that damaged fuel, l as long as it is in a fuel assembly geometry, is still a fuel assembly and is exempt from the double containment requirements as specified in 71.63(b) !
. Solid plutonium in the following fonns is exempt from this paragraph:
)
(1) Reactorfuelelements i (2)...."
. The key determinant in deciding the degree of special handling that is required should be , i the impact on health and safety NEI 5
- tAy'a n pp - : A .. - -- .; ~ . f .
~3.~ 7 13 e - . . q .y - ' -
. It is recommended that special handling l
requirements be based on insuring adequate safety l margins for: l
. criticality control (geometry) l
. source term control (shielding)
. thermal control (geometry)
. Special handling requirement best suited to meet these conditions is a separate non-pressure retaining container that facilitates normal handling procedures and provides retention of fuel particulates
. Containers would be vented with fine mesh -
screens retaining any released fuel ggg fragments @
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- - - - . = _ - - - - -.- . . - . . _-. ..
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. Analyses in EPRI report TR-108237 documents that the use of such a container has no adverse health and safety impact on the public and an insignificant impact on workers
. Benefits of adopting this recommended approach include:
. Health and safety risks are not changed
. Demanding and unnecessary steps of a separately designed, sealed, dried, and inened small canisters are avoided
. Future reopening of a pressurized sealed container is avoided ,
EI 7
.wn:s :- :: e ; p. : +,7? m; . ,y sy .
. Other considerations
. Industry approach is consistent with NRC past practices which have been successfully and safely applied
. The bases for 71.63(b) was heavily driven by the intent to ship plutonium oxide, a very fine powder that is highly dispersible. Spent fuel does not exhibit these characteristics, which is the basis for the exemption of reactor fuel elements in 71.63(b)
. Behavior of damaged rods in pool environments has provided strong evidence as to the manageable behavior of irradiated UO2 material !
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. Other considerations (continued)
. All receiving facilities must be engineered assuring that damaged rods will be encountered, regardless of encapsulation practices at the loading stage
. The insignificant safety benefits of using sealed
- containers compared to the ALARA and cost considerations would produce a highly negative cost / benefit analyses
-. .. . n . ,, .n,..
l 5.3 Bounding nature of l analyses for damaged fuel for both normal and accident conditions Industry Approach / Comments
,s l 10 I
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- 1. The conservative way in which the safety analyses are made lend credibility to the premise that all reasonable normal and l
accident conditions are bounded QI N 11
.,*:'._- , .< . . . ok :10
- 2. Areas of conservatism include:
. Packages are designed to remain sealed for all credible accidents but then are analyzed assuming that containment is breached
. The source tenns used to establish the released amount of activity are inordinately conservative, including the assumed failure fraction of fuel rod cladding
. The assumptions used for the escape mechanisms for
- the various types of radioactive species is very Conservative . . .
/
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5.4 Qualification and Inspection of Damaged Fuel Industry Approach / Comments e
4 NII w
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, , / ._ &_ l ,' z ' p7 . _y,Q ' yp ,_ , ,& Rr,,$.h'.{,. , k'^ * [
. 1. Utilities already have good knowledge of the condition of spent fuel assemblies and fuel
- cladding s . Reactor operating records provide data on whether rod failures have occurred and the approximate extent of damage
. All fuel removed from the core is visually examined as part of normal core refueling operations
. Suspected leakers are further inspected and tested to verify nature and extent of damage - -
NEI 14 w
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- 2. Additional required inspections are burdensome, incur added fuel handling risks, and produce no increase in overall safety
. Visual exams, no matter how thorough, cannot accurately assess all fuel damage
. Ultrasonic and sipping tests also have limitations, panicularly with old fuel, and testing can produce both false positives and negatives NEI 15
- v.,; - -
g :. ps + . . . gn.:,;g
- 3. Other Considerations
. There is no inspection process that can guarantee detection of all failed rods
. Current practices will identify, with only few exceptions, all damaged fuel that requires special handling
. While less detail may be known about older discharged fuel, added inspection requirements will still not assure detection of all damaged fuel
. The consequences of a damaged fuel assembly going undetected have an insignificant . -
impact on health and safety g 16 8
5.5 Retrievability of Damaged Fuel Industry Approach / Comments '
f 17 d'
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- 1. For that fuel designated damaged, assuming the EPRI suggested definition, ease of retrieval is assured :
. Both types of damaged fuel, mechanical and clad penetration, are repackaged in containers that insure ease of handling
. Any bundle that has been handled and loaded at the utility site has, in essence, been verified to be retrievable 18 Y'
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- 2. While it is possible that fuel could change from i undamaged to damaged during the normal l storage or transportation period, this would unlikely make the assembly difficult to retrieve
- 3. The only action that would make the i retrievability process for damaged fuel more difficult would be if the assembly was placed in
- I
, a sealed container i
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i l Dry Storage Round Table j July 1-2,1998 Washington, DC 1
Selective Spent Fuel Loading Industry: John Broschak Consumers Energy ,
, +. . .. c, ..
