ML20126E177
| ML20126E177 | |
| Person / Time | |
|---|---|
| Issue date: | 03/16/1984 |
| From: | Beckwith C NRC OFFICE OF RESOURCE MANAGEMENT (ORM) |
| To: | Dircks W, Rehm T, Solander L Office of Nuclear Reactor Regulation, NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML20126E140 | List: |
| References | |
| FOIA-85-95 NUDOCS 8506150314 | |
| Download: ML20126E177 (13) | |
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f UNITED STATES i
NUCLEAR REGULATCRY COMMISSION
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March 16,1984 MEMORANDUM FOR: Those on Attached List FROM:
C. A. Beckwith Office of Resource Management
SUBJECT:
MINUTES - 33rd DOE-NRC INTERAGENCY MEETING Attached are minutes of the DOE-NRC Interagency Task Force held on Wednesday, March 14, 1984 at Bethesda.
No new items were brought up at this meeting.
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C. A. Beckwith Office of Resource Management
Attachment:
As Stated 8506150314 850320 PDR FDIA AUDIN85-95 PDR p 3-
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Addressees - Memorandum dated March 16, 1984
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,W. Dircks, ED0 T. Rehm, A0EDO L.'Solander, NRR F. Gillespie, RES J. Tomlin, RES A. Muir, IE D. Dambly, ELD
-L. Cooper, ADM J. Holloway, ADM/LFMB W. Besaw, ADM/TIDC-J. Dunleavy, ADM/DS
- J. Beckerley, OPE
.M. Landau, SP H. Faulkner, IP D. Loosley, NMSS P. Baker, NMSS J. Bunting, NMSS C. MacDonald, NMSS C. Heltemes, AE00 A. Newsom, ACRS W. Kerr, SDBU/CR J. Shields, RM/D
'L. Donnelly, RM/B L. Barry, RM
-P. Norry, ADM.
J. Davis, NMSS H. Denton, NRR R. Minogue, RES R. DeYong, IE W. Kerr, SP J. Shea, IP J. Zerbe, OPE G. Cunningham, ELD R. Barber, DOE, PE 221 4
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DOE-NRC COORDINATION MEETIflG
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March 14,1984
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ATTENDEES 00E NRC R. Barber, PE T. Rehm, A0EDO F. Witmer, PE C. Beckwith,.RM S. Ross, MA D. Dambly, ELD G. Redden, CP L. Solander, NRR M. Combs, DP D. Loosley, NMSS-A. Millunzi, NE P. Baker, NMSS L. Cooper, ADM J. Holloway, ADM/LFMB A. Newsom, ACRS 1
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MINUTES 33rd DOE-NRC COORDINATION MEETING March 14, 1984 Agency Lead Item 42 - Nuclear Safety, Research, Development, DOE-Millunzi and Demonstration Act of 1980 NRC-Tomlin This act requires DOE to study the need for a national reactor simulator and an academy to train operators, as well as to identify additional areas for nuclear safety research.
1 c.
Program Plan Update An annual update to the plan was submitted to Congress on January 26, 1984.
NRC had comn,ented on the plan, as agreed.
DOE noted that this is now a recurring annual requirement and should not be closed out.
Action: Place on Continuing Action Items list.
Item 45 - High Level Nuclear Waste Disposal Act DOE-Stein of 1982 NRC-Kearney This act establishes requirements for the disposal of high level commercial nuclear waste.
Status: A DOE Project Office has been set up which reports to the Secretary of Energy.
DOE-NRC working groups are functioning with excellent coordination. An interagency cooperative agreement was signed on June 27, 1983 which laid out procedures for information exchange and staff ir.teraction.
NRC has decided to set up three field offices at DOE locations: the first at Richland, WS, the second at the Nevada operating site, and the third at Battelle, Columbus, Ohio. This is in conjunction with the DOE site offices. DOE support has been requested.
DOE has provided space for the NRC offices as available.
However, additional action is being taken by NRC to establish separate facilities as well.
Action:
Continue to coordinate activities.
Item 51 - License Fee Recovery DOE-Ross
!!RC-Beckwith The rule requiring NRC to collect full costs in the fee recovery process has been completed and sent to the Coninission for final approval.
NRC will require identification of all licensing action costs on future projects on a plant-by-plant, action-by-action basis. The laboratories have been able to implement this requirement.
. Status: A standardized procedure to report this data in the monthly letter status reports has been agreed upon by DOE and is being incorporated into Manual Chapter 1102.
NRC noted the Commission will meet on the topic on Friday, March 16, 1984. Hopefully, a final rule will be approved with little change from the draft document.
Action:
00E to be notified when the rule is approved. It is estimated a May 1, 1984 effective date will be selected.
Item 53 - Changes to 10 CFR Part 20 DOE-Deal (Vallario)
NRC-Tomlin NRC has proposed changes to 10 CFR Part 20.
00E stated that these changes will severely impact DOE programs.
EPA is soliciting guidance to set worker exposure standards.
NRC has recommended a 50 year dose commitment extrapolated from the current exposure rate, but made an exception for long-lived radionuclides such as Plutonium and Uranium.
DOE has recommended against using the dose commitment approach for the reasons stated at the ACRS-NRC-D0E Committee hearings of November,1982.
Eight or nine other issues have been discussed with DOE. Most have been resolved.
Status:
DOE presented its views to the ACRS on September 22, 1983.
The final NRC position has yet to be published for public comment.
