ML20216J815

Application for Amend to License NPF-42,increasing Spent Fuel Storage Capacity & Max Nominal Fuel Enrichment to 5.0 W/O (Nominal Weight Percent) U-235.Proprietary & non-proprietary Holtec Repts,Encl.Proprietary Encl Withheld

ML20216J815
Person / Time
Site:	Wolf Creek
Issue date:	03/20/1998
From:	Muench R WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20216J820	List: ML20216J815 ML20216J829
References
ET-98-0009, ET-98-9, NUDOCS 9803240098
Download: ML20216J815 (26)

Text

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- W$LF CREEK NUCLEAR OPERATING CORPORATION Richard A. Muench Vice Proeident Engmeering March 20, 1998 ET 98-0009 m

U. S. Nuclear Regulatory Commission ATTN: Document Contit>l Desk Mail Station P1-137 1 Washington, D. C. 20555

Reference:

Letter ET 97-0050, dated May 15, 1997, from R. A. Muench,

,: WCNOC t o NRC

Subject:

Docket No. 50-482: Propced Revision to' Technical Specifications to Incr ease the Spent Fuel Pcol Storage Capacity Gentlemens This le* 19t transmjtri an appucc tion for amendoem to Far.iiiey' f 'p e r n t i n.;.

License !;n. flPF-42 icr the Wolf Creek Generating Station (WCCS). Thl+ 4 canze amendment ruquest propor.o. t 1 tc'.:;c the t:chnicM uve t f b ut Um - ' m ,. w. . . t-the speA fuel storago capacity and 2ncicane the ,trij % n c;d 'M entle.b ent tn 5.0 w/o (nominal weignt percent) C 4'35. M i nrii c a t w j ' %: a

3. 3 s of Encksure 1, BORA1, the heut roa absorber for the new nu b , hn (*

licensed by the NRO for Use in numerous nuclear power plant spent icel m m < r .

a1plicatiene.

It' is requested that tha NPC review this 13 cs-nse amendment 7. c q at e t in g conjunction with the review of the Union Clentric license amenament re p n.

Tne issuance date f or t his amendment should be no later t han Notcaber 1990 to support the Union Electric schedule. This approval date will . support the WCCL terack completion date of May 2000 and support the availability vf the Speat Fuel Pool prior to June 2000. 'Ihese dates give suf ficient lead time fcr the removal of the existing racks and installation of the new racks prior tn i he-arrival of Cycle 12 fuel at the WCCS site.

/

The scope of this amendment rcquest focuses on the final configuration of the reracked spent fuel poo'. It is recognized that the transition to the tinal configuration will involve some intermediate stages as existing r.v:kr era [

removed and new tacks installed. WCNOC will prov2de administratave controls that will be implemented during the rerack modification to ensure that fuel storage will be handled in accordance with both current and revised Technical Specifications. It is requested that this license amendment associated with [

the request approve the use of existing and revised specifications during the installation / implementation of the new racks. The new racks and the pool interim configurations will be controlled administratively until the modification in the . spent fuel pool is completed. The administrative controls will assure, that under these ' interim conditions, the intent of the revised specifications will be' met for the new racks. When all of the new racks are installed and the modification is completed, the spent fuel pool will be in compliance with the revised Technical Specifications. At that time the revised Technical Specifications will become fully implemented.

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P.Obox 411/ B rlington, KS 66839 / Phone: (316) 364-8831

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PDR ADOCK 05000482 P PDR

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b ET 98-0009 7c Page 2 of-2

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'Al Description'of Proposed _ Changes / Safety Assessment is provided in Attachment fI. A D No Significant Hazards . Consideration Determination is provided in

. Attachment II. The proposed _ changes have been etaluated using criteria in 10 CFR' 50.92(c), and it has been determined that the changes involve no significant hazards consideration. Spent Tuel Pool configurations prior to, p ' during, and subsequentft o existing rack renoval and new rack installation are bounded.by the analyses. Attachment III is the related Environmental Impact Determination. . Marked up pages , are provided in Attachment IV (for current Technical " Specifications and Bases) and in Attachment V (for Improved

' Technical Specifications and Bases submitted by Reference 2) . Enclosures I a'nd II provide proprietary and non-proprietary versions- of the Licensing Report for Reracking of the WCGS Spent Fuel Pool. '

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is~being provided to the designated Kansas State Official.

IfLyou have any questions concerning this matter, please contact me at (316) 364-8831, extension 4034, or Mr. Michael J. Angus, at extension 4077.

Very truly yours,

/ llAiAO Richard A. Moench RAM /jad; Attachments: 1 - Cescription of Proposed Changes / Safety Assessment II- - No Significant Hazards Consideration Determination III -

Environmental Impact, Determination l IV- . Proposed Carrent Technical Specification Change V --Proposed Improved Technical Specification Change

,Enclosurest' I Licencing Report for Paracking of WCG3 Spent Fuel Fool II Non-Proprietary Version of the Licensing Report for Reracking of WCGS Spent Fuel Pool cc: V. L. Cooper'(KDHE), w/a l W. D.' Johnson (NRC), w/a i E. W. Merschoff (NRC), w/a -

J. F. Ringwald (NRC), w/a K. M. Thomas (NRC), w/a i

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- STATE Or KANSAS )

) SS ,

conuTr or corrEr )- '

Richard A. .Mue7ch, of : lawful.' age, being first. duly' sworn upon oaths says.that he-is:Vice President Engineering of Wolf Creek Nuclear Operating Corporation;

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i that he has read the foregoing document and knows the contento thereofi . that he has , executed that same for and cn' . behalf - of said Corporation - with full --

power and authority'to do.so;-andathat . the' facts therein stated are true and correct to the best of his knowledge, information and belief. . - /'

ny A RichardA./Muench Vice President Engineering SUBSCRIBED and. sworn to before me this day:of df

[,1998.

JULIE A. DALE aC N tary P blic O. La6 Notary Public. plate p Kansas M Appt. Expiree IDI.3019? qq 0c Expiration Date u "O ( 0 s

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1 1

AFFIDAVIT PURSUANT TO 10CFR2.790 I, Alan I. Soler, being duly sworn, depose and state as follows:

(1)- I am Executive Vice President, 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.

