ML20236F896

Safety Evaluation Supporting Amends 102 & 80 to Licenses NPF-68 & NPF-81,respectively

ML20236F896
Person / Time
Site:	Vogtle
Issue date:	06/29/1998
From:	NRC (Affiliation Not Assigned)
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ML20236F894	List: ML20236F891 ML20236F896
References
NUDOCS 9807060005
Download: ML20236F896 (23)

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{{#Wiki_filter:r%q g '4 UNITED STATES g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30e#4001 +,*****,o l j l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l RELATED TO AMENDMENT NO.102 TO FACILITY OPERATING LICENSE NPF-68 AND AMENDMENT NO. 80 TO FACILITY OPERATING LICENSE NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL. VOGTLE FL FCTRIC GENERATING PLANT. UNITS 1 AND 2 i i DOCKET NOS. 50-424 AND 50-425 l

1.0 INTRODUCTION

By letter application dated September 4,1997, as supplemented by letters dated November 20, 1997, and May 19 and June 12,1998, Southern Nuclear Operating Company, Inc., et al. (the licensee) proposed license amendments to change the Technical Specifications (TS) for Vogtle Electric Generating Plant (VEGP), Units 1 and 2. The proposed amendments would change the VEGP Units 1 and 2 TS to allow an increase in the Unit 1 spent fuel storage capacity from 288 to 1476 fuel assemblies. The supplements dated May 19 and June 12,1998, provided clarifying information that did not change the scope of the September 4,1997, application and the initially proposed determination of no significant hazards consideration.

2.0 BACKGROUND

The VEGP Units 1 and 2 spent fuel pools (SFPs) are located in a common area that can accept spent fuel from either VEGP Unit 1 or Unit 2. At the present time, VEGP has a fuel storage I capacity of 288 assemblies in the Unit 1 SFP and 2098 fuel assemblies in the Unit 2 SFP. The i licensee has obtained high-density spent fuel racks, previously utilized at the Maine Yankee i Atomic Power Plant (MYAPP). These spent fuel racks were approved for use at MYAPP in an NRC staff letter and accompanying Safety Evaluation (SE) dated June 16,1982. The spent l fuel racks were observed to perform well in service at MYAPP and did not exhibit swelling or any other type of degradation. The only modification performed on the MYAPP spent fuel l-racks, following issuance of the June 16,1982, NRC staff SE, was the addition of vent and l drain holes in pockets that contain the neutron-absorbing Boral material. I 3.0 EVALUATION The NRC staffs evaluation of the proposed use of the MYAPP spent fuel racks in the VEGP Unit 1 SFP appears in the sections that follow. s 9907060005 900629 PDR ADOCK 05000424 p

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' l 3.1 ' Cnticality Analysis The MYAPP spent fuel storage racks were analyzed for use in the VEGP Unit 1 SFP using the Westinghouse methodology described in " Westinghouse Spent Fuel Rack Criticality Analysis - Methodology," Westinghouse Electric Corporation, WCAP-1416-NP-A," Rev.1, November 1996. The WCAP-1416-NP-A ~ methodology was reviewed and approved by the NRC in an October 25,1996, letter from T. Collins, NRC, to T. Greene, Westinghouse Owners Group. This methodology takes partial credit for soluble boron in the fuel storage pool criticality analyses and requires conformance with the following NRC acceptance criteria for preventing criticality outside the reactor: '(1) The e' ffective neutron multiplication factor (k,) shall be less than 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties at a 95 percent probability, 95 percent confidence (95/95) level as described in WCAP-1416-NP-A; and (2) The effective neutron multiplication factor (k, ) shall be less than or equal to 0.95 If fully flooded with borated water, which includes an allowance for uncertainties at a 95/95 ' level as described in WCAP-1416-NP-A. The reactivity effects of VEGP fuel storage in the MYAPP spent fuel racks were analyzei$ with the three-dimensional Monte Carlo code, KENO-Va, with neutron cross-sections generated with the NITAWL-il and XSDRNPM-S codes. Since the KENO-Va code package does not have - bumup capability, depletion analyses and the determination of small react.'vity increments due to manufacturing tolerances were made with the two-dimensional transport theory code, ) PHOENIX-P. The analytical methods and models used in the reactivity analysis have been l-benchmarked against experimental data for fuel assemblies similar to those for which the l MYAPP spent fuel storage racks are designed and have been found to adequately reproduce the critical values. These experimental data are sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions, which include close proximity storage and 1 strong neutron absorbers. The NRC staff concludes that the analytical methods used are L acceptable and capable of predicting the reactivity of the MYAPP spent fuel storage racks with a high degree of confidence. l-The' existing VEGP spent fuel storage racks, which utilize Boraflex as a neutron poison, have previously been qualified for storage of various Westinghouse 17 x 17 fuel assembly types with l 4 . maximum nominal enrichments up to 5.0 weight percent (w/o) U-235 (enrichment tolerance of ' l

• 0.05 w/o U-235). Because of the Boraflex deterioration that has been observed in many spent fuel pools, th!s criticality analysis for the VEGP spent fuel storage racks neglected the presence l

of Boraflex to allow storage of all 17 x 17 fuel assemblies with nominal enrichments up to 5.0 w/o U-235 using credit for checker boarding, bumup, bumable absorbers, and soluble I boron. The MYAPP spent fuel storage racks, which utilize Boral neutron-absorbing panels, have been analyzed to allow storage of Westinghouse 17 x 17 fuel assemblies 'with nominal enrichments up to 5.0 w/o U-235. The analysis takes credit for the presence of Boral absorber panels on all four sides of each spent fuel rack cell. l

3-The moderator was assumed to be pure water at a temperature of 68 'F and a density of 1.0 gm/cc, and the array was assumed to be infinite in lateral extent. Uncertainties due to tolerances in fuel enrichment and density, storage cell inner diameter, storage cell pitch, stainless steel thickness, Boral thickness and width, Boral wrapper thickness and width, , ; assembly position, calculational uncertainty, and methodology bias uncertainty were accounted for. These uncertainties were appropriately determined at the 95/95 probability / confidence level.' A methodology bias (determined from benchmark calculations), as well as a reactivity bias to account for the effect of the normal range of SFP water temperatures (50 'F to 185 'F), were included. These biases and uncertainties meet the previously stated NRC requirements and are, therefore, acceptable. . For the MYAPP spent fuel racks, an enrichment of 3.5 w/o U-235 was found to be adequate to maintain y less than 1.0 with all cells filled with Westinghouse 17 x 17 fuel assemblies and no solubic boron in the pool water (all-cell configuration). This resulted in a nominal 4 of 0.96852. ' The 95/95 4 was then determined by adding the temperature and methodology biases and the

statistical sum of independent tolerances and uncertainties to the nominal 4 values, as described in WCAP-14416-NA-P. This resulted in a 95/95 % of 0.99985. Since this value is q