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Action:
DOE will be advised on the final rule in time to provide formal comments.
Item 54 - Foreign Nationals at Labs on NRC Work DOE-Gavigan NRC-Tomlin (Faulkner)
It was agreed.o jointly reexamine the policies on assignment, use, clearance, and access to information when foreign nationals are used on NRC work at the labs.
Status:
During the July 1983 meeting, NRC (RES) stated they desire that the operations offices not use foreign nationals on NRC projects without prior NRC approval.
DOE voiced general agreement with this concept.
Action:
Copies of lists of foreign nationals proposed to work on NRC projects are being forwarded to NRC (RES).
Item to be closed.
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. Item 55 - Cooperative Agreements DOE-Witmer (Ross)
NRC and DOE develop a joint procedure for the release of NRC-Savolair.en information on cooperative programs.
(Shomaker)
Status:.The procedure was sent to the DOE field offices for implementation.
In January, SANDIA objected to the procedure, in a Commissirm meeting. The NRC staff has since reconfirmed the procedure with DOE and sent an explanatory policy paper to the Commission.
DOE questioned a recent list of projects to be deleted from the cooperative programs procedures.
Action:
(1) This procedure is being added to tianual Chapter 1102 as a regular NRC-DOE procedure.
(2) RES is to resurvey their projects and reverify the listing provided DOE on March 5, 1984 Item 56 - NRC Use of CPCI:
' DOE-Wagner NRC-Cooper (Dunleavy)
On May 5, 1983, NRC requested DOE to allow the NRC Division of.
Security to process clearances via the DOE system. DOE agreed to this request.
Status: An interagency agreement was written and signed-on March 2, 1984 to implement the process.
Action:
Iten to be classa.
s Item 57 - Changes to 10 CFR 140 00E-Vallario NRC-Saltzman Changes to the criteria for classification of extraordinary
-(Peterson) nuclear occurrences is being proposed by NRC.
This issue effects Price-Anderson issues in which DOE has a vital interest.
DOE has expressed their concerns in writing.
DOE has concerns that this action may necessitate changes to their rules as well.
Status: The pros and cons of both agencies positions were s
discussed at the last meeting.
An NRC paper to the Commission.was supplemented by a second paper noting the DOE concerns and making modifications to the proposed rule. Joint meetings have been held, and NRC-has agreed to change 2 of 3 areas of concern.
' This item is'still under consideration by the Commission.
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Action: A Commission de' cision is needed, after which DOE may' choose to address the issues during the public comment period.
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J Item 58 - NRC Manual Chapter 1102 DOE - Witmer NRC - Beckwith Various changes in procedure for tasking of the laboratories have been agreed upon during the past year.
NRC MC 1102 is being amended accordingly.
This includes cost reporting, changes in equipment reporting,' licensee fee recoupment information, cooperative program reporting, etc.
Status: A joint working group revised the Manual Chapter during the November-January time frame. NRC coordination is now complete, and publication is only awaiting formal 00E concurrence.
DOE noted there are a few comments by their Controller and Office of Defense Programs which should be addressed. Mr.
Beckwith also presented four comments by BNL which were obtained at a visit to the laboratory earlier this week.
Action:
It was agreed the joint working group would meet early during the week of 19-23 March and determine final resolution of the comments by DOE.
Publication can then be made as soon as possible.
New Items:
No new items were presented.
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DOE-NRC CONTINUING ACTIVITIES 1.
Foreign Nationals (Item 2)
' Update listing of current and projected assignments every 6 months.
2.
DOE Institutional-Planning (Item 4)
Complete on an annual, recurring basis.
3.
DOE Nuclear Safety Evalulations (Item 6)
Annual list of DOE requests for NRC safety reviews is to be updated each year.
4.
NRC and DOE Program Updates - Current and Budget Year (Item 7)
Revised NRC program lists, by FIN, are being regularly provided to DOE on a quarterly basis. Both agencies are providing the other with all budget information and submittals as produced.
5.
NRC Evaluation of Laboratories (Item 18)
Annual laboratory evaluations are provided DOE upon request.
A standard format has been agreed upon, as developed by RL and NRC to evaluate PNL.
6.
Safeguards Area (Item 31)
Joint meetings of a users group are being held periodically to continue coordination.
7.
GPP, GPF, and Special Purpose Facilities (Item 38)
This area is reexamined on an annual basis.
8.
NRC Rulemaking (Item 41)
NRC is to advise DOE of proposed rule changes as early as possible before Federal Register publication.
9.
Nuclear Safety, Research, Development, and Demonstration Act (Item 42)
DOE to coordinate their revised program plan each year with NRC prior to submittal to Congress.
- 10. End-of-Year Funding Changes (Item 43)
Close monitoring of year-end funding changes must be accomplished by both agencies on an annual basis.
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"g' U.S. DEPARTMENT OF ENERGY aut 12 su memOrGndum 6Lv TO 4
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. ut c7 Minutes of Meeting with DOT and NRC on OCRWM/ DOE Transportation Issues, June 27, 1984 DISTRIBUTION j
The meeting was held on June 27, 1984 at the Forrestal Building for t'h'e purpose of exchanging information between DOE and NRC on transportation issues.
The agenda and list of attendees are shown in Att.achments 1 and 2 respectively.
4 Bob Baue~r,'As'so[iate'Dir'ector, O'ffic'e of Storage and Systems
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Development, OCRWM, opened the meeting by introducing Ben Rusche,' Director, OCRWM, who reviewed the total systems' approach being knpl.eme.nted for the Waste Managem~ent' program.
