ML20212H528
| ML20212H528 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/18/1999 |
| From: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137U375 | List: |
| References | |
| 1940-99-20334, NUDOCS 9906240014 | |
| Download: ML20212H528 (15) | |
Text
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{ GPU Nuclear,Inc.
l (. U.S. Route #9 South l NUCLEAR Post Office Box 388 Forked River, NJ 08731-0388 l
l Tel 609-9714000 l
June 18,1999 l
1940-99-20334 l
U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Subject:
- Oyster Creek Nuclear Generating Station (OCNGS)
Docket No. 50-219 Facility Operating License No. DPR-16 Technical Specification Change Request No. 261 - Spent Fuel Pool Expansion in accordance with 10 CFR 50.4(b)(1), enclosed is Technical Specification Change Request (TSCR) No. 261.
The proposed amendment modifies the OCNGS Technical Specifications to reflect installation of additional spent fuel pool storage racks. The additional new racks will accommodate an inerease in spent fuel assemblies beyond the existing storage capacity of the spent fuel pool as described in the i enclosed Licensing Report.
I Using the standards in 10 CFR 50.92, GPU Nuclear has concluded that these proposed changes do not l constitute a significant hazards consideration, as described in the enclosed analysis perfonned in j accordance with 10 CFR 50.91 (a)(1). Pursuant to 10 CFR 50.91 (b)(1), also enclosed is the Certifictte /
of Service for this request certifying service to the designated official of the State of New Jersey Bureau of Nuclear Engineering, as well as the senior official of the township in which the facility is located.
The enclosed Holtec International Report HI-981983, Rev. 4, " Licensing Report for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool", contains proprietary information as defined in 10 CFR 2.790 (a)(4). Accordingly,it is requested that this document be withheld from public disclosure. An affidavit certifying the basis for this application for withholding as required by 10 CFR 2.790 (b)(1) is l also enclosed with this letter, along with a non-proprietary version of the report.
duo ( LT9 b NC L O
9906240014 990618 PO R I PDR ADOCK 05000219 L P PDR 2
1940-99-20334 Page 2 This license amendment application has undergone a safety review in accordance with Section 6.5 of the Oyster Creek Technical Specifications.
GPU Nuclear requests NRC review and approval of the proposed change by December 31,1999 to support storage of spent fuel planned for the year 2000.
If any additional information is needed, please contact Mr. David J. Distel, Corporate Regulatory AfTairs, at (973) 316-7955.
Sincerely, {
7LLJ/3 Michael B. Roche Vice President and Director Oyster Creek
/DJD
Enclosures:
- 1. Technical Specification Change Request No; 261 Safety Evaluation, No Significant flazards Consideration, Technical Specification Revised Pages
- 2. Certificate of Service for OCNGS Technical Specification Change Request No. 261 1
- 3. Licensing Report for Storage Capacity Expansion of OCNGS Spent Fuel Pool,11oltec Report Hi-981983, Revision 4, Proprietary -
- 4. Licensing Report for Storage Capacity Expansion of OCNGS Spent Fuel Pool, Holtec Report HI-981983, Revision 4, Non-Proprietary
- 5. Holtec International, Affidavit Certifying R: quest for Withholding from Public Disclosure l
cc: Administrator, Region i OCNGS Senior Project Manager OCNGS Senior Resident inspector
c GPU NUCLEAR, INCORPORATED OYSTER CREEK NUCLEAR GENERATION STATION Operating License No. DPR-16 Docket No. 50-219 Technical Specification Change Request No. 26i This Technical Specification Change Request is submitted in support of Licensee's request to clunge Appendix A Technical Specifications to Operating License No. DPR-16 for Oyster Creek Nuclear Generating Station. As a part of this request, the proposed replacement pages for Appendix A are also submitted.
