IR 07100005/2012017

Staff Exhibit S-1,consisting of 841121 Safety Evaluation Supporting Amends 111 & 105 to Licenses DPR-31 & DPR-41, Respectively.Technical Evaluation Rept TER-C5506-529 & Cover to Transcript of 871005-1217 Meetings in Concord,Nh Encl

ML20149F281
Person / Time
Site:	Seabrook, 07100005
Issue date:	12/01/1987
From:	Office of Nuclear Reactor Regulation
To:
References
OL-S-001, OL-S-1, NUDOCS 8802120151
Download: ML20149F281 (55)

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UNITED STATES

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[ ~ , ",., ' -t,) NUCLEAR REGULATORY COMMISSION W ASHING TON, D. C. 20$55 Uh..c go .h; . IIf f

%.[.# '83 FEB -2 A8 :44 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

~ p J HCRtlhY RELATED TO AMENDHENT NO.111 TO FACILITY OPERATING LICENSM.M-ypylCE AND AMENDMENT NO. 105 TO FACILITY OPERATING LICENSE NO. OPR-41 FLORIDA POWER AND LIGHT COMPANY

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TURKEY POINT UNIT NOS. 3 AND 4

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DOCKET NOS. 50-250 AND 50-251  :

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1.0 Introduction By letter dated March 14, 1984 and supplemented on July 2 and 23, August 14 and 22, September 10 and 28, October 5, 9,18 and 26, and November 16, 198 Florida Power and Light Company (FP&L) submitted an application to increase the storage capacity of the spent fuel pools (SFPs) for Turkey Point, Units 3 and 4, by replacing the existing racks with new storage racks. /vnendment 20 to Facility Operating License DPR-31, dated September 24, 1976, temporarily allowed the storage capacity of the Unit 3 SFP to be increased from 217 to 235 fuel assemblies. Amendment Nos. 23 and 22 for Units 3 and 4, respectively, dated March 17, 1977, increased the SFP storage capacity at each facility to 621 fuel assemblie .1 Discussion

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These proposed amendments will allow the licensee to expand the SFPs from the current capacity of 621 fuel assemblies to 1404 fuel assemblies. This expansion will be accomplished by reracking the existing SFPs with neutron absorbing (poison) spent fuel racks composed of individual cells made of stainless steel. The new spent fuel storage racks will be arranged in two discrete regions within each pool. Region 1 will consist of 286 locations which will normally be used for storage of spent fuel with an enrichment equal to or less than 4.5% U-235 at it's most reactive point in lif Region 2 will consist of 1118 locations and will provide storage for spent fuel assemblies meeting required burnup consideration =

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The existing fuel storage racks (621 locations) have a nominal CQ .g centerline-to-centerline spacing of 13.7 inches. The new Region 1 racks 2 "

will have a 10.6 inch centerline-to-centerline spacing and Region 2 will be --;

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9.0 inches centerline-to-centerline spacing. The major components of the ?. I 23 fuel rack assemblies are the fuel assembly cell, boraflex (neutron , L1 JT absorbing) material and the wrapper. The wrapper covers the Boraflex l

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h material and provides venting of the Boraflex to the pool environmen ??

The existing racks have 636 total storage cells; however due to piping abd  ;

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other interferences the Unit 3 racks have 621 usable cells and the Unit t

"," J.9, %;p racks have 614 usable cells. In the 1986-1987 time frame, the units wilt lose their full-core reserve storage capacity (157 assemblies) and in 1 p: a i

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-2-1990-1991 time frame they will no longer have the capacity to store fuel discharged from the operating units. Since these dates are earlier than the date a federal depository should be available for spent fuel (1998),**

additional capacity for the storage of spent fuel is neede Increasing the SFPs capacity to 1404 cells, as proposed, will allow plant operation with full core reserve in the SFPs to about the year 2005 for Unit 4 and 2006 for Unit 3. These time frames are based on the present FP&L fuel management. The proposed expansion of the SFP storage racks to 1404 cells should be adequate until the federal government begins accepting spent fuel from civilian power reactor ,

2.0 Evaluation

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The "Spent Fuel Storage Facility Modification Safety Analysis Report" provided by the licensee on March 14, 1984, in support of this application for amendments was the basis for the NRC staff evaluation. Supplemental information provided by the licensee is also reflected in the following Safety Evaluation which sumarizes the NRC staff effor .1 Criticality Considerations Each pool will contain racks that provide 1404 designated locations for the storage of reactor fuel. The storage racks will be divided between two regions - one containing 286 locations and one containing 111 The smaller region, having sufficient capacity for approximately 1 1/2 full cores, will be used for the storage of fresh fuel and fuel not suitable for Region The larger region will normally be restricted to fuel having a specified minimum burnup. The licensee proposed that, during installation of the new racks, storage of high reactivity spent fuel (up to fresh 4.5 percent enrichment).be permitted in a checkerboard array with every other location empty. Administrative controls will be used to prevent storage in the empty location The , Region 1 racks will consist of stainless steel cans of 8.75 inch square

--interior dimension and 0.75 inch wall thickness. On the outer surface of each side of the cans Boraflex sheets having a minimum area density of 0.02 grams per square centimeter of B-10 are held in place by a thin-walled stainless steel wrapper pl6te. The rack structure maintains these cans on a 10.6 inch center-to-center spacin The Region 2 rack design consists of stainless steel cans welded together to form a honeycomb type structure. The cans have an interior square dimension of 8.80 inches and are made of stainless stee ** Nuclear Waste Policy Act of 1982, Section 302(a)(5)

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All four sides of interior cans have Boraflex sheets containing 0.012 grams of B-10 per square centimeter of surface area that are held in place by a stainless steel wrapper which is spot welded to the can. The resulting structure maintains the stored fuel assemblies at a center-to-center spacing of 9.0 inche .1.1 Calculation Methods The calculation of the effective multiplication factor, K , for Region 1 makesuseoftheAMPXsystemofcodesforneutroncross-s@Nionpreparation ,

and the Monte-Carlo Code KENO-IV for reactivit This code set has been'

verified against a set of 27 critical experiments that simulate various features of the rack design. A calculational method bias of zero and .

uncertainty of 0.013 based on a 95 percent probability at the 95 percent confidence level (95/95) was inferred from these comparison The calculation of the criterion for acceptable bu'rnup for storage in Region 2 makes use of the concept of reactivity equivalence. Since the KENO-IV code cannot handle burned fuel assemblies it is necessary to obtain the fresh fuel assembly enrichment which yields the same pool K as the burned assembly. Because of the presence of the poison in the Reg b 2 racks, a multigroup transport theory code is more appropriate than diffusion theory for this calculation. The PH0ENIX code was use The calculation proceeds as follows: An end-point of 39.0 GWD/MT burnup for a bundle having an initial enrichment of 4.5 weight percent U-235 is chose . PFj0ENIX is used to calculate the K ,, of such an assembly in the rack geometry (including can and Boraflex absorber). The burnup required to produce the same 4 is calculated for a number of smaller enrichment , The enrichment required to produce the same b without burnup is obtained (in the present case the value is 1.5 weight percent .

U-235). KENO-IV is used to calculate the rack multiplication factor for the 1.5 weight percent enrichment assembl The advantage of this procedure is that only relative multiplication factors are computed by PH0ENIX. The final value of the rack multiplication factor is obtained from the more powerful KENO-IV cod .1.2 Treatment of Uncertainties For the Region 1 analysis the total uncertainty is the statistical combination of the method uncertainty, the uncertainty in the particular KEN 0 calculation, and mechanical uncertainties due to tolerances, spacing, etc. The mechanical uncertainties were treated either by making worst case I

