ML19209A652
| ML19209A652 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/14/1979 |
| From: | Widner W GEORGIA POWER CO. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 7910050111 | |
| Download: ML19209A652 (3) | |
Text
g Georgia Power Company e
230 Peachtree Street 8
Post Office Box 4545
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Atlanta. Georgia 30302 Telephone 404 522-6060
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g Power Generation Department September 14, 1979 Georgia Power i9 MP \\1 nl0 : 3 y
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United States Nuclear Regulatory Commission REFERENCE Office of Inspection and Enforcement RII: JP0 Region II - Suite 3100 50-321/50-366 101 Marietta Street IE Bulletin 79-14 Atlanta, Georgia 30303 ATTENTION: Mr. James P. O'Reilly Gentlemen:
Georgia Power Company hereby submits the following information in response to your letters of July 2, July 27, August 15, and September 7, 1979, which requested verification of the seismic analysis of safety-related piping systems. This documentation package is submitted at this time pursuant to an agreement reached in a telephone conversation with Mr. H. C. Dance, NRC Section Chief, Region II as noted in our letter of August 28, 1979.
Included in this package are:
(1) The sixty day report for Unit I which notes thr.'. t? non-conformances have been identified and assessed to be deviatiots of no significance to plant operations or seismic qualification, ant 143 pack-ages which require further review and assessment.
Individual and collective assessments of deviation significance for the balance of these items is scheduled for completion by the end of October 1979.
(2) The sixty day report for Unit 2 which notes that the assessment of most of the nonconformances is complete. Those completed to date have revealed no desiations of significance to plant operation or seismic qualification, and are included in this submittal for your review.
There are, however, a number of nonconformances in the General Electric Company's scope of supply which have yet to be fully analyzed. These items have been the subject of an abbreviated study by General Electric Company.
On the basis of this examination, and the more extensive inspection and analyses of the as-built condition of this Unit 2 piping made in conjunction with plant construction and testing, it is concluded, as an engineering judgement, that the noted nonconformances will not affect the operability of the main stea= and recirculating systems of the General Electric Company's scope of supply. These remaining items on Unit 2 are scheduled for com-pletion by the end of October 1979 within 120 days of the issuance of IE Bulletin 79-14.
I109 291 7 0401 7910050///
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r Georgia Power A Mr. James P. O'Reilly United States Nuclear Regulatory Commission September 14, 1979 Page Two If you have any questions or comments in this regard, please contact my office.
Yours very truly, i
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W. A. Widner Manager of Nuclear Operations Enclosure WEB /wb xc: Director of the Office of Inspection and Enforcement Director of the Division of Operating Reactors, Office of Nuclear Reactor Regulation 1109 292
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i EDWIN I. HATCH NUCLEAR PLANT - UNIT 1 SIXTY (60) DAY REPORT FOR I. E. BULLETIN # 79-14 INTRODUCTIO3:
As was stated in the thirty day report for Edwin I. Hatch Nuclear Plant-Unit 1 and 2, a physical surveillance of Unit 1 was conducted and the results of this surveillance are being reviewed by Bechtel Power Corporation stress analysts. The follow!ng text is a report of the progress to date, and the schedule for coupletion of the stress ana:ysis review.
DISCUSSION:
The physical plant surveillance program was conducted in accordance with plant procedure HNP-1-10124, and included the seismic category 1 piping systems as marked up on surveillance piping isometric oravings with the following exception:
a)
Submerged main steam relief valve discharge piping was not inspected because inspection would have required drainage, disposal and replenishment of approximately 750,000 gallona of water.
Total packaget generated during surveillance 154 Deviation listing complete 126 Checking of Deviation listings complete 34 Deviations reviewed by stress engineer 11 and dispositions generated (engineering judgement)
Individual and collective assessment of deviation significance for eleven (11) packages vers, performed by qualified stress analysts to determine whether or not the operability of the system might be jeopardized durit.g a safe shutdown earthquake as defined in the requirements.
It was found that deviations are insignificant and system operability is not in jeopardy.
SCHEDULE:
Individual and collective assessment of deviation significance for the rest of the packages will be completed within 120 days (i.e., October 29, 1979).
Our 120 Days Report will include complete listing of all deviations and dispositions.
1109 293
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POWER AUTHORITY OF THE STATE OF NEW YORK INDIAN POINT NO. 3 NUCLE'AR POWER PLANT
.~. O. BOX 215 BUCHANnN. N. Y.10511 TEL EPHO N E: 914 739-8200 September 10, 1979 IP-WDH-5578 Docket No. 50-286 License No. DPR-64 Mr. Boyce H. Grier, Director Office of Inspection and Enforcement Region I U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 I.L. Bulletin No.79-05C and 79-06C
Dear Mr. Grier:
Enclosed is our detailed response to Items in Bulletin 79-05C and 79-06C.
Very truly yours, L
J.
Bayne Resident Manager WDH/rbb Attachement cc:
Office of Inspection ar.d Enforcement U.S. Nuclear Regulatory Commission Division of Reactor Operations Inspection Washington, D. C. 20555 Officu of ';uclear Reactor Regul ation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 1109 294 F
79100so';h.3 I
ATTACIDiENT Short-Term Actions 1.
In the interim, until the design change required by the long-term actions of this Bulletin has been incorporated, institute the fol-lowing actions at your facilities:
A.
Upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, immediately trip all operating RCPs.
RESPONSE
A.
It was determined that immediate tripping of the RCPs af ter a reactor trip and HPI initiation should not be accomplished unless the results of the Westinghouse LOCA analyses indicated that this was the best course of action.
This completed analysis, WCAP-9600, which is referred to in Item 2, indicares that it is not required to trip the RCPs until the reactor coolant system pressure is re-duced to 1250 psia.
B.
Provide two licensed operators in the control room at all times during operation to accomplish this action and other immediate and followup actions required during such an occurrence.
For facilities with dual control rooms, a total of three licensed operators in the dual control room at all times meets the require-ments of this Bulletin.
RESPONSE
B.
Present plant administrative procedures require two licensed
. operators to man the control room post during normal power operation.
1109 295
a 2.
Perform and submit a report of LOCA analyses for your plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip.
For each pair of values of the parameters, deter-mine the peak cladding temperature (PCT) which results. The range of values for each parameter must be wide enough to assure that the maximum PCT or, if appropriate, the region containing PCTs greater than 2200 degrees F is identified.
RESPONSE
2.
A series of Loss of Coolant Accident (LOCA) analyses for a range of break sizes and a range of time lapses between initiation of break and pump trip applicable to the 2, 3 and 4 loop plants has been performed by the Westinghouse Owners' Group.
A report summarizing the results of the analysis of delayed Reactor Coolant Pump trip during small loss of coolant accidents for Westinghouse and NSSS, will be submitted to Mr. D. F. Ross by Mr. Cordell Reed on August 31, 1979.
In the report, maximum PCT's for each break size considered and pump shutoff times have been provided.
The report concludes that if the reactor coolant pumps are tripped prior to the reactor coolant system pressure reaching 1250 psia, the resulting peak clad tempera-tures are less than or equal to those reported in the FSAR.
In addition, it is shown that there is a finite range of break sizes and RCP trip times in all cases 10 winutes or later which will re-sult in PCT's in excess of 22000 F es calculated with conservative Appendix K models.
The oeprator in any event would have at least 10 minutes to trip the RCP's following a small break LOCA, especially in light of the conservatisms in the calculations.
This is appropriate for manual rather than automatic action, based on the guidelines for termination of RCP operation presented in WCAP-9600.
3.
Based on the analyses done under Item 2 above, develcp new guidelines for operator action, for both LOCA and non-LOCA transients, that take into account the impact of RCP trip requirements.
For Babcock &
Wilcox designed reactors, such guidelines should include appropriate requirements to fill the steam generators to a higher level, following RCP trip, to promote natural circulation flow.