(2) The information sought to be withheld is contained in the document entitled Licensing Report for Reracking of the Callaway and Wolf Creek Nuclear Plants, Holtec Report HI-971769.' The proprietary material in this document is delineated by proprietary designation on specific pages or by shaded text identified as being proprietary.

l (3) In making this application for withholding of proprietary information of which it is i

the owner, Holtec International 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 10CFR Part 9.17(a)(4),

i ~ 2.790(a)(4), and 2.790(b)(1) 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 j commercial information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those term:; for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. )

Nuclear Regnbrory Commission, 975F2d871 (DC Cir.1992), and Public Citizen l Health Research Group v. FDA, 704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including I 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

1

AFFIDAVIT PURSUANT TO 10CFR2.790

b. Information 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 Holtec International, its customers, or its suppliers; Information which ' reveals aspects of past, present, or future Holtec

~

d. -

International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be 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 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. Access to such documents within Holtec International is limited on a "need to know" basis.

2

AFFIDAVIT PURSUANT TO 10CFR2.790

-(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, 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 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 f 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.

o 3

AFFIDAVIT 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.

I L STATE OF NEW JERSEY )~

L ) ss:

l COUNTY OF BURLINGTON )

Dr. Alan I. Soler, being duly sworn, deposes and says:

(

l That he has read the foregoing affidavit and the matters stated therein are true and correct L to the best of his knowledge, information, and belief.

I

! Executed at Marlton, New Jersey, this 12th day of March 1998.

Yn, '

Dr. Alan I. Soler Holtec International Subscribed and sworn before me this 4 day of 2 M<'d ,1998.

mn a4y -

f ".9'A C. PEPE NOT/JiY r CUO OF NEWJERSEY W W as,m 4

Attachment I to ET 98-0009 Page 1 of 10 ATTACHMENT I DESCRIPTION OF PROPOSED CHANGES / SAFETY ASSESSMENT

Attachment I to ET 98-C009 Page 2 of 10 Description of Proposed Changes / Safety Assessment '

' Proposed Changes.

This license amendment request' proposes to revise Wolf Creek Generating Station (WCGS)' Technical Specifications by incorporating the proposed changes identified'in Attachment IV to increase the spent fuel - storage capacity and increase the maximum nominal fuel enrichment to 5.0 w/o (nominal weight percent) U-235.

Background

WCGS received its low power operating license on March 11, 1985. At that time, the Spent Fuel Pool was authorized to store no more than 1344 (actual 1340) fuel assemblies. These are stored in 12 spent fuel racks. Current projections, based on expected future spent fuel- discharges, indicate that loss of full-core discharge capability will occur at the end of Cycle 14 in 2005. Operation of WCGS beyond loss of full-core discharge capability is possible for Cycles 15 and 16 to provide an additional three to four years of operation until 2008.

WCNOC has evaluated spent fuel storage alternatives that have been licensed by the NRC and which. are currently feasible for use at the WCGS site. The evaluation concludes that' reracking is currently the most cost-effective alternative. Reracking would provide'an increase in storage capacity which would maintain the plant's capability to accommodate a full-core discharge, through the end of the current plant license in 2025.

DESCRIPTION OF PROPOSED CHANGES WCNOC.is requesting to rerack the Spent Fuel Pool and add racks to the cask loading pit to accommodate a full-core discharge, through the end of licensed plant life (2025). This proposed ' modification will be accomplished by removing the existing racks in the Spent Fuel Pool and replacing them with higher density racks. In order to maximize the storage capacity of the Spent Fuel Pool, truncation of the sparger lines is.also planned. An offset fuel handling tool will be -installed to allow access to some of the storage

-locations that are adjacent to pool walls.

Additional - racks will be placed within the cask loading pit during a later campaign. Since the cask loading pit is deeper than the Spent Fuel Pool, platforms will be installed beneath the cask loading pit racks to allow installation at the same elevation as the Spent Fuel Pool racks.

The. expansion will increase the total storage space from 1340 to 2642 fuel assemblies. This represents a storage capacity increase of 1302 assemblies.

The fif teen racks to be located in the Spent Fuel Pool during the initial campaign will have the capability to store 2363 fuel assemblies. This capacity increase would allow operation until the end of the current license, since WCNOC is projecting a total spent fuel inventory of 2353 assemblies ~ af ter Refuel 27 in the year 2025. The three racks to be located in the cask loading pit during a . later campaign will have the capability to store an additional 279 fuel assemblies. This would increase the total storage capacity to 2642 fuel assemblies to provide additional flexibility.

The new racks will have'a closer assembly to assembly spacing to allow for more fuel storage capability. The ' racks will contain Boral as the active neutron absorber. The Boral absorbers have been sized to sufficiently shadow the active fuel height of all fuel assembly designs stored in the pool.

Attachment I to ET 98-0009 Page 3 of 10 The proposed racks will allow fuel storage for enrichments up to 5.0 w/o U-235 in a Mixed Zone Three Region storage configuration. The storage configuration patterns will be setup using administrative controls to establish storage areas specifically designated for low burnup fuel, including fresh (unburned) fuel. Selected configurations will ensure that a full-core discharge can be accommodated with some allowance for other fuel assemblies that could also require Region 1 storage. Cells reserved for storage of fresh fuel, and spent fuel without any burnup limitations, will be designated as Region 1. Region 2 and Region 3 cells will have associated minimum burnup requirements for unrestricted fuel storage.

Fuel assemblies not meeting the minimum burnup requirements for Regions 2 and 3 may also be stored in a checkerboard storage pattern.

The criticality analysis for the Mixed Zone Three Region and/or checkerboard configuration confirms that the Keff is maintained less than 0.95 without credit for the soluble boron in the Spent Fuel Pool. Calculations show that for the most severe accident condition, a soluble boron concentration of 500 ppm boron, in addition to the Boral contained in the racks, would be adequate to maintain the K eff less than 0.95. In accordance with NRC guidelines, the soluble boron in the Spent Fuel Pool may be credited in accident conditions.

A minimum boron concentration of 2000 parts-per-million (ppm) is maintained in the Spent Fuel Pool. The soluble boron in the Spent Fuel Pool will ensure that K eff is maintained substantially less than the design limitations under all conditions.