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l assumptions (e.g., using thr, minimum rather than nominal value of the boron ,

loading) or by performing sensitivity studies and obtaining a value of the i uncertainty in rack multiplication factor due to uncertainty in dimensions, et In the Region 2 analysis the same uncertainties are considered along with others that are unique to the rack design and usage. These include uncertainty due to particle self-shielding in the boron (actually bias),

uncertainty in the plutonium reactivity and uncertainty in the reactivity as a function of burnup. Including both the plutonium and burnup reactivity uncertainties is conservative since the latter includes the former as one of ,

its component The PH0ENIX code was qualified for burnup calculations by comparing calculated isotopic ratios to measurements made in Yankee-Rowe Core 5, and by comparison of equivalent reactivity burnup between PH0ENIX and the LEOPARD / TURTLE code A set of 81 critical experiments was analyzed to qualify the code for zero burnup conditions. Conservative uncertainties of 5 percent of the reactivity change due to burnup have beea assigned to these parameter .1.3 Results of Analysis Nomal Storage For Region 1, the rack multiplication factor is calculated to be 0.9403, including uncertainties at least at the 95/95 level, when fuel having an enrichment of 4.5 weight percent U-235 is stored therein. . Fuel of either the Westinghouse 15X15 standard or 0FA design may be stored as well as Combustion Engineering 14X14 or 16X16 and Exxon 14X14 designs. Pure water at 1.0 grams per cubic centimeter is assume For Region 2, the rack multiplication factor is 0.9304 for the most reactive irradiated fuel permitted to be stored in the racks, i.e. , fuel with the minimum burnup permitted for each initial enrichment, including at least 95/95 uncertainties, for fresh fuel (4.5 percent enrichment) stored in a checkerboard array in the racks, the effective multiplication factor is 0.8342. Calculation of the remaining uncertainties was not deemed necessary in this case since assuming conservative values for these terms would still for the checkerboard configuration well below the result requiredin0.9 a finalAll K'biculations are obtained for pure water at a density of 1.0 grams per cubic centimeter. Burned fuel of the same designs as allowed in Region 1 may be stored in Region 2. Analyses were perfomed for all allowable fuel types and the proposed curve of burnup versus initial enrichment bounds the results of the calculatio Abnormal Storage Conditions Most abnormal storage conditions will not result in an increase in K of

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the racks. For example, loss of a cooling system will result in an brease in pool temperature but this causes a decrease in the Keff valu .._ - _ ,

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-5-It is possible to postulate events (e.g., a seismic event) which could lead to an increase in pool reactivit However for such events credit may be taken for the approximately 1950 ppm of boron in the pool water. The reduction in the K value caused by the boron (approximately 0.25) more thanoffsetstherNtivityadditioncausedbycredibleaccident .1.4 Sumary of Evaluation, The following discussion suninarizes our evaluation of the proposed re-racking of the Turkey Point SFP #

We have reviewed the assumptions made in the performance of the criticality analyses. These include use of the highest permitted reactivity bundle, pure water moderator at a density of 1.0 gram per cubic centimeter, and an .

infinite array of assemblies. These are consistent with NRC guidelines and are acceptabl We have reviewed the uncertainties which have been include For Region 1, these include variation in poison pocket thickness, stainless steel thickness, cell interior dimensions, center-to-center spacing, boren particle self shielding, and cell bowing. Other parameters, such as boron loading, are taken at their most conservative limits. For Region 2, additional uncertainties due to burnup calculations and calculations of plutonium worth are included. For both regions, calculational uncertainties and biases are included. These uncertainties meet our requirements and are acceptabl We have reviewed the verification of the calculation methods. The KENO-IV code is widely used in the industry for the purpose of calculating fuel rack cri ticalit The set of benchmark critical experiments used to verify the calculations method encompasses the enrichment, separation distance and separating material used in the rack The set of experiments used to verify the PH0ENIX code for the reactivity equivalence calculations is adequate and encompassed the pellet size and enrichment of the fuel proposed for storage in the Turkey Point racks. The uncertainties in the burnup and plutonium worth are verified against Yankee Core 5 isotopics and comparisons with the Westinghouse design LEOPARD / TURTLE code package. We find that adequate verification of the codes used in the criticality analyses has been performe The technique of using reactivity equivalencing to define the storage criterion (burnup as a function of initial enrichment) is, in some form, in widespread use in the industry and is acceptabl For Region 1 racks we have compared the results of the Turkey Point calculation to a generic study and found them to be compatible. Finally the results of the calculation for Region 1 and 2 meet our acceptance criterion of less than or equal to 0.95 including all uncertainties at the 95/95 leve I

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-6-We have reviewed the proposed Technical Specifications 3.17, B3.17, and and find that they are consistent with the assumptions in the safety analysis and are acceptabl .15. Conclusions Based on our review, which is described above, we find the criticality aspects of the design of the spent fuel racks to be acceptable. We conclude that fresh Westinghouse 15X15 fuel of either the standard or 0FA design as

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well as Combustion Engineering 14X14 or 16X16 and Exxon 14X14 designs may be safely stored in Region 1 so long as enrichment does not exceed 4.5 w/o '

U-235. We further conclude that any of these fuel types may be stored in Region 2 provided it meets the burnup and enrichment limits specified in Table 3.17-1 of the Turkey Point Units 3 and 4 Technical Specification During the installation of the new racks, fuel which does not meet this criterion may be stored in Region 2 provided it is stored in a checkerboard arrangement with every other location vacan .2 Materials The safety function of the SFP and storage rack system is to maintain the spent fuel assemblies in a suberitical array during all credible storage conditions. We have reviewed the compatibility and chemical stability of the materials, except the fuel assemblies, wetted by the pool wate The only new material or components to be added during the proposed modification are the nuclear absorber strip The new spent fuel racks to be installed in both regions are constructed entirely of Type 304 stainless steel, except for the nuclear poison materia The existing spent fuel -

liner is constructed of stainless s The high density spent fuel storage racks will utilize Boraflex} ee sheets as a neutron absorbe ,

Boraflex has previously been approved as a neutron absorber and is currently being used in several SFP storage facilities. Boraflex consists of boron

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carbide powder in a rubber-like silicone polymeric matrix. The spent. fuel l

storage rack configuration is composed of individual storage cells

, iMerconnected to form an integral structure. The major components of l the assembly are the fuel assembly cells, the Boraflex material, and the

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stainless steel wrapper around the Borafle The Boraflex absorber will not be sealed within the storage cell and vent paths for any gas generated during exposure will be available to the poo The pool contains oxygen-saturated demineralized water containing boric acid. The water chemistry control of the spent fuel pool has been reviewed elsewhere and found to meet NRC recommendations.

l 2.2.1 Corrosion and Material Compatibility l

l The pool liner, rack lattice structure and fuel storage tubes are stainless

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steel which is compatible with the storage pool environment. In this environmentofoxygen-saturatedboratedwater,thecorrosivedeteriorat{on of the Type 304 stainless steel should not exceed a depth of 6.00 X 10' 2 inches in 100 years, which is negligible relative to the initial thickness .

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Dissimilar metal contact corrosion (galvanic attack) between the stainless I steel of the pool liner, rack lattice structure, fuel storage tubes, and the

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_7 Inconel and the Zircaloy in the spent fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are therefore at similar potentials. The Boraflex is composed of non-metallic materials and therefore will not develop a galvanic potential in contact with the metal components. Boraflex has undergone extensive testing to study the effects of gama irradiation in various environments, and to verify its structural integrity and suitability as a neutron absorbing material. The evaluation tests have shown that the Boraflex is unaffected by the pool water environment and will not be 3 degraded by corrosion. Tests w ge performed at the University of Michigan ,

exposing Boraflex to 1.103 X 10 rads of gama radiation with substantial ,

concurrent neutron flux in borated water. These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of borated water and gamma irradiatio ,

Irradiation will cause some loss of flexibility, but will not lead to break up of the Boraflex.4 Long tem borated water soak tests at high temperatures were also conducted . The tests show that Boraflex withstands a borated water imersion of 240*F for 260 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide. The space which contains the Boraflex is vented to the pool at each storage tube assembl This venting will allow gas generated by the chemical degradation of the silicone polymer binder during heating and irradiation to escape, and will prevent bulging or swelling of the inner stainless steel wrappe The tests have shown that neither irradiation, environment nor Boraflex composition has a discernible effect on the neutron transmission of the Boraflex material. The tests also show that Boraflex does not possess leachable halogens that might be released into the pool environment in the presence of elemental bcron from the Boraflex. Boron carbide of the grade nomally in,the Boraflex will typically contain 0.1 wt.% of soluble boro The test results have confimed the encapsulation function of the silicone '

polymer matrix in preventing the leaching of soluble species from the boron carbid :

To provide added assurance that no unexpected corrosion or degradation of r materials will compromise the integrity of the racks, the licensee has committed to conduct a long term poison coupon surveillance program, which will be representative of the material used in both the Region 1 and Region 2 locations. There will be four sets of coupons, each containing not less than 24 jacketed poison coupons, each set will be designed to be hung on the outside periphery of Region 1 and Region 2 module The initial surveillance of the specimens will be perfomed after approximately five years of exposure to the pool environments. Subsequent surveillances will be based on the ini".ial results to assure acceptable material performance throughout the life of the plan Construction materials will confom to the requirements of the ASME Boiler and Pressure Vessel Code Section II-N I