Selective Spent Fuel Loading a Historically, fuel qualification for storage has been based on the storage systems design basis fuel assembly parameters.
= Increasingly, applicants are relying on additional administrative controls (e.g., targeting specific basket ' -
locations for certain assemblies) g$tI
- a,- ua, . . . - - .~.. ~ . ~ - -~ . . .
Selective Spent Fuel Loading
= All applicants currently adhere to the historical design basis fuel assembly qualification methodology.
= This methodology is inherently very conservative and can limit the capability of designs.
. The result can be a reduction in cask ,
operation ALARA. QEI c
- . ..3 , . -. . .. .,
Selective Spent Fuel Loading
= Certain basket locations contribute more than others to cask operation dose.
Targeting these locations for older, cooler fuel assemblies can significantly reduce dose.
. This practice is not merely an
" administrative control." I
O t
Selective Spent Fuel Loading
. Fuel is loaded into the basket using independent verification similar to reactor core refuelings.
. Any dry fuel system placed into operation must meet off-site dose release limits.
. Off-site dose release is verified ,
through measurement. TEI
. ,::. : ,, 7 : =... .
Selective Spent Fuel Loading
= An overly conservative fuel qualification methodology will increase total life-cycle dose for the management of the nation's spent nuclear fuel.
. Restrictive fuel qualification results in reduced system capacities and -
more cask loadings over time. gIEI l
)
.1 Selective Spent Fuel Loading
= Capacities can be increased with enhanced methods of analysis.
m Industry encourages NRC to view these enhancements as significant improvements in radiation safety performance and not a reduction in overall system safety. ,
't*'
, .+.e
~-
Selective Spent Fuel Loading i
= Industry accepts the responsibility to demonstrate continued acceptab% '
safety margins.
= Industry encourages NRC to be receptive to fuel qualification advancements and enhancements in order to promote total system safety -
l performance. gitI l
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l Dry Storage Round Table July 1-2,1998 Washington, DC i i
) Shielding Aspects of MVDS 4
Industry: l Alstom AutomationPaul Symons '
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QEI ,
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- ,,ge. a g gs p ,p w w a
l Computer Codes i
AEA Technology ANSWERS Shielding l i Service MCBEND - general geometry radiation i transport computer code. Established
- Monte-Carlo techniques for variance l
reduction
- RANKERN - Stochastic point kernel code
} used for gamma ray transpon solutions . -
NEI
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Skyshine Calculations MCBEND validated for use in gamma skyshine calculations RANKERN validated against MCBEND
+ Neutron and gamma skyshine assessed for MVDS Use applications guides for computer codes
-Is volume of air considered sufficiently large?
If RANKERN used, how many orders of scatter?
-Is groundshine a significant effect? -
'1F '
,2,_
';;S ,' {f'} ? N' *;' - * ' '-
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- ^ '.Y
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Streaming Calculations RANKERN/MCBEND/ Hand Calculations
- Validation evidence available for both codes
- Use applications guides (gamma & neutron)
' - Typical MVDS streaming problems include
- Inlet / outlet duct scatter
- Streaming though annular gaps around shield plugs
- Problems l - Inappropriate smearing , . .
- Acceleration problems / misleading results l
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Array Calculations !
RANKERN and MCBEND
. Flexible geometry and source dermitions - ,
ideal for modelling large arrays of storage ;
containers in MVDS l Perform and document preliminary calculations to see how many rows contribute
, w. . ~ ;:. .; s a i
Neutron Measurement
- Not directly involved in neutron i measurements Potential discrepancies between calculation /mesurements
- - neutron multiplication in source term l - modelling deficiencies - full detailed 3-D l model should be used ,. .
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j StaffGuidancewith !
Applicants i
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l WiliamKane l \
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ATTACHMENT 3
o t
. 1' StaffGuidancewithApplicants -
m Pre-application rn.eeti ws with an . applicant will be !
encotragedand x ec rotheapphcant l a Partial or incomplete applications will be retumed to the applicant !
- Review ofRAI response will not stan until complete resparse is l received ;
a Applicant will be asked to state if the up sivpriate Standard Review Plan was followed in c eveloping both their application and safety analysis report, and to identify alldeviations.
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StaffGuidance with Applicants (cont.)
a NRC will dedicate a specific review team for each majorapplication.
a NRC goal is no RAI for any new application or amendment.
a One RAI (pediaps two) will be considered acceptable, but staffwill:
t (<90 day)res finn applicant.
- Si e accordinglyif ond 90 days.
-P onn 2 week review ofresponses to detennine ifreview should proceed.
3 MW9M l
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StaffGuidance with Applicants (cont.)
m Ifmore than two RAIs am needed, staffwill:
-Identify its positions and concems.
