NRC-98-0063, Forwards non-proprietary & Proprietary Versions of Criticality Safety Analysis, Re Request for Exemption from Requirements of 10CFR70.24.Proprietary Version Withheld
| ML20247B504 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/27/1998 |
| From: | Gipson D DETROIT EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20013J030 | List: |
| References | |
| CON-NRC-98-0063, CON-NRC-98-63 IEIN-97-077, IEIN-97-77, NUDOCS 9805080069 | |
| Download: ML20247B504 (39) | |
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Douglas R. Gipson Senior Vice l' resident, Nu(lear Grimration Fermi 2 6400 North Dixie liwy. Newport, Alichian 18166 Tel:3131M rdal fax: 31:LFM4172 April 27,1998 NRC-98-0063 l
U. S. Nuclear Regulatory Commission Attention: Document Control Desk l Washington D C 20555-0001 i
References:
- 1) Enrico Fermi Unit 2 l NRC Docket No. 50-341 l NRC License No. NPF-43
- 2) Detroit Edison Letter, EF2-60,332, I
! dated November 9,1982, l " Revised Application for a License to ;
Possess Special Nuclear Material"
- 3) NRC letter, dated May 31,1983 from
)
R. G. Page to H. Tauber," Issuance of J NRC Materials License No. SNM-1915"
- 4) General Electric (GE) Letter RDW: 98-037, dated April 9,1998,"New Fuel Receipt Criticality Safety Analysis"
Subject:
Request for Exemption from 10CFR70.24, Criticality Accident Requirements This letter is being submitted to request an exemption from the requirements of 10CFR70.24," Criticality Accident Requirements," for Fermi 2. The basis for this /
request is similar to thejustification provided for the exemption to 10CFR70.24 that ,l was granted by the NRC in the May 31,1983 Special Nuclear Material (SNM) ,
License No.1915 for this facility (References 2 and 3). The NRC has taken the .
position that, unless an exemption granted under a 10CFR70 license is explicitly incorporated into the subsequently issued 10CFR50 operating license, the exemption [pC)[ !
expires upon issuance of the 10CFR50 operating license. This position is stated in SECY-97-155," Staff's Actions Regarding Exemptions from 10CFR70.24 for Commercial Nuclear Power Plants," and Information Notice 97-77," Exemptions from the Requirements of Section 70.24 of Title 10 of the Code of Federal
~ ^ [3 Regulations." Information Notice 97-77 recommends that licensees obtain an exemption from this regulation before the next receipt of fresh fuel or before the planned movement of fresh fuel.
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USNRC NRC-98-0063 Page 2 Therefore, Detroit Edison is requesting an exemption to the criticality accident requirements stipulated in 10CFR70.24 specifically for the areas containing Special Nuclear Material (SNM) in the form of calibration sources or not in use in-core nuclear instrumentation, [e.g., source range monitors (SRMs), intermediate range monitors (IRMs), local power range monitors (LPRMs), and traversing in-core probes (TIPS)], and for unirradiated fuel while it is handled, used, or stored onsite.
Enclosure I further describes the exemption request and itsjustification.
General Electric (GE) Letter RDW: 98-037, dated April 9,1998,"New Fuel Receipt Criticality Safety Analysis"(Reference 4, Attachment 2) has been provided by General Electric to support this exemption and is considered by General Electric to be proprietary commercial information as described in 10CFR2.790(a)(4).
Therefore, pursuant to the enclosed affidavit (Attachment 1) it is requested that this information be exempted from public disclosure. Attachment 3 to this letter provides a non-proprietary version of the General Electric (GE) Report.
We have satisfied the good cause requirements outlined in 10CFR70.24(d) and the basis is provided in the enclosed exemption request and justification. We feel the requested ex .nption is authorized by law, will not endanger life or property or the common defense and security, and is otherwise in the public interest.
The next shipment of unirradiated fuel is expected to arrive at the Fermi 2 site in the week of June 15,1998. The fuel will then be inspected and prepared for use in the upcoming refueling outage (RF06), which is scheduled to start in August of 1998.
Detroit Edison requests review and approval of this exemption by June 1,1998 with
- an implementation period of two weeks following the NRC approval. Should you have any questions regarding this exemption, please contact Mr. Norman K.
Peterson, Director, Nuclear Licensing at (734) 586-4258.'
Sincerely, .
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Enclosure 10 CFR 70.24 " Exemption Request and Justification" :
I Attachments cc: A. B. Beach B. L. Burgess G. A. Harris A. J. Kugler l
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USNRC NRC-98-0063 Page 3 I
i I, DOUGLAS R. GIPSON, do hereby affirm that the foregoing statements are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.
h DOUGLAS R. GIPSON Senior Vice President, Nuclear Generation On this M7 day of M ,1998 before me personally appeared Douglas R. Gipson, being first' duly sworn and says that he executed the foregoing as his free act and deed.
1
/kfu Notary Public i'
R06AUE A ARME1TA WS$$en$$
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Enclosure I to NRC-98-0063 Page1 l
I ENCLOSURE 1 i
i 10CFR70.24 EXEMPTION REQUEST AND JUSTIFICATION )
Enclosure 1 to NRC-98-0063 Page 2 10Cl4R79.24 EXEMPTION REOUEST AND JUSTIFICATION Request for Exemption from 10CFR70.24 Requirements t In accordance with the requirements of 10CFR70.14, Detroit Edison requests an exemption from the criticality accident requirements of 10CFR70.24 for Fermi 2, as it applies to Special Nuclear Material (SNM) in the form of calibration sources or not in use in-core nuclear instrumentation, e.g., source range monitors (SRMs),
intermediate range monitors (IRMs), local power range monitors (LPRMs), and traversing in-core probes (TIPS).
An exemption from the criticality accident monitoring requirements of 10CFR70.24 is also requested for areas where unirradiated fuel is handled, used, or stored on site.
New fuel bundles, which are packaged in NRC approved packaging, may be handled, used, or stored in areas not monitored by a criticality accident monitoring system provided they remain in the approved shipping packages. A fully compliant criticality monitoring system is provided in areas where the fuel assemblies are removed from the inner metal shipping containers. Area radiation monitoring instrumentation or procedu alized geometry and moderation controls to preclude critical configurations are provided in those areas where the inner metal shipping containers are removed from the outer wooden shipping containers.
This exemption is necessary to clarify the requirements stipulated in the Fermi 2, Facility Operating License (NPF-43), which invokes 10CFR70 as a whole. This request is also consistent with NRC guidance with regard to criticality accident monitoring requirements. Information Notice 97-77," Exemptions from the Requirements of Section 70.24 of Title 10 of the Code of Federal Regulations,"
recommends that licensees obtain an exemption from this regulation before the next receipt of fresh fuel or before the planned movement of fresh fuel. The NRC further indicated that it would not pursue any further enforcement actions provided licensees obtained the necessary exemption, if warranted.
Granting this exemption will facilitate receipt and processing of new fuel for Fermi 2, which is currently expected to arrive at the site in the week of June 15,1998. The fuel will then be inspected and prepared for use in the upcoming refueling outage (RF06), which is scheduled to start in August of 1998. Detroit Edison requests review and approval of this exemption by June 1,1998 with an implementation period of two weeks following NRC approval. The requested exemptions specified above will in no way affect the health and safety of the public.
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Enclosure 1 to NRC-98-0063 Page 3 Good Cause Justification l
l Section 70.24(d) anticipates that licensees may request relief from these requirements
! and allows licensees to apply for an exemption from Section 70.24, in whole or in i part, if good cause is shown. Detroit Edison believes that good cause exists as l
discussed below.
Scope of Request 10CR70.24(a)," Criticality Accident Requirements," states the following:
"Each licensee authorized to possess special nuclear material in a quantity exceeding 700 grams ofcontained uranium-235, 520 grams ofuranium-233, 450 grams ofplutonium,1500 grams ofcontained uranium-235 ifno uranium enriched to more than 4 percent by weight of uranium-235 is present, 450 grams ofany combination thereof or one-halfsuch quantities ifmassive moderators or reflectors made ofgraphite, heasy water or beryllium may be present, shall maintain in each area in which such licensed special nuclear material is handled, used, or stored, a monitoring system meeting the requirements ofeitherparagraph (a)(1) or (a)(2), as appropriate, and using gamma- or neutron-sensitive radiation detectors which will energize clearly audible alarm signals ifaccidental criticality occurs. This section is not intended to require underwater monitoring when special nuclea, material is handled or stored beneath water shielding or to require monitoring systems when special nuclear material is being transported when packaged in accordance with the requirements ofpart 71 ofthis chapter. "
l Calibration Sources ant' Not in Use In-core Nuclear Instrumentation The major form of Special Nuclear Material (SNM) that is present at Fermi 2 is principally in the form of nuclear fuel. However, there are other quantities of SNM that are used, handled, or stored at Fermi 2. This material is in the form of fissile material incorporated into calibration sources and not in use in-core nuclear instrumentation (e.g., SRMs, IRMs, LPRMs, and TIPS). The amount of SNM contained in the calibration sources and not in use in-core nuclear instrumentation is 3
small and significantly less than the quantities delineated in 10CFR70.24(a). The small quantity of SNM present in the sources and not in use in-core nuclear instrumentation precludes inadvertent criticality.
The total amount of SNM contained in the calibration sources and not in use in-core instrumentation is such that it also meets the " forms not sufficient to form a critical mass" guidance in Section 1.1 of Regulatory Guide (RG) 10.3, " Guide for the Preparation of Applications for Special Nuclear Material Licenses of Less than Critical Mass Quantities." The quantities of SNM specified to be enough for a
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Enclosure 1 to NRC-98-0063 l Page 4 i
l critical mass in RG 10.3 are 350 grams of uranium-235 (U-235),200 grams of uranium-233 (U-233), and 200 grams of plutonium-239 (Pu-239).
l The quantity of SNM in the form of calibration sources and not in use in-core nuclear instrumentation at Fermi 2 is currently well below the amounts for which criticality accident requirements would apply as described in 10CFR70.24(a) and is never
- projected to challenge amounts specified in Section 1.1 of RG 10.3. Since the quantity of SNM in stored calibration sources and not in use in-core nuclear instrumentation is such that an inadvertent criticality cannot occur, Detroit Edison believes that we have demonstrated good cause for granting an exemption to any criticality accident requirements stipulated in 10CFR70.24 for such storage.