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-8-2.2.2 Conclusion From our evaluation as discussed above, we conclude that the corrosion that will occur in the SFP environment should be of little significance during the life of the plant. Components in the SFPs are constructed of alloys which have a low differential galvanic potential between them and have a high resistance to general corrosion, localizeo corrosion, and galvanic corrosion. Tests under irradiation and at elevated temperatures in borated water indicate that the boraflex material will not undergo sionificant degradation during the expected service lif ,

We further conclude that the environmental compatibility and stability of the materials used in the expanded SFPs is adequate bi ~ 1 on the test data cited above and actual service experience in operation reactor We have reviewed the licensee's surveillance program and conclude that the monitoring of materials in the SFPs will provide reasonable assurance that the Boraflex material will continue to perform its function for the life of the pools. The materials surveillance program will reveal any instance of deterioration of the Boraflex that might lead to the loss of neutron absorbing power well before significant deterioration will occur. We do not anticipate, however, that such deterioration will occu ,

We, therefore, conclude that the compatibility of the materials and coolant used in the SFPs is adequate based on tests, data, and actual service experience in operating reactors, and the selection of of appropriate materials and adoption of a surveillance program by the licensee meets the requirements of 10 CFR Part 50, Appendix A, Criterion 61, havir? a capability to permit appropriate periodic inspection and testing of comF ents and criterion 62, preventing critica'<ity by maintaining structural integrity of the componehts and boron poison and is, therefore, acceMable 2.2.3 References - Materials J. S. Anderson, "Bnraflex Neutron Shielding Material -- Product Performance Date," Brand Industries, Inc., Report 748 30-1, (August 1979), J. R. Weeks, "Corrosion of Materials in Spent Fuel Storage Pools."

BNL-NUREG-23021, July 197 , .l. S. Anderson, "Irradiation Study of Boraflex Neutron Shielding ).

Materials," Brand Industries, Inc., Report 748-10-1, (August 1981 4 J. S. Anderson, "A Final Report on the Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials,"

Brand Industries, Inc., Report 748-21-1,(August 1978).

2.3 Structural Design Our evaluation of the structural aspects of the proposed modifications are based on a review performed by the staff's consultant, Franklin Research Center (FRC). The FRC Technical Evaluation Report, TER-C5506-529, is appended 1 l

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to this Safety Evaluation and provides additional details relating to the '

structural evaluatio .3.1 Description of the Spent Fuel Pools and Racks There are two SFPs at Turkey Point, one for each unit. They are constructed of reinforced concrete whose walls and floors are lined with a 1/4 inch-thick water tight stainless steel liner. The fuel assembly storage area is approximately 41'-4" wide by 25'-4" long. Wall thicknesses are 5'-6" on three sides and 4'-0" on the fourth side. The floors of the pools are supported directly cn foundation soi ,

The Region 1 storage racks are composed of individual storage. cells made of stainless steel. The cells within a moduleEach are interconnected by grid .

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assemblies to fom an integral structur rack module is provided with leveling pads which contact the SFP floor and are remotely adjustable from above throughout the cells at installation. The modules are freestanding and are not anchored to floor nor braced to the pool walls. The fuel rack assembly consists of three major sections which are the leveling pad assembly, the lower and upper grid assemblies, and the cell assembl The Region 2 storage racks consist of stainless steel cells assembled in a checkerboard pattern, producing a honeycomb type structure. The cells are welded to a base support assembly and to one another to fom an integral structure without the use of grids which are used in the Region 1 rack This design is also provided with leveling pads which contact the SFP floor and are remotely adjustable from above through the cells at installatio The modules are free standing and are not anchored to the floor nor braced to the pool walls. The fuel rack module consists of two major sections which are the base support assembly and the cell assembl .3.2 Apolicable Codes, Standards and Specifications Load combinations and acceptance criteria were compared with those found in the "Staff Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and amended January 18, 197 The existing concrete pool structure was evaluated for the new loads in accordance with the requirements of the Turkey Point FSAR Section 3. .3.3 Loads and Load Combinations Loads and load combinations for the racks and the pool structure were reviewed and found to be in agreement with the applicable portions of the staff position and the Turkey Point FSAR as identified in Section 2.3.2 of this SE. Additional details are provided in the Appended TER, 2.3.4 Seismic and impact Loads Seismic loads for the rack design are based on the original design floor acceleration response spectra calculated for the plant at the licensing ,

stage. This was based on a 0.05 j 0.159 safe shutdown earthquake (g operating basis earthquake SSE). The seismic loads were applied to the (OBE) an model in three orthogonal directions. Loads due to a fuel bundle drop I

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accident were considered in a separate analysis. The postulated loads from these events were found to be acceptable. Additional, description and details are provided in the appended TE .3.5 Design Analysis of Procedures a. Design and Analysis of the Racks The dynamic response and internal stresses and loads are obtained from a seismic analysis which is performed in two phases. The first phase is a time history analysis on a simplified nonlinear finite element model. The '

second phase is a response spectrum analysis of a detailed linear three dimensional rack assembly finite element model. Two percent damping is used in the seismic analysis for both the OBE and SSE. Further details on the methodology is discussed in the appended TE Calculated stresses for the rack components were found to be within allowable limits. The racks were found to have adequate margins against sliding and tippin An analysis was conducted to assess the potential effects of a dropped fuel assembly on the racks and results were considered satisfactor An analysis was conducted to assess the potential effects of a stuck fuel assembly causing an uplif t load on the racks and a corresponding downward load on the lifting device as well as a tension load in the fuel assembl Resulting stresses were found to be within acceptance limit o. Analysis of the Pool Structures The SFPs are reinforced concrete plate structures supported on compacted limerock fill. The SFP walls are lined with 1/4-in. stainless steel liner These existing structures were analyzed for the modified fuel rack loads using a finite element computer program. Original plant response spectra and damping values were used in consideration of the seismic loadings.

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Design criteria, including loading combinations and allowable stresses, are in compliance with Turkey Point FSAR Appendix 5A and the existing SFPs are determined to safely support the loads generated by the new fuel vack .3.6 Conclusions l Based on the above and appended TER, the staff concludes that the proposed rack installation will satisfy the requirements of 10 CFR 50, Appendix A (GDC 2, 4, 61 and 62), as applicable to structure .4 Installation of Racks and Load Handlina There is spent fuel in both Turkey Point Unit 3 and 4 SFPs. A temporary i

crane will be used to move the racks into and out of the SFPs. The movement of the temporary crane will be over the exclusion areas as defined in the l licensee's Phase I submittal for NUREG-0612. "Control of Heavy Loads at j Nuclear Power Plants." However, the licensee has performed a load drop j analysis which indicates that the consequences of a postulated load drop or l

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temporary construction crane drop would be bounded by the cask drop accident. Furthermore the licensee has re-evaluated the cask drop accident using the assumption that all of the spent fuel in the pool was damaged and the newest fuel in the pool had been cooled for at least 1525 hour0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br /> Technical Specification 3.12 has been revised to require a decay time of -

1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br /> for all fuel in the spent fuel pool prior to cask handling operations. This evaluation is conservative in that not all of the fuel would be damaged in a real cask drop acciden The NRC staff's independent evaluation of the cask drop accident in support of the existing SFP racks dated March 17, 1977, resulted in conservatively ,

estimated two-hour radiation doses at the exclusion area boundary (EAB) of 24 Rem to the tyroid and less than 1 Rem to the whole body. Our independent evaluation of the cask drop for the proposed SFP reracks resulted in .

conservatively estimated two-hour radiation doses at the EAB of 26 Rem to the thyroid and less than 1 Rem to the whole body. The slight increase to the thyroid is insignificant when comparted to the 10 CFR 100 guidelines for the two-hour dose of 300 Rem to the thyroid and the 1 Rem to the whole body, in both cases, is significantly less than the two-hour dose of 25 Rem whole body provided in the 10 CFR 100 guideline Based on the above, the staff concludes that load handling accidents associated with these SFP modifications will not have any adverse consequences as identified in NUREG-0612, are well within the.10~CFR 100 guidelines, and are acceptabl .5 Radiological Consecuences of Accident Involving Postulated Mechanical Damage '

to the Spent Fuel This portion of the staff's review was conducted in accordance with the guidance provided in NUREG-0800 "Standard Review Plan", Section 15.7.4, NUREG-0612,'and NUREG-0554 with respect to the accident assumption For evaluation of accidents involving the spent fuel pool, three types of accidents were considered; a cask drop or tip, a construction accident during rack replacement and a fuel assembly drop while' handling fuel. As noted in Section 2.4 of this SE, the effects of a postulated load drop are bounded by the cask drop acciden .5.1 Cask Droo/Tip Accidents Proposed technical specification 3.12 will require a minimum of 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br /> of decay for all spent fuel stored in either pool prior to cask handling operations. A conservative estimate of damage to stored spent fuel assemblies would be from impact of a cask which is sufficient to damage 91 assemblies (in the appropriate strike sector) and result in the release of their concomitant volatile gap activities. In performing our independent accident radiological consequences analysis, we assumed that the fuel has been discharged from the reactor after operation at a steady-state power level of 2300 MW p for an extended period of time. The calculated (0-2 hr.)

offsite accident Padiological consequences are estimated to be 26 Rem thyroid and less than 0.1 Rem whole body at the Exclusion Area Boundar These consequences are well within the radiological guideline values

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specified in 10 CFR 100. See Section 2.4 of this SE for additional detail Radiological consecuences at the Low Population Zone Boundary (LPZ) are commensurately less than those at the Exclusion Area Boundary (EAB).

