ML20087A064

Safety Evaluation Supporting Amend 175 to License DPR-59

ML20087A064
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Site:	FitzPatrick
Issue date:	12/31/1991
From:	Office of Nuclear Reactor Regulation
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ML20087A063	List: ML20087A061 ML20087A064
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NUDOCS 9201080151
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SAFETY EVALUATION RY THE OFFICE OF NUCLEAR REACTOR REGill ATION RELATED TO AMENDMENT NO.175 TO FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50 333 1.0 INTRODUCTI^N By letter dated May 31, l90, as supplemented October 31, 1990, December 5, 1990, June 26, 1991, July 12, 1991, July 16, 1991, and September 19, 1991, the power Authority of the State of New York (the licensee) submitted a regnest for changes to the James A. FitzPatrick Nuclear Power Plant, Technical Specifications (TS). The requested chenges would revise TS Section $.S.B and the associated Bases. Specifically the number of spent fuel assemblies that can be stored in thespentfuelpool(SFP)willbeincreasedfrom2244to2797.

The FitzPatrick plant SFP was reracked following approval of Amendment No. 55, dated June 18, 1981, with high density racks thus increasing storage capacity to 2244 fuel assemblies. Under the proposed expansion, five new rack modules containing 553 storage locations will be added increasing total storage capacity to 2797 The increased storage capacity will extend the capability for a full-core offload to the uar 1997. This effort is consistent with the objective of the Nuclear Aste Policy Act of 1982 which requires that licensees exhaust all means of storing spent fuel on site.

Supplemental letters of October 31, 1990, December 5, 1990, June 26, 1991, July 12, 1991, July 16, 1991, and September 19, 1991, provided clarifying information that did not change the initial proposed no significant baratds consideration attermination.

2.0 EVALVATION 2.A CRITICALITY ANALYSIS

.:. A.1 Analytical Methodoloal TtecurrentlechnicalSpecification(TS)forthespentfuelstoragepooland existin9 racks (TS 5.5.8) states that the k-effective of the pool shall be less thaa 0.95. The specification further indicates that this k-effective value is satisfied if the maximum exposure depnndent k-infinity of the stored fuel assemblies is less than 1.36. The pro >osed amendment does not change this specification and the pool criterion t1us remains at 0.95 for the new racks. The proposed changes to TS 5.5 B are to revise the maximum number of I

9201000151 911 Ell PDR ADOCK 05000333 I

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. fuel assemblies that may be stored in the pool and to referenc? the May 31, 1990, submittal.

in additfoa, 1s Bases S.S.B is changed to reflect the new maximum number (2797) of stored fuel assemblies and tc. eliminate reference to 3

the 3.2% delta-k margin which was calculated for the existing racks but does not apply to the new racks.

The new rack design comprises a rectanoular array of stainless steel " box" storage cells. A Roral panel it positioned on each interface between adjoining cells. Each Boral panel is sandwiched between the box wall and a stainless steel sheathing welded to the wal' in a manner t.uch that the panel is i

unconstrained.

The design basis fuel employed in the criticality calculat'ont for the new storage racks is an 8x8 BWR fuel rod assemoly with a uniform enrichment of 3.3 w/o U-235 without gadollaium burnible poison.

This represents the most reactive fuel authurized for storage at the FitrPatrick facility.

For the reference design, the pool moderator is assumed to be pure, unborated water at the minimum temperature within the operating rarge (68 'F), corresponding to the nighest reactivity. The Boron-10 contained in the Doral panels was assumed to be uniformly distributed with a minimum areal density corresponding to the icwer limit of the manufacturing tolerance.

The criticality and associated sensitivity calculations were done using both the Monte Carlo code AMPX-KENO (using the 27 group SCALE cros; cections, with NITAW'.)$M0-2f. as the primary method,have been benchmarktd against a number o and the two-dimensional multi-qroup transport code CA These methodologies relevant critical experiments simulating storage racks designed by Holtec and others. These experiments have covereu a range of (,enmetries, material compositions, fuel errichments, and poison sheets. These benchmark calculations have been used to develop methodology bias and uncertainty factors to be added to the nominal k-effective calculations for the FitzPatrick racks.

As part of the sensitivity calculations, the licansee has examined the effect of rack manufacturing tolerances on the computation of k-effective for the rack-fuel system.

