ML20004F980

Safety Evaluation Supporting Amend 55 to License DPR-59

ML20004F980
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Site:	FitzPatrick
Issue date:	06/18/1981
From:	Office of Nuclear Reactor Regulation
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ML20004F979	List: ML20004F978 ML20004F980 ML20004F983
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NUDOCS 8106260392
Download: ML20004F980 (10)

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t-UNITED STATES J"t.,

NUCLEAR REGULATORY COMMISSION a

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WASHINGTON, D. C. 20665

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 55 TO FACILITY OPERATING LICENSE NO. DPR-59 THE POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 l

1.0 Introduction By letter dated July 26, 1978, as supplemented (a) the Power Authority of the State of New York (the licensee or PASNY) proposed to change the spent fuel pool (SFP) storage design for James A. FitzPatrick Nuclear Power Plant (FitzPatrick) from the design which was reviewed and approved in the operating license' review and described in the FSAR. The proposed change consists of increasing the total spent fuel storage capacit9_of the SFP from 760 fuel assemblies to 2244 fuel assemblies.

2.0 Discussion

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The proposed spent fuel storage racks are to be made up of alternating, double-walled aluminum containers. These will be about 14 feet long and i

will have a square cross section with an inner dimension of 6.16 inches.

The nominal pitch between fuel assemblies is 6.625 inches. The outer dimension of the square fuel assemblies that are to be stored in these racks is 5.12 inches. This results in an overall fuel region volume fraction of 0.60 in the nominal ste age lattice cell. A Boral plate is i

to be seal welded in the cavity between the double walls. Thus, in this arrangement there will be only one Boral plate between adjacent fuel assemblies.

In its submittal PASHY states that the minimum amount of i

boron-ten per unit area of Boral plate will be 0.0232 grams per square centimeter. This is equivalent to 1.4 x 1021 boron-ten atoms per square centimeter.

3.0 Description of the Proposal The proposed SFP modification consists of replacing the exis+1ng fuel storage racks with new spent fuel racks to increase the storage capacity from 760 l

to 2244 fuel assemblies. The fuel assemblies are stored in anodized aluminum modules. The modules are interconnected in a group to minimize relative l

displacement and prevent impact. Each module is arranged in a 8 X 10, 8 X 8, or an 11 X 10 array. The fuel assemblies are inserted into cavities that are fomed by a c' ter of cans that are arranged in a checker board (a) Supplemental letters dated May 15, June 22, September 25, October 10, and November 29,1979, April 1, April 31, and October 31, 1980.

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s pattern. The can provides separatfor, and lateral restraint for each fuel Boral (B C) poison materisi.is sealed in cavities within each assembly.

4 can by welding. The cans are constrained by upper and lower castings that are bolted to plates along the perimeter to form a box structure. The lower casting vertically supports each fuel assembly. Each module i's free-standing with no lateral restraints to the wall and is supported by four steel feet that transfer load to the pool floor..The lateral. loads on the racks will be transferred by friction between the feet and the. pool floor.

The new spent fuel racks will be installed on a phased basis to provide additional capacity as required during nonnal refueling outages.

Installa-tion has been sequenced to eliminate any interfacing between the existing racks and the new racks. During periods of phased installation, both groups of racks will be seismically supported. At no time will any object be moved over stored spent fuel in accomplishing these procedures.

3.1 Criticality Analyses As stated in PASNY's July 26, 1978 submittal, the fuel pool criticality calculations are based on unirradiated BWR fuel assemblies with no burn-able poison and a fuel loading of 14.8 grams of uraniunt-235 per axial centimeter of fuel assembly.

The Nuclear Associates International Corporation (NAI) performed the criticality analyses for PASNY. NAI made parametric calculations by using the CHEETAH-B computer program to obtain four-group cross sections for PDQ-7 diffusion theory calculations. The effective boron cross sections for the Boral plates were calculated with the CORC-Blade program. NAI stated that these programs have been extensively tested by using them to make benchmark experiment calculations and core physics calculations for

'several existing operating power reacto-s.

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These computer programs were used to calculate the neutron multiplication l-factor for an infinite array of fuel assemblies in the nominal storage lattice at 20*C with the minimum boron concentration in the Boral, i.e.,

l 0.0232 grams of boron-ten per square centimeter. NAI then performed cal-culations to determine:

(1) the highest neutron multiplication factor as a function of pool water temperature; (2) the effect of a possible reduc-tion in the lattice pitch; and (3) the effect of eccentrically positioning fuel assemblies in the storage lattice. These calculations showed that when all of these effects are accounted for, the maximum effective neutron f ) in the fuel pool will be less than 0.894. The multiplication factor (K accuracy of the diffusio8 (heory method for this storage rack application l

was then checked by calculating the nominal reference case with the KENO-IV j

Monte Carlo program using 123 group cross sections from the GAM-THERMOS library, and it was found that the results of the diffusion theory method are accurate within one percent ak.