BY:
fl b t-e~ s Michael B. lioche Vice President and Director Oyster Creek Swom and Subscribed to before me this I day of ML , 1999 JhWNMS Notary blic ofState ofNew Jersey I'PTTY GOODifEART A N w m lY Kcciti w biy Cmcwwa Expim m, Jersey y g
l' ENCLOSURE 5 1
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Holtec International Affidavit Certifying Request for Withholding from Public Disclosure l
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1 AFFIDAVIT PURSUANT TO 10CFR2.790 I, Charles W. Bullard II, being duly sworn, depose and state as follows:
(1) I am the Project Manager for Holtec International and have been 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.
l (2) The information sought to be withheld is contained in the document entitled !
l " LICENSING REPORT FOR STORAGE CAPACITY EXPANSION OF OYSTER CREEK SPENT FUEL POOL," Holtec Report HI-981983, Revision I
- 4. The proprietary material in this document is delineated by proprietary l designation on specific pages or by shaded text identified as being proprietary.
(3) In making this application for withholding of proprietary information of which it l is the owner, Holtec International relies upon the exemption from disclosure set i forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.790(a)(4), and 2.790(b)(1) for " trade secrets and commercial or l financial information obtained from a person and privileged or confidential"
! (Exemption 4). The material for which exemption from disclosure is here I sought is all " confidential commercial information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemptioli 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Groun v. FDA, 704F2d1280 (DC Cir.1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
I a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's ,
competitors without license from Holtec International constitutes a competitive economic advantage over other companies; 1
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AFFIDAVIT PURSUANT TO 10CFR2.790 i
- b. Information which, if used by a competitor, would reduce his expenditure l l 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 Holtec International, ]
its customers, or its suppliers; i
- d. Infoimation which reveals aspects of past, present, or future Holtec I International customer-funded development plans and programs of ,
potential commercial value to Holtec International; l
- e. Information which discloses patentable subject matter for which it may be I desirable to obtain patent protection.
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a,4.b,4.d, and 4.e, above. )
(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge i and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(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 information in relation to industry knowledge, j l Access to such documents within Holtec International is limited on a "need to j know" basis. j i
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l AFFIDAVIT PURSUANT TO 10CFR2.790 i
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(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.
Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, j and 'others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is ;
classified as proprietary because it contains detailed historical data and analytical l results not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed using codes developed by Holtec International. Release of this information would improve a competitor's position without the competitor having to expend similar resources for the ,
development of the database. A substantial effort has been expended by Holtec International to develop this information.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel . storage 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.
l The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.
The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.
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AFFILAVIT PURSUANT TO 10CFR2.790 Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.
STATE OF NEW JERSEY ) l
) ss: i COUNTY OF BURLINGTON )
Mr. Charles W. Bullard II, being duly sworn, deposes and says:
That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.
Executed at Marlton, New Jersey, this 15th day of June 1999.
h 5 fS M E Mr. Charles W. Bullard II Holtec International Subscribed and sworn before me this / day of b~ ,1999.
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MARIA C. PEPE NOTARY PUBLIO OF NEW JERSEY My Commission Expires Ap 9 25,2000 4
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ENCLOSURE 1 Oyster Creek Nuclear Generating Station l i
Technical Specification Change Request No. 261
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- 1. CHANGE REOUESTED '
GPU Nuclear requests that the following changed replacement pages be inserted into existing Technical Specifications:
Revised Technical Specification Pages: 5.3-1 and 5.3-2 These pages are attached to this change request.
II. DISCUSSION OF PROPOSED CHANGE The Oyster Creek Nuclear Generating Station (OCNGS) spent fuel pool is currently licensed for the storage of 2645 spent fuel assemblies. Technical Specification Change Request (TSCR) No.
261 changes the OCNGS Technical Specification Section 5.3.1.E to identify the resised maximum number of fuel assembly storage locations provided by the installation of four (4) additional new high density spent fuel pool storage racks. These new racks are to be installed in existing available i spent fuel pool floor space and will provide an additional 390 spent fuel assembly storage j
. locations. The existing OCNGS spent fuel pool storage racks will remain in place continuing to )
l provide installed storage locations. No fuel storage racks are being removed from the OCNGS spent fuel pool. The new racks will provide an ultimate storage capacity of 3,035 fuel assemblics in the OCNGS spent fuel pool. This storage capacity temporarily restores the full-core discharge capability at the plant.