less than 1.0 and was determined at a 95/95 probability / confidence level, it meets the NRC J criterion for precluding criticality with no credit for soluble boron and is acceptable. Soluble boron credit is used to provide a safety margin by maintaining 4 less than or equal to j 0.95, including 95/95 uncertainties. The soluble boron credit calculations assumed that the l all-cell storage configuration was moderated by water borated to 350 parts per million (ppm). j As previously described, the individual tolerances and uncertainties, and the temperature and methodology biases, were added to the calculated nominal 4 to obtain a 95/95 value. The - resulting 95/95 4 was.0.94470. Since 4 is less than 0.95 with uncertainties at a 95/95 ~ probability / confidence level, the NRC acceptance criterion for precluding criticality is satisfied . with 350 ppm of boron. This value is well below the minimum SFP boron concentration value of ~ ^ 2000 ppm required by VEGP.TS 3.7.17, " Fuel Storage Pool Boron Concentration," and is, I therefore, acceptable. L The concept of reactivity equivalencing due to fuel bumup was used to achieve the storage of b fuel assemblies with enrichments higher than 3.50 w/o U-235 for the all-cell storage _ l configuration. The NRC has previously accepted the use of reactivity equivalencing predicated l upon the reactivity decrease associated with fuel depletion. To, determine the amount of . soluble boron required to maintain 4 s0.95 for storage of fuel assemblies with enrichments up l .to 5.0 w/o U-235, a series of reactivity calculations was performed to produce a set of (- .' enrichment versus fuel assembly discharge bumup ordered pairs, which all yield an equivalent 4 when stored in the MYAPP spent fuel storage racks. These are shown in proposed VEGP TS Figure 3.7.18-1, "Vogtle Unit 1 Bumup Credit for All Cell Storage," for VEGP Unit 1, and > represents combinations of fuel enrichment and discharge bumup that yield the same rack 4 as the rack loaded with fresh 3.50 w/o fuel. Uncertainties associated with bumup credit include a reactivity uncertainty of 0.01 Ak at 30,000 MWD /MTU applied linearly to the bumup credit

requirement to account for calculational and depletion uncertainties and 5 percent on the

. calculated bumup to account for bumup measurement uncertainty. The NRC staff concludes that these uncertainties conservatively reflect the uncertainties associated with bumup calculations and are acceptable. The amount of additional soluble boron, in excess of the [, L l i t value required above, that is needed to account for these uncertainties is 200 ppm. This results in a total soluble boron credit requirement for the all-cell configuration of 550 ppm. This value is well below the minimum SFP boron concentration value of 2000 ppm required by VEGP ~ - TS 3.7.17 and is, therefore, acceptable. Bumup reactivity equivalencing, as previously described, was also used to determine the allowed storage of fuel assemblies with enrichments higher than 2.45 w/o (VEGP Unit 1) and 12.40 w/o (VEGP Unit 2) but no greater than 5.0 w/o U-235 in the 3-out-of-4 configuration. The amount of soluble boron needed to account for the additional uncertainties assoc!ated with bumup credit in both units was 150 ppm. This is additional boron in excess of the 200 ppm required above, resulting in a total soluble boron requirement of 350 ppm. This is well below the minimum spent fuel pool boron concentration value of 2000 ppm required by TS 3.7.17 and is, therefore, acceptable. Storage of assemblies with enrichments higher than 3.50 w/o U-235 in the all-cell configuration in the MYAPP spent fuel storage racks was determined by crediting the reactivity decrease associated with the addition of integral fuel bumable absorbers (IFBAs). IFBAs consist of neutron-absort>ing material applied as a thin ZrB coating on the outside of the UO, pellet. The 2 l: fuel assembly is modeled at its most reactive point in life. This includes any time in life when l ' the IFBA has depleted and the fuel assembly becomes more reactive. As with bumup credit, - for IFBA credit reactivity equivalencing, a series of reactivity calculations is performed to L produce a set of IFBA rod number versus initial enrichment ordered pairs that all yield the. l. equivalent 4 when the fuel is stored in the all-cell configuration analyzed for the MYAPP spent L fuel racks in VEGP Unit 1. Uncertainties associated with IFBA credit include a 5 percent manufacturing tolerance and a 10 percent calculational uncertainty on the B-10 loading of the IFBA rods. The staff finds these uncertainties adequately conservative and acceptable. The amount of additional soluble boron needed to account for these uncertainties is 250 ppm. l Therefore, the total soluble boron credit required for the all-cell configuration in the MYAPP P spent fuel racks is 600 ppm. However, this is well below the minimum SFP boron concentration I value of 2000 ppm required by VEGP TS 3.7.17 and is, therefore, acceptable. As an altemative method for determining the acceptability of fuel assembly storage based on l, IFBA loading, the infinite neutron multiplication factor (k,,), was used as a reference reactivity L value. When k,is used as a reference reactivity point, the need to specify an acceptable ' ) enrichment versus the number of IFBA rods correlation is eliminated. Fuel assemblies with a reference k,, of 1.431 in the VEGP Unit 1 core geometry at 68 'F have been shown to result in I a maximum 4 s0.95 when stored in the MYAPP spent fuel storage racks; therefore, all fuel assemblies placed in the MYAPP spent fuel racks in an all-cell configuration must have an initial - _ nominal enrichment less than or equal to 3.50 w/o U-235, or must satisfy a minimum IFBA -{ requirement for higher initial enrichments to maintain the reference fuel assembly k,, less than l or equal to 1.431 at 68 'F in the VEGP core geometry. The VEGP Unit 1 SFP was also analyzed assuming a 3-out-of 4 checkerboard storage configuration containing three initially enriched 5.0 w/o U-235 assemblies and an empty cell (or '.a atored non-fuel-bearing component). This resulted in a 95/95 4 of 0.99745 with no credit for soluble boron. This value meets the NRC criterion of 4 less than 1.0 with no credit for soluble boron. The same configuration was then analyzed to obtain the required 5 percent suboritical 4 I' ! i