He further address'edthe~importan~ce o~f transportat' ion'in the Waste, program ~ a,nd.. _ enc _ouraged.furth.er meetings of. thi.s:'t;ype 'and in maintaining op.e.,n commun;ications'between the.agancies.,
1.
Bob Philpott.d.isEussed..t.he.curr'ent draft of the "Propo' sed
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Transportation Business Plan" and some changes being implemented.
Current planning calls for the document to be completed and.made available -to the gublic.in ~ August..This will be followe,d, by a meeting. i.n. October in Albuquerque.
wit'h industry and other ' interested parties to obtain public input to the plan.
NRC representatives fesponded that it was not clear as to how the Business Plan would be used by DOE'and suggested that such an explanation..be included in the document.
DOT representatives sugge.st~ed that the final,"vbrs. ion of the plan discuss the reasons ~..for'r_e'jecting any 'of the proposed
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l options proposed by meeting p~arti~cipants.
S'e'veia' ' attendees' cautioned DOE to be very explicit in the
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l announcement of the business plan workshop about the issues l
that will be discussed during the session.
By doing so.the public will not be misled about what can be gained from-l attending the workshop and DOE will nbt be faced with addressing issues it is not prepared to contend with.
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2.
Dick Cunningham, NRC, discussed a program the UK has embarked on of running semi-scale-tests of casks to verify the results of engineering tests, followed by a carefully selected full scale test designed to increase public confidence in the equipment., Dick stated that, since.the public~has difficulty in understanding conclusions drawn from engineering tests, DOE might want to consider a similar program.
He suggested DOE pursue this further with
_. _.__.UK officials..
He further -stated that this was his personal proposal and not an NRC position.
NRC would continue to certify equipment based upon engineering tests.
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- 3. 'NRC stated,that their comment.s on this mission plan war,e forthcoming and that they had only one comment on the transportation portion.
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/4.
Bob Philpot,t briefly discussed the DOE P.rogram Research a'nd Development Announcement (PRDA) status.
Responses have a
been received and are being reviewed by DOE.
Selection could be complete by the end of July.
5.
Chuck MacDonald, NRC, discussed the resolution of the Sierra Club petition concerning the hazards of dry storage and transport of spent fuel.
The petition was based on an accident at Battelle caused by oxidation of failed spent fuel during dry transport.
The Sierra Club petition was denied, but new guidance has been issued *concerning spent fuel shipments.
Dry shipments of spen't fuel must be conducted in an inert atmosphere (argon or helium) and 4
failed fuel must be canned.-
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6.
Bill Lahs, NRC, discussed Phase II of the Modal Study being conducted by Lawrence Livermore Labs.
The study will address the level of safety afforded by current regulations.
The report will be,available in approximately i
one year.
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7.
John Cook,'NRC, discussed their June 8 proposal to amend 10 CFR 73, eliminating some of. the interim requirements l
for safeguarding spent fuel shipments.
The proposal would eliminate the need for armed guards and advance approval of routes by NRC.
However there would still be a requirement for unarmed escorts, and prenotification'of states.
NRC expects strong state objection to relaxation of the rules.
DOE is concerned about the retention of the prenotification requirement and suggested that perhaps it could be moved to another section of the regulations.
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V8.
George Allen, Transportation Technology Center, (Sandia Labs) discussed work being done on the new generation of casks.
Westinghouse Hanford is developing transportation interface criteria which will help to enhance the standardization of transportation system
- features.
GA Technologies is developing reference packaging descriptions to demonstrate what casks mi,ght look like'in the future.
Included are a 25 ton legal weight
" truck cask,~a 40-ton intermodal cask, a 75 ton rail cask, and a 100 ton rail cask.
9.
At the close 'of the meeting the.following announcements were made concerning -further activities:
The Salt Repository Project Office (SRPO) will hold a' meeting in Augu'stin tolumbus, Ohio with the states in which the potentlal. repository. sites.are. located.
SRPO has
. requested that OCRWM provide representatives to discuss transportation'is'sves. : Bob Philpott suggested that DOT and NRC might want to participate also.
He will provide DOT
, and NRC copiasi ~of ~:some *of the questions the states ar,e asking about. transportation..
It'was suggested that_ FEMA be invited to future meetings.
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- The' next meetings willibe held -prior. to the Bu.siness Plan I
Workshop.
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Chief, Transportation Branch Office of Civilian Radioactive Waste. Management Attachments
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ATTENDEES LIST NAlf ORGANIZATION PHONE Paul Grim DOE-HQ-DP122.2 353-3137 Steve Schneider 00E-HQ-RW-22 353-5545 Pete Bolton Weston 963-6844 Richland Shanklin Weston 963-6853 C.E. MacDonald NRC 427-4122 NRC-RES 443-7874 William Lahs
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John R. Cook.
NRC (55-396) 427-4155 R.E. Cunningham NRC NMSS 427-4485 252-9433 '
' 7 R.E". Philpot-DOE OCSWM~ -'
755-1260
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Lee 5intiaan DOT
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Woody Chu DOT 472-2698
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Dick Mannon DOT 472-2698 Rick Rawl DOT 426-2313 Lynne Fairobent SAI/ DOE-Nevad 556-7357-George ~ Allen SNL/TTC 505-844-5577
- -.j 'Chirli,eMaucF,{~
DOCOCR9M 252-2419 :-_
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AGENDA
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I MEETING ON TRANSPORTATION -PACKAGING ISSUES DOE CONFERENCE ROOM 4E-06 9 FORRESTAL BUILDING 1000. INDEPENDENCE AVENUE, S.W.