To accommodate the proposed increase in capacity and change in maximum fuel enrichment, the WCGS Technical Specifications are required to be modified.

The planned expansion (reracking) will take place during Cycle 11 and will preserve the full-core discharge capability through the end of Cycle 27. The following change is being proposed for Section 3/4.9.12, " Refueling Operations," of the Technical Specification:

Revise Technical Specification Section 3. 9.12 to change the burnup and enrichment parameter limitations for Regions 1 and 2 in the Spent Fuel Pool to burnup and enrichment parameter limitations appropriate for Mixed Zone Three Region and checkerboarding storage configurations in both the Spent Fuel Pool and cask loading pit.

Attachment IV provides a copy of the marked-up current Technical Specifications. Attachment V provide a copy of the marked-up Improved Technical Specifications and Bases submitted by the Reference.

Safety Assessment The planned expansion of the storage capacity initially involves replacing the 12 existing racks in the Spent Fuel Pool with 15 new high-density rack modules -

with a total of 2363 storage cells and in a later campaign installing three new high-density rack modules in the cask loading pit with a total of 279 j storage cells. The new cells will contain a fixed neutron absorber for

, primary reactivity control. To maximize storage capacity, the new racks will l be non-flux trap style racks, which are designed without the usual flux. trap design associated with Region 1 style racks. Region 1 cells will be designated within the new storage racks in accordance with administrative controls, which will develop either Mixed Zone Three Region or checkerboarding storage configurations. Both of these storage configurations are described in detail in Section 4 of Enclosure I. Cells designated as Region 1 can store fuel assemblies up to a nominal 5.0 w/o U-235 enrichment, without restriction on burn-up. Spent fuel storage in cells designated as Region 2 or 3 will be subject to enrichment /burnup limitations.

l l'

q Attachment I to ET-98-0009 Pege 4 of 10

' Administrative controls will also ensure that development of either of the two storage configurations (Mixed Zone Three Region or checkerboarding) will

-develop sufficient Region 1 capacity to accommodate a full-core discharge.

Rack modules in both pools will be free-standing'and self supporting. The new modules will be separated by a gap. of L approximately 1.5 inches from one another. Along the pool walls, a nominal gap will also be provided which varies for.each wall. The- minimum dimension is 3/4 inches and the maximum nominal-dimension is 7.43 inches.

With the expanded capacity, the Spent Fuel Pool cooling system will be required to remove an increased heat load while maintaining the pool water temperature below the design limit. The maximum heat load typically develops from the residual heat in the pool after the last full-core discharge at the end of plant life.

The Spent Fuel Pool thermal performance, criticality, and seismic response have been re-analyzed considering the increased storage capacity and fuel enrichment. The results of these analyses have shown that the pool storage

-systems remain adequate.

The Significant Hazards Consideration (SHC), contained in Attachment II, and the enclosed Licensing Report (Enclosure I) address the safety issues arising from the proposed modification and revisions to the technical specifications.

The scope of the technical analysis supporting this evaluation focused mainly on-the final configuration of the expanded storage space. The transition to the final configuration involving some intermediate stages during the pool reracking is also included in the evaluation.

Mechanical Design Evaluation The new fuel rack designs have been evaluated with respect to the mechanical and material qualifications, neutron poison and poison surveillance requirements, fuel handling qualifications, fuel interfaces, and accident considerations.

The principal construction materials for the new racks will be SA240 Type 304L stainless steel, or plate stock, and SA564-630_ precipitation hardened stainless steel for the adjustable support spindles. The rack designs, material selection and fabrication process will comply with the applicable ASTM Standards A240, A276, A479, A564 and others, for service in the nuclear and the boric acid environments. The governing quality assurance requirements for fabrication of the racks are compatible with the quality assurance and quality control of 10CFR50, Appendix B requirements.

For primary nuclear criticality control in the new racks, a fixed neutron absorber will be used, integrated within the rack structure. The absorber, trade name Boral, is a boron carbide and aluminum-composite sandwich. Boral is chemically inert and has a long history of applications in the Spent Fuel Pool environments where it has maintained its neutron attenuation capability under thermal loads. Boral is manufactured under the control of a quality assurance program which conforms to the requirements of 10CFR50, Appendix B.

The installation of the new rack modules will preserve space for thermal expansion and seismic movement. The support legs on the racks will allow for remote leveling and alignment of the rack modules to accommodate variations in the floor flatness. A thick bearing pad will be interposed between the rack pedestals and the floor to distribute the dead load over a wider support area.

The rack structural performance with respect to the impact and tensile loads, as well 'as the subcritical configuration, has been analyzed. The analysis

. included an accidental drop of a fuel assembly during movement to a storage

Attachment I to ET 98-0009 Page 5 of 10 location and tensile loads (vertical and eccentric) on the rack arising from a stuck assembly in the storage cell. It has been shown that these accidents will not invalidate the mechanical design and material selection criteria *a safely store spent fuel in a coolable and subcritical configuration in any region. The storage rack structural integrity, and thus the fuel configuration, will be maintained. The fuel will retain its structural integrity and remain subcritical.

Criticality Considerations The new spent fuel racks are designed to maintain the required subcriticality j margin when fully loaded with enriched fuel and in unborated water at a i temperature corresponding to the highest reactivity. For reactivity control in the racks, Boral panels will be used. The panels have been sized to sufficiently shadow the active fuel height of all assembly designs stored in the pool. The panels will be held in place and protected against damage by a stainless steel jacket which will be welded to the cell walls. A Boral panel will be mounted between adjacent storage cells and on the exterior wall of peripheral cells.

]

Region 1 cells will be identified and designated within the non-flux trap '

style storage racks through development of pool layouts using either Mixed Zone Three Region or checkerboarding configurations. The remaining cells will be designated as either Region 2 or Region 3 storage cells. The storage of spent fuel in each region will be controlled by the criteria defining the maximum permissible reactivity. Region 1 will store the most reactive fresh Zirconium clad fuel with a maximum nominal enrichment of up to 5.0 w/o U-235, or spent fuel regardless of the burn-up history. This Region has been designed to accommodate a full-core discharge. Regions 2 and 3 storage will also accommodate fuel of up to 5.0 w/o enrichment, but will be subject to burnup limitations.