2.5.2 Construction Accidents For purposes of ensuring that a conservative estimate of damage to stored fuel assemblies from impact of an unspecified object in a non-mechanistically defined construction accident is made, sufficient damage to 157 assemblies (a full core offload) to result in the release of their '

concomitant volatile gap activities was postulated conservatively. The Ifcensee has indicated in their submittals that the reracking operation will take place no sooner than 2150 hours0.0249 days <br />0.597 hours <br />0.00355 weeks <br />8.18075e-4 months <br /> after shutdown for the last batch of spent fuel placed in the SFP. This is to compensate for an 8 ft, water level reduction in the spent fuel pool during rack handling operations. The additional cooldown time compensates for a reduction in pool iodine decontamination factor from 100 to 10 during this period, based upon staff analyses used to determine the Regulatory Guide 1.25 value of 100 for a 23 foot water depth. In performing our independent accident radiological consecuence analysis, we assumed that the fuel has been discharged from the reactor af ter operation at a steady-state power level of 2300 MWg for an extended period of time. The calculated (0-2 hr.) offsite accideHt radiological consequences are estimated to be 45 Rem thyroid and 0.5 Rem whole body at the EAB. These consecuences are well within the cuidelines of 10 CFR 100. Radiological consequences at the LPZ are comensurately less than those at the EA .5.3 Fuel Handling Accident

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The postulated fuel handling accident is not directly related to the -

rereacking application. The fuel handling accident involves the release of the equivalent gap activity of one assembly recently discharged from the reactor for the current fuel exposure of 50,000 Mwd /t.

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In performing our independent radiological consequence analysis for the fuel handling accident, we assumed that the fuel has been discharged from the reactor after operation at a steady-state power level of 2300 MW for an extended period of time. The calculated (0-2 hr.) offsite accid $t radiulogical consequences are estimated to be 30 Rem thyroid and 0.1 Rem whole body at the EAB, well within the guidelines of 10 CFR 100 for the two-hour dose of 300 Rem to the thyroid and 25 Rem to the whole body at the EAB. Radiological consequences at the LPZ are cocinensurately less than those at the EA .5.4 Conclusions The staff concludes that a cask drop /tip or construction accident resulting in damage to either ninety-one 50,000 mwd /t spent fuel assemblies or 157 similar assemblies with at least 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br /> and 2150 hours0.0249 days <br />0.597 hours <br />0.00355 weeks <br />8.18075e-4 months <br /> of cooldown time, respectively, will result in atmospheric radionuclide releases with consequences which are well within the dose guidelines of 10 CFR 10 _

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Additionally, the staff concludes that a fuel handling accident resulting in damage to a recently discharge 50,000 mwd /t spent fuel assembly will result in atmospheric radionuclide releases which are well within the dose guidelines of 10 CFR 10 .6 Occupational Radiation Exposure The occupational exposure for the licensee's plan for the removal and disposal for the high density racks, and installation of the higher density racks is approximately 59 person-rems. This estimate is based on the licensee's detailed breakdown of occupational exposure for each phase of the ,

modification. The licensee considered the number of individuals performing a specific job, their occupancy time while performing this job, and the

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average dose rate in the area where the job is being performed. The spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fue One potential source of radiation is radioactive activation or corrosion products called crud. Crud may be released to the pool water because of fuel novements during the proposed SFP modifications. This could increase radiation levels in the vicinity of the pools. During refuelings, when the spent fuel is first moved into the fuel pool, the addition of crud to the pool water from the fuel assembly and from the introduction of primary coolant to the pool water is greatest. However, the licensee does not expect to have significant releases of crud to the pool water during modification of the pool. Another source of radioactivity in the SFP water is fission product The fission products are released through minute defects in the fuel cladding and are significantly reduced when removed from the reactor vessel and are no longer being irradiated. The purification system for che pool, which has kept radiation levels in the vicinity of the pool to low levels, includes filters and demineralizers to remove crud and radionuclid s. The purification systems will be operating during the modification of the pool FPL's operating experiences has shown that the storage of additional fuel due to reracking will not contribute to the amount of crud released to the pool. If crud deposits should becom significant contributor to pool doses, measures will be taken to reduce such I doses to ALAR l l

The licensee has presented two alternative plans for removal and disposal of ;

the old racks. These are (1) to decontaminate and dispose of as radioactive i

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waste for burial or (2) d? contaminate and dispose of as nonradioactive waste in accordance with existing Turkey Point health physics procedures. The old racks will be rinsed by hydrolasing to remove any loose contamination. This operation will be performed underweter to minimize airborne radioactivity levels. In any event, the disposal methodology will follow ALARA guidelines for each of the alternative Divers will not be used during the reracking operation and no underwater work will be necessary except some simple manipulations which can be perfomed from above the surface of the pool using special tools. If divers ,

are needed, detailed procedures will be developed and submitted to the staff '

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The licensee has taken measures to ensure that personnel exposures during the SFP modifications are ALARA. These measures are described in the licensee's radiation protection program which assures compliance with established procedures to maintain doses ALARA. FPL 's radiation protection program was reviewed prior to the last rerack and was determined adequate and accentable by the staf Based on the manner in which the licensee will perform their modifications, their radiation protection program, including area and airborne radioactivity monitoring, and relevant experience from other operating reactors that have performed similar SPF modifications, the staff concludes ,

that the licensee's SFP modifications can be performed in a manner that will ensure as low as is reasonably achievable (ALARA) exposures to worker We have estimated the increment in onsite occupational dose during normal operations after the pool modifications resulting from the proposed increase in storage fuel assemblies. This estimate is based on information supplied by the licensee for occupancy times and for dose rates in the spent fuel area from radionuclide concentrations in the SFP water. The spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel. Based on present and pro.iected operation, we estimate that the proposed modification should add less than one percent to the total annual occupational exposure of 870 person rem / year / unit (for the years 1970-1982).

2.6 Conclusion The basis of our acceptance of Turkey Point's occupational dose control programs is that doses to personnel will be maintained within the limits of 10 CFR 20 "Standards for Protection Against Radiation", and as low as is reasonably achievchle. Based on present and projected operations in the SFP area, we estinate that the proposed modifications should add less than one percent to the total annual cccupational radiation exposure at both unit The small increase in radiation exposure should not affect the licensee's ability to maintain individual cecupational doses to as low as is reasonably achievable levels and within the limits of 10 CFR 20. Thus, we conclude that storing additional fuel in the two pools will not result in any significant increase in doses received by worker .7 Spent Fuel Pool Cooling and Makeup Systems Each SFP cooling loop consists of a pump, heat exchanger, filter, demineralizer, piping, and associated valves and instrumentation. The pump draws water from the SFP pit, circulates it through the heat exchanger, and returns it to the pit. Component Cooling Water cools the heat exchange Redundancy of this equipment is not required because of the large heat capacity of the pit and its corresponding slow heat-up rate. Nonetheless, a 100-percent-capacity spare pump which is pennanently piped into the SFP cooling system has been installed. This pump is capable of operating in place of the originally installed pump, but not in parallel with the originally installed pump. Also, alternate connections are provided for connecting a temporary pump to the spent fuel pit loo _