WASHINGTON, D.C.
20585
. 7......
...... WEDNESDAY, JUNE 27, 1984 9:00 - 12:00 AM 1.
OCRWM Transportation Business Plan.........................
... DOE 2.
Preliminary Thoughts on Scale Tests for New Packaging..........NRC J
S.
Mission Plan-for 'the Civilian Radioactive Was Management Plan..............................te
.NRC 4.
DOE Program Research and Development Announcement - Status..... DOE
.5.
Resolution of Issue Regarding Oxidation of S Transport.................................w. pent Fuel During
.NRC 6.
Model Study - Update.........
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.NRC
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Safeguard Requirement'for Spent Fuel' Shipments.................NRC 8.
Plans to Update Environmental Study...............~.............NRC l
9.
New Generation Cask' Performance Study Update...................TTC t
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UNITED STATES b*.
NUCLEAR REGULATORY COMMisstON f.fli
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'JAN 141985 FCTC: RHO 71-6639 MEMORAND.UM FOR:
The Files FROM:
Richard H. Odegaarden, FCTC, NMSS
SUBJECT:
SUMMARY
OF MEETING WITH DOE CONCERNING MODEL NO. MH-1A APPLICATION DATED SEPTMEBER 7,1982 Attendees DOE Sandia NRC Charles Mauck Bill Wowak C. E. MacDonald*
Barry Smith W. H. Lake Liz Roybal H. W. Lee C. R. Marotta REA
.C. E.' Williams R.H. Odegaarden '
Herman L. Crawford
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- part-time l
Introduction A meeting was held at Silver Spring,(Maryland on JanuaryDOE) to discuss resp 10,1985 at the f[
request of the Department of Energy request for additional infonnation dated February 24,1983.
a Dis ~cussion DOE has addressed all items identified in our letter of February 24, 1983. A sumary of the DOE response (enclosed) was briefly discussed.
As-built drawings and procedures for' conducting leak tests prior to shipment should be included in the application. A detailed review on the adequacy of the response is pending formal receipt of the safety analysis report for the package'. DOE expects to submit their response in February 1985.
It is expected that our review would be completed within 60 days.
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I Richard H. Odegaarden, Project Manager Transportation Certification Branch Division of Fuel Cycle and Material Safety, NMSS
Enclosure:
As stated t
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ANSWERS TO NRC QUESTIONS OF FEBRUARY 24, 1983. FOR MH-1A CASK I
The MH 1A safety analysis report (SAR) has been revised to reflect the change in format since the SAR was first revised by REA.
Hence, the chapter numbers have been increased by one, eliminating Chapter D.
The SAR has also been revised to reflect the latest edition of 10 CFR, Part 71, effective September 6, 1983, i
In the previous SAR submittal to NRC, only one type of HFBR basket (two tier) was included. Since that submittal, the need arose. for a long basket for uncut HFBR fuel assemblies. A description of this basket and fuel has been integrated throughout the SAR, and an additional criticality analysis has been perfomed as indicated in Chapter 6.
The two types of HFBR fuel assemblies are referred to as "long" and "short" t
HFBR fuel.
For completeness, the NRC questions are repeated herein, and the answer or a brief coment to each question is indented following each question. The answers have been incorporated into the SAR, and the SAR references are included with the answers.
DRAWINGS
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1.
Drawing Nos.1038-1 and 1038-2 do not clearly show the pertinent safety details for trunnions, tie-down lugs, valve boxes, and stiffengrs. Provide drawings to show details of the construction, e.g., dimensions, sizes, locations, weldment types and dimensions, etc.
The MH-1A drawings were revised to reflect the cask as-built condition.
In some instances, for example, the tie-down lugs, additional measurements were perfonned on the casks to obtain the necessary infomation. The REA drawings that were revised are a.
1038-1, Rev. G, Sheet 1 of 1, MH-1A Sectional Views b.
1038-2, Rev. E, Sheet 1 of 1. MH-1A Shipping Cask Assembly c.
1038-9, Rev. 8, Sheet 1 of 2, Material List d.
1038-9, Rev. F Sheet 2 of 2, Mat'e' rial List A further description of cask details of the cover weldment, trunnions, the two valve boxes, and heat transfer fins, which also are structural stiffeners, are presented in a new drawing, REA Drawing ~1038-17 Revision 0.
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As a result of NRC Cuestion 1 in the Themal Section, a fire shield will surrount the bo# of the cask in the fin region. The fire shield is shown in a new drawing REA Drawing 1038-18, t
ce Revision 0.
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The revised drawings are presented in Chapter 2, Section 2.11.2, in the SAR with the other cask drawings.
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2.
Revise the drawings and the material list to reflect the revised.
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. pressure relief valve.
As a result of NRC Question 3 in the Containment Section, the i
pressure relief valve, globe valve, pressure gauge, and associated plumbing have been eliminated from the cask. A luted pipe plug is to be installed in each of the vent and drain ports in the cask.
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' The material list, REA Drawing 1038-9 Revision F Sheet 2 of 2 indicates that during Revision D, Items 50 through 58 were removed from the drawing. These were the valves and plumbing named previously.