Technical Specification Section 5.3 Basis is also amended to clarify that the existing reference to containment air temperature applies to the reactor building since this is the location of the fuel pool. This change is considered editorial in nature.
! OCNGS Technical Specification Section 5.3.1 Basis is also revised to include a reference to the l Holtec Licensing Report HI-981983, " Licensing Report for Storage Capacity Expansion of Oyster l Creek Spent Fuel Pool", Revision 4, dated June 15,1999 (Enclosure 3 of this submittal), which l provides the design basis and safety analysis for installation and use of the new high density spent
, fuel pool storage racks. The report also provides information requested by the NRC position l
contained in "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", published in 1978 with a 1979 Addendum thereto.
! 111. SAFETY EVALUATION JUSTIFYING CHANGE L The four (4) additional new racks are free-standing and self-supporting. The number of storage locations is described below:
Rack Module No. of Cells L 96 M 99 N 88 P 102 Total 390
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Enclosure i 1940-99-20334 Page 2 l
l Each cell is designed for storage of fresh or irradiated fuel assemblics with Uranium-235 initial enrichments up to 4.6 weight percent while maintaining the required subcriticality (Kar 5 0.95).
The high density spent fuel storage rack cells are fabricated from 0.075" thick Type 304L stainless steel sheet material. Boral neutron absorber material strips are emplaced between the cell walls and a stainless steel coverplate. Each storage cell side is equipped with one (1) integral Boral sheet. The cells are welded together in a specified manner to become a free-standing structure j which is seismically qualified without depending on neighboring modules or fuel pool walls for support. Boral panels are installed in the rack wall along both sides of the water gap between adjacent racks. With this configuration, the maximum reactivity of the storage rack is not j I
dependent upon the water-gap spacing between modules. The nominal center-to-center spacings of the cells are 6.106" Each cell is designed for storage of BWR fuel assemblies with K,in the standard core geometry of up to 1.32 and Uranium-235 enrichments up to 4.6% wt. (with credit for burnable poison) while maintaining the required subcriticality (Kars 0.95).
A Boral poison material surveillance program will be implemented which allows access to ,
representative poison samples without disrupting the integrity of the storage system. This program l provides the capability to evaluate the material in a normal use mode, and to forecast changes that l might occur within the storage system prior to occurrence of such changes. This is accomplished utilizing test " coupons" removed at periodic intervals and tested, as well as direct testing of l
installed poison panels in the fuel racks if needed. The periodic testing methods, intervals, and acceptance criteria are described in Section 10 of the OCNGS Licensing Report (Holtec Report HI-981983).
The safety analysis of the proposed rack modules, as described in Holtec Report HI-981983, demonstrates their criticality, thermal-hydraulic, and structural compliance with established requirements. Criticality analysis confirms that the rack design maintains the neutron multiplication factor (Kar) equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in mechanical tolerances, statistically combined, such that the true Karwill be equal to or less than 0.95 with a 95% probability at a 95% confidence level. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under credible abnormal conditions, the reactivity will be maintained less than 0.95. The criteria for spent fuel to be acceptable for storage in the new Oyster Creek high density spent fuel racks are the following:
- Any fuel assembly which has a planar-average enrichment of 3.2% or less, or
. Fuel assemblies with a planar standard cold core geometry (SCCG) K.of 1.32 or less, with an average enrichment of 4.6% or less, or
. Alternatively, any fuel whose enrichment SCCG K. combination falls within the acceptable domain of HI-981983, Figure 4.2.2.