Unirradiated Nuclear Fuel Criticality events from unirradiated nuclear fuel packaged in a NRC approved packaging are precluded due to the construction of the package and the storage configuration of the fuel in the shipping container. The package design ensures that a geometric criticality safe configuration is maintained during transport, handling, storage, and accident conditions. The package design also precludes introduction of any moderating agents due to leak tight construction. NRC approval (i.e.,
represented by issuance of a Certificate of Compliance for Radioactive Materials Packages) of the package design is certified by the NRC that any incident which could occur during transport could not cause an inadvertent criticality accident. The fuel that is received at Enrico Fermi Unit 2 is packaged in a NRC approved shipping package which satisfies the requirements of 10CFR71. The approved shipping package that is received consists of an outer wooden container and an inner metal container. Since the Detroit Edison new fuel handling procedure requires that new fuel only be removed from the inner shipping container in areas where criticality accident monitoring is present (i.e., the Refuel Floor), we believe that we have demonstrated good cause for granting an exemption to the criticality accident monitoring requirements stipulated in 10CFR70.24.
l Criteria for Evaluating 70.24 Exemption Requests As indicated in SECY-97-155, dated July 21,1997, the NRC determined that it is l
appropriate to exercise enforcement discretion in some cases where licensees do not comply with the 10CFR70.24 requirements, since the safety significance of the I failure to comply with these requirements is minimal provided controls are in place to ensure compliance with General Design Criteria (GDC) 62. The NRC also indicated that enforcement discretion was appropriate because it did not recognize the need for an exemption during the licensing process. The NRC does not intend to j take further enforcement action for failure to meet the requirements of 10CFR70.24 j provided licensees obtain an exemption from this regulation before the next receipt
]
Enclosure 1 to NRC-98-0063 Page 5 of fresh fuel or before the next planned movement of fresh fuel. The NRC established and published seven (7) criteria that it is using to evaluate exemption requests to 10CFR70.24. This position, along with the seven (7) review criteria, was reiterated in Information Notice 97-77 issued on October 10,1997. Although some of these criteria may not specifically apply to this exemption request, to assist the NRC in its review of Detroit Edison's 10CFR70.24 exemption request, the criteria is restated below, followed by the Detroit Edison response.
Criterion 1. Plant procedures do not permit more than [1 PWR or 3 BWR] new fuel [ assembly / assemblies] to be in transit between their associated shipping cask and the dry storage rack at one time.
Response
Existing plant procedures restrict the number of new fuel assemblies that can be configured into any close packed array to 3 assemblies, and that such arrays are to be separated from all other fuel by a minimum of 12 inches. The fuel inspection and channeling stand used at Enrico Fermi Unit 2 can only accommodate two bundles at a time. After the bundles are inspected and channeled, they are immediately placed in the spent fuel pool. During transport of an inspected and channeled assembly to the fuel preparation machine for further processing to its spent fuel pool storage location, a replacement assembly from the inner metal container uprighting stand is permitted to be transferred to the irispection and channeling stand. This process ensures that close packed arrays of more than three fuel assemblies are not created, even though a fourth assembly may be either in the Fuel Preparation Machine at a considerable distance from the inspection stand or be in the process of being moved underwater by the Refueling Bridge to its Spent Fuel Pool Storage location.
Therefore, Detroit Edison believes that we have satisfied the intent of this specific criterion.
I Criterion 2. The k-effective ofthefreshfuel storage racksfilled withfuel ofthe l maximum permissible U-235 enrichment andflooded withpure water does not exceed 0.95, at a 95 percent probability, 95 percent confidence level.
Response
This criterion is not applicable at Fermi 2, since procedures preclude the use of the fresh fuel storage racks. Therefore, Detroit Edison believes that we have satisfied this specific criterion.
Criterion 3. Ifoptimum moderation offuel in thefresh storage racks occurs when thefreshfitelstorage racks are notflooded, the k-effective corresponding to this optimum moderation does not exceed 0.98, at a 95 percent probability, 95 percent confidence level.
Enclosure 1 to NRC-98-0063 Page 6
Response
This criterion is not applicable to Fermi 2, for the same reason as stated in response to Criteria 2 above. Procedures preclude the use of the fresh fuel storage racks.
Therefore, Detroit Edison believes that we have satisfied this specific criterion.
Criterion 4. The k-effective ofspentfuel storage racksfilled withfitel ofthe maximum permissible U-235 enrichment andfilled with pure water does not exceed 0.95, at a 95 percent probability, 95 percent confidence level.
Response
The Fermi 2, Technical Specifications (TS) Section 5.6.1 specifically states with respect to the design of the spent fuel storage racks that: "A k,y equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 1.9 percent delta k/k for uncertainties as described in Section 9.1 of the FSAR." Therefore, Detroit Edison believes that we have satisfied this specific criterion since it is included in the current licensing basis. In addition, the fuel vendor's fuel bundle design procedures require confirmation of compliance with specified fuel storage reactivity limits for all new fuel designs which ensures that compliance with the above criterion is maintained.
Criterion S. The quantity offorms ofspecial nuclear material, other than nuclear fuel, that are stored on site in any given area is less than the quantity necessaryfor critical mass. l
Response
Enrico Fermi Unit 2 has a SNM Accountability and Control Program which ensures that non-fuel SNM, typically calibration sources or not in use in-core nuclear instrumentation, (e.g., source range monitors (SRMs), intermediate range monitors (IRMs), local power range monitors (LPRMs), and traversing in-core probes (TIPS)],
are properly stored and accounted for in accordance with the requirements of 10CFR70. These items have been evaluated and determined to contain significantly less SNM by weight than would be required to achieve criticality in any configuration. Therefore, Detroit Edison believes that we have satisfied this specific criterion.
Criterion 6. Radiation monitors, as required by GDC 63, are provided infuel storage and handling areas to detect excessive radiation levels and to initiate appropriate safety actions.
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Enclosure 1 to NRC-98-0063 Page 7 l
Response
l New fuel at Fermi 2 is stored and handled in three areas: 1) the refueling floor,2) the l in plant truck bay and crane bay areas, and 3) the yard areas.
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- 1. On the refueling floor, a fully compliant 10CFR70.24 Criticality Accident
- Monitoring System is in place.
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- 2. In the truck bay and crane bay areas, there are installed plant area radiation monitors (ARMS) that provide an audible alarm and control room alarm if excessive radiation levels are detected. In addition, the Detroit Edison new fuel l handling procedure requires single failure proof rigging methods in compliance l with NUREG-0612 requirements be used when transporting inner metal fuel containers from the first floor of the Reactor Building to the Refueling Floor in the Reactor Building. This requirement ensures that critical container geometry's cannot occur from a postulated load handling incident caused by a single failure, i
- 3. In the outside yard area, Detroit Edison does not believe either criticality monitoring or area radiation monitoring to be necessary for the reasons outlined below and supported by the following discussions.
- a. Proceduralized criticality controls for storage ofinner metal and outer wooden fuel shipping container combinations preclude critical configurations.
The Detroit Edison new fuel handling procedure provides limitations on l storage ofinner metal plus outer wooden fuel container combinations.
l Loaded or empty stored inner metal plus outer wooden fuel containers are to be stacked no more than four containers high. In addition, the procedure requires that outer wooden and inner metal fuel shipping container lids will remain securely closed when not being unloaded. The limitation on stack height is a geometry control, and the physical design of the outer wooden l
container and requirement for the container lids to be secure provides some
- moderation control, in that these requirements preclude all but the most severe flooding of the storage area from challenging the leak tightness of the inner metal fuel containers. GE analysis (Reference 4) has shown that inner l metal plus outer wooden fuel container arrays that meet this geometry l criterion will remain deeply subcritical, even with optimal inter-unit moderation.
- b. Proceduralized criticality controls ior storage ofinner metal fuel shipping containers preclude dical configurations.
The Detroit Edison new fuel handling procedure provides limitations on I storage ofinner metal fuel containers. In addition, the procedure requires that l l l
Enclosure 1 to NRC-98-0063 Page 8 4
outer wooden and inner metal fuel shipping container lids will remain l
securely closed when not being unloaded. Loaded or empty stored inner l metal fuel containers are to be stacked no more than three containers high and i these stacked arrays must be elevated off any floor surface by a minimum of l 3.5 inches. The limitation on stack height is a geometry control, and the l
elevation and secure container lid requirements are moderation controls, in -
L that these requirements preclude all but the most severe flooding of the
( storage area from challenging the leak tightness of the inner metal fuel
! containers. GE analysis (Reference 4) has shown that inner metal fuel container arrays that meet this geometry criterion will remain deeply suberitical, even with optimal inter-unit moderation.
! c. Proceduralized criticality controls for transportation ofinner metal fuel shipping containers preclude critical configurations.
For inner metal fuel containers to be transported, the Detroit Edison new fuel j handling procedure provides a limitation that no more than 18 inner metal l
- fuel containers may be transported at any one time, and that the transported j l inner metal fuel container arrays are to be stacked no more than three containers high. In addition, the procedure requires that outer wooden and inner metal fuel shipping container lids will remain securely closed when not being unloaded. In this case, moderation control is implicitly provideo by elevation off the floor by the cart, trailer or rigging device and the requirement for the container lids to be secure. The geometry controls for
- transportation are the same as that for storage. However, the limitation of a maximum of 18 inner metal fuel containers to be transported at any one time is provided to er.sure such transportation is a factor of at least 10 conservative to that number demonstrated to be safe under hypothetical accident conditions for NRC approved interstate highway transport. GE analysis (Reference 4) has shown that damaged inner metal fuel container arrays that meet this geometry criterion will remain deeply subcritical, even with optimal inter-unit moderation and full water reflection conditions.