REA Drawing 1038-1, Revision G, Sheet 1 of 1, indicates that during Revision E, the aforementioned Items 50 through 58 were removed. A note in Zone E7 on this drawing refers to a new
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drawing, REA Drawing 1038-16 Revision 0, MH-1A Shipping Cask Assembly Modification. Top and Bottom Gauge-Box Housing, which shows a pipe plug installed in a cask port and the o-ring has been removed. Therefore, the only organic seal in the cask is the 0-ring in the top cover.
The removal of the valves and plumbing from the cask.has been reflected in all chapters of the SAR.
b 3.
The fuel basket drawings should clearly state the Al cladding and l
B C-Al mat.rix thicknesses, matr.ix density, and <cmposition.
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REA drawings show the following infomation on the Boral" and cladding in the b'askets:
a.
304 stainless steel. cladding,16 gauge (0.059 inch) thick, over Boral" b.
Overall Boral" thickness of 0.265 inch c.
0.040-inch-thick 1100 alloy aluminum sheet Boral" matrix cladding 0.185-inch-thick'B -C aluminum matrix (also 11g alloy d.
h not shown on drawing)2,ith 35 percent B -C and a B 4
loading of 0.045 g/cm
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The following REA basket ' drawings show the aforementioned infomation:
a.
1038-6, Rev. D, Sheet 1 of 2. Zone H5, NBSR Basket b.
1038-7, Rev. E, Sheet 1 of 2. Zone B4, Short HFBR Basket c.
1038-15 Rev. C. Sheet 1 of 2, Zone D1, Long HFBR Basket,
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d.
1038-3, Rev. C, Sheet 2 of 2 Zone A7 MURR Basket The Boral" as described will be used to fabricate the baskets
3 because of availablility of material, but the criticality b
analysis, which was completed many months ago, assumes that the Boral" thickness is 0.177 inch. The use of thicker Boral" is conservative; _ therefore, the baskets were not reanalyzed for.
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cr.iticality.
Instead, a note was inserted in the criticality section acknowledging that the actual Boral" thickness will be greater than that assumed in the analysis, and that the analysis was -not reworked because of conservatism.
In all other chapters of the SAR, the 0.265-inch thickness is discussed.
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STRUCTURAL i
General 1.
Revise the analyses foe the fuel baskets taking into consideration the following:
MURR Fuel Basket Intemediate Double Bend Plate will not be able to support the web plate force in a side drop condition (water gap, assumed in the
' criticality analysis, will be lost during the side drop).
.The MURR criticality was reanalyzed assuming no water gap. This resulted in an increase of k from 0.820 to 0.845. This result is presented in the SMfin Chapter 6 Table 6.1 ano in
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Section 6.1, Discussion and Results.
HFBR and NBSR Basket t
The 0.25-inch-diameter bottom rod will not be adequate to support the fuel in the basket for the bottom drop condition.
The previous basket drawings showed two 0.25-inch-diameter rods below each fuel assembly. This number has been increased to.five (three additional rods have been added under each fuel assembly).
In actuality, a rod extends continuously across a basket, which spans four fuel assembly compartments. Therefore, a total of 20 rods are installed as indicated in the following REA drawings:
s.
1038-6, Rev. D, Sheet 1 of 2, Zone F3, NBSR Basket b.
1038-7, Rev. E. Sheet 1 of 2. Zone F2, Short HFBR Basket c.
1038-15 Rev. C, Sheet 1 of 2 Zone F3 Long WBR Basket
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The analysis of the rods during the end drop is presented in Section 2.9.3.1 in the SAR.
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2.
The analysis of the tie-downs on page 1-24 is not valid and should
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be revised because the tie-down cable forces do not satisfy the basic equilibrium condition (i.e., balance the applied load).
It is also N-
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noted that the weldme'nt size for the tie-down lufpage 1-26).
s shown on Drawing No.
1038-2 is not consistent with the analysis Analysis' a'nd drawing have been revised.
3.
Show that the cask closure system, including the flange at the top of the cask wall, has adequate regidity to maintain a seal under the accident conditions and under excessive tie-down forces (10 CFR 71.31(d)(3)).
A detailed stress analysis presented in the Appendix, Sections 2.11-6, shows stress values for.the tie-down and drop accidents.
4.
For the external pressure analysis (page 1-34), justify the
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. assumption that the end plate can be treated as fixed edge circular as in reality the end plate is neither circular nor fixed. Also, equations 1.39 and 1.40 are not applicable since they have ignored the forces at the edges where the flat sides meet the curved corner (axial, shear and moment). Bendin2 moment effects at the corner should be considered.
The end plate in question is the top flange, which is welded to the containment vessel and outer shell. Usually, welded connections are considered fixed. To answer the circular vs. rectangular plate issue, a finite ~ element analysis was l (.)
perfomed, which can accommodate generalized geometric shapes.
j 5.
The themal differential expansion analysis presented on page 1-39 '
l 1s not complete. The radial and axial pressure due to lead contraction i
on the containment vessel should be evaluated. Also, justify the assumption that the bolted edge 'of the cask cover is fixed boundary (page 1-41).
Provide an analysis to show how the edge moments are resisted by the bolt.
The resistance of a single bolt ring to resist moments in bolted connections are well recognized. For example, the ASME Boiler and j
Pressure Vessel Code,Section III, Division I, Appendices, l
Articles A-2000 through A-5000, and M. Holmes, Analysis and Design of Structural Connect' ons, John Wiley (1983), Page 145, analyze for the bending moments at bolted connections. The bolts, themselves, usually do not resist the moment, but they are in tension instead. The moment resistance is derived from the prying action of the edge of the flange. The tensile force of the bolt and the resisting force at the edge of the flange fom a force couple, and the distance between the couple is the distance from the bolt to the edge of the flange.