-_ margin' assuming 450 ppm of soluble boron. The resulting 95/95 k,was 0.95104. Since this k, value is less than 0.95,' including soluble boron credit and uncertainties at a 95/95 probability / confidence level, the NRC acceptance criterion is met for the 3-out-of-4-cells storage configuration. Although most accidents will not result in a reactivity increase, four accidents can be postulated - for each storage configuration that would increase reactivity beyond the analyzed conditions. . The first would be a change in the spent fuel pool water temperature outside the normal operstmg range. The second accident would be a misload of an assembly into a cell for which the restrictions on location, enrichment, or bumup are not satisfied. The third would be a drop of an assembly onto an already loaded cell. The fourth accident would be a misload between the spent fuel storage rack module and the spent fuel pool wall. Calculations have shown that the misloaded assembly accident between the rack module and pool wall in the 3-out-of-4 checkerboard results in the highest reactivity increase. The reactivity increase requires an additional 350 ppm of soluble boron above the 450 ppm normal condition requirement for the 3-out-of-4 configuration to maintain k,s0.95. However, for such events, the double' . contingency principle can be applied. This states that the assumption of two unlikely, independent, concurrent events is not required to ensure protection against a criticality accident. Therefore, the minimum amount of soluble boron required by VEGP TS 3.7.17 (2000 ppm) is more than sufficient to cover any accident, and the presence of the additional -soluble boron above the concentration required for normal conditions can be ass'umed as a ~ realistic initial condition since not assuming its presence would be a second unlikely event. i in order to prevent an undesirable increase in reactivity, the boundary between the two different storage configurations in Unit 1 was analyzed. The boundary can be either separated by a vacant row of cells or the interface must be configured so that the first row of cells after the boundary in the 3-out-of-4 storage region uses altemating empty cells and colls containing assemblies at the 3-out-of-4 configuration enrichment of up to 5.0 w/o U-235. Tne interface requirements are shown in proposed VEGP TS Figure 4.3.1-6, "Vogtle Units 1 and 2 Interface Requirements (All Cell to Checkerboard Storage)." On the basis of the preceding review, the staff finds that the criticality aspects of the proposed VEGP license amendment request are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling. The analysis assumed credit for soluble boron, as allowed by WCAP-14416-NP-A. The required amount of . soluble boron for each analyzed storage configuration is given in the following Table 1 summary. " The following storage configurations and U-235 enrichment limits for Westinghouse 17 x 17 fuel assemblies were determined to be acceptable for VEGP Unit 1: Assemblies with initial nominal enrichments no greater tnan 3.50 w/o U-235 can be stored in any cell location. Fuel assemblies with initial nominal enrichments greater than this and up to 5.0 w/o U-235 must satisfy a minimum bumup requirement as shown in proposed TS Figure 3.7.18-1, "Vogtie Unit 1 Bumup Credit Requirements for All Cell Storage," or must have a maximum reference fuel assembly k,,less than or equal to 1.431 at 68 *F. I i l

l Assemblies ~with initial nominal enrichments no greater than 5.0 w/o U-235 can be stored = in a 3-out-of 4 checkerboard arrangement with empty cells (or with cells containing non-fuelbaaring components). A 3-out-of-4 checkerboard means that no more than three fuel assemblies can occupy any 2 x 2 matrix of storage cells. TABLE 1 Summary of Soluble Boron Credit Requirements for Vootle Unit 1 Total Soluble Soluble Boron Boron Credit Soluble Boron Required for Required Storage Required for Reactivity Without Configuration k, s 0.95 Equivalencing Accidents (Ppm) (ppm) (ppm) All Cells 350 250 600 3-out-of-4 Checkerboard 450 0 450 3.2 Hoistino and Control of Heavy Loads The licensee plans to use the overhead bridge crane in the fuel handling building to move the MYAPP spent fuel racks to the SFP, and to install a new temporary gantry crane for moving the racks within the fuel handling building. An offset fuel handling toolis to be installed to reach spent fuel assemblies and move them to storage locations that are adjacent to the pool walls. -3.2.1 Hoistino System As indicated by the licensee, the load handling operations will be conducted in accordance with the guidelines in Section 5 of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," July 1980, as it relates to safe load paths, procedures, crane operator training, inspection and maintenance, and testing. The hoisting system consists of the overhead bridge crane in the fuel handling building, the temporary gantry crane to be installed, and lifting devices. The lifting devices consist of standard lift rigs and spreader bars that will be interposed between the racks and the crane hook during lifts.

m I 7 The overhead bridge crane in the fuel handling building is rated at 125 tons on the main hoist I . and 15 tons on the auxiliary hoist. The maximum weight of the spent fuel racks to be removed is 33,000' pounds and 25,075 pounds for the spent fuel racks to be installed. Because of the safe load path, some areas of the SFP are inaccessible using the fuel handling building's. overhead bridge crane. Therefore, the licensee will erect a temporary 20-ton gantry crane on the existing fuel handling bridge rails to move racks within the SFP area. The loading capacity of both the overhead bndge crane an,d the temporary gantry crane will enable the licensee to handle the storage racks during the rerack operations. Because all of the irradiated fuel will be stored in the Unit 2 SFP and the rerack is to be performed in the Unit 1 SFP, crane and load will not travel over irradiated fuel. Also both cranes will have mechanical stops to restrict crane ~ travel over new or spent fuel. The licensee has committed to further review the load paths to ensure adequate protection of safe-shutdown equipment. To protect safe-shutdown equipment, the licensee will either limit the maximum travel height, upgrade the hoisting system, or use redundant rigging in accordance with the guidelines of NUREG-0612. . The licensee also commits to load test the crane and the lifting devices in accordance with the guidance of NUREG-0612 and ANSI Standard B30.2, " Overhead and Gantry Cranes," 1976. Both the temporary gantry crane and the lifting devices are to be load tested at 125 percent of the maximum service load (rated load) before they are used. However, instead of load testing the crane by transporting the trolley for the full length of the bridge and runway as required by ANSI Standard B30.2, the licensee noted that a rack will be hoisted 6 inches above the floor and held for approximately 10 minutes before starting the rerack operation. This will enable the licensee to avoid transporting the test load over irradiated fuel to do the load test and satisfy the requirements of ANSI Standard B30.2. 3.2.2 Postulated Load Drop Accidents - As indicated by the licensee, the spent fuel racks will not be lifted over or close to spent fuel because the Unit 1 spent fuel assemblies will be moved to the Unit 2 SFP. In addition, load paths will be maintained at a maximum distance from the Unit 2 spent fuel.. Load paths will be established so that the travel and lift heights over safety-related equipment are minimized in accordance with guidelines of NUREG-0612. This will he!p to reduce u.e impact if a heavy load is dropped. The licensee presented a refueling accident analysis performed in accordance with NRC's " Office Technical Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14,1978. In that analysis, the licensee evaluated the consequences of an accidental drop of a spent fuel assembly from the highest lift point during fuel handling operations. The licensee found that no significant safety impacts would result from a dropped fuel assembly. Releases of radioactive material resulting from load drops could not occur due to the absence of irradiated fuel during rack installation in the Unit 1 SFP. Fuel damage could - not result in any increase in the subcriticality margin where K,is less than 0.95 because there is no irradiated fuel in the SFP. Damage to the SFP would not result in water leakage that could uncover the fuel; and the potential for damaging safe-shutdown systems is highly unlikely. ' This enables the licensee to satisfy the guidelines in Section 51 of NUREG-0612. i