RESPONSE
3.
The Westinghcuse Owners' Group has developed guidelines which were submitted to the NRC in Section 6 and Appendix A of WCAP 9600.
The analysis provided as the response to item 2 are consistent with the guidelines in WCAP 9600. All applicable procedures concerning tripping of RCP's at 1250 psia will be revised during the upcoming refueling outage beginning September 15, 1979.
Ii09 296 4.
Revise emergency procedures and train all licensed reactor operators and senior reactor operators based on the guidelines developed under Item 3 above.
RESPONSE
4.
The Owners' Group effort to rcvise emergency procedures covers many issues, including operation of the Reactor Coolant Pumps. The action taken in response to Item 1 is sufficient as an interin measure and no immediate need exists for changing our emergency procedures to in-clude the tripping of the Reactor Coolant Pumps. The expected schedule for revising the LOCA, steamline break and steam generator tube rupture emergency procedures is the following:
Mid-October:
Guidelines which have been reviewed by the NRC will be provided to each ucility.
Appropriate utility personnel associated with writing pro-cedures will meet with the Owners' Group Sub-committee on Procedures and Westinghouse to pro-vide the background for revising their emergency procedures.
January 1, 1980: Plant specific procedures will be revised.
March 1, 1980:
Revised procedures will be implemented and operators trained.
5.
Provide analyses and develop guidelines and procedures related to inadequate core cooling (as discussed in Section 2.1.9 of NUREG-0578, "TMI 2 Lessons Learned Task Force Status Report and Short-Term Recommendations") and define the conditions under which a restart of the RCPs should be attempted.
RESPONSE
5.
Analyses related to inadequate core cooling and definition of con-ditions under which a restart of the RCP's should be attempted will be performed.
Resolution of the requirements for the analyses and an acceptable schedule for providing the analyses and guidelines and procedures resulting from the analyses will be arrived at between the Westinghouse Owners' Group and the NRC staff.
Long-Term Action 1.
Propose and submit a design.nich will assure automatic tripping of the operating RCPs under all circumstances in which this action may be needed.
Ii09 297 s
RESPONSE
As discussed in our response to short-term item 2, we do not believe that automatic tripping of the RCP's is a required function based on the analyses that have been performed and the guidelines that have been developed for manual RCP tripping. We propose that this item be discussed with the NRC staff following their review of the Owners' Group Submittal.
1109 298 1qrv WISCONSIN PUBLIC SERVICE CORPORATION P.O. Box 1200, Green Bay, Wisconsin 54*95 September 17, 1979 Mr. J. G. Keppler, Regional Director Office of Inspection & Enforcement Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137
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Dear Mr. Keppler:
Docket 50-305 Operating License DPR-43 IE Bulletin 79-14 Seismic Analysis for As-Built Safety-Related Piping Systems As a follow-up to our letter of August 9, 1979, this letter is intended to respond to items 2, 3.and 4 of the referenced bulletin. In accordance with bulletin 79-14, the' July 20, 1979 revision, and the August 15, 1979 supplement, an inspection of the safety related piping systems for conformance to seismic analyses for the Kewaunee Plant has been completed.
As discrepancies were found between the existing piping layout /struc-ture as compared to the piping layout / structure documents used in the original seismic analyses, they were evaluated by both Fluor Power Services (FPS), our architect engineers, and our engineering staff. It should be stressed that these discrepancies have not been identified as non-conformances. Most of the differences between the as found and as analyzed piping layout / structures are due to engineering design changes that have analyzed the impact of the change on the seismic analysis. The original design documents would not have been updated to reflect as built conditions unless a complete reanalysis of that piping section was performed. The attached liv:
identifies those discrepancies which have not yet aeen resolved. It appears that for these discrepancies it will be easier and most cost effective to reanalyze these pipe sections than to search for documen-tation which verifies that the existing analysis is adequate for the as found condition. We are proceeding in this manner and expect to have these analyses complete by October 10.
An engineering review of these discrepancies by our AE indicates that there are no substantial problems, but a reanalysis will be performed to justify this position. Until the reanalysis is complete, continued operation is justified because in no case is redundant ECCS systems as required by Technical Specifications affected.
1109 299
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. During the last few months, e number of detailed inspections and examinations have been requeuted or directed by Bulletins issued by your office in regards to piping supports and seismic analysis. We have proceeded with dispatch in regards to each of these issues to assure that the Kewaunee Plant is safely designed, constructed, and maintained. The subsequent revisions to the bulletins which provide additional clarification of the wishes of the NRC staff prompt concern that the bulletins were hastily conceived and not well developed prior to their issuance. Our major concern, however, is in regards to the validity of the issue and the need for comprehensive investigations such as have been demanded by these Bulletins. To date, in regards to Bu_letins 79-02 and 79-14, we have encountered expenses of numbers of hundreds of thousands of dollars and personnel exposure of in excess of 18 Man Rem or one third of the total man rem exposure encountered during the refueling from all sources, while not encoun-tering any safety significant deficiencies or problems.
The direction employed to resolve each of these Bulletins in regards to the Kewaunee Plant has been to assure safety by a comprehensive investigation while attempting to maintain a reasonable control of both in personnel exposure and financial expenses. The bulletins provide direction to review all safety components with little apparent concern for cost effectiveness. A graduated approach to review systems such as employed in Section 11 of the ASME Boiler & Pressure Vessel Code or in Tech Specs in regards to Steam Generator Tube Inspection should be prescribed for Bulletins. When a response to one Bulletin results in an increase of total plant accumulated personnel exposure of more than 10% of the yearly average over the last four years and expenses in excess of a quarter.of a million dollars without discovery of significant difficiencies, a more reasonable approach of selective inspections should be considered.
Very truly yours, A
LLL E. R.
athews Vice President Power Supply & Engineering ERM:ljh 1109 300
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1.0 Auxiliary Coolant 1.1 Rigid hanger installed in place of a spring hanger 1.2 A pipe stand uses schedual 80 pipe in place of schedual 160 pipe 2.0 Containment Spray 2.1 Horizontal restraint omitted 2.2 Hanger location change 3.0 Residual Heat Removal System 3.1 Hanger added 3.2 Axial restraint not installed 4.0 Service Water System 4.1 Hanger location change 4.2 Anchor added 4.3 Horizontal restraint added 4.4 Pipe routing change 1109 301
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"u Florida Power C O R P O R A T BO 8s September 13, 1979 File:
3-0-3-a-3 Mr. J. P. O'Reilly Director U. S. Nuclear Regulatory Commission Of fice of Inspection and Enforcement Suite 3100 101 Marietta Street Atlanta, GA 30303
Subject:
Crystal River Unit No. 3 Docket No. 50-302 Operating License No. DPR-72
- 1. E.Bulletin 79-16
Dear Mr. O'Reilly:
In response to the subject bulletin concerning VITAL AREA ACCESS CONTROLS, the following information is being provided.
Item 1:
Establish criteria for granting unescorted access to each vital area, which shall be based upon the following:
A screening program meeting ANSI N18.17.
a.
b.
The individual has a valid need for access to the equipment contained in each vital area to which access is author ced.
Valid need is based upon assigned duties requiring the performance of specific tasks upon or associated with specific equipment located in each vital area to which access is granted. Valid need to enter one vital area shall not necessarily indicate that the person has a need to enter any other vital area.
Response 1: Criteria have been estab bshed regarding unescorted access to vital areas.
Only personnel who have completed a screening program equivalent to that identified in our Modified Amended Security Plan sections 1.3 and 1.6 will be granted unescorted access.
Each access will be based on a valid need and will be restricted to only those areas specifically associated with the need.