For the above reasons, Detroit Edison believes that neither criticality monitoring nor radiation monitoring is necessary for storage and transportation of unirradiated nuclear fuel in the yard areas.
Detroit Edison maintains a fully compliant criticality accident monitoring system on the Refueling Floor, an area radiation monitoring system in the truck bay and crane bay areas, and procedural controls on geometry and moderation to preclude criticality I
in all areas, including the yard areas. Therefore, Detroit Edison believes we satisfy
- the intent of the conditions specified in this criterion.
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r Enclosure 1 to NRC-98-0063 Page 9 l
Criterion 7. The maximum nominal U-23 : 2nrichment is 5 wt%
Response
i Detroit Edison's nuclear fuel supplier is licensed to 1.andle a maximum of 5 weight l
percent emichment in their fuel fabrication facility. In addition, our enrichment l supplier is only certified, and specified by contract, to enrich natural uranium to 5 weight percent. Therefore, Detroit Edison does not use, or have access to, fuel with enrichments greater than 5 weight percent. Therefore, Detroit Edison believes that we have satisfied this specific criterion.
1 Cost Benefit A considerable amount of resources would be expended to install, maintain, and operate a criticality accident monitoring systen; at Fermi 2, to satisfy the l requirements of 10CFR70.24, without a comparable increase in plant safety. )
l Therefore, installation of a monitoring system designed to meet these requirements 1
! does not appearjustified or necessary.
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Risk to Public Health and Safety Since the quantity, geometry and moderation controls preclude criticality of the SNM l
in question, public health and safety considerations are preserved.
Environmental Assessment I Since all fuel handling activities at Fermi 2 are performed in accordance with ,
approved procedures intended to assure non-criticality and radiation safety, environmental impacts from an inadvertent criticality are not expected. Therefore, granting this exemption will have no significant adverse impact on the quality of the j enviromnent.
Conclusion Based on the above exemption request and supporting justification, Detroit Edison has concluded that operation of Enrico Fermi Unit 2 in accordance with the proposed exemption to 10CFR70.24 is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security.
Detroit Edison considers that good cause for granting an exemption has been demonstrated, and therefore, the requested exemption should be granted in accordance with the requirements of 10CFR70.24(d).
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Attachment 1 to !
NRC-98-0063 Page1 l
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Attachment 1 Affidavit i
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I O
GE Nuclear Energy 7
GeneralElectHc Company l
P. o. Box 780, %1Imington, Ne 28402 I
i Affidavit l
l I, Glen A. Watford, being duly swom, depose and state as follows:
(1) I am Manager, Nuclear Fuel Engineering, Gmeral Electric Company ("GE") and have been l delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in the attachment to the letter, R. D.
Williams (GE) to James M Thorson (DECO), New Fuel Receipt Criticality Safety Analysis, dated April 7,1998.
(3) In making this application for withholding of proprietary information of which it is the ov,ner, g GE relies upon the exemption from disclosure set forth in the Freedom ofInformation Act
- # ("FOIA"),5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4) and 2.790(a)(4) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential"(Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information," and some portions also qualify under the narrower definition of" trade secret," j within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, j respectively, Critical Mass Enerav Proiect v. Nuclear Rsgu'atory Commission. 975F2d871 j (DC Cir.1992), and Public Citizen Health Research Group v. FDA,704F2dl280 (DC Cir. j 1983).
(4) Some examples of categories ofinfonnation which fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention ofits use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;
- b. Infonnation which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers; j
- d. Information which reveals aspects of past, present, or future General Electne j customer-funded development plans and programs, of potential commercial value to General Electric; ;
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- e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
- f. TLe information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.
(5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confida1ce by GE, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The infonnation sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which proside for maintenance of the infomtation in confidence.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the infonnation in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.
(7) The procedure for approval of extemal release of such a document typically requires resiew by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and pota1tial customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) 'Ihe infonnation identified in paragraph (2) is classified as proprietary because it contains details of GE's Criticality Safety Analysis processes and procedures. The development of the methods and procedures used in these analysis was achieved at a significant cost to GE.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The fuel design is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
The research, development, engineering, analytical, and NRC review costs comprise a l substantial investment of time and money by GE.
The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difYicult to quantify, but it clearly is substantial.
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GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent I understanding by demonstrating that they can arrive at the same or similar conclusions. I The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to I undertake a similar expenditure of resources would unfairly provide competitors with a j windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate retum on its large investment in developing these very valuable analytical tools.
State of North Carolina ) gg.
County of New Hanover )
Glen A. Watford, being duly sworn, deposes and says:
That he has read the foregoing affidavit and the matters stated therein are true and i correct to the best of his knowledge, information, and belief.
Executed at Wilmington, North Cara ina, this IN day of A f>n l , 19 9 ?
A /'
len A. \jV ord General Electric Company l 7L Subscribed and sworn before me this // day of dEC
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,19 ff' s )ft. h>ulc%N Nota Public, State of North Carolina My commission expires /0/e///t
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Attachment 3 to j NRC-98-0063 '
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l Attachment 3 Non-Proprietary Version of GE New Fuel Receipt Criticality Safety Analysis L
Criticality Safety Analysis itA-3 Storage and Ilandling at Ileactor Utilities I. SCOPE This analysis is performed to demonstrate criticality safety for the safe storage and transport of GE BWR nuclear fuel designs (8x8,9x9, or 10x10 rod lattice) in RA-3 containers at BWR reactor utilities.
Specifically, this evaluation is performed under GE Proposal No. 283-lilMMW-97RI and subsequent change to Contract No. NS-341230 for the Detroit Edison Company (DECO) Enrico Fermi 2 reactor unit to support exemption from 10CFR70.24 criticality monitoring requirements during onsite storage and transport of up to 18 loaded RA-3 inner containers in areas not covered by a criticality monitoring system.
II. GENERAL DISCUSSION llACKGROUND For more than 25 years, the RA-series shipping containers have been successfully used by General Electric Nuclear Energy Production (GE NEP) to ship BWR fuel bundles from its Wilmington, NC fuel fabrication facility to domestic and international customers. The RA-series containers consist of rectangular steel
' inner' containers transported in wooden ' outer' overpacks. The wooden overpack containers are designed with ethafoam and honeycomb cushioning between the metal inner ar.d the inside walls of the wooden outer. The inner metal container has two internal ethafoam cushioned channel sectior.s each of which can hold a single fuel assembly (8x8,9x9, or 10x10 lattice designs).
The original RA-1 inner container was modified in the 1970's to accommodr.e a longer fuel assembly. j This was accomplished by adding a larger end cap to the body of the inner. The new design was designated l the RA-2. Subsequently, a longer bodied RA-3 (with shorter end cap) was introduced. Corresponding i changes in the wcoden outer container were also made at each stage to accommodate the new inner container design.
GE's latest USNRC Certificate of Compliance No. 4896 is approved for shipment of BWR nuclear fuel ,
bundles (ref. 2). The US NRC Certificate states that the RA-2 and RA-3 containers meet the applicable l safety standards set forth in Title 10, Code of Federal Regulations, Part 71, Packaging and Transportation l of Radioactive Material. The latest USDOT Certificate of Competent Authority No. USA /4986/AF is also valid (ref. 3). The US DOT certificate states that the RA series container meets the regulatory requirements of the International Atomic Energy Agency Safety Series No.6," Regulations for the Safe Transport of I Radioactive Materials," 1973 Revised Edition, as amended, IAEA, Vienna, Austria, and the Unit'ed States of America, Title 49, Code of Federal Regulations, Parts 100-199.
REACTIVITY ASSESSMENT (S)
Prior safety evaluations demonstrate the RA-3 container package (metal inner + wooden outer overpack )
meets 10CFR71, " Packaging of Radioactive Material for Transport." These previous analyses demonstrate safety for both normal and hypothetical accident conditions for the 8x8,9x9, and 10x10 rod lattices as defined in 10CFR71 and show that the GE-1210x10 rod lattice is the ' limiting' or most rea tive fuel bundle design (ref.1). These analyses are or liave been approved by both the USDOT and USNRC as valid safety demonstrations (refs. 2,3) for interstate highway transport.
April 7,1998
1 l
I CSA - RA-3 Storage and llandling of IlWR Fuel (Non P) f Page 2 0f 21 l l t
This analysis uses the GE-12/1410x10 rod lattice, however the prescribed controls may be equivalently applied to other 8x8 (GE-9/10) and 9x9 (GE-11/13) rod lattices fuel designs. The following dimensions apply to the GE-12 rod lattice designs:
Fuel OD =
Clad ID = ,
l Clad OD = '
Rod Array = 10 x 10 Rod Pitch = 0.510" Previous analyses have modeled the fuel by " smearing" the actual fuel volume over the entire inside cladding volume (ref. 4,5). This effectively fills the clad ID with fuel. The fuel density is therefore reduced as a fraction of the theoretical corresponding to the following:
UO2 theoretical density = 10.96 Actual density fraction of theoretical = 0.98 Density reduction factor due to smearing =.
The modeled fuel density is therefore 10.96
- 0.98 * =
g/cc and is also employed in the calculations herein. As was previously analyzed, the plastic rod spacers were modeled as plastic cylinders surrounding each fuel rod. The older plastic separators are used as the basis, since the new cluster separators have been demonstrated to be less reactive than the older individual plastic separator model treatment (ref. 6)in which a thin annulus of high density polyethylene surrounds each fuel rod. The specification for the RA containers are well documented in previous shipping analyses and referenced drawings (refs.1-3) and are not repeated here.
IlWR REACTOR EXEMPTION Section (a) of 10CFR70.24,"Criticahty Accident Requirements," requires that each licensee authorized to porsess special nuclear material (SNM) shall maintain in each area where such material is handled, used, or stored, an appropriate criticality monitoring system. In accordance with Subsection (a)(1) of 10CFR70.24, coverage of all such areas at the DECO reactor units shall be provided by two criticality detectors.