The NRC guidelines and the ANSI standards specify that the margin of safety for criticality be maintained by having the maximum neutron multiplication factor, Ke gf less than or equal to 0.95, including uncertainties, for all normal and accident conditions. The analysis has shown that this criterion is always maintained under all postulated accidents. The accidents and i malfunctions evaluated included a dropped fuel assembly on top of the fuel l rack; impact on criticality of water temperature and density effects; and '

impact on criticality of eccentric positioning of a fuel assembly within the rack.

Thermal Hydraulics and Pool Cooling l

A comprehensive thermal-hydraulic evaluation of the expanded Spent Fuel Pool and cask loading pit has been performed to analyze their thermal performance.

Evaluations performed for the Spent Fuel Pool cooling systems determined the maximum spent fuel decay heat loads which may be accommodated. Heat loads will be maintained below these limiting values by performing refueling outage specific evaluations of the complete pool decay heat load. These outage specific calculations will be required by administrative controls prior to each discharge of spent fuel into the pools. The heat load calculations will be developed in accordance with the provisions of the USNRC Brahch Technical Position ASB 9-2; " Residual Decay Energy for Light Water Reactors for Long Term Cooling" or ANSI /ANS 5.1-1979 and will take into account all past discharges and the predicted heat load for each newly discharged fuel batch.

The time-variant decay heat generated by the upcoming refueling outage core discharge, will be assumed to take place after the shortest period of cooling time allowed by the technical specifications or licensee controlled document and with the highest rate of fuel transfer from the vessel to the Spent Fuel Pool to maximize the heat addition.

' Attachment I to ET 98-0009

, Page 6 of 10

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'Three scenarios were used ' to develop the maximum pool heat loads. The scenarios are based on. current plant practices, and one scenario anticipates

,. Spent Fuel Pool post-LOCA cooling.

l l :In the first scenario, the heat load is from a partial core discharge (up to approximately one half of the core fuel assemblies) 100 hears after reactor L shutdown, plus the additional heat load from fuel assemblies already stored in the pool. One of two, 100 percent capacity cooling loops, is considered l operational. The maximum pool bulk temperature iu limited to less than or equal.to 140*F.

In the second' scenario, the heat load'is assumed to be the maximum a.11owable heat load in the Spent Fuel Pool. Administrative controls assure that the maximum heat load is not exceeded for the full-core discharge case. Again only one cooling loop is operational and the maximum pool bulk temperature is limited to less than or equal to l'10

• F. This scenario also provides the initial conditions for the case of a full-core discharge, followed by a two hour loss of Spent Fuel Pool cooling. In this variation to the scenario, the pool surface temperature is limited to less than boiling during the transient.

, Finally in the post-LOCA scenario, the heat load on the Spent Fuel Pool is very conservatively assumed to be equivalent to that assumed for the partial core discharge scenario. To maximize pool heat, it is assumed the core is

refueled and restarted within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of shutdown, and the resultant decay i heat load on the Spent Fuel Pool is that of approximately half the core plus prior discharges of fuel. It is further assumed that all cooling is suspended i .in the Spent Fuel Fool cooling system, while component cooling water is diverted to the residual heat removal system. At four hours ~ post-LOCA, component cooling water flow is re-established to the Spent Fuel Pool cooling system at 50 percent of its capacity. Analyses have demonstrated that j throughout and subsequent to this Spent Fuel Pool cooling transient, the pool surface temperature is limited to less than boiling.

The local ' water temperature determinations are performed assuming that the Spent Fuel Pool is at its peak bulk temperature. The worst location was identified as the cell with the hottest assembly and the most restrictive

, convective flow. Conservative values for the axial and radial-peaking power factors were used. The local analysis was extended to include the effects of l a partially blocked exit flow, postulated from an accidentally dropped assembly on top of the rack. In all cases analyzed, the heat transfer model

__ conservatively accounted for an additional resistance from the fouling of the l heat transfer surface in the heat exchangers and performance loss due to-( plugged tubes, i

, Under the conditions of forced cooling, the calculated maximum local water l temperature is determined to be 234.6*F in the hottest channel and coincides in time with the highest pool bulk temperature. The maximum fuel cladding temperature at the same location is calculated to be 302*F. These results conservatively assume 50% blockage of the cell inlet hole in the baseplate.

The local boiling point at the top of the fuel, based on the minimum water L level in the pool as required by the technical specifications, is 239'F which indicates that the channel will remain in subeccled flow, thus minimizing the L potential for fuel damage. ,

Under the loss of forced cooling condition, the calculated maximum local water temperature is determined to be at pool boiling conditions in the hottest channel, and coincides 'in time with the highest pool bulk temperature. The maximum fuel cladding temperature is calculated to be 320 *F. These results I conservatively assume 50% blockage of the cell inlet hole in the baseplate.

The local boiling point at the top of the fuel, based on the minimum water level in the pool as required by the Technical Specifications, is 239 'F, and I

Jthe hottest fuel storage channel will experience localized pool boiling.

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i

~ Attachment I to ET 98-0009

! Page 7 of 10 l-l However,-the conservatively high heat flux that was assumed in generating this boiling. condition is far below the critical heat flux for these conditions.

Therefore, neither DNB nor excessive thermal stresses will' be experienced. by the theoretical hottest fuel assembly cladding. In addition, bulk Spent Fuel Pool boiling will not occur.

An evaluation of the Fuel Building's heating, ventilation, and air conditioning (HVAC)' system was also performed for the limiting conditions of normal pool heat load (full-core discharge scenarios). This evaluation has confirmed the adequacy of the.hVAC system to remove the-additional moisture developed from the increase in bulk pool temperature. The building air temperature will be maintained below 110*F, which represents a slight increase from the previously calculated maximum building air temperature of 104*F.

l However, all essential equipment and components have been reviewed to ensure

! that they are qualified for this temperature change.