The parameters considereo included boren loading density, Boral panel width, storage cell lattice pitch, and stainless steel box wall and backing plate thicknesses. The effect of storing a fuel assembly without a surrounding rirconium flow channel was also examired.

In addition one-dimensional axial calculations were performed to evaluate the effect of reduced Boral plate lengths on reactivity.

Neutron absorption oy the structural stainless steel above and below the active fuel was neglected in these computations and fresh unburned fuel with uniform enrichment and ro gadolinium was assumed.

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9 i The effect of water gap spacing between rack : nodules was determined by CASH 0-2E calculations. The noninal gap along the interface totween modules was shown to eliminate the need for Boral panels on the module walls. However, as a precautionary measure against module movement resulting from a seismic event, the rach design provides for Boral panels on the walls of alternate cells along one side of the interface.

The effects on reactivity of abnormal conditions and accidents associated with the spent fuel pool have also been evaluated.

These included increased pool water temperature and void formation, the misplacement of a fuel assembly outside and adjacent to the fuel rack, eccentric fuel assembly positioning within a storage cell, lateral rack motion due to a design basis earthqual'e (discussed above), and a fuel assen61y dropped on top of the rack.

Computations indicated that these conditions resulted in either a neoative reactivity effect or e negligible increate (less than 0.0001 delta-k) in reactivity.

Design basis reactivity calculations resulted in a k-infinity of 0.9297 (bias corrected CASMO). With all known uncertainties statistically combined (a delta-k of i 0.0071), the maximum k-infinity)in the fuel rack becomes 0.937 (95% probability at the 951 confidence level.

This satisfies the design basis requirement of a maximum k-effective of less than 0.95.

Independent verification calculations using AMPX-XEN0 resulted in a k-infinity of 0.924 1 0.008 (95%/9S%, corrected for bias and temperature), which is in agreement with the reference calculation.

2.A.2 Conclusion The basis c?iticality design of the new racks, usir.g baron lined cells to provide the appropriate neutrcn multiplication level for the closer packed array of high density racks, is a cortunonly used concept and has been accepted tor many spent fuel storage pools.

It is an acceptable design concept for i

maintaining criticality levels for the fitzPatrick pool.

The analytical methodologies used to analyze the criticality and reactivity change characteristics of the rocks are standard methodologies, commonly u;ed and approved for other '.u.ensees for such analyses. The CASMO-2E code provides an acceptable methodology for base calculations and for sensitivity calculations, and the AMpX-XENO. SCAL.E code package 3rovides suitable backup and confirmation calculations.

These met. hods have aeen benchmarked against an appropriate selection of critical experiments, with results falling within expected ranges of deviations from the e m rinents. The derivation of the l

uncertainty of the methodology from this aenchurking follows norrnal procedures and also falls within an expected range. Therefore, the staff concludes that the criticality analysis is acceptable.

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The examination of uncertainties attributed to variances in dimensions and rnaterials in the fuel and racks has ', overed an acceptable range of paramcters

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and has used a suitable, standard raethodology for determining the reactivity j

effects and their statistical combination.

Ine examination of the ef'ects of abnorral conottions has covered the standard events relating to changes in temperature and density, seismic movements of racks, and misolacertent and cionping if fuel assernblies.

The staff concludes that these results are acceptable.

2.8 MATERIAL COMpATIBillTY AND CHEMICAt. STABILITY 2.B.1 Discussion Nuclear power plants provide storage facilities or pools for the wet storage of spent fuel assemblies. The safety function of the spent fuel storage pools is to maintain tiie spent fuel assemblies in a subcritical.irray during all credible storage conditions. The M C staff has reviewed the compatibility and chemical stability of the materials wetted in the pool e,ater.

The currently requested expansion would increase the capacity of the went fuel pool to 2797 fuel assemblies. This expansion will be accot.plished by th*

addition of five new mcJules. They wi P be constructed from ASTH A240-Typa 3041 stainless steel with oniy the adjustable suppor t spindles made from A564-Type 630 arecipitation hardened stainless steel.

The neutron abso bing material will J;e Beral with B-10 loading of 0.0135 gm/sq.cm. Boral is a material consisting of a dispersed baron carbide in a 1100 aluminun alloy matrix and clad with 1100 aluminum alloy.

It is in a form of 5 inches wide, 0.075 inches thick, and 144 inches long panels.