. _ _ _ _ - - _ _ - _ _ _ _ - - _ _ - _ _ _ _ _ 1-l The licensee did not evaluate the drop of a rock during installation because irradiated fuel L . assemblies would not be present in the Unit 1 SFP or within the load path. The licensee stated ' l that there is no potential for the load handling accident to result in consequences that exceed the guidelines presented in Section 5.1 of NUREG-0612. As noted, the licensee will prevent accidental load drops by applying defense-in-depth measures described in NUREG-0612. These measures' include plans to train load handling system operators; steps to assure that the lifting devices and rigs are in accordance with the provisions of ANSI N14.6-1978, "Special Lifting Devices for Shipping Containers Weighing 10,000 lbs (4500 kg) or More," 1978; plans to conduct inspections cnd load testings in accordance with ANSI Standard B30.2; and plans to implement procedures to address the handling of specific heavy loads during the entire rarack i: operation. The licensee committed to verify the operability of all cranes and lifting devices before starting ' the raracking operation.' Both the crane / hoists system in the fuel handling building and the i temporary gantry crane would be verified for compliance with design and testing requirements of CMAA Specification No. 70, " Crane Manufactures Association of America Inc., Specification No. 70-Specification for Electric Overhead Traveling Cranes," 1975, and ANSI Standard B30.2. In addition to the testing, the licensee stated that it will develop various load handling procedures to assure compliance with NUREG-0612. The licensee's method of verifying that the hoisting system is functional, coupled with its procedures to minimize operator errors, rigging failures, and inadequate inspections, is acceptable to the staff. L 3.2.3 Conclusions Concernina Holatina and Control of Heavy I anda On the basis of the preceding discussion, the NRC staff finds that the licensee's methods of i handiing heavy loads during the rerack operation, including the licensee's commitment to verify the operability of the crane and lifting systems in accordance with the requirements for design and operation before performing the rerack, and the administrative procedures to improve the '] handling and control of heavy loads are in accordance with the guidelines of NUREG-0612.' 1 These changes enable the licensee to perform its rarack operation in a safe manner. l The licensee's evaluation of the consequences of postulated load drops of spent fuel storage racks and spent fuel assemblies satisfies the guidelines in Section 5.1 of NUREG-0612. The licensee has committed to use procedures and redundant rigging to prevent load drops that . could severely impact safe-shutdown equipment.' On the basis of the preceding discussion, the staff concludes that the temporary gantry crane and the upgraded lifting devices, testing of the hoisting system, operator training, and procedures for inspection and rack removal and l installation will reduce the probability of a load drop in the SFP to an acceptable level. Therefore, the proposed changes to the capacity of the SFP are acceptable from the standpoint of hoistbg and control of heavy loads. l l I i

p .g. l 3.3 u"-61 cc,m=_:.:.=v Cor='e-.::=.s L l The June 16,1982, NRC staff SE, concoming the MYAPP spent fuel racks, addressed the compatibility of the spent fuel rack materials with the SFP environment. The sections from the ' June 16 1982 NRC staff SE that relate to this issue are as stated: i .2.5.1'~ Materials Description I .We have reviewed the compatibility and chemical stability of the materials . (except the fuel assemblies) in the pool water. The proposed new spent fuel l storage racks are fabricated of Type 304 stainless steel with the exception of the adjusting bolts of the rack feet.. These bolts are made from Type 17-4 PH stainless steel. The 17-4 PH stainless steel [will] be heat-treated at 1100*F. The spent fuel storage pool contains high purity water with approximately 2,000 ppm boron as boric acid present in it. Tight controls are placed on impurities in L this water, such as chlorides and fluorides to minimize stress corrosion cracking l (SCC). The new high density fuel rack modules are composed of poison canisters and a, bottom grid. The poison canisters consist of two concentric stainless steel tubes- , with Boral neutron poisonous material in the annulus. Boral is Boron carbide in L an aluminum matrix core,' clad with 1100 series aluminum. '2.5.2 ChemicalCompatibility Leakage of water into the weld sealed Boral cavity due to weld failure is uniikely, since welds are made in accordance with ASME Code [American Society of j Mechanical Engineers Boiler and Pressure Vessel Code) procedures and both inspected and leak checked. Without the presence of water in the cavity, . hydrogen gas resulting from the corrosion of aluminum will not be pmsent. Even if some ' gas should form, the rack design utilizes the inner wall core as the structural member, so that'only the outer skin would bow from gas buildup, thereby preventing the fuel bundle, which is inside the canister, from being wedged and causing any dislocation of the Boral. Ifisolated cases ofleakage should ' occur, any swelling of the cans would not represent a safety hazard. 4 Upon exposure of the Boral plates (B.C/AL matrix) to the spent fuel pool water, galvanic coupling between the aluminum-Boral liner, aluminum binder and the l stainless steel shroud could occur. Deterioration of the Boral plates would be l limited to edge attack by general corrosion and pitting corrosion of the aluminum liner and binder in the general area of the leak. The B,C neutron adsorption s-particles are inert to the pool water and would become embedded in corrosion products preventing loss of the B,C particles. Thus, this small amount of l .c deterioration would have no effect on neutron shielding, attenuation properties or criticality considerations [ Fuel Storage Rocks Corrosion Test Program, Boral-l t l

L Stainless Steel Xn-NS-TP-009, Exxon Nuclear Company, Inc., October 1978, I' Richland Washington]. Boral neutron poison material encapsulated in stainless steel in a borated water coolant environment has been previously reviewed and accepted by us for similar designs in the Salem Nuclear Station and the Zion Nuclear Station. These plants have ongoing material surveillance programs which will timewise, lead the operation of the Maine Yankee Spent Fuel Pool in the unlikely event that any adverse service experience is noted in these surveillance programs there would be sufficient time to initiate corrective action for Maine Yankee. In - addition, the performance of other materials of which the spent fuel pool is constructed have been proven by experience and tests [ Reactor Handbook, Volume 1 - Materials, interscience Publishers,1960] to be stable and to operate n. L satisfactorily at both temperatures and radiation levels in excess of those anticipated in the Maine Yankee Spent Fuel Pool. Based on the above, we l . conclude that a materials surveillance program is not necessary in the case of [ the Maine Yankee Spent Fuel Pool. The pool liner, rack lattice structure, adjusting bolts and fuel storage canisters are stainless steel, which is compatible with the storage pool environment. In this environment of oxygen-saturated borated water, the corrosive deterioration ' ' of the type 304 stainless steel should not exceed a depth of 6.00 X 104 inch in

100 years [A. B. Johnson, Jr., " Behavior of Spent Nuclear Fuel in Water Pool 4

Storage" BNWL 2256, September 1977), which is negligible relative to the initial thickness. Dissimilar metal contact corrosion (galvanic attack) between the j stainless steel of the pool liner, rack lattice structure, fuel storage tubes, i L adjusting bolts and the inconel and the Zircaloy in the spent fuel assemblies will I not be significant because all of these materials are protected by highly L . passivating oxide films and are therefore at similar potentials. - 2.5.3 Concidsions

We conclude that the corrosion that will occur in the spent fuel storage pool environment should be of little significance during the remaining life of the plant -