The analysis in the Appendix, Section 2.11.6, considers the ' bolted connection in detail. The radial stress due to the lead is analyzed in Section 2.11.6.3. The axial stress is calculated in Section 2.6.1.2.
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6.
Under the nomal condition, the contaiment vessel temperature is 202 F and the outer shell taperature is 80 F (page 1-39). For the
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accident condition, the contaiment vessel temperature is 390 F and the outer shell temperature is 560 F (p' ge 1-70). Temperature differential a
across the cask wall should be based on temperature difference between the two shells. Provide an analysis to evaluate the temperature differential stresses in the cask for the nomal and accident conditions.
(Themal analysis needs to be revised, see THERMAL, Item 1).
The themal analysis was reanalyzed, because of incorporation of a l
fire shell., Themal stresses were calculated.
4 7.
The vibration analysis of the cask should be revised ~to consider:
(a)'the effects of the cask cavity, lead and impact limiter; (b) effects on cask attachments such as valves (c) eff2 cts on fuel baskets i
and fuel elements.
The effect of each component was considered separately. The analysis is presented in Section 2.6.5.
8.
Show that the containment vessel will not buckle under the combined loads of axial, bending, and external pressure including those from i
. lead contraction.
Buckling analyses are presented in Sections 2.7.1.1.3 and 2.7.1.2.4.
30-Foot End Drop (1.7.1.1) 1.
Show the derivation'of the effec'tive impact area of 1550 in2 l
used in Equation 1.5.
The areas on top of the cask have been recalculated, and the value
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is 1416 square inches, as shown in Section 2.7.1.1.1 in the SAR.
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- 2.. Justify the adequacy of the use of the redwood crush strength i
derived from static test.
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Static redwood crushing data is equivalent,to4lynamic redwood crushing data, as Sandia detemined by static and dfnamic test documented in Sandia Report SAND 74-0010. Page 59 therein, and 4
Exhibit 2.11-2, letter from W. W. Joseph, Section 2.11.3.4.1.4 in the SAR.
2 The redwood crushing data used for analysis in the SAR came from REA tests on samples of the redwood used in the MH-1A impact limiters. In all instances, the curve value of crushing stress at the appropriate strain was used, instead of an average value. To estimate the maximum crushing distance of an impact limiter, the average curve value of crushing stress minus 15 percent for l
uncertainty was used. To estimate forces on the cask, the greatest crushing stress values were used at the end of crushing, I
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and a band width of +15 percent from the average curve value of l
crushing stress was considered. The analysis is presented in Section 2.7.1 in the SAR.
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3.
The redwood grain orientation is inconsistent. As shown on Drawing N
No. 1038-1, the redwood would be crushed parallel to grain instead of perpendicular to grain as' assumed in the analysis.
The drawing has been revised to be consistent with the analysis.
Also, detailed drawings of the bottom and top impact limiters have l
been included in the SAR for clarity. These drawings are REA Drawings 1038-10 and 1038-11, respectively.
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4.
In dynamic test, the redwood generally bottomed out at 501-60%
compression. Justify the adequacy of the current design which is based I
on redwood compressibility 6.27/9.5 x 100% = 665.
l The compression data for the redwood used in the impact limiters,
are used in the analysis. Bottoming does not occur at a sharply defined deflection.
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5.
The corrugated steel cylinders in the bottom impact limiter may have significant effects on the perfomance of the impact limiter.
i Justification for ignoring the cylinders should be provided.
I In the reanalysis of the bottom impact limiter, the corrugated r
cylinders were not ignored. Inste:d, the perfomance'of the cylinders was evaluated and included in the overall perfomance of 3
.i the bottom impact limiter. The analysis is presented in Sections l[
2.6.4 and 2.7.1.1.2 for the bottom end drop.
30-Foot Side / Corner Drop'(1.7.1.2 and 1.7.1.3)
I 1.
Since the design is based on mixed redwood grain orientations l
(e.g., perpendicular to grain at the flat side, parallel to grain on the corner), detailed drawing'sh'owing redwood grain orientations should be provided.
Redwood properties and construction procedures should also be specified on the drawings.
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The detailed impact limiter drawings are included in the SAR.
The'se drawings are REA Dra' wings 1038-10 and 1038-11. The PAT-2 s
specification, which is the one used by Sandia, is included on the impact limiter drawings.
2.
The portion of the redwood which is not supported by the cask should not be deemed effective in absorbing energy (Figure 1.14, page i
1-54).
The 's reading of stress in a material is well documented, for examp e, Spangler, M. G., Soil Engineering, 3rd ed., Intext (1973, pages 383-385. Thus, ettaer the outside or the inside of 4
en impact limiter may be crushed, depending on where the smaller area is located., This is discussed in Section 2.7.1.2.1, and photographs of drops in SAND 74-0010 substantiate the spreading of stress and the crushing of inside or outside surfaces of the
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1mpact limiters.
3.
The average redwood crush strength used in the corner drop (page
7.
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1-60)" was not justified.
The redwood crushing data used for analysis in the SAR came from l
REA tests on samples of the redwood used in the f#l-1A impact 11spiters.