- Sus mid funher technical review pending certification of a, pp.ication sufficiency by the respective Owner's Group or other Independent thini party review group.
m RAIs will be discussed in a public meeting.
m Infonn potential applicants that woi will not commence until resources have been budgeted.
4 l
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4 SRPRevisions i
Susan Shankman ll I
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SRPRevisions a Changeswillbedone
-On a biennialbasisor
-when a significant change is needed.
m DissaninationofChanges -
a Revisions will not have a fonnal comment period.
6
e Status ofCurrent RulemakingActions Susan Shankman 4
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- STATUS OF CURRENT PART 71 AND PART 72 RULEMAKING ACTIONS
- of .3.,.8 FINAL RULES '
t D6&,iipiki, Rulemaking Number Status (Programmatic Division)
Exempt Vitrified High-Level Waste from Plutonium \
RM M91 Final rule published in 63 FR 32600; effective on 7115/98. I Double Containment Rel _..14. of $ 71.83 i (PRM-71-11) [Part 71] (SFPO)
Adoption of Part 20 Dose Methodology [Part 72] RM M37 PiwJ rule published in 63 FR 13372; comment period closed; final rule t
l (SFPO) in development.
PROPOSED RULES I
Dem, ;yik,n Rulemaking Number Status (Programmatic Division) l Miscellaneous Changes [Part 72] RM M46 Proposed rule published in 63 FR 31364; comment period open until (SFPO) 8/24/98.
Expand Part 72 Scope to Clearly include Certificate j
RM M39 Proposed rule submitted to the Commission for approval in SECY-96-113.
Holders, Applicants for a CoC and their Contractors i and Subcontractors [Part 72]
(SFPO)
Eliminition of 30-Day Delay in Loading Spent Fuel After j
RM M33 Rulemaking plan approved in SECY-98-056; proposed rule in dc;;4imii!.
Prooperational Testing in 5 72.82(e) [Part 72] ,
(SFPO)
Geological and Seismological Characteristics of Spent RM M41 Rulemaking plan approved in SECY-96-126; proposed rule in dc;;4ieiil. -
Fuel Storage Systems [Part 72)
(SFPO)
l Desenphon Rulemaking Number Status (Programmatic Division)
Conform 5 72.48 to changes associated with 5 50.59 RM #508 Commission approved rulemaking in SECY-97-205; proposed rule in (SFPO) d;'. M e 4.
i i
RULEMAKING PLANS UNDER DEVELOPMENT Description Rulemaking
- Number Status (Programmatic Division)
Clarify General vs Site-Specific Requirements and Add RM #438 Rukmoking plan submitted to the Commission in SECY-98-148.
Flexibility [Part 72]
(SFPO)
Rulemaking Actions for Cartificates of Compliance Commission Paper is being developed.
[Part 72]
(SFPO)
Amend Certificate of Compliance for VECTRA RM #518 Rulemaking plan is being L;;;,,pc-d.
Technologies, in, NUHOMS Dry Shielded Canister
[Part 72]
(SFPO)
Fissile Material Exemptions and General License RM #521 Rulemaking plan is being C;;. ,, ped, study of fissile exemption regulatory Limits [Part71]
bases is scheduled to be pehed in July in NUREG/CR-5342.
(SFPO)
Conform Part 71 to IAEA Standard ST-1 and Expand RM #496 Rulemaking plan is being C;; c,pc.d. This rulemaking is being Scope to Clearly include Certificate Holders, coordmaled with US DOT revisions to 49 CFR. ;
Applicants for a CoC, and their Contractors and Subcontractors [Part 71] l (SFPO) i
4 l
PETITIONS FOR RULEMAKING (no rulemaking in process) !
Cek p;vn Rulemaking Number Status (Programmatic Division) '
Allow Interim Storage for Greater than Class C Waste RM M62 Petten was nobced in 61 FR 3619; comment period closed; petlhon urxi.w at an ISFSI under Part 72 (PRM-72-2) [Part 72] review.
(SFPO)
Require Revision Numbers for SARs an CoCs and RM M64 !
Pehhon was nobced in 61 FR 24249; comment period closed; petdion under Require that the SAR be Updated to Reflect the SER review.
and CoC before the CoC is issued (PRM-72-3)
[Part 72] '
(SFPO) f .
Define Degraded Fuel and Retrievability and Require RM M73 Petdion was noticed in 63 FR 12040; comment period closed; pettion under._ :
o Demonstration of Unloading CapabMey sefore review. !
Loading Begins (PRM-72-4)[Part 72] '
(SFPO) - "
Eliminate Double Containment Requirements for RM M71 Petition noticed in 63 FR 8362; comment period extended to 7/31196 in Shipment of Plutonium in 3 71.63 (PRM-71-12) 63 FR 34335.