L [C. Czajkowski, J. R. Weeks, et. al.,." Corrosion of Structural and Poison !~ Materialin Spent Fuel Storage Pools". Paper 163, Corrosion /81, April 6,1981.]. Components in the spent fuel storage pool are constructed of alloys which have a high resistance to general corrosion, localized corrosion, and galvanic corrosion.- We therefore conclude that the environmental compatibility and stability of the materials used in the spent fuel storage pool is adequate based on test data and actual service experience in operating reactors. We find that the L selechon of appropriate materials of construction by the licensee meets the - j requirements of 10 CFR Part 50, Appendix A, Criterion 62, preventing criticality. ' by maintaining structural integrity of components and is therefore acceptable. l As stated previously, the only modification performed on the MYAPP spent fuel racks, following issuance of the June.16,1982, NRC staff SE, was the addition of vent and drain holes in j I l

. ) pockets that contain the neutron-absorbing Boral material. Long-term contact between Boral and borated water is not expected to cause any problems (e.g., swelling or excessive corrosion) as demonstrated by the acceptable performance that the racks demonstrated at MYAPP. l The VEGP SFP environment is very similar to the SFP environment at MYAPP in that they both ' employ the same materials at similar temperatures and radiation levels. Both VEGP and MYAPP SFPs are lined with stainless steel and the pool water is of high quality and borated for criticality control. In addition, both MYAPP and VEGP siore Zircaloy-clad fuel in the spent fuel racks. From the preceding, the NRC staff concludes that use of the MYAPP spent fuel racks in l the VEGP Unit 1 SFP will be acceptable from a materials compatibility standpoint, as was 1 concluded for the use of these racks in the MYAPP SFP. l 3.4 Soent Fuel Coolina System Each VEGP SFP has an SFP cooling and purification system (SFPCPS). Each SFPCPS contains two subsystems; the SFP cooling system, and the SFP purification system. The l SFPCPS is designed to remove the decay heat generated by stored spent fuel assemblies and l to clarify and purify the water in the SFP. The primary safety function of the SFPCPS is to j adequately trrqspof. this heat load to the component cooling water (CCW) system and thereby i maintain the bulk ruol temperature within its specified limit. The system consists of two. i independent cmiing trains. Each train is seismically qualified and safety-related, and contains i one heat exchanger and one pump. Heat is removed from the SFP heat exchangers by the l CCW system. A purification loop, which includes a demineralized and a filter, removes fission i products and other contaminants that may be introduced if leaking fuel assemblies are l transferred to the SFP. A portion of either train of SFP water may be diverted through the i demineralized and filter at a rate of 100 gallons per minute (gpm) to maintain pool clarity and j purity. 3.4.1 Decav Heat Load Limit The decay heat load limit for the refueling cases discussed in the VEGP Updated Final Safety Analysis Report (UFSAR),~ Section 9.1.3, is applicable to both Units 1 and 2 SFPs. In the current UFSAR, a projected fuel discharge scheme was used to determine the loading of the Unit 2 pool, which has a capacity of 2098 assemblies. On the basis of this scheme, it was determined that with the pool essentially at capacity, the addition of a full-core offload would not result in the bulk pool temperature exceeding the licensing basis temperature of 171.1 *F. Slace fuel management (i.e., numbers of assemblies and bumups) may vary from cycle to cycle, the licensee determined that it would not be feasible to assume a single discharge scheme. Instead, in the proposed amendments, the licensee manages the total heat load in order to control the SFP temperature within a specified limit. Accordingly, the licensee performed the rerack heat load analysis in accordance with NRC Branch T ehnical Position (ASB) 9-2,

• Residual Decay Energy for Light-Water Reactors for

'Long-Term Cooling," Revision 2, in order to determine the maximum total heat load to control the bulk SFP temperature to within a limit of 170 *F for nonaccident conditions. This l temperature was chosen since it is within VEGP's previously licensed value of 171.1 *F. Since the SFP cooling systems for Units 1 and 2 SFPs are identical, the licensee plans to apply the

[; q same temperature limit and corresponding heat load to both pools and to manage the pool heat loads by means of administrative controls in plant procedures. The following conservatism are included in the licensee's decay heat load limit calculation: '(1) SFPCPS heat exchanger thermal performance is based on the design maximum fouling l level to minimize the heat rejection capability of the SFPCPS.' l (2) in calculating the SFP evaporation heat losses, the SFP building is assumed to have the maximum' ambient air temperature of 104 *F and 100 percent relative humidity to l minimize the credit for evaporative heat loss. 3.4.2 Maximum Normal Refuelina C=== - The normal practice at VEGP is to unload the entire core'for each refueling outage, which is referred to as the maximum normal refueling case in VEGP UFSAR Section g.1.3. In the VEGP B UFSAR heat load analysis, a full-core offload at 120 hours after shutdown is assumed with a - heat load limit of 54.1 x 10E6 BTU /hr. The bulk SFP temperature peaks at 171.1 *F, then the temperature falls to below 150 'F approximately 400 hours after discharging the entire core to the spent fuel pool. Since a refueling outage takes place about once every 18 months, this is equivalent to approximately 3 percent of the total cycle time. The licensee stated that usually ' more than half of the fuel that was offloaded will be reloaded into the core well before .400 hours. Therefore, the actual time for the temperature to be above 150 *F would be less j than 400 hours. In addition, the concrete walls and floor are several feet thick with a temperature gradient across them. Only a few inches of concrete would experience temperatures above 150 'F for short periods. For long-term durations between refuelings, the i bulk pool water temperature and the concre,te temperature would remain below 150 'F. For the maximum norrr.al refueling csse, the rerack heat load analysis revealed that a steady- ' state heat load of 51.87 x 10E6 BTU /hr would maintain the SFP bulk SFP temperature at j l h 170 'F with a single train of cooling in operation. Operating at or below this heat load will ensure that the SFP bulk spent fuel pool temperature would not exceed its limit of 170 'F for l nonaccident conditions. In the current UFSAR, the temperature of 171.1 *F and the-l l' corresponding heat load of 54.1 x 10E6 BTU /hr are similar to the values in the rerack analyses. Therefore, the licensee concluded that the time referenced above for estimating the maximum time the temperature would remain above 150 'F is bounding for the rerack analyses. The licensee stated that the pool temperatures for long-term operation would remain as described in ' the VEGP UFSAR, which is below the 150 *F aquirement in the American Concrete institute t (ACl) Standard 34g, " Code Requirements for Nuclear Safety-Related Concrete Structures,". 1985. Before offloading a core to the SFP, the licensee will evaluate the impact of the offload on the q total SFP heat load through plant procedures to ensure that the total heat load from earlier offloads and the recent core discharge do not exceed 51.87 x 10E6 BTU /hr The evaluations will include the necessary in-core delay times, prior to offloading spent fuel to the SFP, to ensure that this limit will be met.' l l i L-______--------_-----