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o r Since it will. be possible to inadventently place a fuel assembly between the outer periphery of a loaded rack and the walls of the fuel pool, MI calculated the possible increase in the neutron multiplication factor for g

this event.

In this calculation it was assumed that there was no Boral plate between the external assembly and the adjacent assembly in the rack. NAI found that the increase in the neutron multiplic tion factor will be less than 0.005. Thus, this event would increase the maximum j

possible keff to 0.889.

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_In' order to have proof that the required amount of boron remains in the plates throughout the life of the racks, PASNY's proposal states that.

sealed Boral coupons will be provided for inservice survel'iance.

3.1.1 Evaluation The above described results ' compare favorably with the results of para-metric calculations made with other methods for similar fuel pool storage lattices.. By assuming new, unirradiated fuel with no burnable poison or control rods, these calculations yield the maximum neutron multiplication l

factor that could be obtained throughout the life of the fuel assemblies.

This includes the effect of the plutonium which is generated during the fuel cycle.

In addition to the Quality Assurance Program which is described in the prupool, the NRC. requires an onsite neutron attenuation. test to verify with ninety five percent confidence that there are a sufficient number of the Boral plates in the racks so that the maximum keff will not be greater than 0.95.

We find that all factors that could affect the neutron multiplication factor in this pool have been conservatively accounted for and that the maximum neutron multiplication factor in this pool with the proposed l

racks will not exceed 0.95.

This is NRC's acceptance criterion for the j

maximum (worst case) calculated neutron multiplication factor in a SFP.

This 0.95 scceptance criterion is based on the uncertainties associated with tne calculational methods and provides sufficient margins to preclude criticality in the fuel. Accordingly, there is a Technical Specification which limits the effective neutron multiplication factor in all SFPs to 0.95.

3.1.2 Conclusion l

We. find that when any number of the fuel assemblies, which PASNY described in these submittals, which have no more than 14.8 grams of uranium-235 per axial centimeter of fuel assembly, or equivalent, are loaded into the proposed racks, the keff in the fuel pool will be less than the 0.95 limit.

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We also find that in order to preclude the possibility of the keff in the l

fuel pool from exceeding this 0.95 limit without being detected, it is necessary, pending an NRC review, to prohibit the use of these high density storage racks for fuel assemblies that contain more than 14.8 grams of l

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. uran'ium-235, or equivalent, per axial centimeter of fuel assembly. On the basis of the information submitted, and the keff and fuel loading limits stated above we conclude that the health and safety of the public will not be endangered by the use of the proposed racks.

~3.2 Soent Fuel Cooling The licensed thermal power, for FitzPatrick is 2436 MWth. PASNY plans to refuel this plant annually.

In the annual cycle, about 140 of the 560 fuel assemblies in the core are replaced. To calculate the maximum heat loads in the spent fuel. pool PASNY assumed a 150 hour0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> time interval between reactor' shutdown and the time when 140 fuel assemb P

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transferred to the spent fuel pool and a 250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> time interval-oetween reactor shutdown and tre time when 560 fuel assemblies were transferred to;the spent fuel pool.

For the power history prior to shutdown PASNY assumed an energy prod >ction of 27,558 MWD /MTU with a continuous energy density of 24 MW/MTU. k;th these assumptions PASNY used the method given in the NRC Standard Review Pl a 9.2.5 to calculate the maximum possible heat loads for the modified SFP.

For these cooling times and fuel burnups

'PASNY c lculated the marimum' heat load in the spent fuel pool to be about 6

.10.X 10 BTU /hr for the final refueling which fills the pool, and 24. X 10 BTU /hr for.a full core offload which fills the pool after twelve annual refuelings.

The spent fuel pool cooling system consists of two pumps and gwo heat exchangers.

Each pump is designed to pump 525 gpm (2.63 X 10 pounds per' hour). Both heat exchangers when fed by a single SFP cooling pump are designed to transfer 6.3 X 100 BTU /hr from 125* fuel pool water to

.95 F Reactor Building Closed Loop Cooling Water which flows through the shell side of each heat exchanger at the rate of 467 gpm (2.34 X 105 PASNY stated that when a full core is offloaded into pounds per hour).

the spent fuel pool, the Residual Heat Removal (RHR) system will 'Je used to maintain the fuel pool water temperature at or below 135'F.