[Part 72]
(SFPO) '
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Implementation of ASMIC Code Requirements l
RonRarkhill i 1
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NRC/NEI Workshop 7/l-2/98 4
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FULL IMPLEMENTATION OF ASME CODE REQUIREMENTS, '
INCLUDING STAMPING AND :
SERVICES OF AUTHORIZED NUCLEAR INSPECTOR (ANI)
FOR LWR SPENT FUEL '
STORAGE CANISTERS AND TRANSPORTATION CASKS <
e 9
Objective: Initially request and eventually require full implementation of ASME Code for LWR spent fuel storage canister confinement and spent fuel transportation cask containment for all applications, especially the dual purpose casks.
i Current Fabrication QA Practice: Approval of vendor QA program by NRC.
Vendor ensures fabricator meets their QA program requirements. Inspection of fabrication by vendor, utility and NRC.
j Suggested Change to Current Practice: Ensure ASME involvement in '
fabrication, including ANI and stamping. Same practice as used on standard -
nuclear components.
Reason for Suggested Change: Presence of ANI in fabrication shops should eliminate most of the historical problems (e.g. excessive grinding, undocumented welding, review of radiographs, lack of weld preheat for closure weld).
2
s Applicability: New applications, specifically dual purpose casks, but not to navy nor research reactor fuel.
- Recent Developments: '
i Transportation: New issuance May 1997, ASME Section III, Division 3
" Containment Systems and Transport Packagings for Spent Nuclear Fuel and High Level Waste" '
Storage: Previously could not stamp since not all Code requirements were being met .
(e.g. closure weld configuration, no hydro, lack of volumetric NDE on closure weld). Code Case under development that would permit stamping to ASME Section III, Division 1, Class 1 or 2. i
~
Public Law 104-113, National Technology Transfer and Advancement Act-Enacted 1996 and requires the use of consensus standards nniess there are sound reasons for not doing so.
3
Implementation:
. Short term:
Revise SRP for transportation ofspent fuel to include stamping and services of ANI (draft report for comment published 3/98- NUREG-1617)
~
Discuss apppoach at NEI workshop, July 1-2,1998 Brief CRGR, August 11,1998 Prepare Commission Paper Revise SRPs for storage of spent fuel to include stamping and services of ANI (after code case is approved by ASME and accepted by NRC) nt request -
vendor to obtain the services of an Authorized Inspection Agency (if code case not adopted)
Establish agency peer review group to review Codes and code cases. t Document acceptance / restrictions on use via regulatory guide. (e.g. revise RG 1.84)
Long Term: Rulemaking j I
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. _ _ - _ _ _ _ _ _ _ - _ . - _ _ a
Advantages to Recommending Full Implementation of the ASME Code Fabrication process commensurate with that utilized for nuclear power plant components NRC inspection effort at fabricators reduced to oversight / audit ANI involvement in construction assures quality commensurate with standard nuclear components i
For transportation packages it would place the NRC in compliance with Public Law -
104-113
~
J It is an opportune time since the dual purpose casks are still in the licensing process and the majority ofspent fuel is anticipated to be stored / shipped in these casks j
Stamped components may provide a marketability and public confidence advantage Advantages to Recommending FullImplementation of the ASME Code (cont'd) 5 6
The inspection process of the vendors / fabricators would be more definitive since there would be a clear set of Code requirements for construction with the ANI bei the first line official interpreter of the Code requirements With staniping, the.ANI would more appropriately assume the in-line QC responsibilities.
-Disadvantages to Recommending FullImplementation of the ASME Code -
Slight cost increase due to the ASME accreditation process and obtaining the services of an authorized inspection agency.
Storage cask designs currently do not meet certain Code requirements and could only be stamped when the associated Code Case gets approved by ASME.
i 6
AccidentAnalysis Char"es J.Hauganey 9
4 m@m
RegulatoryRequirements Accident Analysis is required to satisfy many sections within 10 CFRPart 72 l
m 7224(a ,(d 2)&(m)72.90(b),72.92(a), ,
72.94(a , ,&(c),72.104(a),(b c),(i)&(h),72.124 ),&(c),72.1 l 72.122
,72.92(c),(d),&(1). (a),72.126(c) 72236 l
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I AccidentsRoutinelyEvaluated l
l a CaskTipover s CaskDmp
~
a Flood m FireandExplosion a Li s
a losso Shielding nAdiabatic.Heatua(VentBlockage) e Tomadoes and Missiles Generated by Natural Phenomena 11
= - ,
Non-MechanisticFailure a Past practice included an evaluation ofthe consequences ofa Non-Mechanistic Failure ofthe cask confinementboundary a Thatpractice h' as been re-evaluated 12
$3 RS '
l CredibleAccidents l Cmdible Accidents and their Consequences Must Still i
beEvalmted l
m Analysis should seek to identify a credible accident msultingin:
- A faihne of the confinement botmdary
-The transfonnation of the radioactive material into a dispersible fonn
- A mechanism for the release ofsuch materials fnrn the cask
- A mechanism for the offsite dispersion ofreleased material
-Conesponding dose assessments a The staff, will no longer routinely expect that an analysis ofthe conse evaluated. quences ofa non-mechanistic failure be l
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StandardReviewPlan' ProposedRevisions s Rewrite of the accident analysis chapter
. -Clarification of the minimum accidents that must be considered
-Basis and references will be reviewed ar4 manded
-Methods for determinino release
~ frac / don soum:e tennwillbe clarified
- Clarification ofwhat is an off-nonnat events vs accidents
- Reorganization of the material into a stand-alone chapter 14 F
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SpentFuelCladding ~
in Post-Reactor Service CharlesIntenante 15 N ognume -
- -____a
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What are theIssues? .