It is held ar. the Side of the

. cell by a stainless steel picture freme theathing.

The sheathing is telded to the box at the top und bottom and at staggered positions alon9 the longitudinal length.

This design allows pool water free entry to the cavity, and the gases produced by radiolysis and/or water-aiuminum reaction are free to escape, thus

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preventing swelling and bulging due to pressure buildup. The spent fuel pool contains air-saturated demintralin.d water with conductnity of less than l

5AS/cm and chloride contents of less tonn 500 ppb.

The licensee proposed a surveillance program to monitor parforir:ance of the Boral in the spenc fuel pool.

For that pur>ose, ten specially desioned test l

coupons will be placed in locations where tiey will be exposed to the typical I

spent fuel pon1 environment.

Each coupon will have a Boral specinien encased in L

a jacket of a material identical to that used in the racxc, and the position and tolerances will be simi'ar as that in the actual fuel cell. The jacket will have provisions for easy opening without disturbing the Boral specimcn.

l The coupons wfil be removed at scheduled irtervals and e u mined for loss of

hysical and neutron absorbing properties.

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5 2.B.2 Evaluation The low carbon sustenitic stainless " teel in the spent fuel racks is compatible with the high purity, detrineralized, air-saturatM water and the radiation environment of the spent fuel poal. Oxygen dissolved ir water will help to passivate the stainless steel.

In this environment, austenitic stainless steci will exhibit only extremely low rates of corroi, ion These corrcsion rates are negligible for even the thinnest stainless steel riements of rack asserblies, r,alvanic attack between stainle n stsel, Zirceloy in the fuel assemblies, and Boral will not be signifiant since the conductivity of water in the pool is relatively low and the materials arc protected by passivating oxide films. The concentrt. tion of chloride is maintained below the limit et which significant initiation of stress corrosion cracking could occur.

Roral has undergone extensive testing to study the effects of gamma irradiation in variot.s environments and to verify its structural integrity and suitability as a neutron absorbing material.

It has been qualified for 1.0E11 rads of aamma radiation while f.taintaining its neutron attenuation capability. Tests have shown that Boral does not possess leachable halogens that could be released into the pooi environment in the presence of radiation.

Similcr findings have been nade regarding the leaching of elemental boron from the Boral.

Surveillance coupens containing Boral <1ll provide tim related information of the actual behovior of Borel in the spent fuel pool. The staff reviewed the description of the proposed surveillance program for monitoring the Boral in the spent fuel pool and concludes that the program is adequate to reveal deterioration that might lead to loss of neutron absorbing capability during the life of the spent fuel racks. The staff does not anticipate that such de+erioratien will occur,

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but in case it does, it would be gradual.

In +he unliL.y event of Boral l

deterioration in the pool environment, the monitoring program will detcet such deterioration and allow the licensee time to take suitable corrective actions.

2.0.3 Conclusions l

l Based on the above discussion, the staff concludes that corrosion of '5e l

proposed fuel storage racks due to the sxnt fuel pool environment should be of little significance during the life of t1e facility. The surveillance program proposed by the licensee would reveal any deteriorations in neutron absorbing capability of Boral and if a significant degradation is found, the licensee would have sufficient time to take the appropriate correctivn measures.

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The staff finds that the selection of appropriate materials of construction, ar.d the development of a proposed Boral surveillance program meet the requirements of If1 CFR Part 50, Appendix A, Genert1 Design Criterion 61, I

regarding the capability to permit appropriate periodic inspection and tetting l

of components, and General Criterion 62 regarding prevention of criticality ty the use of neutron absorbers and by maintaining structural integrity of components and are, therefore, acccptable.

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2.0 THERMAt/ MECHANICAL LOAD C0hS1pIRAT10NS 2.C.1 Spent Fuel pool Cooling No modificatiens to the spent fuel pool (SFP) cooling system are necesscry with the proposed expansion.

Theref cre, the spent fuct pool cooling system was only reviewed against the reouirements of General resign Criterion (GDC) a for decay heat removal and GDC 2 for makeup durinc loss of all cooling er defined in Standard Review plan, Section 9.1.3, for storrge of 2797 fLel assrshlies.