However, exemptions may be requested pursuant to 10CFR70.24(d), provided that the licensee believes that good cause exists for the exemption. In particular, Reg Guide 8.12, Revision 2,' Criticality Accident Alarm System," states that it is appropriate to request an exemption from 10CFR70.24 if an evaluation determines that a potential for criticality doer not exist, as for example where geometry controls (e.g.,
spacing) is used to preclude a critical excursion.
DECO is hereby requesting an exemption from criticality monitoring requirements of 10CFR70.24 described above for receipt, storage, and handling of RA-3 contamers until the 8x8,9x9, or 10x10 nuclear fuel bundles are removed from the inner container at the refuel floor (at which point criticality monitoring is fully compliant with 10CFR70.24).
Specifically, the exemption request covers the following activities:
The fuel shipment arrives onsite in undamaged RA-3 " packages"(loaded inner + outer RA-3 containers).
Receipt normally consists of up to 15 loaded RA-3 packages to comply with USDOT interstate highway weight restrictions. The RA-3 pac!: ages are unloaded at the Ferrni 2 facility and may be stacked in arrays of up to 5 packages high. (This analysis demonstrates that an infinite planar array ofloaded inner + outer packages stacked 5-high remains subcritical smder optimalinterunit H2O moderation conditions, with one foot of water above and onefoot of water or 18" ofconcrete beneath the planar array}.
April 7,1998
r CSA - RA-3 Storage and liandling of BWR Fuel (Non l')
l l' age 3 oI21 Up to 15 loaded metal inner containers are unloaded from the outer uooden RA-3 containers and are stored or transported in a maximum stacked arrays of up to 3 inner metal containers high.(This analysis demonstrates that an infinite close-packedplanar array ofloaded inner packages stacked 3-high remains subcritical under optimal interunit H2C moderation conditions with onefoot of water above and onefoot
, of water or 18" ofconcrete beneath the planar arrav}.
l l The 15 RA-3 inner containers (in maximum stacked arrays of up to 3 high) are then transported to the l reactor building tefueling floor. (This analysis demonstrates that an infinite close-packedplanar array of loaded inner RA-3 containers stacked 3-high remains subcritical smder optimal interunit H2O moderation fully reflectedconditions).
CRITICALITY SAFETY CONTROLS The following controls are necessary to meet this analysis:
l Control #1 (Geometry)- Only approved RA containers may be used to safely store GE BWR fuel l
bundles. RA-3 inner with outer containers may be stored up to 5-high, otherwise unlimited in x-y plane.
RA-3 inner containers (without outers) may be stored up to 3-high, otherwise unlimited in the x.y plane.
RA-3 inner containers may also be used for ONSITE TRANSPORT between unloading or storage areas and the reactor building refueling floor (up to 18 inner containers may be transported,3 high maximum stack height).
Control #2 (Moderation)- RA-3 inner and outer containers are approved for onsite fuel storage and transport. Only approved fuel bundle moderators such as cluster separator (s) and plastic wrap are permitted >
beyond container structural cushioning materials. RA-3 inner with and without outer container (s) must be l
stored with the RA container lids secure. Stored inner container (s) must be elevated a minimum of 3.5" off the floor (e.g, using a standard 4"x4" timber). When in transit, this elevation requirement may be provided by the conveyance vehicle (e.g., fann cart).
The following assumptions are made for the RA-3 container arrays:
Worst Credible Contents Form: UO2 Density: 10.96
- 0.98 i U235 Enrichment: 5.00 %
Structure: Carbon Steel .
Moderation: Optimum Boundary Conditions Top: 30.48 cm water Bottom: 30.48 cm water (or 45.72 cm ORC)
Sides: Mirror Interunit Water: Optimum IV. MODEL l
The GEMER Monte Carlo models constructed for this analysis consist principally of an inner and outer I i
wooden container meeting RA-3 design specifications identified in referenced Certificate of Compliance ,
(ref. 2) drawings, as amended. The inner container modeled may be constructed of either carbon steel (RA-3). Due higher neutron absorption properties of stainless steel, the carbon steel was conservatively selected l
[
I to represent the inner container structure. 1 April 7,1998
CSA - RA-3 Storage and llandling of BWR Fuel (Non P)
Page 4of21 l Two GE-12 fuel bundles were placed in each carbon steel inner container. The fuel was assumed 5.00 wt.
% U235 uniform enrichment over the full axiallength. For conservatism, no gadolinium rods were l modeled. The inner container was then modeled by itselfin an in6 nite 3-high planar array; or placed into the outer wooden container and modeled in an infmite 5-high planar array. The planar array was achieved l by modeling mirror reDection on the x faces and the z faces. To represent the shipping accident conditions, the two fmite faces (1 y) were fully reflected using 12-inches (30.48 cm) full density water on top and bottom. To represent the storage accident condition, the finite face (+y) was fully reflected using 12-inches (30.48 cm) full density water, while the finite face (-y) was fully reflected using 18 inches (45.72 cm) of oak ridge concrete.
RA-3 inner container analysis - model filename convention:
a-yyy.in where, a = ' inn' for inner container storage (3-high); wt.ter reflected array a = 'out' for inner + outer container storage (5-high); water reflected array a = 'innc' for inner container storage (3-high); concrete bottom, water top reflected array a = 'oute' for inner + outer container storage (5-high); concrete bottom, water top reflected array yyy = percent interunit H2O (x10)
For example, the file name inn-125.in represents an infinite 3-high planar array ofinner containers at l 12.5% interunit 1120. Sample mixtures used are provided in Attachment 1. Sample input files and I associated geometry plots are shown in Attachments 3,4 V. CALCULATIONAL RESULTS I
These calculations were performed with the GEMER Monte Carlo neutronics program on the microcomputers at the GE NEP facility (ref. 7). GEMER (" Geometry Enhanced MERIT") is a modined version of the Battelle Northwest Laboratory's BMC Monte Carlo Code which has been combined with the geometry handling subroutines in KENO IV, and later improved to include the more capable GEKENO geometry treatment (ref. 7). Cross section sets in GEMER are processed fami the ENDF/B-IV library in 190 broadgroup and resonance parameter formats except for thermal scattering in water which is represented by the Haywood Kernel in the ENDF/U-IV library.
In GEMER, the resonance parameters describe the cross sections in the resonance energy range and Monte Carlo sampling in this range is done from the re onance kernels rather than from broad group cross sections (i.e., explicit treatment of resolved resonances). Thus there is a single unique cross section set associated with each available isotope and dependence is not placed on Dancoff(Dux shadowing) correction factors or effective scattering cross sections. The cross section library includes fission, capture, elastic, inelastic, and (n,2n) reactions. Absorption is implicitly treated by applying non-absorption probability weights at each collision point.
The calculational bias of GEMER may be represented by the following function of the hydrogen-to-U235 ratio (or W/F ratio):
I f
l Bias !s conservatively applied over its negative range and assigned a value of zero over its positive range.
Attachment 2 contains a summary of the W/F ratio as a function ofinterunit H20. The resulting GEMER bias using above equation is then applied.
l April 7,1998
1 CSA - RA-3 Storage and llandling of BWR Fuel (Non P) l Page S of 21 Calculations were nominally .un with 110 generations 1000 neutrons each. The first 10 generations were ,
skipped before starting the statistical output processing. Calculations were performed on the
{
microcomputer identified as:
1 organization: GENE, NEP, CRIT.S AFETY.ENG, WILMINGTON,NC 1 system: PAULSON, PENTIUM,200MHz hardware: Compaq, DeskPro, Serial No.6710BBND178 The fol owing validated code / cross-section library version was used:
PROGRAM NAM: C:\ PROGRAMS \GEMER.EXE PROGRAM VER: 00 PROGRAM DAT: IC81 PROGRAM TIM: 3F5C l PROGkAM SIZ: 4541898 j LIBRARY NAM: C:\XSEC. LIB \GEMLIB i LIBRARY DAT: 1 AB3 LIBRARY TIM: 6E2F LIBRARY StZ: 41021424 1
Figure I A shows the summary calculations for both the RA-3 inner (3-high infinite array) for both the l water reDected condition and the concrete bottom reflected condition. Figure 1B shows the same two I curves for the RA-3 inner + outer package (5-high infinite array) as a function ofinterunit H20. Table 1 provides a listing of the modified kiist.dat output which includes the Keff + 30 - bias results.
It is noted the maximum reactivity of the inner container system occurs at 10-12.5% interunit 1120, which ,
agrees with previous internal storage result for the GE12 lattice (refs. 4,5) . In addition, the results indicate l the inner souter package array is over-moderated, that is, as interunit water is added, the system reactivity decreases. In either case, the 18-inch concrete reflected system is slightly more reactive than the water-reflected counterpart, thus planar array storage is also confirmed safe.
l April 7,1998 l
~l CSA RA-3 Storage and llandling of HWR Fuel (Non P)
Page 6 of 21 l FIG. lA: HA-3 INNER CONTAINER - GE12 LATTICE, NO GD203 e.,6e LEGENO atactos cattTr stonnct e tantes s.nti* nJo stro n tuutas 8-nl6n ORC tt9L e.990
= ACC30 TNT LIMIT x e [e.9 l
- . I fx N 1
/
E-tFF k)#
7 l .....
I i I l ..... l l l e.e90
...:e
- .i. s. 7. ne ise ... 23.
INTERUNIT n20 X1.
FIGe.,6e 1B: RA-3 INNER +00TER PACHAGE - GE12 LATTICE, NO GDZO3 LEGEND ;
REncfoe cut tTE STopact
/
i e inn +0uTes ngsn n20 Atr l l u !NN+00TeS-MIGN CAC REF I
. 94. - ACCIDENT LIMIT s e.95 I
l e.92e l
e.96.