! License Renewals and Applications for Licensing of HigherBurnupFuels ;
a Title 10 CFR-Eneq!y '
l -Part 50: Wet Storage
-Part 72:DryStomge
-Part 71: Transportation
-Part 60: Disposal i
- Part 63: Disposal that is Specific to Yucca Mountain l
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10 CFR,Part 72-Storage 1 l
1 m Materials mquimnents forhigher bum up fuels am not addressed directly a NRC limits creep by limiting service tanpemtures a These limits am set, in part, using modeling and reactor service expenence l
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17 ,
4 I
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i Values ofManyImportantParameters IncreasewithBurnup a 1. Heat output in storage increases:
Incmase in hwiature and Pressum of fuel rods.
m 2. Radioactivity increases--more fission products a 3. Corrosionincreases-oxidethickensandC[H]
increases a 4. EmbrittlementisafunctionofC[H], hydride orientation, and yield stress. t 18
(
1
Failure Mechanisms for Fuel Cladcing a Creepoccursunderslowstrains
- reep is gC using a Diffusion-Cmtmiled Cavity-a Flow- ASME Code 9proach (Y.S.,UIS)
-This appmach protects namst fadum frtrn loads that may lead to excessivedefonnation a Fracture-Potentialmechanism forembrittled materials 19 i
i M MeH M@*
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Failure Criteria forCreepBehavior m SRP Limits Temperatures using a DCCG Model:
"StageIII" Behavior.
m Short-Tenn Creep Limit =~0.1%.
a Long-Tenn Cree 3 Limit =~1%.
m Linuted " Stage I" Behavior is desired for consentative storage, a Creep Data is Needed to Support This Approach.
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Failure Criteria forFlowBehavior l i
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" ASME Code spyivach uses flow criteria Y.S., UTS, with SafetyFactors g @highTemI$Y.S.maybelow¬ m en a Data on effects ofimportant parameters on strength l would support the flow critena approach i
21 l
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FailureCriteria forFracture Y.S. andUTS vis-a-visK c andK, i orK(d) a Fractum is to be avoided in Handling, Operational Events and Transportation a Fractum ,K,isproportional to stress and flaw size, .
-Strength may
~
inemased byHydrogen andFadiatinn.
and at lowerTemperatures.
-Fracture ess is decmased as level ofStmngth inemase
-Hydmgen rittles materials (decmases fracture toughness, K, and increases Y.S.)
a Fmeture Toughness Data Support these Failure Criteria N
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l l Infonnation for Use in Bounding Models :
i
! Considerin Thickness,g effects ofAlloy Camposition, Oxide C H,RwiiationEfrects,Tergature a Conservativecreepcurve(s) a Strengthssimate a Fracturetoughnessestimates l
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InternalNatural Convection l
SteveHogsett !
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eneham
Whatis theIssue?
m SameapplicationsreceivedbytheNRCareapplym' g for the credit ofnatural convection within the ~
tTueI canister as a mode ofheat transfer. Although method departs from previously approved applications, the NRC welcomes new and different approaches to analyze a package's heat transfer given that the appropriate expenmental data is provided and used as a benchmark for the mixed modes ofheat transfer.
25 t
4
InternalNaturalConvection RegulatoryBasis:
a 10CFR7224(cX3),7224(d),72.122(hXI), '
72.128(a)(4), 72236(f), 72236(g), 72236(h)
.\
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i l NUREG-1536 i Chapter 4.0 SectionV-4(a) l m Convection by natural circulation should be limited to l
that between the extemal surface ofthe cask and the ambient envirunuient. The staffhas not pmvicusly ,
approved specific thennal models for natum! l cuculation mtemal to the cask... Applicants seelo'ng '
l NRC approval ofspecific intemal convection models should propose a comprehensive test pmgram to demonstrate the adequacy of the cask design and validation of the convection models.