The only exce'ptions being Operations and Security personnel who will require access throughout the plant in order to perform their job fh 1109 302 7 910 0 50 r*l/d'4 a ^' o '
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General Office 32oi Tnirty fourtn street soutn. P O Box 14042. st Petersburg Rorida 33733 813 - 866 5151
Mr. J. P. O'Reilly Page Two September 13, 1979 4
Item 2:
An access list will be established for each area not to exceed 31 days. An individual will be on the access list only for the duration of the task to be performed.
If an individual has a valid need for unescorted access for a single entry or for intermittent occasions during this period, a separate daily access list shall be prepared.
All access lists shall be approved by the station manager (or equivalent) or his designated representative.
Response 2: Access lists will be established with a duration not to exceed 31 days as identified in our Modified Amended Security Plan section 5.3.1.3.
All access lists will be approved by the Plant Manager or his designated representative.
Item 3:
Individuals will be removed from the access list immediately upon termination of need.
If an individual has not entered the vital area during the effective period of the access list 0 t to exceed 31 days) the need for access should be reassu.ed prior to extending the authorization. To ensure that these actions are taken, the access list shall be reviewed and reapproved at least every 31 days.
Response 3: Individuals will be removed from access lists upon termination of their need for access.
To assure that access authorizations are not carried beyond the time frame associated with a valid need, the access list shall be reviewed and reapproved at least every 31 days.
Item 4:
Void access authorizations for all personnel not satisfying the criteria in 1.a and 1.b and where appropriate, reprogram the key card system and reissue key cards that are coded :o implement the above vital area access authorization program.
Response 4: Presently badged personnel who have not completed an ANSI N18.17 type screening program or who do not have a valid need for access to vital areas will not be allowed access until these criteria are met.
Item 5:
Develop reasonable alternatives so that the number and frequency of access to vital areas can be minimized consistent with safe operations.
Response 5: The number and frequency of accesses to vital areas are presently being minimized consistent with safe operation of the plant.
Item 6:
Establish emergency procedures where, during an emergency, additional authorized personnel, meeting criteria in 1.a and 1.b, can move freely throughout the vital areas with their entry and exit being recorded.
Upon securing from the emergency, the entry / exit recorc will be reviewed, and' normal accc::s control will be reestablished.
1i09 303
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l Mr. J. P. O'Reilly Page Three September 13, 1979 Emergency Procedures will be established to allow Response 6:authorized emergency team personnel to move freely throughout the vital areas.
Normal access control will be reestablished after a review of the entry / exit record and upon securing from the emergency.
Item 7:
Prevent tailgating by one or more of the following:
Establish procedures that require authorized personnel a.
to prevent other personnel, including those authorized unescorted access, from tailgating.
Ensure all authorized personnel are trained in the procedure, and establish a management program that ensures that the procedure is properly performed.
b.
Acquire equipment, such as turnstiles, to prevent tailgating.
Ensure that such equipment will not deny access or egress under emergency conditions.
Station a guard, watchperson or escort at the vital c.
area access portal.
This alternative would be most useful when there is a large number and frequency of access, such as occurs with containment during refueling.
d.
By any other means that achieve this objective.
Response 7: Procedures have been established that require authorized This is personnel to prevent tailgating by other personnel.
included in our general employee training which is given annually.
Item 8:
Assign corporate responsibility for management oversight of VA access control and require personal involvement to ensure that all intermediate levels of management are properly discharging their responsibilities in this regard.
Response 8: The Nuclear Plant Manager is responsible for site security.
Any oversight of vital area access controls will be brought to his attention and corrective action will be taken.
Audits of this activity will be performed by the plant staf f Compliance Section, FPC Quality Program personnel with appropriate corrective action being taken where necessary.
Item 9:
Conduct routine functional tests of the electronic access control system, including each key card reader, to verify (i) its operability and proper performance, and (ii) the accuracy of the data recorded.
This test should be incorporated into the seven-day test required by 10 CFR 73.55(g).
Ii09 304
Mr. J. P. O'Reilly Page 4 September 13, 1979 Response 9: A weekly (no greater than every 7 days) f unctional test is performed on the electronic access control system.
This includes a check for proper performance and operability of ecch key card reader and the accuracy of the data recorde4.
Procedures and training required to Laplement the above 9 items are under development. We expect to be in total compliance by January 31, 1980.
Should you have any questions concerning our responses to the above items, please contact this office.
Very truly yours, FLORIDA POWER CORPORATION O
i W. P. Stewart Manager, Nuclear Operations WPShewR01 D47 cc:
U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D.C. 20555 NRC Office of Inspection and Enforcement Division of Reactor Construction Inspection Washington, D.C. 20555 File Code: 3-0-3a-3 1109 305
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Portland General ElectricCcmpany
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Docket 50-344
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.Licenten NPF-1 Mr. R. H. Engelken, Director Nuclear Regulatory Comtnission Region V Suite 202, Walnut Creek Plaza 1990 N. California Blvd.
Walnut Creek, CA 94596
Dear Sir:
In accordance with IE Bulletin 79-09, we have determined that no GS-type AK-2-type circuit breakers are used or planned for use in safety-related systems at the Trojan Nuclear Plant.
If you have any questions, please contact me.
Sincerely,
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C. Goodwin, Jr.
Assistant Vice President Thermal Plant Operation and Maintenance CG/SML/4sb5A26 c:
Mr. Lynn Frank, Director State of Oregon Department of Energy
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PACIFIC GAS AND ELECTR,IC C O M PANY 32362WJE l
77 BEALE STREET,31ST FLOOR
- SAN FRAN CISCO, C AliFORNI A 94106 (415) 781 4211 J O H N C. M O R RlSS E Y
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Dear Mr. Engelken:
This is in response to your letter dated April 17, 1979 which enclosed I.E. Bulletin No.
79-09 and requested information concerning the use of General Electric Type AR-2 circuit breakers.
There are none of the subje_ct breakers in ur>e or planned for use in safety related equip-ment in Humboldt Bay Power Plant Unit No. 3.
Ver truly yours, r
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Direccor Office of Inspection and Enforcement Division of Reactor Operations Inspection U. S. Nuclear Regulatory Commission Washington, D. C. 20555 1109 307 7D10050(.'
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,1 Department of Energy 21 u.'s. Pg379A Oak Ridge Operations j
P.O. Box E Oak Ridge, Tennessee 37830 September 14, 1979 4
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Mr. Chas. E. MacDonald, Chief Transportation Certification Branch U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. MacDonald:
Confirming my September 14, 1979, conversation with Mr. Odegaarden, enclosed are eight copies of:
DOE-0R Certificate of Compliance No. USA /9853/BF.'
a.
b.
Report ORNL/ENG/TM-15, " Safety Analysis Report for Packaging:
The Unirradiated Fuel Shipping Container."
c.
Report ORNL/CSD/TM-77, " Nuclear Criticality Safety Assessment of ORR, NBS, and HFBR Fuel Element Shipping Package."
d.
ORNL internal report dated September 10, 1979, "Re-evaluation of ORR Shipping Package."
e.
00E-0R SARP Review Report.
The shipping containers covered by the Certificate of Compliance and SARP are to be used for fuel elements fioricated by Texas Instruments, Inc., at their Attleboro, Massachusetus, facility. Due to the impor-tance of the reactors utilizing fuel elements fabricated by TI, which operates under an NRC lincese, we respectfully request NRC assign this package for a very high priority review.
7-Sincerely, La lock, Chief AD-464:LGB Transportation Branch cc:
R. F. Garrison, ECT, HQ., w/3 cys ea.
i W. H. Travis, MS-33, w/o encl.
i J. A. Rounsaville, ER-lll, w/o encl.
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September 10, 1979 J. H. Evans, Building 1000 (4-6331)
Re-evaluation of ORR Shinpino Pa + ace in March of this year it w!s pointed out that the density of the phennlic foa : used in nuclear criticality saf ety evaluation of the ORR package un that corresponding to a confined forming, whereas, free forming was ir. plied by the pacLacn wei hts.