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Structural compliance is demonstrated by analysis showing the free-standing modules will not l
affect the stored spent fuel assemblies under all postulated seismic events, and the primary stresses i in the module structure will remain below the ASME Code allowables. The structural qualification j includes analytical demonstration that the suberiticality of the stored fuel wi4 be maintained under l l accident scenarios such as fuel assembly drop or accidental misplacement of fuel outside a rack. '
Thermal-hydraulic analysis confirms that fuel cladding will not fail due to excessive thermal stress, !
and the steady state pool bulk temperature will remain within the limits prescribed for the spent fuel pool. Thermal-hydraulic analysis confirms that spent fuel pool water bulk temperatures are kept below 125 F, as required by existing OCNGS Technical Specification Section 5.3.1.D, during l nonnal refueling offload (Case ii - normal refueling batch transferred to pool six days after reactor shutdown using augmented fuel pool heat exchanger with one pump in operation) and f'ill-core offload (Case iii - full core transferred to pool thirty-one days after reactor shutdown using augmented fuel pool heat exchanger with one pump in operation) discharge scenarios. The peak bulk pool temperature for Case i (abnormal partial core ofiload six days after reactor shutdown l using the smaller capacity shell-and-tube heat exchanger with one pump) is 168 F. In this l scenario, the spent fuel pool water temperature alanns (120 F) would alert operators that the 125 F bulk pool temperature is being approached and thus action can be taken to either utilize the augmented pool cooling system which will maintain pool temperature below 125 F as shown in Cases (ii) and (iii) or to stop core offloading until pool temperature is restored. Approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is available before reaching the Technical Specification limit of 125 F and approximately 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> is available before reaching the analyzed peak bulk pool temperature. The maximum bulk temperatures demonstrate compliance with the latest USNRC acceptance criteria. Conservative time-to-boil analyses, assuming forced pool cooling becomes unavailable, show results (minimum 7.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />) comparable to other BWR pools with high density storage. Thus, sufficient time is available to respond to existing spent fuel pool water temperature alamis (120 F) to restore pool cooling with either the shell-and-tube heat exchanger or the augmented heat exchanger. Peak clad temperature analysis confirms that nucleate boiling does not occur anywhere in the spent fuel pool when conservatively considering the maximum decay heat input Case (iii) scenario.
These thermal-hydraulic analysis cases represent bounding scenarios considering conservative
! decay heat loads and fuel pool heat exchanger heat transfer capability. Actual conditions during l fuel discharge evolutions will determine the allowable fuel assembly discharge rate and time after l shutdown for fuel assembly transfer to the spent fuel pool, and will remain bounded by the bulk l pool temperatures determined in Cases (i), (ii), and (iii) of Holtec Report Hi-981983.
The structural adequacy of the new high density spent fuel racks is demonstrated under all loadings postulated for normal, seismic, and accident conditions at OCNGS. The analyses are performed in accordance with NRC Standard Review Plan and the OT Position Paper. The analysis shows that the fuel racks do not impact the pool walls under any seismic conditions. The whole pool multi-rack analysis shows that impacts between adjacent racks at the baseplate level may occur during postulated OBE or SSE loads. Impact loads arc much less than the yield stress of the material.
Therefore, the analyzed rack-to-rack impacts do not affect the configuration of the stored fuel.
Analysis also confirms that there is no possibility of a new rack overturning due to a scismic crent j occurring during the installation process.
r Enclosure 1 1940-99-20334 -
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1 The design of the new racks has been evaluated under all credible fuel drop events in the spent fuel pool as specified in the NRC OT Position Paper. These evaluations confirm the integrity of the racks and the ability to maintain safe storage configuration. In addition to postulated fuel 1 assembly drops, a rack drop accident during installation is postulated to confirm pool structure l integrity. This evaluation confinns that the pool structure would not sustain significant damage from the postulated rack drop. The analysis shows that the rack pedestals would pierce the pool 4 liner with localized concrete cracking. Any leakage resulting from such localized damage would be detectable and capability is provided to makeup the loss ofinventory to the pool. Structural analysis of the spent fuel pool demonstrates that for the bounding load combinations the structural integrity is maintained when the pool is assumed to be fully loaded with 3,035 spent wi assemblics, which is the maximum pool capacity after rack installation.