The Fit: patrick spent fuel pool cooling system consists of two pur.ps and two heat exrhangers for normal decay heat removal. Bothheatexchangers, phen supplied by a single SFP cooling punp, are designed to transfer 6.3x10 BTU /hr from 125 'F fuel pool water to 95 'F reactor building closed loop cooling water which flows thtough tSe the11 side of each beat exchantier. The spent fuel pool cooling system can be supplemented by the use of the RPR system in the spent fuel cooling assist mode. The Rhr assist rnode is available during plant shutdowns. When the temperature of the spent fuel pool exceeds the peak efficiency temperature of 100 F for the spent foal pool cleanup systm, the filters and demineralizers can be bypassed.

0 The licensee calculated decay heat loads of 13.1x10 BTU /hr after a normal discharge of spent fuel during refueling. This is based on a proposed storage capecity of P797 spent fuel assemblies. The heat load value was compared to Branch Technical Position 9-2 and found to ne conservative. The calculated pool temperature rises to a maximum of Irss than 150 'F at 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> after shutdown for a normal discharge of 208 fuel asseinblies.

For a sing's active failure af ter the plant has starteC up, the muimum fuel pool temperature would be maintained lest than 150 #F.

This max; mum temperature is above the guidtline of 140 *F for a normal discharge; hwever, it is acceptable becaase it is well below the boiling temperature.

l When a full core is offloaded into the spent fuel pool, the Residual Heat Removal (RHR) System will be usea to maintain the fuel pool temperature at or below 135 'f.

The use of the RHR assist mode for toolin; when the full core is unloadedwatacceptedinaSafetyEvaluationdatedJune18,19gl. The decay heat load for a full-core offload is calculated to be 25.79x10 BTV/hr at 238 hours0.00275 days <br />0.0661 hours <br />3.935185e-4 weeks <br />9.0559e-5 months <br /> af ter shutdown. This heat load results in a calculated maximum temperriure of 133 'f with RHR assist in operation. The maximum terperature of the pool for the abnormal condition of full-ccre offload with RHD assist is acceptable because if is below the boiling temparature, l

Hakeup for the SFP is manually transferred from the seismic Category I condensate storage system to the skimer nurge tanks to make up any pool losses.

i Capability exists to add water from t,ake Ontario to the pool through the RHR system in the event of loss of normal makeup system and when pool water level l

is threatened due to heavy pool water inventory loss.

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Based on th9 above, the decay hett removal for normal and abnormal conditions, and the makeup capablHty are acceptela.

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?.C.2 H_eavy load Handlino A spent fuel storage rack is considered to be a heavy inad because is weighs more than a spent futi ass 2mbly and its hardling tool. The licensae indicated that lifting and installation of'the opent fuel racks will be performed in accordance with the guidelines of NUREG-0612. All load handling will follos clearly estab!ished safe load handling paths. The crane operator will be ghen special training and will be required to follow specific load handling procedures. The 1)fting crane and the rig will meet the NUREG-0612 stress and inspection criteria.

In a Safety Evaluation dated January 3,1984, the licensee's provisions for handli.19 and control'of heavy loads at the FittPatrick Nuclear power Plant were found to meet the guidelines of NUREG-0612.

i-Based on the above, the staff finds *; hat heavy load handling will be performed

-in accordance with tne guidelines of NUREG-0612 to ensure that an unacceptable release of radioactivity or criticality accident uf11 not result from a heavy-load drop, and is therefore acceptable, t

2.0 !TP.UCTL"1AL DESIGN This evaluation addresses the adequacy of the structural ano seismic aspects of the application submitted by the licensee in support of their increa.ted rack capa.:ity in the spent fuel 2001 The primary areas of review are focused on the structu al integrity or tie fuel, fuel cells, rack moduler, and the spent fuel pool floor and walls under the postulated loads (Append;x 0 of SRp 3.8.4) and fuel handling accidents.

.l 2.0.1 Structu N 1-Analysis Sjent fuel Storage Pool t

lim spent fuel pool is a reinforced concrete structure and is designed as a Seismic Category I structure. The pool is approximar.ely 31 feet wide,10 feat long a,.d 37 feet deep with a 5 feet thick slab. Wetted surfaces of the pool are lined with stainless steel to ensure water tignt integrity.

'no concrete strength capacities were compared with anticipated loads on the l

concrete' structure from high density rack dynamic 1 cads as weli as other loadings speciffed in the Standard Review plan and the margins were found to be acceptable.