The change in foam density from 0.201 g/cc uwd in the analyse, reported in OdNL/CtD/TM-77 to 0.05 q/cc, typical of free feming, does not affect the conclusions given in the referenced docu'"ent The density reduction does result in an increrr in the infinite neutron inul >; lication of packages but the value re:aains balou a k eff of 0.9.
Tht package, there-fore, neets the requirenunts of thn Fissile Class I category.
The attachnent summarizes the benchmari.ing and the calculationc performn a.
If I can be of further help, please let me know.
t n h(
J.)T. Thomas, Chairman A.
Criticality Conmittee JTT/bbf Enclosures cc:
H. M. Butler, Bldg. 3546 (4-4340)
G. H. Burger, Cldg. 354G (4-4339)
. R. Mouring, lildg.1000-C215 (4-6397) 1109 309 7910050 h3 14101 1
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-5 L317..i r
11101
- .11'" 1.
6.,
1-
,. f/
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' f.
9.
a-6 0-1310) 2.M " 1.'
l.
3-
- 1. W I?lu3 2..',3?"
- 3. H13)-0 2.173s1-6 2;10' l. r ? l: u.
7.; 171-6 1.3153 M Ce etry b'criptien a fuel re.
m c f p u k m;e (c. te for i.1_.1_i r e.. J.. _i r t ro j i r.,
Gr -.'ri
$ > t v-
.x
.y 2
1 Co' ti !
l 3.111',
4. 01 '.
n.1 f.J 2
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- 3. " 11 4, '. :. ]
33 v
3 f ; oi1 7
3.w11 4.0/57 c) 7 4
( abii !
5 3.7/03 4.214, n.0J/
5
P o il l'
3.M60 4.? L B. 341 Pac kr;^ Je t ript. ion e
'3.
rc,o j 3 3 1
.,-rr ent R -
-.e q i nn 3 -." r v
'S < t r., ro 1
C",
! Wy 3
11.57 '
.y = 12 ':
.i - "?.344 2
Cylirter 6
r - 21.39.3
.* - Ni.3;;
4*
- P.3.344 3
(ylir ier 9
r 23.3^0
+: -
,nl.
,. e %.200 4
C yl i r. i -
l?
r
- 31. r'a g
.f l>n 11,
- 1 n.9' s
Cyli,J r l'
r - 11.3?O
+ z - l 3. : i,
z - l';3.M 6
Cuboi 1 Virtable. = 31. 03 tye 31. * +
= 111.00 1109 310 i
s 300RORIBEL I? c n c i.n_n_ _r_k.
The typic 31 fuel elec:ent considered is the li
- lo He1 to 350 g U(93.2) per element.
The in:criar of the region define in + h.
aluminua sich plates was a homogenorus ni: '.nre of the alu ;inu:n, uraniu o ice uw water v.hich resul ted in nu.aber don':itie ; qiven above.
Confirmation or th, adequacy of this fuel element represntation un had by recalculatio'n of t'.. ) of the configurations calculated ir. OEL/CFD/i';-63 (1973) where the elment was described in detail.
The two calculations were those of Tale 4 in U;-58.
A comparison of tLe calculation; is nude in lable 1.
Table 1.
Comparison of Hornooncous and Detailed representation of HFI;i. Fuel Elements Arrangement TF1-58 Detailed Homogeneous 4<4/1 1.32
- 0. 0 P 1.137 + 0.005 4
4 3xl 1.054 4 0.010 1.059 0.0CG This co:parison with the validate 1 calculaticnal L.nthod reported in OR';L/rSD/Ti1-58 shows good o';reement. An additional calcalation or a 3 s M array in water q1ve a k-eff of 1.00; + 0.003.
The latter arran+. nt is that used in tne par.kage evaluation and gives on indication of the effecti,<ne n of the pad.oje as a re fl ec tor.
Packane caltulations A sutrv rnel ; ingle p utop! with nine elements d" scribed aLove gave a k-eff of 0.715 1 0.005.
The L-eff of an infinite nu ;ber of pacLages withnut water between pacLages was 0.089 f 0.006.
The k-ef f of an infinite nuuber of package <. with water between the packages was 0.791 A 0.007.
These calculations with ninc-fuel elements my be considered as producinq an upper limit te any k-eff thu m v be achieved wit h ;evcn elements.
The lower k-ef f for seven elmv:nts is suf f icient to consider allowance in the mass per element as 110 ut u ra n i um
==
Conclusion:==
Any finite number of packaqus in transport cannot result in a k-eff larger than that corresponding to the k of 0.9.
The nuclear criticality safety assessment of the ORR shipninq packaqe meets the requirements of fissile Class I package in transport.
1109 $ll 16101 i
U.S. DEPARTMENT OF ENERGY (11 77)
CERTIFICATE OF COMPLIANCE I'*
For Radioactive Materiais Packages Ia. Certificate Nummer Ib. Revision No.
Ic. Pacnage identification No.
Id. Pa;e No.
Ie. Total No. Pages.
9As1 0
USA /9853/BF 1
3
- 2. PREAMBLE 2a.
Th a certificate is issued to satisfy Sections 173.333a.173.394,173.395,and 173.396 cf the Department of TranspJrtation Hazardot,s Materials Regulatens I,J CFR 170-183).
N The pacmaging and contents described in item 5 below, meets the safety standards set forth in Subpart C of Title 10. Code of Federal Regulations. Part 71,"?ackaging of Radioacts Material for Transport and Transportation of Radioactive Mater <al Under Certain Condit ons."
fhis certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transporution or other appiicante reguistory agencies, including the governrnent of any co antry througn or into wnicn the pacmage will be transported.
- 3. This certificate is issued on the basis of a safety ans" ss report of the package design or apphCation-II) Prepared by (Name and address):
(2) Title and Identification of report or apphcation:
(3) Date:
Oak Ridge National Laboratory Safety Analysis Report for Pack-Post Office Box X aging: The Unieradiated Fuel Shipping Oak Ridge, Tennessee 37830 Container Report: ORNL/ENG/TM-15
- 4. CCNotTIONS This certificate is conditional upon the fuifilling of the requirements of Suopart 0 of 10 CFR 71, as apphcaD e,and the conditions specified in item 5 beiow.
- 5. Description of Packaging and Authorized Cor: tents, Model Numoer, Fissue Class. Other Conditions, and
References:
a.
Packaging:
(1) Model: ORNL Unirradiated Fuel Shipping Container (2)
Description:
Packaging for unirradiated fissile material as fuel elements. The fuel ele-ments are positioned in a basket consisting of seven square cavities fabri-cated from 16 gauge plate and a base fabricated from eleven gauge plate.
The plate is Type 300 stainless steel. Eight 3/8" nuts and bolts retain the basket lid, which is made from 0.125" thick aluminum, in place. The basket is positioned inside a cylindrical outer shell. The outer shell and lid are fabricated from eleven gauge plate and the base is 1/4" thick plate.
The plate for the shell is Type 300 stainless steel. The outer lid is held in place by si:: 5/8" nuts and bolts. The basket is supported on 2" x 6" timbers inside the outer shell. The remaining space around the basket is filled with phenolic foam insulation.
There are different types of packages. Table I describes the details of each design.
g
}
b.
Contents:
(1) Type and form of material N nrmb-is enriched to s93': 235U and is in the oxide form.
It is Ga. Care of issuance: JUN 4 1979 i sb. E oiration cate:
FOR THE U S. DEPARTMENT OF ENERGY la. Address (of CQE lssuing OffiC9}
lb. $ignature, Name. and Tit >e tof OCE Aporoving Officia0 Slam _ -. b wa 4 U. S. Department of Energy Willia = H. Travis, Director Post Office Box E Safety and Environmental Control Oak Ridge, Tennessee 37830 Division
& n, n n e. n,/,A[ ')
--1 Ml '1
.. vuvw n
7
Page 2 - Certificate No. 9853 - Revision 0 contained in fuel plates as reactor fuel elements.