The structural adequacy of the pool structure is demonstrated, as required by the USNRC OT Position Paper, based on the proposed total installed capacity of 3,035 fuel assembly storage l locations. The structural analyses performed demonstrate that the safety factors in the pool slab i and the undergirding concrete beams exceed 1.0 for all postulated loadings, including the weight of l a shipping cask (100 tons). The integrity of the pool liner during postulated seismic events is maintained.
Although the design basis fuel handling accident for OCNGS is a fuel assembly drop onto the top of the core, a radiological analysis of a potential fuel assembly drop in the spent fuel pool was performed. This analysis utilized Regulatory Guide 1.25 assumptions and considered bounding initial enrichments and assembly burnup. The resulting radiological consequences are well within the guidelines of 10 CFR 100 and Standard Review Plan 15.7.4 as described in Section 9 of the OCNGS Licensing Report.
Installation of the racks will be performed in compliance with NUREG-0612 and site-specific procedures as described in Section 11 of the OCNGS Licensing Report. Crane and fuel bridge operators will be trained. The lifting device designed for handling and installations of the new racks at OCNGS is in compliance with the provisions of ANSI N 14.6-1978 and NUREG-0612.
ALARA program procedures will minimize the total man-rem received during the rack installation project. Installation of these new racks does not require any of the existing racks to be removed or displaced.
IV. NO SIGNIFICANT HAZARDS CON _ SIDERATIONS GPU Nuclear has determined that this Tech .k.,1 Specification Change Request involves no i significant hazards condition as defined by ;JFC in 10 CFR 50.92.
- 1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or the consequences of an accident previously evaluated. The following previously analyzed accident scenarios have been considered as part of the analyses required to support the installation of the high density spent fuel storage racks:
Enclosure 1 1940-99-20334 Page 5 :
I a) Spent Fuel Assembly Drop - The criticality acceptance criteria, IGrs 0.95,is maintained for postulated abnormal occurrences such as a fuel assembly mistoading or assembly drop. The radiological consequences of a fuel handling accident in the spent fuel pool remain well within the guidelines of 10 CFR 100 and Standard Review r tan 15.7.4.
b) Loss of Spent Fuel Pool Cooling System Flow - The spent fuel pool cooling system will continue to provide acceptable cooling of the stored assemblies. i l
Approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is available before reaching the Technical Specification
- limit of 125 F and approximately 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> is available before reaching the
- analyzed peak bulk pool temperature. Therefore, sufTicient time is available to respond to the spent fuel pool water temperature control room alarm (120 F) and to provide an altemate means of cooling in the event of a failure in the cooling system. Therefore, the proposed change has no afTect on this accident scenario.
l c) Seismic Event - The new racks are designed and fabricated to remain functional during and aner a Safe Shutdown Earthquake under all loading conditions.
Analysis has demonstrated that no rack-to-wall impacts occur. Analyzed potential rack-to-rack impacts demonstrates the stored fuel configuration remains unaffected. Spent fuel pool structural analysis demonstrates that for the bounding factored load combinations, including the weight of a shipping cask (100 tons),
structural integrity is maintained when the pool is assumed to be fully loaded with 3,035 spent fuel assemblies. Therefore, the proposed change has no affect on this accident scenario.
d) Spent Fuel Cask Drop - Structural analysis of the spent fuel pool demonstrates l that the pool structure remains adequate for the loadings associated with normal l operation and the condition resulting from the postulated cask drop accident.
Accordingly, the proposed modification does not increase the probability of occurrence or the consequences of an accident previously evaluated.
i 2. Operation of the facility in accordance with the proposed amendment would not create the l possibility of a new or different kind of accident from any accident previously evaluated.
i Administrative controls during rack installation will preclude the movement of a new rack directly over any fuel. The new racks will be lined using the 100-ton overhead crane
- which has a suflicient safety factor such that potential single failure mechanisms need not l be considered. The lifting device designed for handling and installation of the new racks is in compliance with NUREG-0612. A postulated rack drop analysis demonstrates that the pool structure would not sustain significant damage from the postulated rack drop. The analysis shows that the rack pedestal would pierce the pool liner with localized cocerete cracking. Any leakage resulting from such localized damage would be detectable and capability is provided to make up the loss ofinventory to the pool. No unproven l
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technology is involved either in the installation process or in the analytical techniques utilized to evaluate the planned fuel storage expansion. The basic technology for fuel pool expansion has been developed and demonstrated in over 80 applications for fuel pool capacity increases previously approved by NRC. The proposed modiDeation has been evaluated in accordance with the guidance of NRC Position Paper, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", April 14, 1978, and Addendum dated January 18,1979. Therefore, this change has no afTect on the possibility of creating a new or different kind of accident from any accident previously evaluated.