I l- Seismic and Structural Evaluation A complete re-evaluation of the mechanical and civil structures, to address the structural. issues resulting from the expansion of the pool storage capacity, has been performed. The analysis considered the loads from seismic, thermal, and mechanical forces to determine the margin of safety in the structural integrity of the fuel racks, the Spent Fuel Pool and liner, the

! cask loading pit and liner, the Fuel Building, and the cask loading pit platforms. The loads, load combinations, and acceptance criteria were based on the ASME Section III, Subsection N F, and on NUREG-0800, Standard Review Plan (SRP) Section 3.8.4, Appendix D.

a. The storage rack evaluation l-l The' final configuration of the pool will consist of free standing and self-supporting non-flux trap style rack storage modules. The seismic analysis is performed using a whole pool multi-rack l analysis. It was based on the simulation of the Safe. Shutdown Earthquake and the Operating Basis Earthquake in accordance with SRP 3.7.1 requirements. Separate models were developed for each of the two distinct storage locations; Spent Fuel Pool and cask loading pit. The rack modules were analyzed as completely full using a conservative assembly weight of 1647 pounds. This weight is conservative, since it considers the additional mass of a rod cluster control assembly . to be stored integrally with every l assembly.

The results indicate that the maximum seismic displacements result in no impacts with the pool walls and some inter-rack impacts.

The resultant member and weld stresses in the racks are all below l the allowable stresses, with a safety factor of at least 1.31.

l This minimum calculated safety factor is associated with the pedestal support female thread shear stress. The minimum safety factor for the cell membrane material and associated welding is 2 26. The racks will remain functional during and after a Safe Shutdown Earthquake.

Fatigue analysis was performed on the' storage racks to determine the cumulative damage factor resulting from twenty operating basis earthquakes followed by one design basis earthquake. This analysis showed that the margin is greater than 2 for far.igue within the rack components.

Attachment I to ET 98-0009 Page 8 of 10 The. rack analysis provides pedestal to bearing pad impact loads resulting from lift-off and subsequent resettling during dynamic events. The pool floor stresses were evaluated for these impact loads and determined to remain within alloweble limits even when considering the worst case pedestal location with respect.to leak chases.

In addition to the seismic evaluations, the storage racks ' were also analyzed for all postulated accident conditions. A fuel-handling accident involving a fuel assembly dropped from the spent Fuel Pool Bridge ' Crane highest possible lif t point would not compromise the integrity'of the rack. Permanent deformation of the rack would be limited to the top region only. This is acceptable'since the rack cross-sectional geometry at the active fuel height is not altered. Thus, the functionality of the rack is not affected.

In the event of a stuck fuel assembly in the rack, the resultant load on the members will not affect the rack structural integrity to maintain the fuel storage qualifications.

The platforms are designed to support the storage racks in the cask loading pit and maintain the elevation of the top of these racks level with those in the Spent Fuel Pool. The platforms were designed in accordance with ASME Section III, Subsection NF based l on maximum calculated pedestal loadings from the supported storage

f. racks.

b. Pool and Fuel Building structural evaluatAon

'The Fuel Building consists of cast-in-place monolithic reinforced concrete interior and exterior walls. It is structurally isolated from other structures. The Fuel Building is designed as a seismic Class I structure. The Spent Fuel Pool and cask loading pit are cast-in-place steel lined reinforced concrete tank structures that provides space for storage of spent fuel assemblies.

The pool structure and appropriate portions of the Fuel Building have been analyzed using a 3-D finite element model seismically accelerated with motion applied at the base mat level and pool rack and hydrodynamic loads applied. .The individual loads and load combinations used were in accordance with NUREG-0800, SRP Section 3.8.4 and based on the " ultimate strength" design method.

The primary loads considered were:

the dead weight of the concrete structure, cranes, a Spent Fuel Cask, fully loaded racks, and the water, l

seismic motion consistent with the original plant design for 2% damping of the Operating Basis Earthquake and 4% damping of the Safe Shutdown Earthquake cases applied at the base mat level, hydrostatic pressure force lateral to the walls, 1

hydrodynamic coupling forces applied to the lower portion of the wall and water slosh pressures on top portion of the wall, bounding thermal loads from a full-core discharge and a loss of cooling, producing the largest temperature gradient across the thickness of the wall and the slab, l

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. Attachment I to ET 98-0009 Page-9 of 10 reactive forces on the Spent Fuel Pool wall from the Fuel Building structure, and' seismically induced rack / pool wall impact loads.

In addition ~ to the loads described above, the pool structure and liner were also analyzed for mechanical icids under accident conditions. Analyses were also performed on liner fatigue considering both temperature and seismic cycles. The result of the analyses performed.on the Spent Fuel Pool, cask loading pit and Fuel Building indicate that under all postulated loadings the structural components, floor slabs, pool walls, supporting columns, liner and its anchorages will be subjected to stresses or strains within' acceptable limits.

Radiological Considerations Radiological consequences of accidents in the Spent Fuel Pool building have been evaluated. The events considered were a fuel handling accident and a gate drop. For the fuel handling accident, the radiological dose analysis assumes that only the rods in the dropped assembly are damaged if the assembly is dropped into the Spent Fuel Pool The consequences of the accidental drop of a fuel assembly in . the Spent Fuel Pool have been re-evaluated for the proposed change. The results show that the postulated accident of a fuel assembly striking the top of the storage racks will not result in damage to assemblies other than the dropped assembly. Therefore, the analysis assumptions for the amount of fuel damage resulting from a fuel handling accident are not affected by the reracking of the pool. As a result, the radiological dose at the exclusion area boundary remain within the limits documented in the WCGS Updated Safety Analysis Report for the licensed core parameters. Cycle specific calculations, using core specific parameters, continue to ensure that the radiological dose at the exclusion area boundary remain within the limits documented in the WCGS Updated Safety Analysis Report.

The gate drop accident concluded that no damage to fuel would occur.

Deformation of the upper portion of the rack provides sufficient energy dissipation to preclude penetration to the depth of the upper end fuel fitting. Therefore, fuel will not be impacted.

A rack drop involving radiological consequences 'is precluded, since i ' rack movement during the removal and installation phase will' follow safe lo.. paths that prevent heavy loads from being transported over the stored spent fuel.

Thus, there are no credible radiological consequences from this accident.

- There has been no steady long-term increase of radiological conditions in the Spent Fuel Pool resulting from the radionuclides within the fuel as more spent fuel is added to the pool. The radiological conditions within the building

- are typically dominated by the most recent batch of the spent fuel from a full-core discharge. The radioactive inventory of the older fuel that will increase with the expanded storage capacity will be insignificant compared to that of the recent discharge.