Makeup water for the SFP is obtained from the seismic Category I Condensate Storage System, which has two 200,000 gallon storage tanks.

3.2.1 Evaluation We find that PASNY's calculated peak heat loads for the modified pool with a storage capacity for 2244 fuel assemblies are conservative and acceptable.

We also find that the maximum incremental heat load that will be added by increasing the number of spent fuel assembifes that are to be stored in this pool from 760 to 2244 will be 2.0 X 10 BTU /hr. This is the differ-ence in peak heat loads for full core offloads that essentially fill the present and the modified pools.

-S-We calculate that with both pumps operating, the spent fuel pool cooling system can maintain the fuel pool outlet water te erature below I?S*F for a peak annual refueling heat load of 10. X 10 BTU /hr. ' We fini that when the RHR system is aligned with the spent fuel pool cooling system, the combined system will have sufficient capacity to keep the SFP oytlet water temperature below 135'F for a full core heat load of 24. X ".00 BTU /hr.

Assuming an initial maximum average fuel pool water temperature of 125 F, the minimum time to achieve bulk bo'.ing after any credible accident will be about nine (9) hours. This assumes no cooling of the SFP dLring that time interval.

In order to preclude actual boiling, it is PASNY's inten-tion to use a single RHR train to cool the SFP when an accident makes i

the normal SFP cooling system inoperable. During normal power operation the RHR system.is unavailable for SFP cooling since both trains of this system must be available for the LPCI. post accident ECCS function.

In order to make the RHR system available, PASNY has instituted procedures which allow prompt reactor cooldown. Cold shutdown conditions are achieved in a timely manner (i.e., less than 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />) at which time the train of the P.HR system can be lined up to the SFP. This will preclude pool boiling.

Although SFP boiling will not occur, to be conservative such an occ'ur-rence has been considered. Assuming the same 125'F SFP water temperature, bulk boiling will likewise occui within nine hours. Such boiling will result in a maximum evaporation rate of fifty gallons per minute. Within this time an equivalent makeup rate can be established from the condensate storage system.

In addition, we also find that under bulk boiling con-ditions the fuel temperature will not exceed 350*F.

In sum, we find that makeup capability will preclude fuel element uncovery and that the maximum temperature does not unacceptably effect fuel el? ment integrity or surface corrosion.

3.2.2 Conclusion We find that the present cooling capacity in the FitzPatrick SFP will be sufficient to handle the incremental heat load that will be added by i

the proposed modification. We also find that this incremental heat load will not alter the safety consideratiuns of SFP cooling from that which l

we previously reviewed and found to be acceptable. We conclude that there is reasonable assurance that the health and safety of the public will not be endangered by the use of the proposed design.

'3.3 Installation of Racks and Fuel Handling About 429 out of the 760 storage spaces that are presently in the SFP are filled with spent fuel assemblies. Thus, about 45 percent of the pool will not have fuel assemblies in it at the time PASNY is proposing to l

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6-change the racks.

In this regard PASNY states that a portion of the south end of the pool has already been cleared of existing racks. PASNY also states that vacant racks in the south end of SFP will be cleared prior to installation of the new racks, i.e., additional racks will be removed.. PASNY also states that during the installation of the new racks:

administrative controls will be placed on the reactor building crane to insure that the racks cannot be lifted or carried over spent fuel.

3.3.1 Evaluation Since about forty five percent of the pool will not have fuel assemblies in it wnen the-racks are changed, PASNY should have no difficulty in keep-ing the racks that are being moved away-from the spent fuel that is presently in the pool.

After the racks are installed in the pool, the fuel handling procedures in, and around, the pool will be the same as those procedures that were in effect prior to the proposed modifications.

3.3.2 Conclusion We conclude that there is reasonable assurance that the health and safety of the public will not be endangered by the installation and use'of the proposed racks.

3.4 Evaluations 3.4.1

' Spent Fuel Handling Evaluation The NRC staff has underway a generic review of load handling operations in the vicinity of SFPs to determine the likelihood of a heavy load impact-ing fuel in the pool and, if necessary, the radiological con:equences of such an event. We have concluded that the likelihood of a heavy load handling accident is sufficiently small that the proposed modification is acceptable and no additional restrictions on load handling operations in the vicinity of the SFP are necessary while our review is underway The consequences of fuel handling accidents in the SFD 3pg not changed from those presented in the Safety Evaluation (SE) dated November 1972.