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The staff, therefore, c)ncludes that the Fitzpatrick spent fuel pool will continue to support the additional loads caused by additional fus1 d ring no mal, severe environmental, and accident ::onditions and maintain its integ*ity.

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-b-l Refuelino Acciderits 1

The following three accidents were evaluated by the lice:deu:

(a) a fuel assemt4 is dropped f rom 6n elevation PP above a stortte location and impacts the base of the module, (b) a fuel essembiv is dro? ped from an elevation 24" above the rack ord hits the top of the rack, and (c) the same as (b) except that the f'Jel ansembly 1:, assumed to be dropped in an irclined manner on the top of the rack.

The licensee found that the above po.tulated accidents would rtot lead to adverse conditioris, incloding unacceptable damooe to the fuel.

Furthnrrnore, tb licentee found that the rack cross-sectione s geometry would not be altered during these accidents.

The NRL staff has evaluated the licensee's analyses and cor.cludes that its l

f ndingr 1re acceptable.

i hck Motu'ies The five racks to be edded to the pool are seismic Category ! equipment and are, therefore,) required t9 remain functional during and after a Safe Shutdown Earthouake (SSE.

ihey are neither anchured tc the pcol floor nor the pool wall and are rot structurally intercnnnected.

Each rack rnedule is )rovided witi) leveling 'eads M ich support the rack and are in contact with t:e spent fuel pool (100.

The license? performed dynamic analyses that concluded that the rack modules would r.ot develop enough kenetic energy during a SSE te damage the spent fuel pool liner c,r tNr rack modules themselves.

Furthermore, analysis performed by the licensen also found that there will be no rack-to-rack, or rack to-pool wall impact during a SSr.

Racks C1 and C2 have core inoving space than the other three rack';.

For those two racks, pottutial for e tip over is t.reater than for translatienai movement.

The licensee's calculation, based on the DYHARACK code, indicated thc.t such a possibility does rot exist. However the code was benchmarked to an incomplete verificationprocessandthetheoretlcalaspectofthehighlynoniinearcalculation has not Leen doturc7ted adequately to address the staff concerns on potential numerical error and instability. The staff, for this reason, did not :.olely rely upon the results of ti,e DYNARACK analysis to make a safety evaluation of the racks. The licensee indicated that their vendor, Heltec, has performed an experiment that demnstrated that rack-to-wall impact is unliioly. However, the licensee has r.ot documented this exoeriment formally and, consequently, the staff has not reviewed the basis of the experimental findings. Therefore, the stai'f rnade the following independent assessment to suoplement licensee's calculation and design idequacy conclusion.

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9 Based on the geometry and aspect ratio of the racks, the f*C stof f has determined that, were one to tip over, it would be more likely to occur in the east-west direction that the north-south direction.

Simplified assessment of the rack lift-off potential, based on rack overturning stability considerations in conjunction with a conservative use of the floor response spectrum at elevation 236 feet, indicates that there is a possibility of one side of the supporting legs to lift off the pool floor. The acceleration r.eeded to lift the rack in the eest-west direction was found to be approximate'y 0.29 (hori:.ontal) when a cortined excitachn by the full horizontal acceleratian ar.d two-thirds of the horizontal accelerotion as the vertical input was assumad to act on the rack simultaneoucly, and no resistance from water against the lifting was considered. The staff also evaluated the actual safety margin against overturning of the rack.

It was found that the conservatively assessed safety f actor against nyerturning the rack is approximately 1.1, which is consistent with the provision of SRP Sectica 3.B.5, and, therefore, is acceptable. This evaluation was based on the conrervation of energy principle whereby the kinetic energy resulting from the maximum velocity of the rack induced by an earthquake is equated to the potential ener)y that is needed to raise the rack to position where the center of gravity of the rack moves beycad a line cent.ecting the two supporting legs of the rack.

Pack stresses due to the horizontal inertial force corresponding to the SSE were also found to be small and acceptable.

H(wever, for increased safety margin, the NRC staff requires that racks C1 and C2 be either kept empty or loaded in such a way that the center of gravity of a partially loaded rack be maintained at least a 0: stance of one half the rack width in the east-west direction away from the rack boundary which is closest to the pool wall (i.e., minimize the poteritial for rack tip toward the east poolwall).