(2) Fissile Class:
I 1109 313
Page 3 TABLE I SUIRIARY DESIGN DATA CONTAINERS AND FUEL ELEMENTS Container Fuel Elements Gross Outside Inner Basket Nominal Maximum 235U Per Weight Length Diameter base Length Cavities
- Dimensions
- ype (Ib.)
(In.)
(In.)
(In.)
(In.)
(In.)
(In.)
Fuel Element (g) 1.
A. ORR" 580 56 5/8 24 1/2 29 x 29 38 J/4 4x4 A. 38 3/8 x 3.033 x 3.397 370 B. BSR B. 34 3/8 x 2.996 x 3.397 370 c
PCA II.
IlFBR 700 75 1/2 24 1/2 29 x 29 57 1/2 4x4 57 1/4 x 2.8'.8 x 3.218 370 III.
NBSRR" 850 87 1/8 26 30 1/2 69 3/16 4 1/2 68 51/64 x 3.125 x 3.8125 370 Y
x 30 1/2 4 1/2 a.
Oak Ridge Research Reactor b.
Bulk Shielding Reactor c.
Pool Critical Assembly d.
liigh-Flux Beam Reactor e.
National Bureau of Standards Research Reactor
- 7 Cavities per basket; 1 fuci element per cavity Vh g3 pA r-c:a
[4
%4)
(rw
3.
ORNL/CSD/TM-77 Contract No. W-7405-eng-26 COMPUTER SCIENCES DIVISION Nuclear Criticality Safety Assessment of ORR, NBS, and HFBR Ftel Element Shipping Package J. T. Thomas 4
UNION CARBIDE CORPORATION, NUCLEAR DIVISION operating the Oak Ridge Gaseous Diffusion Plant Oak Ridge National Laboratory Oak Ridge Y-12 Plant Paducah Gaseous Diffusion Plant for the DEPARTMENT OF ENERGY 1109 315 7910050 / 5 7 141C1 0
- - = - - ~ - - -
. _ _... ti.i__.._ _. _ _.._ _ _
TABLE OF CONTENTS ABSTRACT..............................
1 I.
INTRODUCTION.........................
1 II.
METHOD OF ANALYSIS......................
2 III. DESCRIPTION OF CODE INPUT 2
IV.
RESULTS OF CALCULATIONS 3
V.
CONCLUSIONS 4
J 1109 316
~
Nuclear Criticality Safety Assessment of ORR, NB3, and HFBR Fuel Element Shipping package J. T. Thomas ABSTRACT A fuel element shipping package employing a borated phenilic foam as a thermal insulating material i-designed to t.ansport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, or the National Bureau of Standards reactor.
The nuclear criticality safety evaluation demonstrates that the require-ments for a Fissile Class I package are satisfied by the design.
I.
INTRODUCTION The nuclear criticality safety of a shipping pack.ge designed to transport as many as seven plate-type fuel elements is examined by calculational techniques.
Three distinct packages are proposed, one for each of three sites having a light-watar reactor.
These are The Oak Ridge Research Reactor, The Brookhaven Fast Beam Reactor and the reactor at The National Bureau of Standards.
The three packages have similar neutronic characteristics, the same materials of construction but differ slightly in their dimensions.
The gross characteristics of a typical package are shown in Fig.1.
The analysis is performeo assuming there are nine fuel elements present in the package rather than the seven proposed.
This modification is made because it facilitates the geometric description of the package in the calculatio,. _ _ results in an overestimate of the neutron coupling between package, and of the k of the arrays of packages.
eH 1i09 317
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' ~ ~C '*d. C 4G 73. t tt94
_. ~
--lQ l
..--~..
.~.s.-.
--..- - - - -. -. - - - - - -~
9
'O
/
jo\\
j/
FUEL r' ~"Ert f-
- ~ ~ ~
yp 7 P00R BREM
~ AMC
~
.h
~h i 7' / t,
i y
FUEL 840XET o
,/
ojs o/
/'1
\\c
, /
d
,w.
uo i-g
_m n.
/.:
\\.
I.
'Y.
l N
\\
r h.rseuC J/ rch 45.
!} /
\\ ;c vc:ars ij sa cra -
a'.CC A j L
4,
,r g2;gr; y;
- o o.
a 1
4 I
i i
\\
a
- Y ai 2 m f
s' 51 ;
curc= s-st-fI s.1.,- l (CAREON S*iE.)
rn l o
J
\\
n ruc a.sxtr
- l r
,il / ^ % >:n e.v.
/
m.e:
um.
35 i
(
I f
j ru u.
j
'3l r 2=6 ? ?l bij
--- yl ' w L i
L a'.
N N
m
,/
r :sE e
[J e
l il z e, 8,D ci-us e ms.c.s
.s:-csi 29 Fig. 1, Proposed Package Design 1109 3P8
2 I). METHOD OF ANALYSIS Th'e neutron multiplication factors of the shipping package and of l
2 arrays of packages were calculated by the KENO IV code and the Hansen-Roach neutron cross section sets.
This combination of code and cross section sets has been validated by calculation of critical experiments with fuel elements of the same fissile materials and configurations. The results of the validation with fuel elements are reported in Refs. 5 and 6.
The results of calculations of critical experiments with the borated phenolic foam are reported in Ref. 7.
The conclusion of the compa.rison of calcula-tions and experiments is that systema calculated to have a keff of 0.98 should be regarded as having a potential for criticality.
III.
DESCRIPTION OF CODE INPUT Each plate in a fuel element is described in the code by a box type and these are arranged to form a fuel element within a region of the steel grid.
One-half the thickness of the steel forming the 3 x 3 matrix is associated with each element.
When the matrix of fuel elements is described in the code, the correct steel thickness is specified between fuel elements, however, only one-half the steel thickness is represented for the outer surface of the matrix.
This description is conservative in that it will result in larger calculated k
's than would be measured.
eff The geometry description is given in Table 1.
A fuel element and its associated section of the matrix is formed by stacking box types in the order 9, 7, 5, 3, twelve l's, 2, 4, 6 and 8.
The different box types are required to preserve the different water channel thicknesses and the non-fuel beaming end plates of the HFBR element.
The materials occupying the 1109 319
3 geometric regions specified in this case is representive of a calculation in the. damaged package evaluations. The materials and their number densities are iaentified and reproduced in Table 2.
The end boxes of the fuel elements are represented as a smeared density which preserves the mass of aluminum.
This is material 6 in region 4 of boxes 1 through 9 in Table 1.
The reflector region description of Table 1 specifies the geometry in the agion between the 3 x 3 steel matrix to the outer container of carbon steel. The material mixture in regions 0 through 4 nonnally would be 8, the undamaged borated phenolic foam insulation, but in the case depicted, region 4 represents the result of damage to the insulation by exposure to fire.
The principal damage to the package will be a charring of the borated 8
phenolic foam insulation.
Actual tests,9 show that an average char depth from the outer surface will not exceed 6.4 cm.
An overestimate of neutron coupling between damaged package would be observed if a larger char depth is assumed.
A value of 7.6 cm. was therefore used in the evaluation of the camaged package.
IV.
RESULTS OF CALCULATIONS The computed multiplication factors for the undamaged package are presented in Table 3.
It is evident that the absence of water from the fuel region of-the package results in very low values for k This is eff 5
consistent with previous results,6 with these fuel elements.
There is a large increase in keff.when water occupies the fuel region, however, km remains well below a value of unity.
Table 4 summarizes calculations of the damaged package condition.
Again, one finds k,well below unity.
1109 320
i t
l i
s I
s 1
i r
Table 1.