27 l
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NRC Concems forThennalReview l
\
l m Appwpriateconvectiondata .
m Appc ' ate benchmark ofcodebeingused e analysis with only radiation and conduction '
s Effecto vacuumdryingdnvingmaxheatload a Clearly present the key parameters ofthe convection model: geometric and non-dimensional I
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NRC Concems forThennalReview (cont) e Assess the role ofnatural convection during the fire i accident
! l m Analyze both vertical and horizontal storage conditions a Consider the po.ssibilities offlow blockages in the basket (e.g. canned fuels with screen covers e Detailed description ofmodel and all assump) tions a Consider the effect offission gases diluting the heat
- convecting covergas l
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l l Fission Gases and Cover Gas SteveHogsett x
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Whatis taeIssue?
s A recent analysis by NRC staffhas myealed that the introduction ofheavy fission gases into a lighter canister cover gas strongly decreases the thennal conductivity of the gas nuxtum. In the future, applicants will neec to address this issue to account for its negative effect on the thennal perfonnance ofthe cask.
31 i
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! FissionGas Concems RegulatoryBasis:
= 10 CFR 7224(c)(3), 7224(d), 72.122(hX1), i 72.128(a)(4), 72236(f), 72236(g), 72236(h) 1 32 i
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NRC Concerns forThennalReview:
I a Reduction of the thennal conductivity ofthe mixed canistergas a Tempestum and pressure rise within the canister a App:y a suitable method for the mduction of the thennal conductivity with mspect to tanpemtum a Consider conservative mixed cover gas pmperties and re-evaluate for peak temperature and canister pressurization 33 esmeene w MM M e
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SuggestedChange to SRP l
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a Under the conditions where any ofthe cask ~
ent tempemtures are close (within 10%) to their 1 values during an accident or the MNOP is within 1 ofits design basis pressure, or any other s conditions, the applicant shouldbyconsider,pecial analysis, the potential impact ofthe fission gas in the canister to the cask component temperature limits and the cask intemalpressurization.
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Low-VelocityImpact Test and Cask !
l TipoverAnalysis DavidTang 35
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Low-VelocityImpactTest and Cask
! TipoverAnalysis i
a NUREGCR-6608
-SNL and LLNL testdata
-Ripd fnrn totalresponse 1
i
-Bilet ismodel ' '
-Fnun illetanalysismodeltocask-pad-soilsystemmodel i a Issues a SRPchanges '
s Futureconsiderations l
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- SUREG/CR-6608 ' '
S'SL andLLNL TestData s SNL: 13 billet end drop tests l 8 LLNL: 12 billet end drop, side drop and tip over tests l 5 Test confi gurations and accelemtion traces l 8 Rawtest cata on diskettes
-June 4,1997,M. Witte, LLNL letter to NRC l i
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6 NUREG/CR-6608 Rigid-Body Response finn Total Response a Totalresponse= rigid-body + vibratory a Iow-passfiltertotalresponsetoobtainrigid-body response a Use billet modal properties for selecting filter frequencies .
-Flexural modes: side drop and tip over
-Longitudinalmodes: end drop 40 l
I
ISSUES s DYNA 3Dfiniteelementmodelingofbillet-pad-soil test configuration a Principalmodelingfeatums
- Billet-to-pad and pad-to-soil sliding surfaces
- Infinite soil dcmam modeled with non-reflection boundmies 41 l
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NUREG/CR-6608 From Billet Analysis Model to Cask-Pad-Soil System Model a Retain
-Slidag surfaces
- Soil non-reflecticn boundaries, as appropriate a Replace F. E. model ofthe billet with that ofthe cask u Use site-specific or bounding pad and soil properties 44 6m 6 -
69 N
l CaskTipoverAnalysis Unfiltered and filtered at350Hz MaxAcceleration=66.7g i
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j tan eau eme ass saa ans aan aus aan aam M EE8 M F W 29. b rk enskaf and Shared at asses, anah* reseg 45 1
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Issues ,
a General ptuposecomputercodes '
s Boundinganalysisandsite-specificevaluation a Energy dissipation thm shielding materials a Rigidmaterialfonnulation 46 l
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SRP Changes m Curmnt: '
- Flexibility ofpad and soil included in modeling
- Analyticalmodel shall be valid *A m Willadd:
-Reference toNUREGCR6608
- An acceptable approach to use the billet test data to validate a billet-pad-soil model and... to calculate cask deceleration loads.