Since.the new storage racks will be located in closer proximity to the Spent Fuel Pool walls, an increase in the adjacent radiological doses is expected.

Radiological analyses have.shown that the dose levels adjacent to all. Spent Fuel Pool areas will remain within acceptable levels, f

Attachment I to ET 98-0009 Page.10 of-10 Supporting Analysis For supplNaental information on the WCGS Spent Fuel Pool Reracking License Amendment Request, refer to the enclosed Licensing Report. Two versions.of the report are enclosed. The version included as Enclosure I contains complete documentation for all sections of the report, including some information which is considered proprietary pursuant to 10CFR2.790. WCNOC requests that this version be ' withheld from public viewing. The version

included as Enclosure II is identical, except that proprietary information has been removed and replaced by a note of explanation at each location where information has been omitted. WCNOC offers this additional version for the purposes of public review.

Schedule Regarding the proposed schedule for' this amendment, . it - is requested that issuance be no later than November 1998 in conjunction with the Union Electric license amendment request. This approval date will support the removal of the existing WCGS racks and also to support the scheduled rerack completion date of May 2000, which is prior to the date that Cycle 12 fuel will begin arriving on site.

Conclusion Based on the above discussions and the No Significant Hazards Consideration Determination presented in Attachment II, the proposed changes do not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report; or create a - possibility for an accident or' malfunction of a different type than any previously evaluated in the safety analyses report; or reduce the margin of safety as defined in the basis for any technical specification. Therefore, the proposed changes do not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.

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Attachment II to ET 98-0009 Page 1 of 7 ATTACHMENT II NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

I Attachment II-to ET 98-0009 Page 2 of 7

[ No Significant Hazards Consideration Determination i

I Proposed Changes l This license amendment request proposes to revise Wolf Creek Generating Station (WCGS) Technical Specifications by incorporating the proposed changes identified in Attachment IV-to increase the spent fuel storage capacity and increase the maximum nominal fuel enrichment to 5.0 w/o (nominal weight-percent) U-235.

l Application of Standards H The following Standards identified in 10 CFR 50.92 have been used to determine whether ' the proposed changes involve a Significant Hazards Consideration.

Each of the identified proposed changes is evaluated against the three Standards.

Standard I - Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated i

In the analysis of the safety issues concerning the expanded Spent Fuel-Pool storage capacity, the following previously postulated accident scenarios have been considered:

a. A spent fuel assembly drop in the Spent Fuel Pool

b. Loss of Spent Fuel Pool cooling flow l

! c. A seismic event

d. Misloaded fuel assembly l I l The probability that any of the accidents in the above list can occur is not significantly- increased by the modification itself. The probabilities of a seismic event or loss of Spent Fuel Pool cooling flow are not influenced by the proposed changes. The probabilities of accidental fuel assembly drops or misloadings are primarily influenced by the methods used to lift and move these loads. The method of handling Avads during normal plant operations is not significantly changed, since the same equipment (i.e., Spent Fuel Pool Bridge Crane) and procedures will be used. A new offset handling tool will be required to access some storage rack cells located adjacent to the pool walls.

The grapple mechanism, procedures, and fuel manipulation methods will be very similar to those used by the spent fuel handling tool. Therefore, this tool does not represent a significant change in the methods used to lift or move fuel. Since the methods used to move loads during normal operations remain

, nearly the same as those used previously, there is no significant increase in l'

the probability of an accident.

During rack removal and installation, all work in the pool area will be controlled and performed in strict accordance with specific written

! procedures. Any movement of fuel assemblies required to be performed to l support the modification (e.g., removal and installation of racks) will be performed in the same manner as during normal refueling operations. Shipping cask movements will not be performed during the modification period.

Accordingly, the proposed modification does not involve a significant increase in the probability of an accident previously evaluated.

The consequences of the previously postulated scenarios for an accidental drop

of a fuel assembly in the Spent Fuel Pool have been re-evaluated for the l

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Attachment II to ET 98-0009 Page 3 of 7 proposed change. The results show that the postulated accident of a fuel

-assembly striking the top of the storage racks will not distort the racks sufficiently to impair their functionality. The minimum suberiticality margin, K eff less than or equal-to 0.95, will be maintained. The structural damage to the Fuel Building, pool liner, and fuel assembly resulting from a fuel assembly drop s+riking the pool floor or another assembly located within the racks'is primarily dependent on the mass of the falling object _ and the drop height. Since these two parameters are not changed by the proposed modification, the structural damage to these items remains unchanged. Cycle specific calculations, usinc core specific parameters continue to ensure that the radiological dose at the exclusion area boundary remain within the limits documented in the WCGS Updated Safety Analysis Report. Dose levels remain i

well within the levels required by 10 CFR 100, paragraph 11, as defined in Section 15.7.4.II.1 of the Standard Review Plan. Thus, the results of the postulated fuel drop accidents remain acceptable and do not . represent a significant increase in consequences from any of the same previously evaluated accidents that have been reviewed and found acceptable by the NRC.

The consequences of a loss of Spent Fuel Pool cooling have been evaluated and found to have no increase. The concern with this accident is a reduction of Spent Fuel Pool water inventory from bulk pool boiling resulting in uncovering fuel assemblies. This situation would lead to fuel failure and subsequent significant increase in offsite dose. Loss of Spent Fuel Pool cooling at WCGS is mitigated in the usual manner by ensuring that a sufficient time lapse exists between the loss of forced cooling and uncovering fuel. This period of ;

time is compared against a reasonable period to re-establish cooling or supply an alternative water source. Evaluation of this accident usually includes determination of the time to boil. The time allowed for operator action is much less than the onset of any significant increase in of fsite dose, since once boiling begins it would have te continue unchecked until the' Spent Fuel Pool surface was lowered to the point of exposing active fuel. The time to boil represents the onset of loss of Spent Fuel Pool water inventory and is commonly used as a gage for establishing the comparison of consequences before and after a refueling project. The helt up rate in the Spent Fuel Pool is a nearly linear function of the fuel decay heat load. The fuel decay heat load will increase subsequent to the proposed changes because of the increase in the number assemblies and higher fuel burnups. The methodology used in the thermal-hydraulic analysis determined the maximum fuel decay heat loads which are allowed by maintaining the current time allowed for operator action (i.e.,

more than two hours to boil during complete loss of forced cooling).