In all instances, the curve value of crushing stress at the appropriate strain was used, instead of an average value. To estimate the maximum crushing distance of an impact limiter, 'the average curve value of crushing stress minus 15 percent for i
uncertainty was used. To estimate forces on the cask, the greatest crushing stress values were used at the end of crushing, j
and a band width of +15 percent from the average curve value of l
crushing stress was considered. The analysis is presented in Section 2.7.1.3 in the SAR.
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4.
It is not clear how the impact limiter crush angles "a" and "b" were detemined. Provide the derivations of Equations 1.61 and 1.63.
The drop accident equations and analysis are explained in Sections i
2.7.1 for each type of drop.
5.
The rationale that the redwood crush strength varies according to the cosine of the. cask rotation angle (page 1-64) was not adequately l
' justified.
,'i The cosine of the angle was not used in the reanalysis of the
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impact limiters.
Instead, the exact curve for crushing wood at
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.Various angles with respect to the parallel-grain directinn was extracted from JPL Report 32-120 (January 13,1961), Page 7.
6.
Provide an analysis of the cask for 30-foot drop on the top corner.
Provide a free body diagram of the impacted cor.ner to show all forces acting in that corner and equilibrium condition is satisfied.
The ANSYS input data shows the locations and directions of forces applied to the corner of the ca'sk (Section 2.11.6.5.1, Table 2.11-24).
7.
The impact limiter is a square with rounded corners not a circle as assumed in the analysis. Also, the redwood grain orientations are different at the flat portions and the corners. Assess the difference for the cask to be impacted on the corner of the flat portion and the rounded corner.
A detailed analysis of impact on a corner with a flat side is presented in Section 2.11.6.4.
Since a cylinder (rounded edge and side of the cask) is stiffer than a flat side, the rounded corner can withstand the corner drop accident with less deflection than l
that for the flat side.
8.
It is noted that the lifting trunnions, tie-don lugs and valve boxes are not fully covered by the impact limiter. Provide an analysis 4
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for the cask impact directly on these attachments.
Impact was assumed on these components, and the loss of wood in
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these regions was considered in the analysis in Section 2.7.1.,
40-Inch Puncture. Test (1.7.1) 1.
Provide a'n analysis for puncture cirectly on the valve box cover.
Equation 1.69 is not applicable because the cover plate on the valve box does not have lead backing.
The valves and piping were removed fran the valve boxes.. and these were replaced by pipe plugs flush with the surface of the cask.
Therefore, it is not necessary to show that the valve box is not damaged. The existence of the valve boxes makes puncture on these locations less severe than puncture at other places on the outer shell, because of the increased metal thickness under the puncture pin. Also, puncture on the valve boxes is less severe than in the center of the cask, because the valve boxes are near the ends of the cask and the cask tends to rotate when struck with the
' puncture pin. This is discussed in Section 2.7.2.
2.
The package should be evaluated for " beam like" stresses due to puncture test.
In the revised SAR,'the cask is analyzed for beam-like stresses for the puncture accident, as presented in Section 2.7.2.
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THERMAL
. 1.
Our independent themal analysis (HEATING, NUREG/CR-0200),
indicates considerably more lead melt than.is reported in the application. Provide a themal protection system to preclude lead' melt for the 1/2-hour fire test.
A fire shell has been designed for the cask.
Consequently, the cask has been remodeled and reanalyzed. The results are presented in Chapter 3.
Lead does not melt with the fire shield 4. stalled.
2.
.Since the cask is loaded under water, residual water may remain in the closed cask. Revise the estimates of maximum pressure for nomal and accident conditions by including the partial pressure of water.
The operating procedures have been revised to test for water vapor with a 5 m Hg. vacuum. The analysis of the operation of this procedure is presented in Section 7.5.3 in the SAR. The water vapor pressure has been estimated and included in the total pressure in the cask.
-CONTAINMENT i
1.
Identify and provide specifications for the new pressure relief l
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valve (3.1.2.2) that will replace the bronze valve described in the L /
text and shown on the drawings.
As a result of Question 3 in this.section, all valves and plumbing, including the pressure relief valve, were removed from the cask. Therefore, this question in not applicable now.
I 2.
Revise Paragraph 3.2.1 to reflect the actual leak test to be used.
This paragraph, now called Paragraph 4.2.1, has been revised to l
show the required sensitivity required. From this value, the cask user can select an appropriate leak test from ANSI N14.5.
4 3.
Show that the pressure gauge does not violate containment for l
nomal and accident conditions of transport.
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The pressure gage and plumbing have been removed from the cask, rather than furnish an analysis. This change is reflected throughout the SAR.
4.
The description of the valve arrangement on the vent line, as it is given in several sections, is not consistent. ' The description given on page 0-4 agrees with Drawing Nos. 1038-1, Rev. A and 1038-9, Rev.
0.
The description given on pages 3-1 and 3-3 are inconsistent and confusing.
As a result of the action taken as discussed in the previous question, Question 4 is not applicable now.
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5.
The metal filter on the pressure relief valve should be deleted to I
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-,,,, -,-_.. -n-. - - - -, - - - - -
r,_w---
,-_--,--,a,-weewmmr
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preclude the possibility of plugging and not pemitting the valve to -
function as anticipated. The filter appears to serve no useful function.
l As' a result of the action taken as discussed in Question 3 in this section, the valves, plumbing, and filter have been removed from '
j the cask. Therefore, this question is not applicable now.
i CRITICALITY 1.