47
- .- =. . . . . . . - . - - . . . - . - . . - . - - - . . . - - --
N
Future Considerations m Numerical simulation sensitivity analyses ofcask systans
-Lower bound decelerations as functions of
-Majormsk aanknes
- Pad concretesurngth and depth
- Soiland foundationppues a Cask designed to lower bound decelerations
-Bounding cask-pad-soil attnbutes
- Staff acceptance without funher revkw 48
e Cask Seismic Stability
-High Seismic G's Davic Tang 49 l l
=
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l CthrentApproach SRPNUREG-1536 l
i e ANSI 57.9
-Safety factor 1.1 nahM sliding /ovemuning
-Imphedno sliding i
a Notipoverwillresult !
e Impact between casks should be precluded or should be considered an accident 50
, 4 i
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RegulatoryBases a 72.212. SAR reviewed against site parameters a 72.102. Standardized design earthquake, .25g
-Minimtun site-specificZPA, .10 g n- Ap)pendix Soi A ofPart 100, siting criteria
-stmeture interaction (SSI) efrects
-Expected duration ofvibratory motion a R.G.1.60. Design response spectmm 51 0
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Issues .
m Staticapproach-nocasksliding/upli&g
-Moderate g level
- Site specific SSI analysis -amplified g level
~
. m Slid lihgbynonlineardynamic analysis "higi seismicGs addresses onlypad seismicGs
-No acceptance criteria
-Validation ofdynamic analysis appmach a Rulemaking on probabilistic seismic hazard appmach l
52 i
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e Future Considerations a Shake table testing ofscaled cask models
-Data base for validating analysis appmach
-Testing to detennine lower-bound cask perfonnance 53 i 1
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s Structural Analysis of Storage Pad DavidTang M
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l CuuentApproach
, DraftSRPNUREG-1567 l
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m Table 7-1 Load and load combinations a ANSI 57.9 !
-Strength offounation sections 1
-Soil capacity ;
-Overtummg/slidingresistance
-Dead load
-Live load l
-Lateral soilpressure
-Accident loads
- Etc.
55 t
-,-p W'* - -- *** *8 m m em e W-- _ sammuum u I
Structural Analysis ofStoragePad .
1 i
a RennlatoryBasis
- 72.712 ). Staticload ,
- 72.102 .
Soilliquefactionpotetial
- 72.10 . Soil .uo u,
-IP 60851, ' Design tmlofISFSICompmmts" t
abe
- Pad designed to Category I to preclude cask tip over !
- Pad classified as important to safety to facilitate bolting ofcasks to the storage pad
- Rulemaking on graded appmach to seismic design 56 i
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Failed Fuel e
MarissaBailey 9
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Damaged orFailec FuelIssues Canning andDouble Containment srnaaoemmtumm com 8?# Lee voacemcmanaaeas gma mg s
== MSMiFS M*""'
a Canning (handling,retrievablity,criticalitycontml)
-Known or susoected cladding defects greater than hairline cracks or le lea WD N Summ e m emessee
SelectiveLoading of l Storage Casks TimMcGinty 65
_pe m
e Whatis theIssue?
Part 71 and 72 opGudzed loading applications have beenreceived a SelectiveLewiina
- Such requests meohpmprietmy a Requestshaveincluded
-selective loadingpattems
- optimized within the design basis fuel paraneters f.
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i
SelectiveLoading m Typical design basis fuel qualification parameters
-Size
-Weight
-Enrichment (min andmax)
-Bum up
-Cooling tirne
-Heat generation
- etc.
m in the license and/orcertificates a '
ized loadingpattems are effectivelyprecluded 67 gummu m eenummmm t
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~
RegulatoryBasis a 10 CFRPart 71
- 7133(b), 7135(c), 71.47, 71.55, 71.59, 71.87(f), 71.91(aXii) a 10 CFRPart 72
- 7224(cX3), 7224(d), 72.44(c), 72.122(h)(1), 72.124, 72.236 68
Transportation SRP Status m Bounding analysis per SRP typically provided
-Thenalheatloads
-Sourt:e tenns
-Reactivity 1
m SRP evaluates spyropriate special 'ontmis for
-Enading
-T
-Hand u Industryhasgenerally
-Demonstrated adec uate inading controls
-Pmvented mis-kw iny 69 1
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Transportation SRP Status continued a NRC recently approved a "prefemntial" loading configuration a Approvalwasbasedon
- Additional administrative controls
-Analysisofcmdiblemis-Inading (invokedbyCoC) a Draft SRP does not cummtly address optimized loading configumtions 70 e mm mm_- e e -- ._ _
Storage SRP Status a NRC has not yet approved any fonn ofselective loading a SRP establishes mviewmethodologies for bounding cases a Technical specifications provide acceptable fuel parametes 71 m m we e v' N - M
I
!' Storage SRP Status :
continued a Operatingpmceduresdefine '
- Appiyih spt fuel candidate selection criteria
-Emdmg witma a NRC must find reasonable assurance for
- Increased reliance on =*niniorative controls
. -Areas subject tosingle failures ,
a SRP does not specify guidance for optimized configurations f
1 I
J I
J
l NRC Concems a Clear and concise fuel selection criteria
, a Adding additional complexity to the review disciplines a Increased reliance on administmtive controls m Incmasedpotentialformis-loadings a Charactenzingtheconsequencesofmis-loadings a On-site calcu ations to qualify fuel assemblies a Establishing review acceptance criteria 73 l l l
1
WhatCanYouDo?
s Coordinateindusty'saproach !
u Focus on simplicity anc measurability n Identi what additional analysis needs to be considemd a Identi what' additional administmtive contmls am need a Establish pmposed acceptance criteria a Propose technical specifications a Presentyourcase 74 !