Geometric Representation of Package.
l l
!l 110x 7 WE I
i st E G IO N MIXTURE
+
i e
i OsDOID I
- x
= 2. e4 H S E 0 0
-x =- 2. 8 4 8 ?.E 00
+Y
=
- 2. N7 =,0 E-0 2
-Y
=-2. 89 50E- 02
+2
=
2.9027E 04
-2 =-2.9027E 04 1
5 2
0.sHO I D 2
+x
= 3.IO64E 00
-x =-3.1064E 00
+Y
=
- 6. 40 00 E- 0 2
-Y a-6. 4 0 00E-02
+Z
= 3.0862E 01
-2 e-3.0562E Ob s
3 GJHOID 5
+x
= 3.1064 E 0 0
-x
=-3 3064E 00
+Y
= 1.N540E-0*
-Y m-l.8540E-Ol
+2
= 3 0162E 08
-2 =-3 0162E 08
?
4 OsHOID 2
+x
= 3.58 5 = E 00
-x s-3.5884E 00
+Y
= 1.E640E-05
-Y m-l.8540E-01
+Z
= 3.08 62E 0 8
-2 =-3.0862E 05 l
5 01008D 6
+x
= 3. 5M l 4 E 00
-x =-3.5N14E 00
+Y
= 1.RS40E-01
-Y a-l.8540E-01
+2
= 8 802 7E 01
-2 =-8.8027E On 3 1
6-OtHOlb S
+x
= 3.7F0 0 E 00
-x
=-3.7 70 0C 00
+Y
= 1.H540E-05
-Y
==1.8540E-ol
+2
= 8.8027E 01
-2
=-8.8027E 01 l
7 070080 to
+x
= 3. 84 6 0 F 00
-x m-3.8460E 00
+Y
= 3 0540E-03
- Y =-l. 8 5 4 0E-01
+2
= 8.R344E OR
-2 =-8.8344E OS i
l i
00X T W8E 2
i i
f i
NFGION
=
~
l
~ l ~ GJt t01 D
~
l -
~ ~ ~+x~ ~ ' -= 2. fM 85 E~00
- X "=- 2.~84 8'$E"OO
+Y ~
~2 [F"a s"J E-0 2
- Y12. 8 9 50 E-02
- Z
=
~~
~
~
2.'90 2 7E~01 12'=-2.9027E'OR
{
2 OJHOln 2
+x
= 3.1064E 00
-x =-3.IO64E 00
+Y
= 6.4wo0E-02
-Y a-6.4 0 00E- 02
+2
= 3.0862E 01
-2 =-3.0162E 05
{
5' ~
+x
= 1.1064E 00
-x =~3.1064E 00
+Y
= 1.PS40E-01
-Y
=-2.0060E-01
+2
= 3.08 62E 01
-2
=-3.0862E On 4-3 QJnOlu 3
l 4
OJHOID 2
- x
= 3.Se l 4 E 00
. _ -x =~3.5814E_
00
+Y
= 1.th40E-Ol
-Y
=-2.0060E-Ol
+Z
= 3. 0162E 01
-2 =-3.0162E 03 t
t 6
OISOID 5
+x
= 3. 770 0 F 00
-x m-3.7700E 00
+Y
= 1.f640E-01
-Y =-2.0060E-On
+Z
= 8.8027E 01
-2 =-8.8027E 05
}
7 GJ00 l O 10
+x
= 3. fM 6 0 E 00
-=
s-3.8460E 00
+Y
=
1.H540E-03
-Y m-2.0060E-Ol
+Z
= 8.8344E 01
-2
=-8.8344E On
(_
REGION I
l j
1 CUtiOID l
+x
= 2. 84 8 5 E 00
-x m-2.8465E 00
+Y
= 2. Fn 50E-02
-Y a -2. 8 9 50E- 02
+Z
= 2.9027E 01
-2 a-2.9027E 01 l
2 OsHOID 2
+x
= 3 1064F 00
-x s-3.lO64E 00
+Y
= 6. 40 00E-02
-Y a-6. 4 0 00E-02
+2 5 3.0362E 0 8
-2 =-3.0562E 01 f
t 3 usHOID 5
+x
= 3.106 4 E 00
-x s-3.1064E 00
+Y
=
- 2. 00 60E-O S
-Y a-l. 8 5 40E-0 3
+2
= 3 0162E 08
-2 =-3.Ol62E on l
4 OJ00lO 2
+x
= 3.59 8 4 E 00
-x s-3.5314E 00
+ v = 2. 00 60 E-01
-Y
=-l. 8 5 4 0E - 01
+Z
= 3 0162E 08
-2 =-3.0162E on t
5 OsnotD 6
+x
= 3.5R l 4 F 00
- x =- 3.5 814 E 00
+Y
= 2. 00 60 E-O n
-Y s-l.8540E-01
+2
= 8 8027E nl
-2 =-8.8027E 04 p'
6 OstsO R D 5
+x
= 3. 770 0 E 00
-x m-3.7707E 00
+Y
= 2. 00 60F-01
-Y a-l.8540E-Ol
+Z
= 8 802 7E 05
-2 =-8.8027E On
[
7 OJHOID to
- x
= 3.H460E 00
-x s-3.0460E 00
+Y
=
- 2. 00 60 E-O l
-Y =-3.8540E-OS
+Z
= 0 8344E 08
-2
=-8.0344E 01
~
g __ i 1109 321 l;
i h
[
T I
e i
c e
,\\
Table 1.
Geometric Representation of Package (continued).
i 3
l 1
f tpA 1 vpE 4
I' t
NFGinN e
b I
ru s so l o I
+x
= 2. tw n% C 00
- ~4 =-2.8485E 00
+Y
= 2.to*>0E-02
-Y =-2.6950E-02
+2
= 2.9027E 01
- Z =-2.90 27E 01 i
?
OfDORO 2
+x
= 3.1064 E 00
-x m-3.1064E 00
+Y
=
- 6. 40 00E-0 2
-Y
=-6. 4 0 00E- 02
+Z
= 3 0862E 01
-2 e-3.0462E 01 i
I l
'~ 3' G8001D
$ ~ ' ' ' + r ' = 3. IO6 4 E' 00
- x s-3.~ 10 6 4 E 00
+Y
= 2. 00 t.0 E-O l
-Y
=-2.1340E-JI
+2
= 3.0162E 08
-Z m-3.0462E Ol I
+
4 CutsOID 2
ex = 3.58 8 4 E 00
-x s-3.5814E 00
+Y
= 2. 00 60 E-01
-Y
=-2.1340E-On
+2
= 3.0162E 01
-2 =-3.0862E 01-j '
i 6
I -
5 02t+010
' ~ 6 ex = 3. 5M 14 C 00
-x =-3.5814E 00
+Y
= 2. 00 60E-01
-Y
=-2.1340C-01
+Z~~= 8 8027E 01
-Z =-8.'8027E'~05
~
6 GD0010
+x
= 3.77000 00
-x m-3.7700E 00
+Y
= 2. 00 60E-0 3
-Y a-2.1340E-OI
+ Z. = R. 802 7E 0 8
-Z =-8.8027E 0l; 7
OJODID 10
+x
= 3. H4 6 0 E 00
-x m-3.H460E 00
+Y
= 2. 00 60E-01
-Y a-2.1340E-01
+Z = 8 834 4E 0 8
-2 =-8.8344E 01.
l i
j.
h t
REGION I
G8t1010 1
+x
= 2. 04 6 *i F 0 0
-x =-2.8485E 00
+Y
= 2. to *>0E-0 2
-Y
=-2. 8 9 50E-02
+Z
= 2.902 7E 01
-Z e-2.9027E 01; j
2 09001D 2
+x
= 3 1064 C 00
-x m-3.3064E 00
+Y
=
- 6. 40 00E-0 2
-Y a-6. 4 0 00E-02
+2
= 3.0362E 01
-2 =-3.0862E Oli 3
Of tn I D S
+x
= 3.10 6 4 85 0 0
-x m-3.IO64E 00
+Y
= 2 1340F-Ol
-Y a-2.0060E-01
+Z = 3. 016 2E 01
-Z m-J.0162E Ol*
t 4
O 3HO I D e
?