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The existing cooling systems for the SFPs are not safety grade and there are no connections to the shutdown cooling system or other safety related cooling systems. Therefore in accordance with the Standard Review Plan Section 9.1.3, we assumed that all pool cooling would be lost following a safe shutdown earthquake. Assuming the loss of cooling, boiling would occur after 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the nonnal heat load condition and after 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the maximum heat load condition for the new racks. This would result in a boil off rate of 37.0 and 72.0 gpm, respectively. The licensee has comitted to upgrade the SFP cooling systems such that they will remain functional after a safe shutdown earthquak The SFPs will be analyzed and modified, as necessary, to assure that the cooling function is not lost as the result '

of the seismic event. The design, procurement, and construction associated with this upgrade will be completed by the end of the second refueling outage after issuance of approval for the re-racking of the SFP The structural considerations of the thermal loads imposed by a pool water temperature of 212*F on the steel liners and the corecrete have been reviewed by the Structural Engineering Branch. The resulting tensil stress is 38 ksi versus the allowable value of 36 ksi. However, realizing the self-relieving nature of the thennal stresses and further acknowledging that the section in general remains elastic, pool function and structural integrity are maintained. See Section 3.4.3 of the appended TER for further details. The radiological effects have been reviewed by the Accident Evaluation Branc An independent accident evaluation of the radiological consequences of SFP boiling was perfonned. The offsite radiological consequences were found to be a small fraction of the 10 CFR 100 guidelines, provided that sufficient make up water capacity is availabl The proposed rerack will result in no significant change in the time to boiling under the presently authorized storage. Until the upgrade is complete the amount of fuel that will be stored will be less than the capacity of the existing racks. Multiple alternate means of makeup water are available until seismically upgraded. Temporary connections can be provided from the fire water system or from the primary water storage tan Additionally, there are two firehouses nearby such that, should a safe I shutdown earthquake occur before the upgraded cooling system is operational, ;

fire engines could be available in less than an hour and provide makeup l water to the pools. Thus, adequate time is available to provide the I necessary makeup water, j 2.7.1 Support Systems The SFP cooling system heat exchangers are cooled by the component cooling water systems. The component cooling water system heat exchangers are ;

cooled by the service water systems. The licensee proposed no modifications l to these two systems as part of this spent fuel pool expansion projec l These systems were a reviewed as to their adequacy to remove the additional heat load and were found to be capable of removing the additional hea .7.2 Decay Heat Loads The licensee's calculated spent fuel discharge heat load to the pools, which ,

was detennined in accordance with the Branch Technical Position ASB 9-2, '

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"Residual Decay Energy for Light Water Reactors for Long Term Cooling", and the Standard Review Plan Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System", indicates that the expected maximum normal heat load following the last refueling will be 17.9 MBTU/Hr. This heat load will result in a maximum bulk pool temperature of less than 143* This nonnal pool temperature (143'F) is higher than the acceptance criteria of 140*F as defined in the Standard Review Plan, however, it is acceptable because the heat load calculations considered each reload to consist of one half of a core instead of the actual reloads being thirds of a core. Had the calculations been performed using the third core reloads, the pool temperature would have been less than the 140* The expected maximum ,

abnormal heat load following a full core discharge is 35.0 MBTU/Hr. This abnormal heat load results in a maximum bulk pool temperature of less than 183*F which is below boiling (212'F) and within the acceptance criteria identified abov .7.3 Conclusions Based on the above, we have concluded that the proposed overall SFP modifications are acceptable with respect to the storage rack capacities, the SFP cooling system capabilities, support system capabilities, the heat loads and pool water temperature .8 Radioactive Waste Treatments The Turkey Point plant contains radioactive waste treatment systems designed to collect and process the caseous, liquid, and solid wastes that might contain radioactive material. The radioactive waste treatment systems were evaluated in the Safety Evaluation dated March 1972, in support of the issuance of the Operating Licenses. There will be no change in the conclusions.given regarding the evaluation of these systems because of the proposed spent fuel pool rerac .8.1 Conclusion Our evaluation of the radiological considerations supports the conclusion that the proposed installation of new spent fuel storage racks at Turkey Point, Unit Nos. 3 and 4, is acceptable based on the fact that previous conclusions relating to the radioactive waste treatment systems, as found in the Turkey Point Unit Nos. 3 and 4 Safety Evaluation, are unchanged by the installation of new spent fuel storage rack .0 Significant Hazards Consideration Coments The request for these amendments was individually noticed on June 7, 1984 (49 FR 23715) followed by a monthly notice on July 7, 1984 (49 FR 29925).

Coments, request for a hearing and petition for leave to intervene were filed on July 9,1984, by the Center for Nuclear Responsibility and M Joette Lorion. We have addressed the relevant coments in the text of this l Safety Evaluation. The petitioners contend: l

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"A.1 The Comission has traditionally held, in a series of case law that expansion of the spent fuel facility constitutes a significant safety hazards consideration."

Under the Comission's regulations in 10 CFR 50.92, an initial determination -

that the proposed amendments involve no significant hazards consideration was made based on a determination that on the operation of the facilities in accordance with the proposed amendments would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a ,

significant reduction in a margin of safety. Section 4.0 of this Safety Evaluation contains the Final No Significant Hazards Consideration Determination based on our evaluation and the fact that the reracking technology in this instance, has been well developed and utilized (over 100 similar applications have been approved) and the K,ff of the SFPs will be maintained less than or equal to 0.9 "A.2 Acceptance criteria for criticality will not be met and thus, FPL will not be able to ensure that the fuel storage facility will always be subcritical by a safe margin in both nonnal operating and accident Conditions."

This contention is addressed in Section 2.1 (Criticality Considerations) of this SE. The criterion for the neutron multiplication factor (K ff) for storage of spent fuel is less than or equal to 0.95 including ali uncertainties at the 95/95 probability confidence level. As noted in Section 2.1, this criterion is met for all normal and abnonnal conditions for storage of the spent fuel in the proposed configuration at the Turkey Point facilitie "A.3 The recitation and notice in 48 (sic) Federal Register Notice 23715 Vol. 49, No. Ill, June 7,1984, that the established acceptance criteria for criticality in the spent fuel pool shall be kept at or below K 0 untrue as evidence by 48 (sic) Federal Register Notics 25360, TNume.95 49, N is 120, June 20, 1984."

This contention is incorrect. As noted above in response to contention A.2, the K for the SFPs is maintained equal to or less than 0.95 including allubrtaintiesatthe95/95probabilityconfidencelevel. The June 20, 1984 Federal Register Notice (49 FR 25360) was related to a separate action addressing the existing new fuel (unirradiated) storage racks which are not affected by these proposed amendment "A.4 In light of the fact that the utility, FPL, wants to operate the facility with a K of 0.98 (FR 25360), as above referenced, places the proposedundertakThintheSignificantSafetyHazardsCategory,andthere can be no issuance of a license amendment to expand the spent fuel facility without a public hearing required by the Atomic Energy Act of 1954." ,

In support of contentions A.1 - A.4, the petitioners note the position taken l

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by the Comission in Policy Issue SECY-83-337, STUDY ON SIGNIFICANT SAFETY

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HAZARDS, August 15, 1983:

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"A K of greater than 0.95 may be justifiable for a particular appibtionbutitwouldgobeyondthepresentacceptedstaffcriteria and would potentially be a significant hazards consideration." page 5- This contention is factually incorrent. As inoicated in responses to contertions A.1 through A.4, the SFPs for Turkey Pont Units 3 and 4 utilize current and accepted technology and the K,ff will be maintained less than or equal to 0.9 .0 Final No Significant Hazards Consideration *

The standards used to arrive at a proposed determination that a request for

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amendments involves no significant hazards consideration are included in the Comission's regulations,10 CFR 50.92, which state that the operation of the facilities in accordance with the proposed amendments would not (1)

involve a significant increase in the accident previously evaluated; or (2)create probability or consequencs the possibility of a newoforan different kind of accident from any accident previously evaluated; or (3)

involve a significant reduction in a margin of safet The proposed SFP expansion amendments are very similar to the initial SFP expansions, identified in Section 1 of this SE, in which many of the same issues were raised and resolved when the initial expansions were approve Each specific aspect of this request was reviewed in detail and was very much a repeat of the initial expansion review. The knowledge and experience gained by the NRC staff in reviewing over 100 7,imilar requests was also utilized. The current expansion request does r.ot use any new or unproven technology in either the construction process or in the analytical techniques necessary to support the expansion request. The same postulated q accidents were looked at again and the same precautions have been proposed '

by the licehsee during the installation In addition, the neutron multiplication less than 0.95facter (K'uSc)ertainties.of the pools will be maintained equal to or including Accordingly, the staff has detennined that the request for amendments to i

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expand (reracking to allow clo:er spacing) does not significantly increase  !

the probability or consequences of accidents previously evaluated; does not '

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create new accidents not previously evaluated; and does not result in any i significant reduction in the margins of safety with respect to criticality, j cooling or structural consideration '

The following evaluation in relation to the three standards demonstrates that the proposed amendments in support of the SFP expansions do not involve a significant hazards consideratio First Standard - Involve a significant increase in the probability or consequences of an accident previously evaluate The following potential accident scenarios have been identified: A spent fuel assembly drop in the spent fuel poo !