K-EFF ese l
si i
Il
.. 6. i l
l l
'l l
..e9e ii l
e.e2e
.. se se se Te ee 8
intriumarn20 xte April 7,1998 1
I
CSA - RA-3 Storage and llandling of BWR Fuel (Non P) ,
l Page 7 of 21 l Table 1. RA-3 Container Array Calculation Summary File ID KEFF SIGMA BlAS K+3S-Il filST SKIP DATE ELAPSED LOST inn-000 0.83082 0.00312 0.00000 0.84018 46000 64 10/14/97 26.53 319 inn-025 0.8681 0.00254 0.00000 0.87572 46000 64 10/14/97 24.58 208 inn-050 0.88725 0.00213 0.00000 0.89364 100000 l0 10/14/97 23.43 178 inn-075 0.89828 0.00207 0.00000 0.90449 94000 16 10/14/97 22.53 172 inn-100 0.90654 0.00209 0.00000 0.91281 86000 24 10/14/97 21.23 152 inn-125 0.91695 0.00207 0.00000 0.92316 86000 24 10/14/97 20.82 151 inn-150 0.90972 0.00215 0.00000 0.91617 93000 17 10/14/97 20.35 136 inn-200 0.90023 0.00233 -0.00138 0.90860 100000 10 10/14/97 19.8 112 j inn-300 0.88195 0.0027 -0.00397 0.89402 51000 59 10/14/97 19.28 l00 inn-500 0.85631 0.00265 -0.00861 0.87287 92000 18 10/14/97 19.4 83 in n-full 0.89011 0.00282 -0.01732 0.91589 100000 10 10/14/97 19.97 64 innc-025 0.89364 0.00217 0 00000 0.90015 100000 10 10/15/97 24.75 191 in nc-075 0.92448 0.00198 0.00000 0.93042 98000 12 10/16/97 21.35 157 innc-100 0.93404 0.00236 0.00000 0.94112 71000 39 10/16/97 21.45 164 innc-125 0.92901 0.00224 0.00000 0.93573 97000 13 10/16/97 30.07 152 innc-150 0.92435 0.00213 0.00000 0.93074 81000 29 10/16/97 19.55 115 inne-200 0.91616 0.00194 -0.00138 0.92330 71000 39 10/16/97 19.07 108 out-000 0.87502 0.00209 0.00000 0.88129 81000 29 10/14/97 27.7 225 out-025 0.86869 0.0027 0.00000 0.87679 31000 79 10/14/97 26.8 153 out-050 0.84857 0.00188 0.00000 0.85421 94000 16 10/14/97 26.48 122 ou t-075 0.8366 0.00223 0.00000 0.84329 81000 29 10/14/97 26.25 143 oute-000 0.88795 0.00206 0.00000 0.89> l 3 93000 17 10/16/97 26.77 228 i
oute-025 0.87936 0.00431 0.00000 0.89229 21000 89 10/I6/97 26.4 159 oute-050 0.86177 0.00367 0.00000 0.87278 21000 89 10!!6/97 28.18 158 oute-07f 0.84155 0.00199 0.00000 0.84752 97000 13 10/16/97 31.85 127
- GEMER bias calculation based on W/F ratio (Attachment 2).
l l
l April 7,1998
CSA - RA-3 Storage and Handling of HWR Fuel (Non P)
Page 8 of 21 l VI. SAFETY DURING UPSET CONDITIONS USNRC and IAEA Regulations 10CFR 71.59, Standards For Arrays of Fissile Material Packages adopts IAEA safety series no. 6 regulations mr demonstration of safety in the U.S.. In particular, subsection (a) states that a Hssile material package mt,t be controlled by either the shipper or the carrier during transport to assure that an array of such packages remains subcritical. To enable this control, the designer of a fissile material package shall j derive "N" based on all the following conditions being satis 6ed, assuming packages are stacked together in l any arrangement and with close full reflection on all sides of the stack by water; l
(1) Hve times "N" undamaged packages with nothing between the packages would be subcritical l
(2) two times "N" damaged packages, if each package were subject to the tests speciGed in section 71.73 (hypothetical drop, puncture, thermal, and water immersion accident conditions) would be subcritical with optimum interspersed hydrogenous moderation.
Loss of Geometry and/or Moderation Conditions ;
Referer.ce analysis clearly demonstrates that under normal conditions, an infinite array ofloaded l
(undamaged) RA-3 containers is subcritical(Keff < 0.95). This atisGes the above "5N" undamaged l stipulation. In this case, the undamaged container is over-mod _ rated. By removing moderator from the j undamaged container model, the system becomes more rear. ve (conservative). The reference safety 1 demonstration shows that by removing the cluster separators, excluding the ethafoam, and only using half density honeycomb in the model, the infinite array Keff result is 0.9424 (ref.1.3).
Reference analysis also clearly demonstrate that N=130, and that a close packed array of 260 (damaged) containers in a 20x13x1 close packed array also remains subcritical(Keff < 0.95) under fully reRected, optimal interspersed hydrogenous moderation condition. The reference safety demonstration shows that for the limiting 10x10 rod lattice, using the limiting Gad rod distribution, the maximum reactivity result under ,
hypothetical accident conditions is 0.9494. This satisfies the above "2N" accident condition stipulation. l This analysis supports the above results, and shows that a 3-high in6 nite planar array ofinner containers remains subcritical under optimal interunit H20. This work also demonstrates that a 5-high in0 nite planar array of inner + outer RA-3 packages remains suberitical under optimal (zero interunit 1120) moderation conditions.
In summary, the proposed DECO onsite storage and transport of up to 18 inner RA-3 containers in arrays of l up to 3 high and loaded with fuel of GE lattice designs to the reactor building refueling floor is safe - and quali0es for exemption under 10CFR70.24(d). Good cause has been demonstrated that this limited number of containers will remain deeply subcritical- even under loss of geometry and/or moderation conditions. If the reference shipping container results are used for comparison purposes, the onsite transport limit of 18
' inner' RA-3 containers is a factor of-14 less than the number demonstrated safe under hypothetical accident conditions (2N=260) for interstate highway transport. Clearly, controlled onsite transport of 18 loaded inner containers poses less risk to onsite workers and the public.
- - Vll. SPECIFICATIONS AND REQUIREMENTS FOR SAFETY April 7,1998
1 CSA - RA-3 Storage and llandling of BWR Fuel (Non P) l Page 9 0f 21 The design specification (s) include:
U235 Enrichment: 5.00 wt. %
Uranium Form: GE Fuel Bundles (8x8,9x9,10x10 rod lattice)
Geometry: RA-3 inner / Outer Containers The criticality safety monitoring exemption requirements for DECO onsite storage and transport of RA-3 package containers includes:
- 1. ADM: RA-3 container lids must remain securely closed when not in use (e.g, when not unloading).
- 2. OPR: Loaded or empty RA-3 inner + outer container storage array is limited to 5-high , and is otherwise unlimited in the x-y plane.
- 3. OPR: Loaded or empty RA-3 inner container storage array is limited to 3-high , and is otherwise unlimited in the x-y plane. Inner storage arrays shall be elevated off Goor a minimum of 3 1/2 inches.
- 4. CPR: A maximum of 18 loaded RA-3 inner containers (limited to 3-high) may be transported ONSITE in a controlled manner (e.g., using ' farm carts', or equivalent).
Vill. CONCLUSIONS Safe storage and handling criteria are established by this analysis. Calculations demonstrate that limiting the stack height geometry of the loaded RA-3 inner / outer container (s) results in acceptable margin of safety. In all cases, the accident condition k-effective is less than 0.95 as required by the BWR Reactor Facility Commercial Operating License.
In summary, the proposed onsite transport of 18 inner RA-3 containers is safe - and qualiGes for exemption under 10CFR70.24(d). Good cause has been demonstrated that this limited number of containers will remain deeply suberitical- even under loss of geometry and'or moderation controls.
1 Analysis by: /-
Date: J/7/98 Lon E. Paulson VeriGed by: . Date: J/7/98 fjohn T. Taylor 1
J l
i
-- -. REFERENCES
):
April 7,1998 I 1
CSA - RA-3 Storage and llandling of !!WR Fuel (Non P)
Page LO of 21 l
- 1. " Request for Renewal of NRC Certit. te for the Model RA-2 and RA-3 Shipping Container -
Consolidated Application," RJ Reda, Manager Facility Licensing, to CR Chappell. USNRC Package Certi0 cation Section, September 10,1997.
1.1 Section 8.0, Appendi;I," Criticality Safety Analysis for the RA-3 shipping container with generic 8x8 fuel assemblies with Cluster Separators". Original CSA dated 6/8/95.
1.2 Section 8.0, Appendix J," Criticality Safety Analysis for the RA-3 shipping container with generic 9x9 fuel assemblies with Cluster Separators". Original CSA dated 6/1/95.
1.3 Section 8.0, Appendix K," Criticality Safety Analysis for the RA-3 shipping container with generic 10x10 fuel assemblies with Cluster Separators". Original CSA dated 5/22/95.
- 4. CSA "RA-3D Shipping Container with Generic 10x10 Fuel Assemblies, RE Stachowski,12/4/92.
- 6. CSA "Use of Cluster Separators,' CR 96.0091, JT Taylor,3/28/96.
- 7. "GEMER - Microcomputer Version Users Guide," JT Taylor,4/20/94
- 8. NRC Information Notice 97-77, Exemptions from the Requirements of Section 70.24 of Title 10 of the Code of Federal Regulations," October 10,1977.
LIST OF ATTACilMENTS
- 1. sample mixtures: treename - nuclide ID's and number densities
- 2. water-to-fuel: W/F ratio cal.tation vs. interunit H2O density
- 3. inn-125.in sample input file,2D & 3D plots w/ fission distribution
- 4. out-025.in sample input Gle,2D & 3D plots w/ fission distribution l
I Attachment 1. sample mixtures: treename vs. nuclide ID's and number densities April 7,1998 1
CSA - RA-3 Storage and flandling of BWR Fuel (Non P) l Page 11 of 21 1.\CSXSEGUO2\GUO2-50.TD 3 293 0 0 Guo2-50.TD 16 4.60580E-02 1 2.\CSXSEC\NOLAGNOU-0.WAT 0.05 2 293 0 0 WATER I
! l 6.6866E-02 16 3.3433E-02 l 3.\CSXSEC\NOLAGNOU-0.flDP 2 293 0 0 It!GII DENS. POLYETliYLENE 1 8.2938E-02 I 12 4.1469E-0?