- 3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. Analysis has demonstrated that the established criticality acceptance criteria, Kerrs; 0.95 including uncertainties, is l maintained with the racks fully loaded with fuel of the highest anticipated reactivity. ;
Thermal-hydraulic analyses demonstrate that the pool bulk temperatures are maintained below 125 F for the normal refueling of00ad and the full-core offload discharge scenarios using the augmented fuel pool heat exchanger, and that the maximum local water temperature along the hottest fuel assembly is below the nucleate boiling condition value. The maximum bulk pool temperatures for each of the analyzed scenarios confirms that adequate time is available to provide an alternate means of cooling in the event of a failure in the cooling system. The rack materials used are compatible with the spent fuel pool and the spent fuel assemblies. The structural analyses have demonstrated that the proposed change maintains spent fuel pool structural integrity and margins of safety. The new racks are designed and fabricated to remain functional during and after a Safe Shutdown Earthquake. Therefore, this change has no affect on the margins of safety related to nuclear criticality, thermal and structural integrity, and material compatibility.
The proposed amendment is considered to be in the same category as example (x) of amendments that are considered not likely to involve significant hazards consideration as provided in the final NRC adoption of 10 CFR 50.92 published on Page 7751 of the Federal Register Volume 51, No. 44, March 8,1986. This example indicates that an amendment involving the expansion of the storage capacity of the spent fuel pool is not likely to involve a significant hazards condition as follows:
Criterion (lh The storage expansion method consists of either replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies or placing additional racks of the original design on the pool floor if space permits.
Proposed Amendment:
The OCNGS storage expansion method consists of the addition of four (4) new high density racks utilizing a neutron absorber, similar to the design of the existing installed high density racks, to available pool Door space.
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Enclosure i 1940-99-20334 Page 7 Criterion (2h The storage expansion method does not involve rod consolidation or double tiering.
Proposed Amendment:
The amendment application does not involve consolidation of spent fuel. The OCNGS racks are not double tiered and all racks will sit on the spent fuel pool floor.
Criterion (3h The Kerrof the pool is maintained less than or equal to 0.95.
Pronosed Amendneg e The design of the new racks integrates a natron absorber, Boral, within the racks to allow closer storage of spent fuel assemblies while ensuring that Kerrremains less than 0.95 under all operating cenditions with pure water in the pool.
Criterion (4h No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary tojustify the expansion.
Pronosed Amendment:
The rack vendor has successfully participated in the licensing of numerous other racks of a similar design. The construction process and the analytical techniques utilized for the OCNGS pool expansion are substantially the same as in the other completed rack installation projects in the industry. Thus, no new or unproven technology is used in the construction or analysis of the j OCNGS high density spent fuel racks. )
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V. ENVIRONMENTAL CONSIDERATIONS t 1
GPU Nuclear has reviewed the proposed license amendment against the criteria of 10 CFR 51.22 for environmental considerations. The proposed changes do not significantly increase the types !
and amounts of effluents that may be released offsite nor significantly increase individual or !
cumulative occupational radiation exposures. Based on the foregoing, GPU Nuclear concludes {
that the proposed changes meet the criteria delineated in 10 CFR 51.22 (c)(9) for a categorical i exclusion from the requirements for an environmental impact statement.
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VI. IMPLEMENTATION .
i GPU Nuclear requests that the amendment authorizing this change be issued by December 31,1999 1 to support storage of spent fuel planned for the year 2000. The amendment shall become effective immediately upon issuance.
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