Fuel handling eouipment, specifically a channel storage rack and fuel preparation machines, occupies part of the space within the distance t,etween the racks and the pool wall. Based on the proximity and dimensions of this equipment, the !$C staff has concluded that in the unlikely event cf a rack tip-over during an earthquake and subsequent impact with this equipment, the rack wculd not develop sufficient kinetic energy to damage itself to such an extent that the basic fuel assembly integrity would be ecmpromised.

Finally, the FitrPatrick plant is located in a low-seismic-activity zone.

The imC staff believes that ground motion capable of le3M ng to significant dynamic excitation of the rack is highly unlikely.

2.0.2 Conclusion Based on tbn review and evaluation of the licensee's submittn's, and the staff's indep* indent assessment, it is concluded that the spent fuel rack modules and the spent fuel pool are capable of withstanding the abnortaal loading associated with a SSE in combination with other applicable loads.

Furthemare, the design of the spent fuel : nodules and the spent fuel pool are l

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' in conformance with the applicable acceptance criteria established in the Standard Review Plan and are consistent with the current licensing practice.

Therefore, the NRC staff concludes that the structural aspects of the additional racks are acceptable.

Maintenance of uniform gaps between the racks and between the racks and pool well is desirable from a structural point of view since it minimizes a potential for impact. Therefore, the staff requires the licensee to institute a surveillance program that inspects and maintains rack gaps after an earthquake equivalent to or larger than an Operating Basis Earthquake (OBE), if any occurs. The surveillance should also include inspection of rack and fuel intt.grity for any darace.

In addition, she st6ff requires that racks C1 and C2 be either kept g

empty or loaded in such a way that the center of gravity of a partially loaded

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rack be maintained at least a distance of one half the rack width in the east-west direction away from the rack ooundary which is closer to the pool wall.

2.E RADIATION PROTECTION AND ALARA CONSIDFRATIONS 2.E.1 OcculaponalExposureControls The Spent Fuel Pool rack addition sill f all under the responsibility of the FitzPatrick Rediological and Environmental Services Department and will require pre-job briefings, man-rem estimates, and exposure tracking.

Radiation, coniamination, and airborne surveys will be )erformed prior to any work in the pool and radiological conditions along wit 1 protective clothing requirements will be stated on the applicable Radiation Work Permits.

Storing additional spent tuel in the pool will increase the anount of corrosion and fission product radionuclides introduced into the pool water.

Specifically, activated corrosion products such as Co-58, Co-60, Fe-59, and Mn-54 may be released to the pool from the surface of the spent fuel assemblies and fission products such as Cs-134, Cs-137, Sr-89, and Sr-90 may be released to the pool water through defects in the spent fuel cladding.

However, the additional activity introduced to the fuel pool from the increase in stored fuel assemblies should not increase radiation dose rates above the fuel pool.

Furthermore, the spent fuel stored in the new racks will be shielded by approximately 24 feet of water resulting in negligible dose rates above the fuel pool.

fhe collective occupational dose for the proposed modification of the SFP is estimated by the licensee to be about 2 person-rem.

Based on previous experience with related activities at similar facilities, the staff believes that the licensee's estimate is low aho thai. 011ective doses for these activities will more likely fall in the range af 4-6 person-rem.

- 11 The licensee has indicated that the removal of irradiated material currently stored in the spent fuel rool where the additional r:cks will be installed is estimated to require collective doses c' about 13.5 p rson-rem. The licensee has further stated that this U.,5 pcrson-rem is not directly attributed to the new rack installation.

Even if this exposure were included in its entirety, and '.hc staff value of 4-6 person-rem were used to estimate occupational iadiation exposures for the rack installation, the total additional collective dose of 17.5-19.5 person-rem is a : mall fraction of the 1987-1989 average annual occupational dose for FitzPatrick. This small increase in co'.lective radiation dose should not affect the licensee's ability to maintain individual occupational duses within the limits of 10 CfR part 20, and is os low as is b

reasonably achievable. Normal radiation control procedures shcald preciude any significant occupational oposures.

Based on present and projected operations in the SFP area, we est%te that the proposed expansion of the SFp should sdd less than 3% to the total annuel oct.upational rediation dose at the facility, based on the average collective dose reported by the licensee for the 1987-1989 period.

Therefore, we conclude that the proposed storage of additional fuel in the modified SFP will not result in any significant increase in doses received by workers.