Therefore, the allowed operator action time remains unchanged from the previous design basis. In the unlikely event that all Spent Fuel Pool cooling is lost, sufficient time will still be available subsequent to the proposed changes for the operators to provide alternate means of cooling before the onset of pool boiling. Therefore, the proposed change represents no increase in the consequences of loss of Spent Fuel Pool cooling.

The consequences of a design basis seismic event are not increased. The consequences of this accident are evaluated on the basis of subsequent fuel damage or compromise of the fuel storage or building configurations leading to radiological or criticality concerns. The new racks have been analyzed in their new configuration and found safe during seismic motion. Fuel has been determined to remain intact and the storage racks maintain the fuel and fixed poison configurations subsequent to a seismic event. The structural capability of the pool and liner will not be exceeded under the appropriate combinations of dead weight, thermal, and seismic loads. The Fuel Building structure will remain intact during a seismic event and will continue to adequately support and protect the fuel racks, storage array, and pool moderator / coolant. Thus, the consequences of a seismic event are not increased.

Attachment II to ET 98-0009 Page 4 of 1 Fuel misloading accidents were previously postulated occurrences. The consequence of this type of accident has been analyzed for the worst possible storage configuration subsequent to the proposed modification and the consequences were found to be acceptable because the reactivity in the Spent Fuel Pool remained below 0.95. After the proposed modification, the worst case postulated accident condition, for the Mixed Zone Three Region configuration, occurs when a fresh fuel assembly of the highest possible enrichment is inadvertently loaded into a Region 2 storage cell. Further, after the proposed modification, the worst case postulated accident condition, for the checkerboard configuration, occurs when a fresh fuel assembly of the highest possible enrichment is inadvertently loaded into an empty storage cell. In both postulated accident scenarios, credit is allowed for soluble boron in the water, and the Spent Fuel Pool reactivity is maintained below 0.95. Therefore, there is no increase in consequences due to the modification.

Therefore it is concluded that the proposed changes do not significantly increase the probability or consequences of any accident previously evaluated.

Standard II - Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated To assess the possibility of new or different kind of accidents, a list of the important parameters required to ensure safe fuel storage was established.

Safe fuel storage is defined here as providing an environment which would not present any significant threats to workers or the general public. In other words, meeting the requirements of 10 CFR 100 and 10 CFR 20. Any new events which would modify these parameters sufficiently to place them outside of the boundaries analyzed for normal conditions and/or outside of the boundaries previously considered for accidents would be considered a new or different accident. The criticality and radiological safety evaluations were reviewed to establish the list of important parameters. The fuel configuration and the existence of the moderator / coolant were identified as the only two parameters which were important to safe fuel storage. Significant modification of these two parameters represents the only possibility of an unsafe storage condition.

Once the two important parameters were established, an additional step was taken to determine what events (which were not previously considered) could result in changes to the storage configuration or moderator / coolant presence during or subsequent to the proposed changes. This process was adopted to ensure that the possibility of any new or different accident scenario or event would be identified.

Due to the proposed changes, an accidental drop of a rack module during construction activity in the pool was considered as the only event which might represent a new or different kind of accident.

An installation accident of a rack dropping onto stored spent fuel or the pool floor liner is not a postulated event due to the defense-in-depth approach to be taken, as discussed in detail within Section 3.5 of the enclcsed Licensing Report (Enclosure I). This approach is similar to that taken previously for lif ting a pool gate with the Spent Fuel Pool Bridge Crane. A new temporary hoist and rack lifting rig will be introduced to lift and suspend the racks f rom the bridge of the Cask Handling Crane. These temporary lift items have been designed in accordance with the requirements of NUREG-0612 and ANSI N14.6 with respect to redundancy in load path or safety margin. The postulated rack drop event is commonly referred to as a " heavy load drop" over the pools.

Heavy loads will not be allowed to travel over any racks containing fuel assemblies, thus a rack drop onto fuel is precluded. A rack drop to the pool liner is not a postulated event, since all of the lifting components (except for the Cask Handling Crane) either provide redundancy in load path or are designed with safety margins greater than a factor of ten. Nevertheless, the

Attachment II to ET 98-0009 Page 5 of 7 l

analysis of a rack dropping to the liner has been performed and shown to be acceptable.

However, the question of a new or different type of event is answered by determining whether similar heavy loads have been carried over the pool.

As stated above, pool gates have been previously lifted witnin the Spent Fuel Pool.

The pool gate and the storage racks are both designated as "similar.

heavy loads" and the safeguards taken to preclude these accidents are All movements of heavy loads over the pool will comply with the applicable administrative controls and guidelines (i.e., plant procedures, NUREG-0612, etc.).

Therefore, the rack drop does not represent a new or different kind of accident The proposed change does not alter the operating requirements of the plant or of the equipment credited in the mitigation of the design basis accidents. The proposed change does not affect any of the important parameters required to ensure safe fuel storage. Therefore, the potential for a new or previously unanalyzed accident is not created.

Standard III - Involve a Significant Reduction in the Margin of Safety The function of the Spent FLel Pool is to store the fuel assemblies in a suberitical and coolable configuration through all environmental and abnormal loadings, such as an earthquake or fuel assembly drop. The new rack design must meet with compatible all applicable requirements the Spent Fuel Pool. for safe storage and be functionally WCNOC capacity in has theaddressed the safety issues related to the expanded pool storage following areas:

a. Material, mechanical and structural considerations

b. Nuclear criticality

c. Thermal-hydraulic and pool cooling The mechanical, material, and structural designs of the new racks have been reviewed in accordance with the applicable provisions of the NRC Guidance entitled, " Review and Acceptance of Spent Fuel Storage and Handling Applications."

The rack materials used are compatible with the spent fuel assemblies and the Spent Fuel Pool environment. The design of the new racks preserves the proper margir. of safety during abnormal loads such as a dropped assembly and tensile loads from a stuck assembly. It has been shown that such loads will not invalidate the mechanical design and material selection to safely store fuel in a coolable and subcritical configuration.