3.4.2 Structural and Mechanical Evaluation The design and fabrication of the racks are in accordance with " Aluminum Con-struction Manual - Second Edition, Nov.1971, Specifications of Aluminum Structures"; " Aluminum Standards and Data - Aluminum Association, 5th Edition, Jan. 197C" "S+. eel Construction Manual AISC (7th Edition), June 1973, American i

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1 Institute of Steel Construction"; and "ASME Boiler f. Pressure Vessel Code, l

Section III Subsection NA, Appendix I and XVII, 1974 Edition *. The loads, load combinations and acceptance criteria used for the rack design are con-sistent with Sections 3.8.4.II.3 and 3.8.4.II.5 of the Standard Review Plan for steel structures. The materials and fabrication processes are essentially the same as those used at the Yankee Nuclear Power Station, which has performed sati.sf actorily for over.10 years.

The seismic design of the racks is based upon a nonlinear dynamic analysis using the ANSYS computer program that was developed by Swanson Analysis Systems, Inc.

Seismic excitation along three orthogonal directions was used in the design.

Floor acceleration time histories corresponding to SSE and OBE group acceleration lev'els of 0.15g and 0;08g were imposed..The analysis includes the effects of friction between the rW;k ali floor, gaps between the' fuel assembly and can, rack l

uplift, and fluid coupling due to the constrained water within the rack structure. No benefit was taken for the damping effect of water surrounding the rack. A low value of friction was used to maximize the predcted sliding displacement, while a high value of friction was used to maximize horizontal rocking displacement at the top of the rack. Under the most severe loading conditiens the racks slide a maximum of 1.472 inches. A distance of 3.0 inches will be maintained between the rack and any rigid object within the pool. In l

addition, a distance of 6.M in~ches will tie maintained between any rack and the poo.1 Walls. Fuel assemblies were conservatively assum'ed to impact with the cans all at the same time. The integrity of the fuel ~ cladding will be maintained under these conditions. Furthermore, rack to rack impact was considered although the racks are constrained to mi. nim pirelative motion. The worst case 6f two fully loaded racks impacting undfr a loading of highest sliding was evaluated assuming that all momentum is transferred from one rack to the other'. The maximum rack-to-rack impact forces have beenscalculated to be 81,000 pounds for the SSE and 64,000 pounds for the UBE. The forces occur at the top grid only and are included in the stress analysis of this member. Since the upper fitting of the fuel assemblies is not attached to the top grid, these impact loads are not directly transmitted to the fuel assemblies.

All three components of earthquake have been conservatively considered in the rack design. As explained in the licensee's submittal, the time history analysis l

was done for only two components of earthquakes which were the ma-imum horizontal (X - direction and Y - direction). However, the forces computed from this planar time-history model were applied on the detail (3-0) model simultaneously in both the X-Y and Z-Y planes. These resultant loads were then combined by SRSS to ob-tain the overall loads. This method, in effectively considering all three com-ponents of earthquake, doubles up on the vertical (Y-direction laading).

The new racks have been analyzed to determine the effects of a dropped fuel In assembly impacting at critical locations on the upper and. lower castings.

all cases there is no change to the center to center spacing of the fuel and there is no dislocation of the Boral neutron absorbing material.

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i 8-3.4.2' Conclusion The licensee performed a review of the load carrying ability of the SFP f

structure and found that the existing structure is capable of supporting the increase in overall loading as a result of the proposed fuel pool modification. The steel liner and concrete floor slab were also evaluated for the effects of rack impact due to rocking to conform to the bearing stress and punching shear stress allowables of the American Institute of Steel Construction Specification for Steel Structures and the American Concrete Institute Building Code Requirements for Reinforced Concrete (Ai 318-71). The temperature limits established in the FSAR for the poA remain the same and therefore the effects of temperature gradients 1

on the pool structure will remain unchanged.

3.5 Occupational Radiation Exposure We have reviewed the licensee's plan for the removal and disposal of the low density racks and the installation of the high density racks with respect to occupational radiation exposure. The occupational ex-posure for this operation is estimated by the licensee to be about 6 man-rem. We consider this to be a reasonable estimate because it is based on the licensee's detailed breakdown of occupational exposure for

. The licensee considered the number c/

- each phase of the modification.

individuals performing a specific job, their occupancy time while per-forming this job, and the average dose rate in the area where the job.