4.\CSXSEC\UO2\CUO2-47. GAD 4 4 293 0 0 Guo2-47. gad I i
l _
l 16 4.7492E-02 l 5.\CSXSEC\NOLAGNOU-0.CS f 2 293 0 0 CARBON STEEL I
< 12 3.9210E-03 l
26 8.3491E-02 6.\CSXSEC\NOLAGNOU-0.ETil 2 293 0 0 ETilAFOAM, FULL DENSITY I 3.0300E-03 l 12 1.5150E-03 1 7,\CSXSEC\NOLAGNOU-0.IINY l 3 293 0 0 llONEYCOMB
! l 3.0131E-03 l 12 2.0929E-03 16 1.22197-03 8.\CSXSEC\NOU\GNOU-0.WAT O.125 2 293 0 0 WATER I 6.6866E-02 16 3.3433 E-02 9.\CSXSEC\NOU\GNOU-0.WAT 2 293 0 0 WATER 1 6.6866E-02 l 16 3.3433E-02 10.TCSNSEC\NOLAGNOU-0.WD 3 293 0 0 WOOD, FULL DENSITY I 2.1334E-02 =
l 12 1.1858E-02 16 8.59330-03 11.\CSXSEC\NOLAGNOU-0.CS 0.85 2 293 0 0 CARBON STEEL i 12 3.9210E-03 26 8.3491 E-02 j 12.\CSXSEC\NOLAGNO U-0.ETil 0.50
[
2 293 0 0 ETilAFOAM, FULL DENSITY I 3.0300E-03 12 1.5150E-03
, 13.\CSXSEONOLAGNOU-0.ZR I
I 293 0 0 95% FULL DENSITY ZlRC I
401 4.07091 E-02
. - ., Attachment 2. W/F Calculation vs. Interunit 1120 Density April 7, .1998
CSA - RA-3 Storage and llandling of BWR Fuel (Non P)
Page 12 of 21 l WIF Calculation: RA 3 inner Container w/ GE-12 Bundle The GEMER validation demonstrates that the analytic bias increases as the H/U235 ratio (or water-to-fuel) ratio increases. The maximum W/F ratio for a given interunit H2O condition may be computed as follows:
- 1. Calculate effective moderator area within the assembly: I A (H2O) = [A(assy) A (rod + plastic)]* max, H2O density EA (H2O) = A (H2O) + A (plastic) l l
- 2. Calculate effective area of moderator outside assembly, but inside the inner container. l l l IA (H2O) = [A(inner) - 2*A(assy) - A(structure)]* max. H2O density 3 Calculate total effective moderator area TOT-EA (H2O) = 2*EA(H2O) + IA (H2O) j 4. Calculate total fuel area l A(uo2)=N*pi(inner clati radius)"2 l l l Computed Values: GE12 Bundle in RA 3 I rho mix:
rho-h20= 0.025 0.05 0.075 0.1 0.125 0.15 0.2 l assembly:
A(assy) = 167.8061 167.8061 167.8061 167.8061 167.8061 167.8061 167.8061 A(rod) = 73.8550 73.8550 73.8550 73.8550 73.8550 73.8550 73.8550 A(plastic)= 31.3422 31.3422 31.3422 31.3422 31.3422 31.3422 31.3422 A(wat. rod)= 8.5775 8.5775 8.5775 8.5775 8.5775 8.5775 8.5775 A(H2O)= 1.5652 3.1304 4.6957 6.2609 7.8261 9.3913 12.5218 EA(H2O)= 32.9074 34.4727 36.0379 37.6031 39.1683 40.7336 43.8640 inner:
A(inner)= 1301.00 1301.00 1301.00 1301.00 1301.00 1301.00 1301.00 A(assy)= 167.8061 167.8061 167.8061 167.8061 167.8061 167.8061 167.8061 A(struct.)= 32.0000 32.0000 32.0000 32.0000 32.0000 32.0000 32.0000' lA(H2O)= 23.3347 46.6693 70.0040 93.3386 116.6733 140.0079 186.6772 l
total EA:
1 TOT EA(H2O)= 89.1495 115.6146 142.0797 168.5448 195.0099 221.4750l 274.4052 i
i two fuel s.sy, A(uo2)= 115.5002 115.5002 115.5002 115.5002 115.5002 115.5002 115.5002 water-to-fuel:
l W/F = TOT-EA(H2O) / A(uo2) l W/F = 0.7719 1.0010 1.2301 1.4593 1.6884 1.9175 2.3758 1 WTFR-H2O = 0.0696 0.0884 0.1065 0.1239 0.1406 0.1567 0.1871 H/U235= 44.02449 57.09368 70.16287 83.23206 96.30125 109.3704 135.5088 G-BIAS = 0.003595 0.002848 0.002113 0.00139 0.00068 -1.93E-05 -0.001382 Attachment 3. inn-125.in sample input file,2D & 3D plots w/ fission di.;tribution l April 7,1998
CSA - RA-3 Storage and llandling of BWR Fuel (Non P) l Page 13 of 21 inn-125.in
- 97. D A,,,CY L,UO2,5.00%, WTFR=.05,CS,125,R,C E I10 1000 10 0 0 0 0 0 I
0 293 0 0
\C9SEC\UO2\GUO2-50.TD
\CSXSEC\NOU\GNOU-0.WAT 0.125
\CSXSEC\NOU\GNOU-0.HDP
\CSXSEC\UO2\GUO2-47. GAD l
\CSXSEC\NOUiGNOU-0.CS '
\CSXSEC\NOU\GNOU-0.ETH l \CSXSEC\NOU\GNOU-0.HNY
\CSXSEC\NOU\GNOU-0.WAT 0.125
\CSXSEC\NOU\GNOU-0.WAT
\CSXSEC\NOU\GNOU-0.WD
\CSXSEC\NOU\GNOU-0.CS 0.85 l
\CSXSEC\NOU\GNOU-0.ETH 0.50
\CSXSEC\NOU\GNOU-0.ZR KENO GEOM 0 /* # OF REGIONS OR ZERO l 0 /* # OF BOX TYPES OR ZERO I 1 /* # OF BOXES IN X DIRECTION 1 /* # OF BOXES IN Y DIRECTION I /* # OF BOXES IN Z DIRECTION I /* BOUNDARY CONDITION OPTION 0 /* STARTING SOURCE OPTION 1 /* COMPLEX EMBEDDED OPTION 0 /* # OF PRINT PLOTS
-1.0 - 1.0 0.0 0.0 -1.0 -1.0 BOX TYPE 1 /* 5% fuel rod, no gad, hd poly wrap CYLINDER 1 441.960 0.000 16* 5 CYLINDER 13 441.960 0.000 16*.5 CYLINDER 3 441.960 0.000 16*.5 CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 2 /* 85% carbon steel basket, vertical section between bundles CUBOID 11 0.15900 -0.15900 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 8 0.15900 -0.15900 14.4930 -13.4470 441.960 0.000 16*.5 CUBOID $ 0.15900 -0.15900 14.6520 -13.6060 441.960 0.000 16*.5 BOX TYPE 3 /* left side vertical basket, left side inner container l CUBOlD 5 -2.06300 -2.38000 5.60300 -4.55700 441.960 0.000 16*.5 l CUBOID 8 -2.06300 -2.38000 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 11 -1.90500 -2.38000 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 8 -1.90500 -7.14300 14.4930 -13.4470 441.960 0.000 16*.5 l CUBOID 5 -1.90500 -7.30200 14.6520 -13.6060 441.960 0.000 16*.5 BOX TYPE 4 /* right side vertical basket, right side inner container CUBOID 5 2.38000 2.06300 5.60300 -4.55700 441.960 0.000 16*.5 CUBOID 8 2.38000 2.06300 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 11 2.38000 1.90500 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 8 7.14300 1.90500 14.4930 -13.4470 441.960 0.000 16*.5 l CUBOID 5 7.30200 1.90500 14.6520 -13.6060 441.960 0.000 16*.5 l BOX TYPE 5 /* complete inner container cuboid l CUBOlD 8 23.0200 -23.0200 14.1290 -14.1290 442.119 -0.159 16*.5 BOX TYPE 6 /* single fuel bundle + cs basket cuboid
~ -. CUBOID 8 14.4824 -2.9816 4.1386 -14.4344 441.960 0.000 16*.5 April 7,1998
{
CS A - RA-3 Storage and llandling of BWR Fuel (Non P)
Page M of 21 l l CUBOID 8 14.4824 -2.9816 8.7426 -19.1974 441.960 0.000 16*.5 i l CUBOlD 5 14.4824 -2.9816 8.9016 -19.3564 441.960 0.000 16*.5 BOX TYPE 7 /* 5% fuel rod, no gad, NOT USED CYLINDER 1 0.48770 441.960 0.000 16*.5 CYLINDER 13 0.55120 441.960 0.000 16*.5 CYLINDER 3 0.61390 441.960 0.000 16*.5 CUBOID 2 0.61390 -0.61390 0.61390 -0.61390 441.960 0.000 16*.5 BOX TYPE 8 /* water rod CYLINDER 2 1.16840 441.960 0.000 16*.5 CYLINDER 13 1.24460 441.960 0.000 16*.5 BOX TYPE 9 /* one quarter fuel bundle cuboid CUBOID 2 3.23850 -3.23850 3.23850-3.23850 441.960 0.000 16*.5 BOX TYPE 10 /* one quarter fuel bundle cuboid CUBOID 2 3.23850 -3.23850 3.23850 -3.23850 441.960 0.000 16*.5 BOX TYPE 11 /* upper and lower carbon steel support CUBOID 5 5.08200 0.00000 0.31700 0.00000 441.960 0.000 16*.5 BOX TYPE 12 /* 85% carbon steel horizont:1 basket CUBOID 11 14.8424 -2.9816 0.15900 0.00000 441 960 0.000 16*.5 BOX TYPE 13 /* 4.7% U235 fuel rod,2.85% gd203, NOT USED CYLINDER 4 0.44700 441.960 0.000 16*.5 CYLINDER 13 0.50550 441.960 0.000 16*.5 CYLINDER 3 0.57290 441.960 0.000 16*.5 {
CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 I BOX TYPE 14 /* one quarter fuel bundle cuboid I CUBOID 2 3.23850 -3.23850 3.23850 -3.23850 441.960 0.000 16*.5 BOX T"E 15 /* one quarter fuel bundle cuboid CUBO 2 3.23850 -3.23850 3.23850 -3.23850 441.960 0.000 16*.5 BOX TYPE 16 /* carbon steel end ofinner container CUBOID 5 23.0200 -23.0200 14.1290 -14.1290 0.15900 0.00000 16*.5 BOX TYPE 17 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 18 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 19 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 20 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 21 /* global unit: 3 high planar array (inner containers)
CUBOID 0 23.0200 -23.0200 70.6450 -14.1290 442.119 -0.159 16*.5 CUBOID 9 23.0200 -23.0200 101.125 -44.609 442.119 -0.159 16*.5 CUBOID 0 23.0200 -23.0200 101.125 -44.609 442.119 -0.159 16*.5 21 111 111 11I 1 BEGIN COMPLEX
/* embed water rod into quarter boxes (for speedup)
COMPLEX 17 8 0.64770 -0.64770 0.00000 1 1 10.00.00.0 COMPLEX 18 8 -0.64770 -0.64770 0.00000 1 1 10.00.00.0 COMPLEX 19 8 0.64770 0.64770 0.00000 1 1 10.00.00.0 COMPLEX 20 8-0.64770 0.64770 0.00000 1 1 10.00.00.0
/* embed fuel rods into quarter bundle (for speedup)
COMPLEX 9 l -2.59080 -2.59080 0.00000 5 5 1 1.29540 1.29540 0.0 l
I /* overlay water rod in quarter bundle
! COMPLEX 9 17 1.29540 2.59080 0.00000 1 1 10.00.00.0 COMPLEX 9 18 2.59080 2.59080 0.00000 1 1 1 0.0 0.0 0.0 COMPLEX 9 19 1.29540 1.29540 0.00000 1 1 10.00.00.0 April 7,1998
CSA - RA-3 Storage and llandling of IlWR Fuel (Non P) l Page 1.5 of 21 COMPLEX 9 20 2.59080 1.29540 0.00000 1 1 10.00.00.0
/* embed completed { lower-left} quarter bundle into inner COMPLEX 6 9 2.51230 -8.81430 0.00000 1 1 10.00.00.0
/* embed fuel rods into quarter bundle (for speedup)
COMPLEX 10 1 -2.59080 -2.59080 0.00000 5 5 1 1.29540 1.29540 0.0
/* overlay water rod in quarter bundle COMPLEX 10 17 -2.59080 -1.29540 0.00000 1 1 10.00.00.0 COMPLEX 10 18 -1.29540 -1.29540 0.00000 1 1 10.00.00.0 COMPLEX 10 19 -2.59080 -2.59080 0.00000 1 1 10.00.00.0 COMPLEX 10 20-1.29540 -2.59080 0.00000 1 1 10.00.00.0 l' embed completed { upper-right} quarter bundle into inner COMPLEX 6 10 8.98930 -2.33730 0.00000 1 1 10.00.00.0
/* embed fuel rods into quarter bundle (for speedup)
COMPLEX 14 1 -2.59080 2.59080 0.00000 5 5 1 1.29540 1.29540 0.0 ,
/* embed completed (upper-left} quarter bundle into inner COMPLEX 6 14 2.51230 -2.33730 0.00000 1 1 10.00.00.0 l'
/* embed fuel rods into quarter bundle (for speedup)
COMPLEX 15 1 -2.59080 -2.59080 0.00000 5 5 1 1.29540 1.29540 0.0
/* embed completed (lower-right} quarter bundle into inner COMPLEX 6 15 8.98930 -8.81430 0.00000 1 1 10.00.00.0
/* embed cs supports around bundle j COMPLEX 6 11 3.20940 3.82170 0.00000 1 1 10.00.00.0 i COMPLEX 611 20940 -14.4345 0.00000 1 1 10.00.00.0
/* embed es horize baskets around bundle COMPLEX 6 l'. 20000 -14.1174 0.00000 1 1 10.00.00.0 COMPLEX 6 12 0.00000 3.66260 0.00000 1 1 1 0.0 0.0 0.0 I
/* embed bundles into inner (LHS then RilS) l COMPLEX 5 6 -14.6414 5.22740 0.00000 1 1 10.00.00.0 l COMPLEX 5 6 3.14060 5.22740 0.00000 1 1 10.00.00.0 ;
/* embed es vertical baskets around bundle (center,left, right) '
COMPLEX 5 2 0.00000 -0.52300 0.00000 1 1 10.00.00.0 COMPLEX 5 3 -15.7180 -0.52300 0.00000 l 1 10.00.00.0 COMPLEX 5 4 15.7180 -0.52300 0.00000 1 1 10.00.00.0
/* cmbed cs container ends ofinner container COMPLEX 5 16 0.00000 0.00000 -0.15900 1 1 10.00.00.0 COMPLEX 5 16 0.00000 0.00000 441.960 1 1 10.00.00.0
/* embed inner container into global unit: 3-high planar array COMPLEX 21 5 0.00000 0.00000 0.00000 1 3 1 0.0 28.2580 0.0 END GEOM DEFAULTS =YES END GEMER April 7,1998
CSA - RA-3 Storage and llandling of IlWR Fuel (Non P)
Page isof 21 l GEMPLOT: inn-125 10/14/97 up: +Y across: +X units: DM slice: 20 i.
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-0.76 X O.76 April 7,1998
I CSA - RA-3 Storage and llandling of IlWR Fuel (Non P) l Page ,ll of 21 AttacInnent 4. out-025.in sample input,2D & 3D plots w/ fission distribution out-025.in
- 97. D A,,,CYL,UO2,5.00%,WTE R=.05,CS,025,R,C E I10 1000 10 0 0 0 0 0 0 293 0 0
\CSXSEC\UO2iGUO2-50.TD l \CSXSEC\NOU\GNOU-0.WAT 0.025 l
\CSXSEC\NOU\GNOU-0.HDP
{
! \CSXSEC\UO2\GUO2-47. GAD l
l \CSXSEC\NOU;GNOU-0.CS l
\CSXSEC\NOU\GNOU-0.ETH l
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\CSXSEC\NOU\GNOU-0.WAT 0.025 I
\CSXSEC\NOU\GNOU-0.WAT
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\CSXSEC\NOU\GNOU-0.CS 0.85 i
\CSXSEC\NOUTGNOU-0.ETH O.50
\CSXSEC\NOU\GNOU-0.ZR KENO GEOM 0 /* # OF REGIONS OR ZERO O /* # OF BOX TYPES OR ZERO 1 /* # OF BOXES IN X DIRECTION 1 /* # OF BOXES IN Y DIRECTION 1 /* # OF BOXES IN Z DIRECTION I /* BOUNDARY CONDITION OPTION 0 /* STARTING SOURCE OPTION I /* COMPLEX EMBEDDED OPTION 0 /* # OF PRINT PLOTS
-1.0 -1.0 0.0 0.0 -1.0 -1.0 BOX TYPE 1 /* 5% fuel rod, no gad, hd poly wrap CYLINDER 1 441.960 0.000 16*.5 ,
CYL.INDER 13 441.960 0.000 16*.5 CYLINDER 3 441.960 0.000 16*.5 [
CUBOID 2 0.64770-0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 2 /* 85% carbon steel basket, vertical section between bundles CUBOID 11 0.15900 -0.15900 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 8 0.15900 -0.15900 14.4930 -13.4470 441.960 0.000 16*.5 CUBOID 5 0.15900 -0.15900 14.6520 -13.6060 441.960 0.000 16*.5 BOX TYPE 3 /* lett side vertical basket,left side inner container CUBOID 5 -2.06300 -2.38000 5.60300 -4.55700 441.960 0.000 16*.5 CUBOID 8 -2.06300 -2.38000 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 11 -1.90500 -2.38000 9.57200 -8.36700 441.960 0.000 16*.5 .
CUBOID 8 -1.90500 -7.14300 14.4930 -13.4470 441.960 0.000 16*.5 !
CUBOID 5 -1.90500 -7.30200 14.6520 -13.6060 441.960 0.000 16*.5 i BOX TYPE 4 /* right side vertical basket, right side inner container I CUBOID 5 2.38000 2.06300 5.60300 -4.55700 441.960 0.000 16*.5 !