+x
= 3. 55' 14 E 00
-x s-3.5aI4E 00
+Y
= 2 1340E-08
-Y
=-2. 0 0 60E -0 3
+Z
= 3 03 62E 03
-2 =-3.0862E 01
...a.-
f 6
010010 5
+x
= 3. 7700 E 00
- x m-3. 770 0E 00
+Y
=
i' 2 1340E-on
-Y a-2.0060E-On
+Z
= 8.8027E 08
-2 =-8.3027E 08 7
OrHOID 10
+x
= 3. t% 6 0 F 00
-x m-e 3. 8 4 6 0E 00
+Y
=
g 2 13 40F-O S
-Y a-2. 0 0 60E -O f
+Z
= 8. 834 4E 08
-2 =-R.8344E 08 i
g t
004 7 YPE 6
HEGION i
l 1
U nin. D l
+x
= 2. P4 HS E 00
-x =-2.84PSE 00
+Y
= 2.rN50E-02
-Y a-2 8 9 50E-02
+2 = 2.902 7E ' O l ~~ -2 =-2.9027E*0l' 2
01D0113 -
2
+4
= 3. IO64 E 00
-x m-3.1064E 00
+Y
= ti. 40 00E-0 2
-Y a-6. 4 0 00E-02
+Z
= 3 0162E 01
-2 =-3.0162E 01 3 Ositoin 5
+x
= 3.1064 E 00
- x m-3.1064 E 00
+Y
=
2.13 4 0 E-O l
-Y m-I.9050E-08
+2
= 3. 0 8 6 2E on Z =-3 0862E 01 l
4 O ttiO I D 2
+x
= 3.5R 14 F 00
-X m-3.5884E 00
+Y
= 2.1340F-01
-Y a-l. 9 0 50E-0 8
+Z = 3.0162E On
-2
=-3.0862E 04:
O n+113 u 6
+x
= 3.59 8 4 E 00
-x =-3.5814E 00
+Y
=
2.3340E-01
-I m-l.9050E-01
+Z
= 8.8027E 08
-2
=-8 8027E 01 i
6 08H01D 5
+x
= 3. 7/00 E 00
-x m-3.7700E 00
+Y
= 2.1340E-05
-Y m-l. 9 0 50E-31
+Z
= 8.802 7E. 01
-2 =-8.8027E On; f
7 G7001D to
+x
= 3. n4 6 0 E 00
-x
=-3. 8 4 6 0E 00
+Y
= 2.1340E-01
-Y =- 8. 9 0 5 0E -01
+Z
= 8.9344E 01
-2 m-8.8344E 01
[.
I i
I109 322 i
i I
l I
j-(
Table 1.
Geometric Representation of Package (continued).
i n
o now T vp E 7 ~
g HFGION 1
0U008D 1
+x
= ?.84RS E 00
-M'=-2.0485E 00
+Y~~ = 2.tr>505-02
-Y a-2. 8 9 505- 02
+2 = 2.902 7E Ol'
-2 m-2.9027E Ol ~~ ~~ I 2--OuHoln 2
+ x - a - 3. 8 064 E-0 0 x -- =-3 3 0 6 4 E-00 --. - = - 6. 40 00 E-0 2 Y-=-6. 4 0 00E-0 2 --- + Z - = - 3. 0 3 6 2E- 01 ----Z - =-3. 0 8 62 E-O n 3 02n010 S
+x
= 3.106 4 E 00
-x
=-3 1064E 00 '
+Y
= 1 40 50E-o n
-Y a-2.13402-03
+ Z 's 3.0862E 08
-Z m-3.0862E 05 l
l 4
Q 0010 2
+x
= 3. 56 8 4 E 00
- x
=-3. 5 A 14 E 00
+Y
= t. 90 50 E-0 8
-Y a-2.1340E-01
+2
= 3.0to2E 01
-2 =-3.0162d 01 i
I 5 OlnOID 6
+x
= 3. 5H l 4 E 00
-x m-3.5614E 00
+Y
=
- 1. 90 50 E-01
-Y a-2 1340E-01
+Z
= 8.802 7E On
-2 =-8.8027E 08 6
0un080 5
+x
= 3.7 F00 6 00
-x m-3.7 70 0E 00
+Y
=
7 0J0010 to ~
- 1. 90 50E-01
-Y a-2 1340E-On
+Z
= 8.802FE 01
-Z m-8.8027E 03 g
- X
= 3.6460E 00
-x m-3.'8 4 6 0 E 00
+Y'= 1. 90 50E-0 5
-Y a-2 13 4 0E-03
+Z = 8.834 4E Ol'
-2 =-8.8344E 08 ~j DOM 7 VD E 8
t l
+
NFCION f
j l
Q1HO I O 2
+M
= 3 106 4 E 00
-x m-3. 3 06 4E 00
+Y
= 6. 40 00 E-0 2
-Y
=~6. 4 0 00E-02
+Z
= 3. 03 6 2E 01
-Z =-3.0462E On 2
08001D 5
+x
= 3 106 4 E 00
-x m-3.1064E 00
+Y
=
3
- 2. 54 00E-01
-Y a-2.5 4 00E-01
+Z
= 3.0162E OR
-2 =-3.0862E 01 j
3 C0001D 2
+x
= 3 5H I
- E 00
-x =-3.5014E 00
+Y
= 2. 54 00F-O l
-Y a-2. 54 00E-01
+2
= 3.0162E On
-2 =-3.0362E on 4
CuhnIn 6
+x
= 3 5814 E 00
- x s-3.5814 E 00
+Y
= 2. 54 00 E-O l
-Y a-2.54 00E-OI
+Z
= 8.8027E 01
-2 =-8.802FE 03' g
1 i
5 OsMulD 5
+x
= 3.7F00E 00
-x m-3.7700E 00
+Y
= 2. 54 00E-O S
-Y
=-3.0540E-Ol
+Z
= H.802 7E 01
-Z m-8.8027E On 6 (UHnID to
+x
= 3. 64 6 0 E 00
-X m-3.8460E 00
+Y = 2. 54 00E-0 5
-Y a-3. 814 0E-0 8
+Z
= 8.8344E 03
-2 =-8.8344E 01 e
hot 7YPE o e f
e l
I OJHo t D 2
ex = 3. IO64 E 00
-x =-3.1064E 00
+Y
= 6. 40 00 E-0 2
-Y
=-6 4 0 00E-02
+Z
= 3.0162E 01
-Z m-3 0862E 03, l
l
(
I 2
0)MOln S
+x
= 3.106 4 F 00
-x m-3.1064E 00
+Y
= 2. 54 00E-01
-Y a-2 54 00E-01
+Z
= 3.Ol62E 01
-Z e-3.Ol62E 01 3 CubO I D 2
+x
= 3.5314 E 00
-x =-3.5884E 00
+Y
= 2. 54 00 E-0 3
-Y m-2 5 4 00E-01
+Z = 3.0162E Ol'
-2 =-3.0162E 01 l
4 (UNOlp 6
+x
= 3.5M I 4 E 00
-x m-3 5H14E 00
+ v = 2. 54 00 E-O n
-Y
=-2 5 4 00E-O n
+Z
= 8.8027E 01
-2 =- 8. 8027E 0 5 ~~ ~"
5 OJDolfD' 5
+x
= 3. 7 70 0 E 00
- x
=-3. 7 70 0E 00
+Y
= 3. 05 4 0E-01
-Y a-2.54 00E-01
+Z
= 8.8027E 08
-2 =-8.8027E os; 6
CUOulD 10
,x
= 3. 846 0 C 0 0
-x m-3.0460E 00
+Y
= 3 8140E-08
-Y
=-2. 5 4 00E- 01
+2
= 8.H344E On
-2
=-8.8344E OB l
I i
HEGION i
l CONE ODY O
+x
= 1 153eF 01
-t
=-1.153HE 01
+Y
= 1.2192E of
-Y a-l. 2192E 01
+F
= 8 834 4E 01
-2 m-8 8344E 03) 2 CY L I Nnt il 8
HAD !ys = ?. 33HO C 01
+7
= 0.H344F 01
-Z m-8.n144E ol
)