Revise Tables 4.2 (now 5.2) and 5.2 (now 6.2) to limit the HFBR l
fuel' to 300 grams of U-235/ element rather than the 351 grams used in the tables. This revision is necessary because the number density (Table 5.3, now 6.3) and geometry (Figures 5.11 and 5.12, now 6.11 and i
6.12) used in the criticality analysis correspond to 300 grams of U-235/ element.
i l
The desired U-235 loading is 351 grams per fuel assembly, and the
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number densities have been corrected consistently through the criticality analysis, specifically in Table 6.3.
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- 2. ' Resolve the difference in the length of the HFBR element between Tables 4.2 (now 5.2) (22 inches) and 5.2 (now 6.2) (22.75 inches).
i
.J The correct length is 22.75 inches as used in the criticality l
analysis. The use of a shorter length (22 inches) in the
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shielding analysis is conservative Lgreater source density which causes dose rate values to be,redicted higher than actual l.
Therefore, the values remain as presented in the SAR.
A paragraph explanation, as presented herein, is included in Section 5.3, Model Specification, in the SAR.
OPERATING PROCEDURES 1.
paragraph 6.1 (now 7.1) should be modified to require a visual inspection of the Boral" plates in the fuel baskets for the NBSR. HF8R, and MURR fuel assemblies for each shipment. If swelling of the neutron poison plates is observed, the basket should be taken out of service.
Section 7.1, procedure Number 19. for loading NBSR. HFBR, and MURR fuels, requires visual inspection of the Boral".
If bulging is observed, the cause of the bulge must be detemined and repaired.
A test for the presence of Boral" is required for final inspection.
2.
The appropriate corrections should be made in the referencing of previous or subsequent procedural steps in Tables 6.2 through 6.5.
To avoid confusion, these tables have been eliminated, and complete procedures are presented for typical fuel in each of the.
types of baskets to be shipped.
3.
On page 6-15 Step s, clarify the instruction to connect the water
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supply to the relief valve.' This instruction is not compatible with the Lunkenheimer relief valve.
Because of NRC Question 3, Containment, the valves have been removed from the cask. Therefore, this question is not applicable now.
4.
Expand Step 1.b., Page 6-20, to include non-destructive examination
- of welds which make up the steel envelope that contains the'Boral" pl ate. Inspection of the welds is necessary to preclude in-leakage of water which could subsequently result in bowing or rupture of the plates. This inspection should also be included in Section 7.0 (now 8.0), Acceptance Tests and Maintenance Program.
The MURR, long and short WBR, and NBSR basket drawings refer to notes on the material list. Note 7 on REA Drawing 1038-9 covers liquid penetrant inspection of all new welds, which include the new baskets, except the MH-1A basket, which is existing.
Therefore, the new basket welds will be liquid penetrant inspected to ensure that water does not leak between the Boral" plates and stainless steel cladding.
In Section 7.5.2, Step b (the same place NRC questioned),
additional procedures have been included for visual inspection of m
the stainless steel cladding during the annual inspection. A visual inspection is also included before each fuel loading (NRC
(
Question 1, Operating Procedures). Therefore, if leakage occurs, the damaged basket will be taken out of service until the leak has been repaired and the presence of Boral" assured by testing.
l Moreover, liquid penetrant inspection of the small basket spaces is believed to be impossible due to the size and length of the cavity that must be inspected, but the visual inspections and test i
I for Boral" discussed previously will ensure the presence of I-critical control material.
5.
There is no instruction to remove the bolts holding the bottan impact limiter to the cask.
If the bottom impact limiter stays with the cask during the loading operation, describe the method used to l
remove contaminated pool water which can collect at the interface of the cask and impact limiter.
The bottom impact limiter is not intended to be placed in the pool. Specific instruction to remove the bottom impact limiter is contained in Steps 6 and g in the revised package loading procedures for each type of fuel.
Either delete from the Operating Procedures the option of lifting a 6.
fuel basket containing irradiated fuel assemblies into and/or out of the cask or provide safety and structural analyses demonstrating the
.4 operation can be safely accomplished.
The option has been deleted from the procedures. Loading the cask is an on-site function and not a transportation function. The option was inadvertently included in the previous version of the
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12 1.
SAR, but it has been removed from this revision.
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ACCEPTANCE TESTS AND MAINTENANCE PROGRAMS The. program should give a specific replacement schedule for all 1.
' seal s.
The seals consist of a silicone o-ring in the top of the cask ~ body i
which seals the cover and two luted p pe plugs, one each.in the vent and drain ports. Luting material is placed on the threads of the plugs every time they are installed (every shipment) as
. required in the operating procedures. Step 38, in the 9
one-basket-section procedures, later in the multi-section basket The 0-ring is to be replaced annually..Section 8.2.1, procedures.
or as needed according to physical damage during use, Section 7.1,
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Step 17.
The program should consider the need to perform a periodic bench 2.test of the pressure relief valve to verify pressure setting.
The pressure relief valve has been removed from the cask.
Therefore, the question is now not applicable.
l IAEA APPROVAL If the package will be used for foreign shipments (Section 0.1 Page
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0-2) provide a section in the Safety Analysis Report showing all the requirements of Safety Series No. 6. Regulations for the Safe Transport of Radioactive Materials,1973 Revised Edition (as Amended), are met.
The reference to the shipment of fuel from foreign countries to the United States has been deleted. Only shipments within the contiguous United States are planned.
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