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4 A spent fuel cask dro . A construction acciden The probability of any of the first four accidents is not affected by the racks themselves; thus reracking cannot increase the probability of these accidents. As for the construction accident, FPL does not intend to carry ,

any rack directly over the stored spent fuel assemblies. All work in the spent fuel pool area will be controlled and performed in strict accordance with specific written procedures. Details on the precautions and  ;

requirements related to the installation and load handling during the SFP expansion activities and the licensees compliance to the requirements of NUREG-0612 "Control of Heavy Loads at Nuclear Power Plants" are provided in our SE dated August 29, 1984 Accordingly, the proposed expansion does not significantly increase the probability of an accident previously evaluate The consequences of (1) a spent fuel assembly drop in the SFP and (4) a t

spent fuel cask drop and (5) a construction accident are discussed in detail in Sections 2.4 and 2.5 of this S As noted in Section 2.4 of this SE, a load drop analysis was perfonned and indicates that the effects or consequences of a postulated load or temporary construction crane drop are bounded by the cask drop analysis. The consequences of the cask drop accident analysis results in a slight increase from the previous analysis for the existing racks in the estimated two-hour radiation dbses at the EAB of 2 Rem to the thyroid with no change to the estimated doses to the whole body. The estimates resulting from our current aralysis of 26 Rem to the thyroid and 1 Rem to the whole body are significantly less than the two-hour dose of 300 Rem to the thyroid and 25 Rem to the whole body at the EAB provided in 10 CFR 100 guideline The postulated fuel handling accident is not directly related to SFP expansion request as stated in .'ection 2.5.2 of this SE. The results of our analysis assuming fuel exposure of 50,000 Mwd /t and steady-state power level of 2300 MW g results in 30 Rem thyroid and 0.1 Rem whole body at the EAB, well within the 10 CFR Part 100 guidelines identified above. There will be no significant increase in the consequences in that the fuel handling accident is not directly related to the SFPs storage capacity but is dependent on the release of the equivalent gap activity of a sinole assembly recently removed from the reactor.

, Section 2.3.4, and 2.3.5 and the Appended TER of this SE indicate that the postulated loads from a seismic event will not result in failures to the racks or pool structures, thus their integrity will be maintained. Neither the staff nor the license could identify any new means of losina cooling water. Therefore, since the integrity of the racks and SFP will be maintained there will be no significant change in the consequence of a l

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seismic event as the result of this amendment than previously evaluated seismic event .

As stated in Section 2.7 of this SE, the proposed rerack will result in no significant change in time to boiling under the presently authorized storag The existing SFP cooling systems are not seismic Category 1, however, the licensee has committed to upgrade the systems to assure functional capabilit Adequate time is available to provide the necessary makeup water from either on-site sources or fire engines from a nearby fire house. Thus, the time availt.ble and alternate means of providing mckeup water to the SFP result in no significant increase in the consequences of loss of flow from that ,

previously evaluate Therefore, based on the above, the probability or consequences of previously analyzed accidents will not be significantly increased as the result of the proposed SFP expansion Second Standard - Create the possibility of a new or different kind of accident from any accident previously evaluate The proposed SFP expansions have been evaluated in accordance with the guidance of the NRC position paper entitled, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate Industry Codes and Standards as identified in this SE. In addition, several previous NRC SEs for SFP expansions similar to this proposal have been reviewed. Neither the licensee nor the NRC staff could identify a credible mechanism for breaching the structural integrity of the SFPs which could result in loss of cooling water such that cooling flow could not be maintained or any other accidents not previously evaluated that might result from these amendment As a result of this SE and these reviews, the proposed SFP expansions do not, in any way, create the possibility of a new or different kind of accident from any accident previously evaluated for the Turkey Point SFP Third Standard - Involve a significant reduction in a margin of safet The NRC staff safety evaluation review process has established that the issue of margin of safety, when applied to a SFP modification, will need to address the following areas: Nuclear criticality consideration . Thermal-Hydraulic consideration . Material, Structural and Mechanical Consideration The established acceptance criteria used to assess the adequacy of SFP facilities assure maintenance of the necessary margins of safety. The staff's SE addresses the three areas identified above. The current request is very similar to the first request for expansion in that it raises the

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same issues that were raised and resolved in the first request. Whereas each aspect of this request was of course reviewed in detail, the review process and scope was very much a repeat of the first expansion. In both reviews, the established criteria have been me With the criteria met, the necessary and intended safety margins are maintained and there is no significant reduction in margi The criterion used in addressing nuclear criticality considerations for the storage of spent fuel is that the neutron multiplication factor (Klessff ) isthan or equal probability confidence level . '

As noted in Section 2.1 of this SE, the criterion is met for all nomal and abnormal conditions for the storage of spent fuel in the proposed configuration. The proposed amendments, therefore, do not significantly reduce a margin of safety for criticalit The criteria used in addressing thermal-hydraulic considerations for the storage of spent fuel are the methodologies and assumptions identified in Branch Technical Position ASB 9.2 and the SRP Section 9.1.3 to assure the temperatures for the SFP do not exceed 140*F under nomal conditions during reloads and not exceed 212'F (boiling) during abnomal conditions following a full core discharg As noted 2.7 of this SE, the criteria are met for the nomal third of a core reload and for the abnormal full core discharge conditions for bulk pool temperature The proposed amendments, therefore, do not significantly reduce the margin of safety for spent fuel coolin The criteria used in addressing material, structural and mechanical consideratipns are that the compatibility and chemical stability of the materials wetted by the SFP water be demonstrated and no significant corrosion occur. The structural and mechanical design of the SFP and storage racks maintain the fuel assemblies in a safe configuration through all environmental and abnormal loadings using the codes, standards and specifications identified in Section 2.3.2 of the S As noted in Section 2.2 of this SE, the corrosion that will occur in the SFP .

environment will be of a little significance for the life of the plant and the !

environmental compatibility and stability of the materials used is adequate l

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based on test data and actual service experience in operating reactor As noted in Section 2.3 of this SE and the Appended TER, the structural and rrechanical design of the SFPc and storage racks can withstand the environmental and abnomal loading and the SFP structure can sustain the higher density floor loadings with adequate margin. The proposed amendments, therefore, do not significantly reduce the margin of safety with regard to materials, structural, and mechanical integrit '

As the result of this SE and these reviews, the proposed SFP expansions do not result in a significant reduction in a nargin of safety with respect to criticality, cooling or structural consideration .

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Based on the foregoing, and the fact that the reracking technology in this instance has been well developed and demonstrated (100 similar applications have been approved), the Commission has concluded that the standards of 10 CFR 50.92 are satisfied. Therefore the Comission has made a final determination that the proposed amendment does not involve a significant hazards consideratio .0 Environmental Considerations A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51. The Notice of Issuance of Environmental Assessment and Finding of ,

No Significant Impact was published in the Federal Register on November 16, 1984 (49 FR 45514).

6.0 Conclusion We have concluded based on the considerations discussed above, that: (1)

these amendrrents will not (a) significantly increase the probability or consequences of accidents previously evaluated, (b) create the possibility of a new or different accident from any previously evaluated or (c)

significantly reduce a margin of safety and, therefore, the amendments do not invohe significant hazards considerations; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted

- in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and necurity or to the

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health and safety of the publi Dated: November 21, 1984

, Principal Contributors:

D. Mcdonald, Project Manager M. Wohl, Accident Evaluation Branch J. Lee, Meteorology and Effluent Treatment Branch -

J. Mins, Radiological Assessment Branch E. Branagan, Radiological Assessment Branch S. Kim, Structural and Geotechnical Engineering Branch B. Turovlin, Chemical Engineering Branch J. Ridgley, Auxiliary System Branch L. Kopp, Core Performance Branch ,

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R. Samworth, Environmental and Hydrologic Engineering Branch  !