.. The NRC staff performd a confirmatory decay heat load calculation and verified that the proposed decay heat IJad limit was acceptable. ' Also, the NRC staff verified that the long-term bulk SFP temperature of less than 150 'F was within the limit specified in ACI Standard 349. Therefore, the staff finds that the licensee's VEGP Unit i rarack analysis for the maximum normal refueling case is acceptable. l 3.4.3 uwimum Emernancv Core U,4="dCase The maximum emergency core unloading case for VEGP Unit 2 as described b the current UFSAR Section 9.1.3 assumes that the entire core is unloaded into the SFP at 150 hours after the emergency shutdown of the reactor, it also assumes that 84 assemblies from the most recent refueling, with a decay time of 36 days, and 1821 assemblies from earlier refuelings are - present in the SFP. For this case, the decay heat load limit is 58.13 x 10E6 BTU /hr to maintain the SFP temperature below 182 'F. In the rorack analysis, the licensee did not perform a new analysis for the maximum emergency core unloading case for VEGP Unit 1. Since the VEGP Unit 2 SFP has a capacity of 2098 assemblies, the heat load and temperature analysis for the VEGP Unit 2 SFP bounds the VEGP Unit 1 SFP with a capacity of 1476 assemblies. The licensee concluded that the Units 1 and 2 SFPCPSs are identical, so the bounding analysis for the emergency core unloading case -. applies to both units as it does in the VEGP UFSAR. The licensee plans to revise the UFSAR to reflect that the SFP temperature will be limited to 182 'F by controlling the heat load, and that a heat load evaluation will be performed before an emergency unloading of the spent fuel . assemblies.' nThe NRC staff finds that the Unit 1 maximum emergency core unloading case is bounded by the Unit 2 analysis with a single train of cooling and a bulk SFP temperature of less than 182 'F. This case is conservative since Standard Review Plan (SRP) Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System," only requires that bulk SFP boiling should not, occur .with two trains of cooling, Therefore, the staff finds that the licensee's Unit 1 maximum emergency core unloading case for the proposed amendments is acceptable. 34.4 Effects of SFP Bothng in the event that all forced SFP cooling becomes unavailable, the SFP water temperature will rise and eventually reach the normal bulk boiling temperature of 212 'F. The licensee determined that the minimum time to reach the boiling point is 2.90 hours by assuming that the decay heat load and the bulk SFP temperature limit are at their maximum calculated values. Since the SFPCPS has two independent trains, which are seismically qualified and safety l related, the probability of a complete loss-of-cooling event coinciding with the instant that the ' SFP water has reached its peak value is unlikely. J The licenses calculated the bolloff rate of the VEGP Unit 1 SFP at the decay heat load limit to be 5.347 x 10E4 lb/hr (approximately 111.5 gpm)J The primary source of makeup water for the SFP is the refueling water. storage tank (RWST), which serves as the seismic Category 1 i makeup water source that can be pumped or gravity-fed into the discharge line from SFP l L i

-_ - --.-_-- - _ - Pump A. The RWST has a total capacity of 715,000 gallons that can be provided to the SFP at a rate of 200 ppm within 2.90 hours. On the basis'of its review, the staff finds that the makeup rate to the SFP neeeds the bolloff rate, and the time in which the makeup water can be provided to the SFP occurs within the minimum " time-to-boil," as recommended by Regulatory Guide 1.13, "Epent Fuel Storage - Facility Design Bases," and SRP Section g.1.3; therefore, the staff finds thatihe licensee's j. time-to-boil analysis is acceptable. 3.4.5 Conclusion Concerr.lr,c Sonnt Fuel Pool Canlina -I On the basis of the preceding evaluation, the NRC staff's confirmatory decay heat load I [ calculation and the licensee's fulfillment of the commitments documented in Section 5.0, herein, the NRC staff concludes that the thermal-hydraulic aspects of the proposed amendments for increasing the capacity of the Unit 1 SFP from 288 to 1476 assemblies are acceptable. i 3.5 Radiological Assessment The NRC staff has evaluated the radiological aspects of the licensee's proposed reracking of the VEGP Unit 1 SFP as descibed in this section. 3.5.1 Occupational Radiation Exposure The staff has reviewed the licensee's plan for the replacement of the VEGP Unit 1 spent fuel racks with respect to occupational radiation exposure. As stated above, for this modification tim licensee plans to remove the two existing SFP rack modules and replace them with 26 rack modules that were previously licensed by the NRC for use at MYAPP, The licensee will then decontaminate t.he two SFP rack modules removed from VEGP and will take them from the site. On the basis of experience gained from SFP reracking operations peiformed at other. plants, the licensee estimates that it can perform the proposed raracking for approximately 4.3 person-rem. This dose estimate is based on the licensee's detailed review of the . anticipated work activities, durations, and expected dose rates associated with each of the activities associated with the SFP roracking. In order to achieve this dose, the licensee plans to closely monitor and control work, personnel traffic, and the movement of equipment to minimize contamination and to assure that i exposures are maintained as low as is reasonably achievable (ALARA). All activities will be 0 governed by radiation work permits, and all personnel will be provided with electronic personnel dosimeters. 1 Each diver will be monitored, using multiple teledosimetry devices to ensure accurate recording of their doses. These teledosimetry devices will transmit diver dose and dose rate data to a l. computer, which will display the data on a monitor near the SFP. These data will be monitored I continuously by a technician. Divers will be able to perform underwater area surveys using a remoto-readout radiation-monitoring instrument capable of measuring dose rates as high as j g 1000 rem /hr. I I I

f 15- - The hcensee will remove all spent fuel assemblies and all known sources of high radiation from the Unit 1 SFP before sending divers into the SFP. In addition, the licensee will close the weir gates connecting the Unit 1 SFP with the cask loading pit and the Unit 1 transfer canal. The

licensee will perform an extensive underwater radiation survey of the Unit 1 SFP before allowing divers access to the SFP to remove the old SFP storage rarAs. All divers will be fitted with a tether to control their movements in the SFP. The licensee will also use cameras to monitor diving operations.

l The 26 SFP rack modules that will be installed in the Unit 1 SFP wese obtained from MYAPP. These racks will be unpacked in a contamination control area having high-efficiency particulate air filtered ventilation. The racks will be surveyed, checked for hot spots, and decontaminated, if necessary, before installation in the Unit 1 SFP. The licensee will monitor and control personnel traffic and equipment movement in the SFP E L area to minimize contamination and generation of radioactive wastes. The licensee will use a combination oflong-handled and diver-controlled tools to facilitate SFP rack module removal and installation. The use of diver-controlled tools will reduce the need for decontamination during remoto-tool handling. During reracking operations, there is the potential for an increase in radioactivity, concentrations in the SFP from crud spalling from spent fuel assemblies during movement. ' in order to f ' minimize the effects of spelling in the SFP, the licensee will move all spent fuel to the Unit 2 SFP and will clean the racks to be removed before removing them from the Unit 1 SFP.' The . licensee also plans to use an underwater vacuum to minimize any potential radiological effects

of spalling and to maintain water clarity in the Unit 1 SFP.