The methodology used in the criticality analysis of the expanded Spent Fuel Pool meets the appropriate NRC guidelines and the ANSI standards (GDC 62, NUREG-0800, Section 9.1.2, the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," Regulatory Guide 1.13, and ANSI ANS 0.17). The criticality analysis for the Mixed Zone Three Region and/or checkerboard configuration confirms that the Keff is maintained less than 0.95 without credit for the soluble boron in the Spent Fuel Pool. Calculations show that for the most severe accident condition, a soluble boron concentration of 500 ppm boron, in addition to the Boral contained in the racks, would be adequate to maintain the K eff less than 0.95. In accordance with NRC guidelines, the soluble boron in the Spent Fuel Pool may be credited in accident conditions. A minimum boron concentration of 2000 parts per-million (ppm) is maintained in the Spent Fuel Pool. The soluble boron in the Spent Fuel Pool will ensure that Keff is maintained substantially less than the design limitations under all conditions. The margin of safety for subcriticality is maintained by having the neutron multiplication factor equal to, or less than, 0.95 under all accident conditions, including uncertainties.

Attachment II-to ET 98-0009 Page 6 of 7 This criterion is the same as that used previously to establish criticality safety evaluation acceptance and remains satisfied for all analyzed accidents.

The thermal-hydraulic and cooling evaluation of the pool demonstrated that the pool can be maintained below the specified thermal limits under the conditions of- the maximum heat -load and during all credible accident sequences and seismic events. The bulk pool temperature will not exceed 207 'F during the worstisingle-failure of a cooling pump. Localized pool boiling is predicted to occur ~in the worst - single failure of a cooling pump in the hypothetical worst case storage cell, immediately following the completion of a full-core discharge. This cell is very conservatively modeled to contain the hottest spent fuel assembly, with maximum flow resistance includi ng 50% blockage- of both the inlet' and outlet flow areas. However, bulk

• boiling will not occur, nor will fuel cladding . experience DNB or exce- thermal stresses.

The fuel will not undergo any significant heat up after an accidental drop of a fuel assembly on top of the rack blocking the flow path. A loss of cooling to the pool will allow sufficient time (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) for the operators to intervene and line up alternate cooling paths and the means of inventory make-up before the onset of pool boiling. Therefore the allowed operator action time remains unchanged from the previous design bases. In the unlikely event that all pool cooling is lost coincident with the completion of a full-core discharge, sufficient time will still be available subsequent to the proposed changes for.the operators to provide an alternate means of cooling before the onset of bulk pool boiling. Therefore, the accepted margin of safety remains l the same.

Thus, it is concluded that the changes do not involve a significant reduction in the margin of safety.

The NRC has-provided guidance concerning the application of standards in 10 CFR 50.92 by providing certain examples (51FR7751, March 6,- 1986) of amendments that ' are considered not likely to involve a Significant Hazards Consideration. The proposed changes for WCGS are similar to Example (x): an expansion of the storage capacity of Spent Fuel Pool when all of the following are satisfied:

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(1) The storage expansion method consists of either replacing existing racks l with a design that allows closer spacing between stored spent fuel I assemblies or placing additional racks of the original design on the Spent Fuel Pool floor if space permits.

The WCGS reracking modification involves replacement of the existing racks with a design that will allow closer spacing of the stored fuel assemblies and placement of additional racks of the same configuration'in an adjacent pool.

(2) The storage expansion method does not involve rod consolidation or double l

tiers.

The WCGS reracking does not involve fuel consolidation. The racks will not be double - tiered; no fuel assemblies will be stored above other assemblies.

'(3) The Kgff of the Spent Fuel Pool and cask loading pit is maintained less than, or equal to, 0.95.

The design of the new racks integrates a neutron absorber, Boral, within the racks to allow closer storage of spent fuel assemblies while ensuring that K egg remains less than 0.95 under all conditions. Additionally, the water in the Spent Fuel Pool and cask loading pit does contain boron as L

Attachment II to ET 98-0009 Page 7 of 7 further assurance that K,ff remains less than 0.95. The boron that is contained in the pool is not credited, except in the accident condition.

(4) No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary to jurtify the expansion.

The rack vendor has successfully participated in the licensing of numerous other racks of a similar design. The construction process and the analytical techniques of the WCGS Spent Fuel Pool expansion are substantially the same as in the other completed rerack projects. Thus, no new or unproven technology is used in the WCGS reracking.

Conclusions Based on the above discussions, it has been determined that the requested technical specification revisions do not involve a significant increase in the probability of consequences of an accident or other adverse conditions over previous evaluations; or create the possibility of a new or different kind of accident or condition over previous evaluations; or involve a significant reduction in a margin of safety. Therefore, the requested license amendment does not involve a significant hazards consideration, i

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l Attachment III to ET 98-0009 Pags 1 of 2 ATTACHMENT III ENVIRONMENTAL IMPACT DETERMINATION

Attachment III to ET. 98-0009 Page 2 of 2-Environmental Impact Determination This. license amendment request proposes to revise Wolf Creek Generating Station (WCGS) Technical Specifications by incorporating the proposed changes identified in Attachment IV to increase the Spent Fuel Pool storage capacity and increase the maximum fuel enrichment to 5.0 w/o (nominal weight percent)

U-235, 10 CFR 51.22(b) specifies the criteria for categorical exclusions from the requirement for a specific environmental assessment per 10 CFR 51.21. This amendment request meets the uriteria specified in 10 CFR 51. 22 (c) ( 9) as specified below:

(i) the amendment involves no significant hazards consideration As demonstrated in Attachment II, the proposed changes do not involve any significant hazards consideration.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite The proposed changes do not involve a change to the facility or operating procedures that would cause an increase in the amounts of effluents or create new types of effluents. Radiological consequences of accidents in the Spent Fuel Pool building have been re-evaluated and have been shown to be acceptable.

(iii) there is no significant increase in individual or cumulative occupational radiation exposure The proposed changes do not significantly increase the individual or cumulative occupational radiation exposures. Radiological consequences of accidents in the Spent Fuel Pool building have been re-evaluated and have been shown to be acceptable.

Based on the above, it is concluded that there will be no impact on the environment resulting from this change and the change meets the ' criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.21 relative to requiring a specific environmental assessment by the l Commission.

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