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ATTACHMENT 1 TECHNICAL EVALUATION REPORT

EVALUATICN OF SPENT FUEL PACKS STRUCTURAL ANALYSIS

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FLORIDA POWER AND LIGHT COMPANY TURKEY POINT UNITS 3 AND 14

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NRC DOCKET NO. 50-250, 50-251 FRC PROJECT C5506 FRC ASSIGNMENT 26  !

NRC TAC NO. 54480, 54481 l

i NRC CONTRACT NO. NRC-C3-81 130 FRC TASK 529 Prepared by l Franklin Research Center 20th and Race Streets Philadelphia, PA 19103 FRC Group Leader: R. C. Herrick ;

Prepared for Nuclear Regulatory Comrnission Leed NRC Engineer: S. B. Kim Washington, D.C. 20555

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- October 25, 1984 This report was prepared as an account of work sponsored by an agency of the United States 3 Govemment. Neither the United States Gover.iment nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or \, .

responsibility for any third party's use. or the results of such use. of any information, appa-ra*as, product or process disclosed in this report. or represents that its use by such third party would not infringe privately owned rights, o

Prepared by: Reviewed by: Approved by:

RC&ddweW ]hd*1 E Principal Author Project Manager Department Director (icting)

Dater / U- N Datet /0 WNI Date: 10/lI/OY FRANKL!N RESEARCH CENTER DIV1510N OF ARVIN/CALSPAN

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20th and Rue Sireets. Phlia.. Ps. 19103 (215) 448 1000

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TER-C5506-529 SUMMARY OF DESIGN STRESSES AND MINIMUM MARGINS OF SAFETY Normal & Voset Conditions  ;

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, Design Allowable Margin Stress Stress of Svooort Pod Assembiv -

J Support Pod

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Shear 3504 23150+ 5.61

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i . Axial and Bending 10288 23150* 1.25 j Bearing 7631 23150* 2.03

Support Pod Screw Shear 6974 9260 .33 ;

, Support Plate  ;

Shear 4403 9260 1.10 t Weld Shear 16556 21000* .34 ,

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' Cell Assembly Cell  :

Axial and Bending .877 l.0 i .!! !

j Cell to Base Plate Weld

Weld Sheer 15482 21000 .36 j Cell to Cell Weld q Weld Sheer -

18389 23150+ .26 ,

t Cell Seam Weld '

l Weld Shear 1751' 2174 " .25 !

! Cell to wrapper Weld i t

Weld Shoor 10279 18520 " .80 l

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. " SSE Stress is Limiting  !

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t Allowable per Appendix XVll-2115 Eq (24)

tt Design Lood and Allowable Lood in Lbs is Shown

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TER-C5506-529 2 and 3 were provided by the Licensee (5) and the support plate weld shear stress and allowable stresses were subsequently changed as discussed belo Tables 2 and 3 provide the final data which were found to be acceptable during the revie For Tables 2 and 3, the allowable shear stress in the weld of Item 1.3, 9 Support Plate, was changed to 21,000 psi to be in accordance with the allowable weld stress of Table NF-3292.1-1 of the ASME Code.* For Table 3, i

the weld shear stress for Ites 1.3 was changed to 16,556 psi, recognizing that the support plate compressive load is carried in metal-to-metal contact and is not dependent upon the wel .4 REVIEW OF SPENT FUEL POOL STRUCTURAL ANALYSIS 3. Scent ruel Pool structural Analysis The spent fuel pool is a reinforced concrete plate structure supported on compacted limerock fill. The spent fuel pool walls are lined with 1/4-in stainless steel liner. The Licensee presented an analysis to demonstrate the I structural integrity of the spent fuel pool for the postulated loading

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conditions for the new high density rack l

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3. Analysis procedure The Licensee used the finite element method for the analysis of the spent

fuel pool. The struc are was modeled with three-dimensional solid elements l and the ANSYS computer code. By approximating symmetry along the long ( nor th-sou th) diretion of the pool, only half of the pool was modeled. The l boundary conditions on the plan of symmetry were adjusted to represent syneetric and non-symmetric leading conditions. The liner plate was not considered to provide structural resistance in the pool analysis. The soil i medium was represented by vertical compression spring elements. The thermal 1 ef fects were obtained by imposing a uniform uneraal gradient across solid elements.

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• American Society of Mechanical Engineers, Boiler and Pressure vessel Code,Section III, Division 1 Subsection NF,1980 Edatio .

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TER-C5506-529

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The following critical loading combinations were considere . Y = 1. 2 5 (D+ P+ L) with and without T Y = 1.25 (D+P+L) with and without W Y = 1. 2 5 (D+P+L+E) with and without T Y = 1.0 (D+P+L+E') with and without T where Y = required yield strength of the structure '

D = weight of the structure plus permanent loads P = hydrostatic pressure of pool water L = weight of loaded fuel racks in pool E = design tarthquake load. 0.059 horizontally, 2/3 (0.059)

vertically E'= maximum earthquake load, 0.159 horizontally, 2/3 (0.159)

vertically T = thermal load (inside f ace of walls 180*F, exposed face 30'F, and bottom face of slab 50*F)

W = wind loa As a result of this analysis, the Licensee stated the following: Seismic analysis for the new racks showed that these racks do not uplif t during the seismic event and, thereferi, no additional amplification f actors for impact were considere . The analysis showed that the seismic loading created a more severe effect than the combined effect of torne,do, wind, and c

depressurizatio I The resulting stresses in the elements caused by mechanical loads were evaluated by computing the capacities of individual sections and  ;

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comparing the capacities to the actual normal forces and moment l I

A For the combinations of mechanical and thermal loads, the sections l were analyzed following the approach shown in * Commentary to ACI 349-R-80.* A separate analysis was conducted to determine the ef fects of thermal, hydrostatic, and hydrodynamic loads on the functionality of the liner. The analysis showed that there was no loss of f unctio The results of the structural analysis were summarized in the Licensee's Table A (5), reproduced here as Tables 4-a and 4-b.

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Table 4- Sgent ruel hiol Load Containations and Stresses

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MECilAlalCAL LOADS MECliAtJICAL & IIEftMAL o

1.25 (D e P e L) 1.25JD e P * L) * E I.25 (D e P e L) E e T (I) (1) (2) (3)

• N M Mm/M i1 M Mm Mm/M flebor Stress ffY Mnr ft .bor Location (K/ft) K-f t/f t K-ft/ft (K/ft) K-f t/f t K-ft/ft Stress .95 1 .7 27 is = 12.8 ksi(5) 2.81 Base Mot 1 Eos Wall .36 2 .3 -43 1.47 fy = 142 psi (6) f.fM (4)

(iv = 82 psi) 1.00(4) (fv = I42 psi) 1.04 (4)

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Y 568 4.66 6 .0 is = 35.l ksi I.03 East Woli 3 I's = -9.6 ksi (Pool)

iJoeth Wall 1 %6 -123 f.27 -13.! -140 -151 1.08 is = 27.1 ksi l's = -2.65 ksi 1.33-3 .99 2 .8 -182 2.39 is = 35.3 ksif7) 1.02 South Wall 1 l's = l.4 ksi

209 9,46 .5 218 is = 9.6 ksi 3.75 Middle Wall 2 .1

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Ps = 9.0 ksi is = Stress in tension steel N

li = Applied normal force on section M = Applied mom nt on section l's fv

= 5 ress in compression : teel

= Conctele simor stress

{ Mm = Manimum elastic moment E (ne.fitIve sl<p indicates compressive stress)

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TER-C5506-529

- Table 4- Notes for Table 4-a Maximum elastic moment for a section with normal force N impo' sed on i (g)

B sed on a crocked onolysis per the methodology discussed in Reference 2, (2)

reinforcing steel stress is obtained directl ,

(3) Ove to the self relieving nature of thermal loods on reinforced concrete, the ratio of maximum moment copocity to octual moment connot be uniquely determined. As on alternative, the ratio of dFy to computed reinforcing steel stress is provide Since structural integrity is maintained beyond the allowable stress for thermal loading, the octual sofety factor is grecter than the ratio reporte (4) Where sheer stresses control, the ratio provided is that of allowable sheer stress (conservatively token as 145 psi) divided by f (3) This stress represents the maximum stress found in the top loyer of reinforcing steel in the thinner center section of the base mot. The top steel in this.crea is important for transfer of the tensile loads imposed by the laterol water pressure from the pool. The bottom steel in the center portion of the base mot of the pool is used primarily for ercek contro Since the base mot rests directly on competent fill material, 7 tresses in this bottom (secondary) steel resulting from thermal loods have no odverse effect on the ceility of the pool to transfer load. Therefore, the stress in the bottom steel is.not included in Table .