CUBOID 8 2.38000 2.06300 9.57200 -8.36700 441.960 0.000 16*.5 i CUBOlD 11 2.38000 1.90500 9.57200 -8.36700 441.960 0.000 16*.5 CUBOID 8 7.14300 1.90500 14.4930 -13.4470 441.960 0.000 16*.5 l l CUBOID 5 7.30200 1.90500 14.6520 -13.6060 441.960 0.000 16*.5 BOX TYPE 5 /* complete inner container cuboid 1
-- - CUBOID 8 23.0200 -23.0200 14.1290 -14.1290 442.'19 -0.159 16*.5 l
April 7,1998 l
)
i CSA - RA-3 Storage and llandling of BWR Fuel (Non P)
Page 18 0f 21 l BOX TYPE 6 /* single fuel bundle + cs basket cuboid CUBOID 8 14.4824 -2.9816 4.1386 -14.4344 441.960 0.000 16*.5 l CUBOID 8 14.4824 -2.9816 8.7426 -19.1974 441.960 0.000 16*.5 CUBOID 5 14.4824 -2.9816 8.9016 -19.3564 441.960 0.000 16*.5 BOX TYPE 7 /* 5% fuel rod, no gad, NOT USED i CYLINDER I 0.48770 441.960 0.000 16*.5 CYLINDER 13 0.55120 441.960 0.000 16*.5 l CYLINDER 3 0.61390 441.960 0.000 16*.5 CUB 01D 2 0.61390 -0.61390 0.61390 0.61390 441.960 0.000 16*.5 BOX TYPE 8 /* water rod CYLINDER 2 1.16840 441.960 0.000 16*.5 CYLINDER 13 1.24460 441.960 0.000 16*.5 l BOX TYPE 9 /* one quarter fuel bundle cuboid CUBOID 2 3.23850 -3.23850 3.23850 -3.23850 441.960 0.000 16*.5 BOX TYPE 10 /* one quarter fuel bundle cuboid CUBOID 2 3 23850 -3.23850 3.23850 -3.23850 441.960 0.000 16*.5 BOX TYPE 11 /* upper and lower carbon steel support CUBOID 5 5.08200 0.00000 0.31700 0.00000 441.960 0.000 16*.5 BOX TYPE 12 /* 85% carbon steel horizontal basket CUBOID 11 14.8424 -2.9816 0.15900 0.00000 441.960 0.000 16*.5 BOX TYPE 13 /* 4.7% U235 fuel rod,2.85% gd2o3, NOT USED CYLINDER 4 0.44700 441.960 0.000 16*.5 CYLINDER 13 0.50550 441.960 0.000 16*.5 CYLINDER 3 0.57290 441.960 0.000 16*.5 CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 14 /* one quarter fuel bundle cuboid t CUBOID 2 3.23850 -3.23850 3.23850 -3.23850 441.960 0.000 16*.5 BOX TYPE 15 /* one quarter fuel bundle cuboid i CUBOID 2 3.23850 -3.23850 3.23850 -3.23850 441.960 0.000 16*.5 BOX TYPE 16 /* carbon steel end ofinner container CUBOID 5 23.0200 -23.0200 14.1290 -14.1290 0.15900 0.00000 16*.5 6
BOX TYPE 17 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 18 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 19 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 l6*.5 BOX TYPE 20 /* water rod - quarter section CUBOID 2 0.64770 -0.64770 0.64770 -0.64770 441.960 0.000 16*.5 BOX TYPE 21 /* outer container cuboid CUBOID 0 33.0200-33.0200 26.8290 -26.8290 453.549 -11.589 16*.5 CUBOlD 10 34.2900 -34.2900 28.0990 -28.0990 454.819 -12.859 16*.5 BOX TYPE 22 /* upper and lower horizontal honeycomb regions i CUBOID 7 26.0350 -26.0350 3.8100 -3.8100 442.119 -0.1590 16*.5 i l BOX TYPE 23 /* left and right vertical honeycomb regions CUBOID 7 2.5400 -2.5400 14.1288 -14.1288 442.119 -0.1590 16*.5 !
l BOX TYPE 24 /* upper and lower horizontal ethafoam regions l CUBOID 12 26.0350 -26.0350 2.5400 -2.5400 442.119 -0.1590 16*.5 .
BOX TYPE 25 /* left and right vertical ethafoam regions CUBOID 12 0.6350 -0.6350 14.1288 -14.1288 442.119 -0.1590 16*.5 BOX TYPE 26 /* outer container END honeycomb regions ;
CUBOID 7 32.385 -32.385 24.7650 -24.7650 5.7150 -5.7150 16*.5 l BOX TYPE 27 /* global unit: 5-high planar array (inner + outer containers)
- - . CUBOID 0 34.2900 -34.2900 252.891 -28.0990 454.819 -12.859 16*.5 1 f i April 7,1998 l
1
CSA - RA-3 Storage and llandling of BWR Fuel (Non P) l Page l_9 of 21 CUBOID 9 34.2900 -34.2900 283.371 -58.5790 454.819 -12.859 16*.5 27 111 111 111 1 BEGIN COMPLEX
/* embed water rod into quarter boxes (for speedup)
COMPLEX 17 8 0,64770 -0.64770 0.00000 1 1 i0.00.00.0 COMPLEX 18 8 -0.64770 -0.64770 0.00000 1 1 10.00.00.0 COMPLEX 19 8 0.64770 0.64770 0.00000 1 l 10.00.00.0 COMPLEX 20 8 -0.64770 0.64770 0.00000 1 1 10.00.00.0
/* embed fuel rods into quarter bundle (for speedup) j COMPLEX 9 l -2.59080 -2.59080 0.00000 5 5 1 1.29540 1.29540 0.0 '
/* overlay water rod in quarter bundle COMPLEX 9 17 1.29540 2.59080 0.00000 1 1 10.00.00.0 COMPLEX 9 18 2.59080 2.59080 0.00000 1 1 10.00.00.0 COMPLEX 9 19 1.29540 1.29540 0.00000 1 1 10.00.00.0 COMPLEX 9 20 2.59080 1.29540 0.00000 1 1 10.00.00.0
/* embed completed { lower-left) quarter bundle into inner COMPLEX 6 9 2.51230 -8.81430 0.00000 1 1 10.00.00.0
/* embed fuel rods into quarter bundle (for speedup)
COMPLEX 10 1 -2.59080 -2.59080 0.00000 5 5 1 1.29540 1.29540 0.0
/* overlay water rod in quarter bundle COMPLEX 10 17 -2.59080 -1.29540 0.00000 1 1 10.00.00.0 COMPLEX 10 18 -1.29540 -1.29540 0.00000 1 1 10.00.00.0 COMPLEX 10 19 -2.59080 -2.59080 0.00000 1 1 10.00.00.0 COMPLEX 10 20 -1.29540 -2.59080 0.00000 1 1 10.00.00.0 ,
/* embed completed (upper-right) quarter bundle into inner '
l COMPLEX 6 10 8.98930 -2.33730 0.00000 1 1 1 0.0 0.0 0.0
/* embed fuel rods into quarter bundle (for speedup)
COMPLEX 14 1 -2.59080 -2.59080 0.00000 5 5 1 1.29540 1.29540 0.0
/* embed completed (upper-left} quarter bundle into inner I COMPLEX 6 14 2.51230 -2.33730 0.000')0 1 1 10.00.00.0 l' embed fuel rods into quarter bundle (for speedup)
COMPLEX 15 1 -2.59080 -2.59080 0.00000 5 5 1 1.29540 1.29540 0.0
/* embed completed (lower-right) quarter bundle into inner COMPLEX 6 15 8.98930 -8.81430 0.00000 1 1 10.00.00.0
/* embed cs supports around bundle COMPLEX 6 11 3.20940 3.82170 0.00000 1 1 10.00.00.0 COMPLEX 6 11 3.20940 -14.4345 0.00000 1 1 10.00.00.0
/* embed es horizontal baskets around bundle l
COMPLEX 6 12 0.00000 14.I174 0.00000 1 1 10.00.00.0 COMPLEX 6 12 0.00000 3.66260 0.00000 1 1 10.00.00.0
/* embed bund!cs into inner (LilS then RHS)
COMPLEX 5 6 -14.6414 5.22740 0.00000 1 1 10.00.00.0 COMPLEX 5 6 3.14060 5.22740 0.00000 1 1 10.00.000
/* embed cs vertical baskets around bundle (center,left, right)
COMPLEX 5 2 0.00000 -0.52300 0.00000 1 1 10.00.00.0 COMPLEX 5 3 -15.7180 -0.52300 0.00000 1 1 10.00.00.0 COMPLEX 5 4 15.7180 -0.52300 0.00000 1 1 10.00.00.0 l /* embed cs container ends ofinner container j COMPLEX 5 16 0.00000 0.00000 -0.15900 1 1 10.00.00.0 COMPLEX 5 16 0.00000 0.00000 441.960 1 1 10.00.00.0
/* embed inner container into outer wooden container COMPLEX 21 5 0.00000 0.00000 0.00000 1 1 I0.00.00.0 l' embed honeycomb regions around inner container
.- -. COMPLEX 21 22 0.00000 -23.0188 0.00000 1 1 10.00.00.0 April 7,1998
CSA - RA-3 Storage and llandling of BWR Fuel (Non P)
Page H of 21 l COMPLEX 21 22 0.00000 23.0188 0.00000 1 1 10.00.00.0 COMPLEX 21 23 -30.4800 0.0000 0.00000 1 1 10.00.00.0 COMPLEX 21 23 30.4800 0.0000 0.00000 1 1 10.00.00.0
/* embed ethafoam regions around honeycomb wrap COMPLEX 21 24 0.00000 -16.6688 0.00000 1 1 10.00.00.0 COMPLEX 21 24 0.00000 16.6688 0.00000 1 1 10.00.00.0 COMPLEX 21 25-27.3050 0.0000 0.00000 1 1 10.00.00.0 }
COMPLEX 21 25 27.3050 0.0000 0.00000 1 1 10.00.00.0 {
/* embed END honeycomb regions to complete construct ofinnner+ outer !
COMPLEr 21 26 0.00000 -2.06400 447.834 1 1 10.00.00.0 COMPLEX 21 26 0.00000 -2.06400 -5.8740 1 1 10.00.00.0 l
/* embed inner + outer container into global unit: 5-high planar array l
COMPLEX 27 2l 0.00000 0.00000 0.00000 1 5 1 0.0 56.1980.0 l END GEOM
- DEFAULTS =YES END GEMER l
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l April 7,1998
CSA - RA-3 Storage and llandling of HWR Fuel (Non P) l Page 2I of 21 GEMPLOT: out-025 10/14/97 up: +Y across: +X units: DN slice: 20 I I
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-1.13 X 1.13 April 7,1998
Attachment 2 to NRC-98-0063 Page1 Attachment 2 i
Proprietary Version of l'
GE New Fuel Receipt Criticality Safety Analysis l
l 1
i
.