?.F.? Accident Analyses ses The staff, in the Safety Evaluation Report issued t' arch 4,1970, addressed the safety and environmental aspects of a fuel handling accident. A fuel handling accident may be viewed as a " easonably foreseceble" design basis event which the pool and its associated structures, systems, and components (includiag the racks) are designed and constructed to prevent. The envirenuental impacts of the accident were found not to tre significant.

The staff has reviewed the accidental fission product releases that cotid occur at Fitzpatrick in conjunction with the proposed expansion of the spent fuel stcrage capacity. The staff finds that neither the reracking operations nor the increased capacity of spent fuel storage resulting from the proposad modification affect the calculated consequences of postulated accidents.

Likewise, the proposed rack addition does not create the possibility of new type of accident not previously analyzed.

The radiological ansequen..s resulting from postulated accidents have been previously analyzec and found acceptat,le as specified in the applicatie regulation at 10 CFR Part 100.

3.0 HME, CONSULTATION In-accordance with the Commission's regulations, the New York State officia'l was. notified of the proposed issuence of the amendment. The State official had 1

no comments.

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A 4.0 THVIRONPCHTAL CON 5IDIP.ATION t ursuant to 10 CFR 51.21, 51.32 and 51.35, en environwntal assessinent and finding ot r.o significant itapact have beon prepred and published (56 FP 66460) in the Fnderal Register on Occesaber 23, 1991.

Based upon the environnental assestr@ht, the'Conrassion has deterinined that the istuonce of this afnendmert will rot have a significant of fect on the quality of the human environment.

5.0 00HClUS10E l

The Concitsion published a Hotice of Consideration of Issuance of !xendtnent to r

Fit 1ity Operation license and Opportunity for Maring in the rtdtral.i 4

Re ister 1

on July 24,1990{56rR30051). No requests for hearing vere rece N ar:d U I ittte of New Yort did not hase any coments.

i Pa have coorluded therei'reasonableassurancethatthehealthandsafetbased on the considerations discussed a not be endangered by operaticn in the aroposed manner, y of the public willand(2)suchactivities will be conducted in compliance with t1e Contissio1's regulations and the iss;,..cc of'this arr.endrnent will not be inimical to the coman defanse and seu ity or to the health and safety of the public.

r Principal Contributor:

11. Abelson K. Eccleston S. Kim X. Parcrewski A. Dimrer Date:

!beeniter 31, 1991 s

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S ph' A f *,.l December 31, 1991 5

Docket No. 50 333 in 4 DISTRIBUTION:

See aHHhid sheet Mr. Ralph E. Beedle

/,bh Executive Vice President. Nu'. Mar Generation Dower Authority of the State t 1 New York Q

83 Main Street ff

,hite Plains, New York 10601

SUBJECT:

ISSUANCE OF AMENDMENT FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (TAC NO. M76937)

The Comissica has iseued the enclosed Amendment No 175 to facility Operating License No. DPR.59 for the James A. FitzPatrick Nuclear Power Plant. The amendment corsists of changes to the Technical Specifications (TS) in response to your application tiansmitted by letter deed May 31, 1990, and supplementcd by

'4 letters datea October 31, 1990, December 5,1990, June 26,1991, July 12,1991, July 16, 1991, and September 19, 1991.

Tl e amer

'nt revises the Technical Specifications to allow for the expansion of the spunt fuel pool storage capacity fron the current 2244 fuel assemblies to the proposed 2797 fuel assemblies.

As previously discussed with your staff, during the implementation of this amendment, the NRC staff expects you to adhere to the surveillance and loading re uirements specified on pages 4 and 10 of the enclosed Safety Evaluatior. Any deviation from these requiretients must be rey wwed by and have prior approval of the NRL staff.

/. copy of the re

.d Safety Evaluation is enclosed. A Notice of Issuance will be included in the: Tommission's nn t regular biwee Q Federal Register notice.

i Sincerely.

~

OrdGINA1 DON 2 gy Brian C. McCabe,, Project Manager Project Directora+.e I-1 Divisior of Reactor PrMecu. I/II Office of Nuclear Ret.c hr Regulation i

Enclostices:

1, Amendment No.175 to DPR.59 2.

Safety Evaluation cc w/ enclosures:

See next page

• See previous concurrence

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