The licensee estimates that the increased number of fuel assemblies stored in the Unit 1 SFP l ' may result in a small increase in doses in the areas adjacent to the sides of the SFP, although [ any increase will not be enough to change any existing radiation zone designations. To !~ minimize any potential dose rate increases from the increased storage of spent fuel, the licensee plans to control the placement of freshly discharged fuel so it is not placed in SFP rack positions adjacent to the sides of the SFP. Dose rates on the fuel pool level are primarily due to radionuclides in the pool water. During normal operations, dose rates in this area are generally 2.5 mrom/hr or less. The staff finds these dose rates to be acceptable and in accordance with SFP dose rates at other plants. The licensee does not expect the concentrations of airbome radioactivity in the vicinity of the SFP to increase because of the expanded SFP storage capacity. However, there will be a monitor in the ares to continuously monitor airbome radioactivity levels. In addition, the plant effluent radiation monitoring system will monitor any gaseous releases. . On the basis of its review of the licensee's proposal, the NRC staff concludes that the VEGP Unit 1 SFP rack modification can be performed in a manner that will ensure that doses to the workers will be maintained Al. ARA. The staff finds that the projected dose for the work . approximately 4.3 person-rom) is in the range of doses for similar SFP modifications at other ( plants, and it is acceptable. l a

p p [i 3.5.2 Sohd Radmactive Waste ' Spent resins are generated by the SFP purification system in order to minimize the generation of spent resins, the licensee will vacuum, inspect, and remove any debris from the floor of the SFP before installing the SFP rack modulesc Since the number of fuel assemblies handled in the pools annually at VEGP will not increase with the expanded storage capacity, the licensee concludes that the additional fuel storage will not result in a change of the amount of solid L redweste generated l L The existing spent fuel raci s in the VEGP Unit 1 SFP will be removed from the site by a K salvage company. After usable material has been salvaged, the volume of the remainder will L . be reduced and disposed of at the Bamwell, South Carolina, facility, in a worst-case scenario, L no salvageable material and no volume reduction, the resulting material would represent 44 percent of the expected solid waste volume associated with VEGP Units 1 and 2 for 1998; however, this volume is not significant when viewed over the 40-year, operationallifetime of the VEGP facility. - 3.5.3 Design-Basis Accidents i in the VEGP UFSAR, the licensee evaluated the possible consequences of the following three hypothetical accidents involvmg fuel in the SFP: a fuel handling accident in the fuel handling ~ builoing; a fuel handling accident in the containment with the airlock closed; and a fuel handling accident in the containment with the airlock open. The licensee evaluated these hypothetical accidents to determine the thyroid and whole-body doses at the exclusion area boundary (EAB), low-population zone (LPZ), and control room. The proposed reracking of the VEGP Unit 1 SFP did not affect any of the assumptions or data used in evaluating the dose i consequences of any of these hypothetical accidents. L - The NRC staff reviewed the licensee's analysis and performed confirmatory calculations to check the acceptability of the licensee's doses. !n performing these calculations, the staff used the assumptions in Regulatory Guide 1.25, " Assumptions Used For Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage E Facihty for Boiling and Pressurized Water Reactors." The NRC staff performed an assessment for the fuel handling accident with the most limiting dose consequences. : For a fuel handling accident in the containment with the airlock closed, the radionuclides release to the environment . will be mostly contained. For a fuel handling accident in containment with the airlock open, the l-release to the environment will be assumed to be released directly to the environment and, therefore, this accident will be bounded by the fuel handling accident in the fuel handling buildmg For a fuel handling accident in the fuel handling building, the radionuclides release is assumed to be released directly to the environment with no filtration; therefore, the NRC staff performed an assessment of a fuel handling accident occurring in the fuel handling building. i - For this accident, the staff assumed that the cladding of all of the fuel rods in a single fuel l assembly (264 rods) plus an additonal 50 rods (for a total of 314 rods) would be perforated if i the fuel assembly were dropped during handling. The damaged fuel assembly is assumed to contain freshly offloodec' fuel with a minimum of 100 hours of decay. The other parameters that the staff utilized in its assessment are presented in Table 2. i l I

l l ' L \\ l TABLE 2 ASSUMPTIONS USED FOR CALCULATING RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT IN THE FUEL HANDLING BUILDING Parameters Malut Power Level, MWt 3565 Number of Fuel Rods Damaged (Single Assembly + 50 Rods) 314 Total Number of Rods in Core 50,052 Shutdown Time, hours 100-L Power Peaking Factor 1.7 L Fission-Product Release Fractions (%)* lodine 12 Noble Gases 30-Pool Decontamination Factors

• lodine 100 Noble Gases.

1 - lodine Forms (%)* Elemental 75 . Organic 25 Filter Efficiencies for Control Room (%) 99 Control Room Flow Rates (ft*/ min)' Recirculation (emergency)i 15,600 Intake (emergency / normal) 1,500/3,000 l Unfiltered inleakage (emergency / normal) 15/15 Atmospheric Dispersion Factors, X Q (sec/m') /

Exclusion Area Boundary (0-2 hours).

1.8 x 104 ' Low Population Zone (0-6 hours) 3.1 x 104 i Control Room (0-4 hours) 5.7 x 104 i Core Fission Product inventories per TID-14844

• Regulatory Guide 1.25 1

l'

l l ! The staff's calculations confirmed that the thyroid doses at the EAB, LPZ, and control room from a fuel handling accident in the fuel handling building meet the acceptance criteria, and that l the licensee's calculations are acceptable. The results of the staff's calculations are presented s l in Table 3. For a fuel handling accident in the fuel handling building, the staff calculated a dose { of 54.5 rem to the thyroid at the EAB and 9.4 rem to the thyroid at the LPZ. The acceptance criterion at the EAB and LPZ for these accidents is contained in SRP Section 15.7.4 of NUREG-0800 and is 75 rem to the thyroid dose (25 percent of 10 CFR Part 100 guidelines of 300 rem). For the same accident, the staff calculated a dose of 5.0 rem to the thyroid of the f '~ l control room operator. For this calculation, the staff assumed that the control room emergency { ventilation system did not initiate until 2 minutes into the fuel handling accident (as per accident j description in the VEGP UFSAR). The acceptance criterion for the control room operator dose j l is 30 rem to the thyroid (SRP Section 6.4 of NUREG-0800). The NRC staff, therefore, finds the l proposed reracking of the VEGP Unit 1 SFP to be acceptable with respect to potential ) radiological consequences as a result of a hypothetical fuel handling accident. TABLE 3 q THYROID DOSES FROM FUEL HANDLING ACCIDENT. IN THE FUEL HANDLING BUILDING l AT VOGTLE. UNIT 1 NALUES CALCULATED BY NRC STAFF) l FUEL HANDLING ACCIDENT l i AREA DOSE (REM-THYROID) l I EAB* 54.5 l LPZ* .9.4 l Control Room ** 5.0 l l

• Acceptance Criterion = 75 rem thyroid

' ** Acceptance Criterion = 30 rem thyroid 3.6 StructuralEvaluation The NRC staff has reviewed the use of the MYAPP spent fuel racks in the VEGP Unit 1 SFP to assure the structural integrity and functionality of the racks, the stored fuel assemblies and the SFP structure subject to the effects of the postulated loads (Appendix D to SRP Section 3.8.4) and fuel handling accidents. 3.6.1 Storage Racks l The 1476 storage cells will be contained in 26 fuel storage racks, which are seismic Category I I I equipment, and are required to remain functional during and after a safe-shutdown earthquake 1 (SSE). The licensee, with assistance from its contractor, Holtec Intemational, performed structural analyses for the spent fuel storage racks.