3 CYLI NDEM 9 HAOlub = 2. 31H O E ol
+ 2 = 9.54 0 0E on
-Z m-9. 54 00E 0 8 r
4 CYL I NDf"4 1?
HADIOS = 3.100 0 E 01
+Z
= I.03tlE 02
-Z a-l. 02'm F 02 5
CV L I No r H lt H AD fits = 3 3370F Og
+Z
= 1.0343E O2
-Z m-3. 03 %A E 0 2 l
Q 0700I0 0
+x
= 3 3.900E 03
-X n-3.lJSOE 05
+Y
= 1.0360E 0?
-Y
=-l. 0 3 60E 02
+Z
= 1 1380E 02
-Z m-l.13 HOE 02
. _.1.1.09.__4,)!I i
tg,t i
l i
....-4.-....we-..
_,e-.-en..
.-s
==-.-_w-Table 2.
Description of Materials for C0de Input.
MIXTURE NUCLIDE DEN SITY 1
-92502
- 1. 3 0 703 E-03 '
l hh d8[$
_oj kJranium Oxide and Aluminum 1
92834
- 1. 41 241 E -04 1
8100
- 7. 4 7 30 0 E -03 1
13100
- 5. 2 3 867 E-0 2 2
13100
- 6. 0 2 72 6 E -0 2 A_ bim i,,.m 3
6100
- 1. 7 3 76 5 E -0 2 '
3 1101
?. 6 0 64 7E -0 2 Not used in study 3
17100
- 0. 6 8 925 E -0 3 4
5_0.P 5._ O_0_00 0 E -0 2,
5 502
- 1. 0 0 00 0F 00)
Water 6.
13100 1 81 19 0E-0 4 )
Aluminum 7
13100
- 1. 8 1 19 0 E -0 4 '
7 11n1
- 6. 3 P_3o_45_-0 2 Alumimtm ar d Wa*pr 7
N100 3.16 19 7 E -0 2,
8 6100 4.412005-03' 8
1101 5.- 5 3 E0 0 E -0 3 a
a 1.0_0
- 3..O.o e0 0 E -0 3 Borated phenolic foam 8
5100
- 2. 5 3 60 0 E-04 8
14100
- 1. 5 0 400 E-0 4 8
11100
- 1. 6 3 c0 0 E -0 5 n
17.100
- 2. 3 2 80.0 E_-0 9 e
13100 6.905005-06 8
12100
- 9. 8 8 30 0 E -06 8
20100
- 6. 0 5 60 0 E-05 d o
6.1_0_Q
- 6. Q2.96_0E.-0_3 '
9 1101
- 7. 2 0 010 E-33 9
8100
- 3. 8 9 47 0 E -0 3 Borated h
E enolic foam and 9
5100 3.1 1 90 0 E -0 4 o
14100 1 32.70_OE-Q.4 -
9 11100 1.44 '00E-05 Wood 4
17100
- 2. 0 5 30 0 E -0 5 9
13100
- 7. 9 3 40 0 E-06 9
12.10.0 R._7_1.7.0_O F --O 5 9
20100
- 5. 3 4 10 0 E -0 5 '
10 200
- 1. O O OOOE 00 3 Stainless Steel 11 100
- 1. 0 0 00 0E 00)
Carbon Steel 12 6 10.0
- 6. 0 3 0S O E_-03 '
72 1101 0.0 1P 8100
- 0. 0 1P 5100
- 3. I 100 0 E -04 12 1
4 5-haned oomed phmlic 12 17100
- 0. 0 12 13100
- 7. 9 3 4 0 0 E -0 6 foam 1.2 1210.0 P._7_130 0 E -0 6._
12 20100
- 5. 3 4 10 0 E -0 5.
+.
,_.-m 1109^$24
Table 3.
Computed Neutron Multiplication Factors for the Undamaged Package Number of Packages Conditions kff IU Single No water present 0.033 0.002 Single Water in full region 0.666 0.006 Single Water in full region, package 0.688 0.009 closely reflected by water Infilite array No water present 0.092 0.003 Infinite array Water in fuel region 0.739 0.008 Infinite array Water in fuel region, water filling 0.682 0.007 void between packages i109 325
.~.
Table 4.
Computed Neutron Multiplication Factor for the Damaged Package l iumber of
+
_ Packages Conditions eff -
Single Water in fuel region 0.669 0.008 Single Water in fuel region, package closely 0.670 0.008 reflected by water Infinite array Water in fuel region 0.812 0.007 Infinite array Water in fuel region, water filling 0.812 0.007 void between packages i109 326
4 V.
CONCLUSIONS The evidence of this study shows that the package loaded with seven HFBR fuel elements meets the nuclear criticality safety requirements of a Fissile Class I package.
In view of the comparative calculations of various similar fuel elements with different fissile material loadings S
reported,6, it may be concluded that the package may also be used as a Fissile Class I package for the ORR and the NBS fuel elements with fissile 2 ss / element.
material loadings at least as large as 350 g V
Fuel elements of similar construction used at the Oak Ridge National Laboratory, such as, the PCA reactor (140 g assU/ element) and the BSR reactor (200 g 23sU/ element), may also be shipped in the container.
il09 327
REFERENCES 1.
L."M. Petrie and N. F. Cross, KENO-IV: An Improved Monte Carlo Criticality Program, ORNL-4938, Oak Ridge National Laboratory (1975).
2.
G. E. Hansen and W. H. Roach, Six and Sixteen Group Cross Sections for Fast and Intermediate Critical Assemblies, LAMS-2543, Los Alamos Scientific Laboratory (1961).
3.
J. K. Fox and L. W. Gilley, Critical Experiments with Arrays of ORR and BSR Fuel Elements, Neutror. Physics Division Annual Progress Report for period ending September 1, 1958, ORNL-2609, Oak Ridge National Laboratory (1958).
4.
E. B. Johnson and R. K. Reedy, Jr., Critical Experiments with SPERT-D Fuel Elements, ORNL/TM-1207, Oak Ridge National Laboratory (1965).
5.
J. T. Thomas, Nuclear Criticality Safety of the Fuel Element Fabrication Facility at Attleboro, Massachusetts, ORNL/CSD/TM-55, Oak Ridge National Labaratory (1978).
6.
J. T. Thomas, Nuclear Criticality Assessment of Oak Ridge Research Reactor Fuel Storage, ORNL/CSD/TM-58, Oak Ridge National Laboratory (1978).
7.
D. W. Magnuson, Critical Three-Dimensional Arrays of Neutron Interacting Units:
Part III Arrays of U(93.2) Metal Separated by Various Materials, Y-DR-83, UCCND Y-12 Plant, 1972.
8.
A. J. Mallett and C. E. Newlon, Protective Shipping Package for 5-inch-Diameter UF Cylinder, K-1716, ORGDP, 1967.
6 9.
A. J. Mallett and C. E. Newlon, New End-Loading Shipping Container for Unirradiated Fuel Assemblies, Proceedings of Second International Symposium on Packaging and Transportation of Radioactive Materials, CONF 681001 USAEC, 1968.
I109 328 141t1