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l
O ShieldingIssues Car"Wi6ee andMichaelWaters 75
-- . = -- .....
l .. _
4 ShieldingIssues m TmxisinNewCaskDesigns
-Oatimintion ofContets Sum-upvmuscoolingrune
-IncreasnigNeutron SourceCwuments
-LargerCaskAnays m CurmntShieldingIssues
-Streaming -
-NeutronMeasurernents
--SkyshineAnalysis
-ArrayCalculations 76 eumspe emusee a
em l
i Streamincr v a hene:Perfonningproperradiationshraning calculations m Regulations:
- 10CFR72.104(a) - Dose linit to n:al individual is 25 mmn/yr
-10CFR72.104(b -ALARA
- 10CFR71.47(b)) - 200 mmn/hr vehicle surface; 10 mmn 2m a SRP: ,
-NUREG-1536,5.V3.a -Treat streamingpaths conservatively 77 i
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__ i m
ecomepeee e some -
i
Streaming J
s CurmntStatus:
-Chiilations difbilt j -Had musotrope
-Streamingmaybelimiting factor a ActionNeeded:
-Include adequate designmargin '
-Measumnent and feedback to calculations
-Lignuvemethods 78
N-reutron. Measurement u Issue:Perfonningpmperneutronmeasumnents a Regulations:
- 10CFR72.44(c)(3) - Surveillance to confinn operation within !
limits <
-10CFR71.87(j) - Detennine that extemal radiatim limits are l met l m SRP: l 1
l
-NUREG-1536, 8.IV.5 - Identify surveillance and monitoring j procedures
~
79 t'
M ese m esege. e.
l l -
1 .
NeutronMeasurement a CurrentStatus I
-Neutron ComponentIncreasing
-ThennalSpectraMom Common ;
m FutureActions
-nwCahbration
-Incidentneutron energy
- Surface flux energydistribution
- Compam Neutron Calculations and Neutron Measuments 80 l ._. . _ . . _ . . _ . . . _ _
_ _ . . . _ . _ _ _ _ _ . . . . _ _ =..._... .__
l
SkyshineAnalysis a 1ssue:Perfonningproperskyshinecalculations a Regulations:
- 10CFR72.1 Dose linit to real individual is 25 mmn/yr
- 10CFR72.1 -ALARA a SRP:
-NUREG-1536,5.V.4.c & 10.V3
- Perfonn cmservative dose rate venus distance calculations for cask aray.
81 t
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l
- . . . . . . . . . . _ - . - = . --
e _
O
. SkyshineAmlysis m CurmntStatus
-Future dose rates could -- vach regulatory limits due to optimized cask loadingsker array sizes and/or decreased site boundarydistances. ,
- Current stomge sites operate well-below ofTsite regnhtory dose limits.
m ActionsNeeded
-Validation ofecmputercodes
- Closer examination ofmodeling assumptions as dose rates roach latorylimits are shine calculations and sk,/ shine measurunents i
9 O
.. . m e en eusman empe eese m e enumeuseep gnee 0
- ArrayCalculations ,
m Issue: Side-flux assumptions used for cask array dose '
rate ca]cn12tions a Reenlations:
- 10dR72.1 Dose linit to mal individual is 25 mmn/yr.
- 10CFR72.1 -AIARA a SRP: -
-NUREG-1536, SN.4.c & 10.V3
- Perfonn conservative dose rate vasus distance calmiaHons for cask amty.
83 l
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ArrayCalculations l
a CurmntStatus
-Past calculatinns have assumed that side flux fian inner-rows do not contnhlte to offsite dose rates. ,
-NRC has found that second and third rows in cask array can
~
si cantly contribute to offsite dose rates at close distances ;
m).
l t storage sites have large site boundaries and operate well- i below offsitedoselimits. '
I E SRP Options: l
-NRC ; glans to incorporate guidance in SRPs to address inner-row l mode ing assumptions.
84 1
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emen eseme I 1
IS:SER-ROW SIDEFLUX -
Radial View 0
6 &
9-4
- _ essee ones>
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\
f FollowupActions a Comparison ofShielding Calculations and Measunments
-Description offuelloaded
-Measumnent points and values (survey diagnun)
-Neunons
- Gamma
- Sataming m NeutmnMeasurementsandCalibmtionMethods
-Sointe
-Moderation a Comparison ofSkyshine Calculations and Measuienents
-Neutrons
-he 86 l
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- g e*