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was being perfo_rmed. To ensure that the modification will be perforced in a manner consistent with as low as is reasonably achievable (ALARA) occuoationel exposure, the licensee will remove unnecessary radioactive equipment and material from the fuel pool prior to the installation work, will store the spent fuel at the opposite end of the pool from where the installa-tion work is being performed, and will pre-plan procedures necessary for the removal of the old racks and installation of the new ones. The exist-ing low density racks will be decontaminated upon removal from the pool, packaged and shipped intact to a disposal site as low concentration ra-dioa:tive waste.

We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies on the basis of in-formation supplied by the licensee for dose rates in the spent fuel area from radionuclide. concentrations in the SFP water and deposited on the SFP walls. The spent fuel assemblies themselves will contribute a negligible

-amount to dese rates 'in the pool area because of the depth of water shield-ing the fuel. The occupational radiation exposure resulting from the.addi-tional spent fuel in the pool represents a negligible burden. Based on present and projected operations in the spent fuel pool area, we estimate modification should add less than one percert to the that the proposed total annual occupational radition exposure burden at this facility. The small f.4rease in additional exposure will not affect the licensee's abil-ity to maintain individual occupational doses to as low as is reasonably L

Thus, we conclude that achievable and within-the limits of 10 CFR Part 20.

storing additional fuel in the SFP will not result in any significant in-crease in doses received by occupational workers.

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'3.6:

Radioactive Waste Treatment The plant contains waste treatment systems designed to collect.and process the gaseous, liquid and. solid wastes that might contain radioactive

- material. The waste treatment systems were evaluated in the Safety Evaluation dated' November 1972. Th'ere will be no change in the waste-treatment-system or.in the conclusion given in Section 8.2 of the

~ ' evaluation of this system because of the proposed modification.

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- Materials f

.The purity of SFP water is maintained by a combination of filtering and ion exchange process to assure compatibility with the aluminum spent fuel racks. Significant corrosion of the. rack structure or nuclear fuel components is highly unlikely to occur. The consequences of a S al storage cavity weld leak have been evaluated and found to be % igible.

However, a' vacuum and pressure test-is performed to assure, e integrity E

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Evaluation-

'The criteria used in t% analyses, design and construction of4 the'new spent.

fuel racks to account. for anticipated loadings and postulated conditions.

that,may.be imposed upon the structure during their service lifetime are I

in conformance with established criteria, codes, standards, and specifica-

tions ' acceptable-to the NRC staff.

The use of these criteria provide reamnable assurance that the new fuel pool structures will with '

? stand the s:ecified design conditions without impainnent of structural j

integrity or the perfomance of required safety functions.

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'3.7.2, SunMry.

Our evaluation supports the corclusion that the proposed modification to the FitzPatrick SFP is acceptable because:

(1)' The increase in occupational radiation exposure to individuals c a to the storage of additional fuel in the SFP woul<1 be negligible.

(2) The potential consequences of the postulated design basis accident for the SFP, i.e., the rupture of the fuel pins in the equivalent of one fuel. assembly and the subsequent release of the radioactive i

inventory within the gap, are acceptable l

i (3) - The likelihood of an accident involving heavy loads in the vicinity.

c of the SFP is sufficiently small that no additional restrictions on loc.d movement are necessary while our generic review 'of the issues is underway.

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, 4.0 Technical Specification As indicated in the criticality analysis of this Safety Evaluation and in the licensee's referenced submittals the maximum average Uranium-235 content is specified in Technical Specification 5.5 to be 3.3 w/o.

Therefore, fuel, assemblies that are bound by the fuel assembly designs described in the licenseet referenced submittals may be stored in the

-spent fuel pool. This will result in satisfying the facility design criteria of keff' (dry) <0.90 and (flooded) <0.95.

As indicated in the spent fuel pool cooling L alyses of this Safety Eval-uation, the RHR system may be used to augment the spent fuel pool cooling system. Since such RHR usage would make the LPCI system unavailable, cross'-tieing the RHR and SFP cooling system is allowable only during plant shutdowns. Therefore, in accordance with Technical Specification 3.5 the RHR system may be used for SFP cooling when reactor coolant temperature is below 212 F.

5.0 Conclusion We have concluded, based on the considerations discussed above, that:

(1) there /

reasonable assurance that the health and safety of the public wil. not be endangered by operation in the proposed man ~ner, and

.(2) such activities wil'. be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Dated:

June 18, 1981 i

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