(6) As shown in Figure 6, this section occurs in the 3 foot wide by 18 inch thiek section of the east wall between the two canal wo!!s. Beecuse of the short span of this section, and the large ratio of section thickness to span length, the section does not resist foods in the fashion of o shcIlow beams sheer stresses control the section capacity. Since sheer stirrups are provided, the ellowable sheer stress in the concrete exceeds 148 psi. The reinforcing steel on the outside foce of this section is used only for ercek control and is not needed to resist mechanieel loods. Therefore, the flexural stresses in this reinforcing steel are not included in Table (7) This represents on overage stress (total force on the total section) over the top 10 feet of the outside fece horizontal reinforcing steel. The result

, indicates that the section in general remains beiow the minimum specified yield stress. However, o maximum stress of 38 ksi has been coleulated for tne reinforcing steel in the top element of the well. Realizing the self-relieving nature of the thermal stresses and further acknowledging that the section in general remains elastle, pool function and structurci integrity I are maintaine Additionally, in accordance with the Turkev Point Uedated TSAR. Apeendix 5A,Section II. limited vieldine is allevable 8 h, p

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provided the deflection is checked to ensure that the affected Class l I svstems and equipment are net stressed bevond their allevables, j No Class I systems er eeutpment are attached to this section of val I

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TER-C5506-529 3. Summary of Results ,

The results of the analysis listed in Table 4-a show that the stress levels snder critical loading combinations remain within the spectried allowable values, but with one exception. The review showed that:

' The average bearing stress under the pool slab is below the allowable pressure of 10 kst for the compacted limerock fil . The maximum tensile stress in steel is shown to be 35.3 kai compared to the allowable value, Fy = 36.0 ks . The shear stress in concrete controls the design in the 18-in-thick section of the east wall between the two canals. The ratio of the allowable shear stress to the maximum shear stress is shown to be 1.0 The exception to stresses within the allowable values concerns the tensile stress in the steel of the south wall, which, in accordance with note 7 of Tables 4-a and 4-b, was computed to be a maximum of 38 ksi. For use in

Table 4-a and for comparison to the allowable valuef the Licensee averaged the maximum stresses in the steel over the upper 1C ft of wall to yield an average of 35.3 ksi which was cenpared to the allowable value of 36 kai. Where this procedure may be questioned, the Licensee also cited Appendix 5A, Section !!

of Turkty Point's updated TSAR which states that limited yielding is allowable under certain accident conditions. This was reviewed and considered to be acceptabl In addition, the Licensee's response (10] to USNRC Question No. S regarding the ef f ects of 212'F water in the spent fuel pool concludes that stresses for the thermal load remain within the original design allowable For simultaneous occurrences of seismic and thermal conditions, the Licensee cerorted (10] that localized steel stresses were slightly higher than the allowable stress of 36 kai, and justified their magnitudes by the FSAR statement cited in the paragraph above that would permit local thermal stress yielding under certain accident condition After considering this review, evaluation showed t. hat the 212'T pool water terperature resulted from a coeling system pipe break during a seismic

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TER-CS506-529 event. Thus, considering the hours it would take to raise the pool water torperature to 212'r and increase the thermal gradient in the pool structure, -

the short duration seismic event would have been long past so that the structural considerations would remain to be those of thermal and deadweight ,

only. The Licensee's response to USNRC Question No. 8 (10) indicates that .

an11ysis showed this to be 38 ksi versus the allowable value of 36 kai and was justified by statements in the FSAR as discussed abov This review concludes that the spent fuel structure is acceptable for the higher density loadin .5 TVIL ASSEMBLY DRCP ACCTCENT ANALYSIS

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With respect to accidental dropping of a fuel assembly, the Licensee provided the following:

"In the unlikely event of dropping a fuel assembly, accidental deformation of the racn will not cause the criticality acceptance

criterion to be violate For the analysis of a dropped fuel assembly, three accident conditions are postulated. The first accident condition conservatively assumes that the weight of a fuel assemoly, control rod assembly and handling mechaInism of 3,000 pounds Lapacts the top end fitting of a stored fuel assembly from a drcp height of 3 feet. Calculations will show that the impact energy is absorbed by the dropped fuel assembly, the stored fuel assembly, the cells and rack base plate asseacly. If in the unlikely event that two adjacent cells are crushed together for their fuel length, critically, calculations show that k gg g,0.95. Under these f aulted conditions, credit is taken for dissolved boron in the ,

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water, and the critically acceptance criterion is not violate The second accident condition is an inclined drop on top of the rac Results will be the same as for the fire' conditio The third accident assumes that the dropped assembly (3,000 lbs) falls straight through an empty cell and impacts the rack base plate drom a drop height of 201 inches. The results of this analysis will show that the impact energy is absorbed by the fuel assee.bly and the rack i base plate. Criticality calculations 'shown that k,gg g,0.95 and the critically acceptance criterion is not violated.'

l This statement was found to be acceptable during the revie _ _ _ _ _ _ - - - _ _ _ _ . _

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TEAaC5'406-529 REFERENCES i *

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t 1. Florida Power & Light Company -

l Licensing Report on Turkey Point Units 3 and 4

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spent Fuel Storage Facility Modifications, Safety Analysis Report j NRC Docket Nos. 50-250 and 50-251 2. OT Position for Review and Acceptance of Spent Fuel Storage and Mandling *

Applications, U.S. Nuclear Regulatory Commission January 18, 1979

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3. Franklin Research Center Technical Evaluation Report, 'Ev.tluation of Spent Fuel Racks Structural Analysis for Duke Power Company, McGuire Nuclear Station Units 1 and 2*

August 10, 1984  !

4. Florida Power & Light' Company ,

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Response to FRC's Request for Information October 5, 1984 5. T1crida Power & Light Company Response to FBC's Questions October 1, 1984 6. R. J. Fritz

'The Ef fect of Liquids on the Dynamic Motions of Immersed Solids *

Journal of Engineering for Industry pp. 167-1,73, February 1972 7. D. F. De'Santo l

'Added Mass and Hydrodynamic Damping of Perforated Plates Vibrating in 1 Water *

ASME, Journal of Pressure vessel Technoleyy Vol. 103, p. 175. Kay 1981 8. C. B. Gilscre

' Seismic Analysis of Freestanding Fuel Racks *

Presented at 1982 orlando Pressure and Piping Conference ,

9. ASME Boiler and Pressure Vessel Code,Section III Division 1, Subsection NT, 1980 Edition, Table NF-3292.1-1 10. Florida Power & Light Company Response to USNRC Question No. 8 tegarding the ef frcts of a sustained pool water temperature of 212'T on the pool and cooling system I

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I TER-CS$06-529 i CONCLUSICHS

Based upon the review and evaluation, the following conclusions were r eached

) o The limitations of the modeling technique employed for hydrodynamic coupling of fuel assemblies within a fuel tack cell and of fuel rack modules to other rack modules and the pool walls indicate that the ,

l modeling technique contributes known accuracy only for the condition in which the displacements are small compared to the available i clearance space. An the Licensee's reported dispiscenents are small, an acceptable use of the hydrodynamic coupling was employed.

o computed displacerents are small relative to clearance between rack modules or between rack modules and the spent fuel pool walls. Thus, the use of two-dimensional dynamic rack module analfris was satisfactory for dispAacemen o while the methodology employing two-dimensional nonlinear models and linear three-dimensional models correlated by load correcting f actors

to introd. ace the nonlinear impacting load characteristics to the these-dimensional lineer eodel was not considered to be fully acceptable wittput furthe.t validn' tion as a stress analysis method, a

) detailed step-by-step'rsview of the stress analysis coupled with 1 adcational load tabulations requested and rupplied indicates that,

] with the constrvat Lsma noted to be present, the stress analysis is acceptable, i s

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o The spent fuel pcel structure has design margin to sustain the higher density floct loadings.

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3- - . . NUCEAR REGULCORY COMVZSSION . - - - - .

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NRC STAFF'S 1 ,

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IN THE MATTER OF: DCON1'T NO: 50-443-OL 50-444-OL-Off-Site PUBLIC SERVICE COMPANY OF ergency NEW HAMPSHIRE, et a < Planning i

ao 1 (Seabrook Station, Units 1 and 2) g?

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L I,00ATION: CONCORD, NH DATE: OCTOBER 5, 1987 through DECE"d a 17, 1987

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