1 -ig-The licensee used a computer program, DYNARACK, for dynamic analysis to demonstrate the structural adequacy of the VEGP spent fuel rack design under the combined effects of ~ earthquake and other applicable loading conditions. The proposed spent fuel storage racks are free-standing and self-supporting equipment (not attached to the floor of the ' storage pool). A nonlinear dynamic model consisting of inertial mass elements, spring elements, gap elements, ; , and friction elements,' as defined in the program, was used to simulate th,ee-dimensional (3-D) f dynamic behavior of the rack and the stored fuel assemblies,' including frictional and hydrodynamic effects. The program calculated nodal forces and displacements at the nodes,. and then obtained the detailed stress field in the rack elements from tho' calculated nodal - H

forces.

1 1 1 Two model analyses were performed; the '3-D' single rack (SR) model analysis and the 3-D whole-pool multi-rack (MR) model analysis. For the 3-D SR analysis, two rack geometries were considered for de calculation of stresses and displacements: l. (1) ~ 5.2 ft (W) x 7.7 ft (L) x 14.8 ft (H), and (2) 6.8 ft (W) x 7.7 ft (L) x 14.8 ft (H) where W, L, and H are defined as width, length, and height of a rack, respectively. Each rack was considered fully loaded, half-loaded, and almost empty, with three different coefficients of friction between the rack and the pool floor (p=0.2, 0.5, and 0.8, respectiv'ely) to identify the worst-case response for rack movement and for rock member stresses and strains. l In the 3-D MR analysis,26 free-standing racks were considered to investigate the fluid-structure ' interaction effects between racks and pool walls, as well as those among the racks. iThe seismic analyses were performed utilizing the direct integration time-history method. One set of three artificial time histories (two horizontal and one vertical acceleration) were generated from the design response spectra defined in the UFSAR~ The licensee demonstrated the - adequacy of the single artificial time history set used for the seismic analyses by satisfying . requirements of both enveloping design response spectra as well as by matching a target L power spectral density function compatible with the design response spectra as discussed in SRP Section 3.7.1. The licensee performed 85 3-D single-rack model analyses. The results of the a' nalyses show L that the maximum displacements of the racks at the top and the baseplate comers are about l '4.81 inches and 2.67 inches, respectively, indicating that there is adequate safety margin against overtuming of the racks and, thereby, the structural integrity and stability of the racks are maintained. In addition, the calculated stresses in tension, compression, bending, L . combined flexure and compression, and combined flexure and tension were compared with corresponding allowable stresses specified in ASME Code, Section ill, Subsection NF. The ^ comparisons show that all induced stresses under an SSE loading condition are smaller than the corresponding allowable stresses specified in the ASME Code, indicating that the rack design is adequate. In the 3-D MR analyses,26 fully loaded racks were considered and were subjected to the [ service, upset, and faulted loading conditions (Level A, B, and D Service Limits). The results of the MR analysis indicate that the calculated stresses on a rack are higher than those obtained from the single-rack analyses. However, all calculated stresses for the MR analyses are - (smaller than the corresponding allowable stresses of the ASME Code. L l L - _-__-_ - - _

l: f The licensee also calculated the weld stresses of the rack at the connections (e.g., baseplate-to-rack, baseplate-to-pedestal, and cell-to-cell connections) under the dynamic loading conditions. The licensee demonstrated that all the calculated weld stresses are smaller than the corresponding allowable stresses specified in the ASME Code, indicating that the weld L connection design of the rack is adequate. Based on (1) the licensee's comprehensive parametric study (a.g., varying coefficients of L fncten, different geometries and fuel loading conditions of the rack), (2) the adequate factor of safety of the induced stresses of the rack when they are compared to the corresponding allowable values given in the ASME Code, and (3) the licensee's overall structural integrity conclusions supported by both SR and MR analyses, the staff concludes that the rack modules will perform their safety function and maintain their structural integrity under postulated loading l; . conditions and, therefore, are acceptable. 3.6.2 Spent Fuel Storage Pool The SFP structure is made of reinforced-concrete and is designed as seismic Category 1..The dirr.ensions of the VEGP pool structure are approximately 34 feet wide, 50 feet long, and ~ 40 feet deep. The intemal surface of the pool structure is lined with 0.25-inch-thick stainless i l steel plates to ensure watertight integrity. I i K The pool structure was analyzed by using the finite element computer program ANSYS to 4 - demonstrate the adequacy of the pool structure under fully loaded fuel racks with all storage locations occupied by fuel assemblies. The fully loaded pool structure was subjected to the load combinations specified in the VEGP UFSAR. ' The May ig,1998, supplement shows the predicted factors of safety varying from 1.01 to 1.3g - for bending moments and axial forces of the concrete walls and slab. In view of the calculated factors of safety, the staff concludes that the licensee's pool structural analysis demonstrates the adequacy and integrity of the pool structure under full fuel loading, thermal loading, and SSE loading conditions. Thus, the SFP design is acceptable. 3.6.3 Fuel Handhng Accident 4 . The licensee evaluated the following two refueling accident cases: (1) drop of a fuel assembly with its handling tool, which impacts the baseplate (deep drop scenario) and (2) drop of a fuel assembly with its handling tool, which impacts the top of a rack (shallow drop scenario). l l The analysis of accident drop case 1 shows that the load transmitted to the liner through the rack structure is properly distributed through the bearing pads located near the fuel handling j -area; therefore, the liner would not be ruptured by the impact as a result of the fuel assembly i l drop through the rack structure. The analysis of accident drop case 2 shows that damage will be restricted to a depth of 13.5 inches below the top of the rack, which is above the active fuel 1 L region. The NRC staff reviewed the licensee's findings in the September 4,1997, application 'and concurs with them This is acceptable on the basis of the licensee's conclusions about i . structural integrity, supported by the parametric studies. [ {, l

. 3.6.4 Conclusions Concomina Strur*nral Eva!"dian On the basis of its review and evaluation of the licensee's application, as supplemented, the NRC staff concludes that the licensee's structural analysis and design of the spent fuel rack modules and the SFP structures are adequate to withstand the effects of the applicable loads, . including the effects of the SSE. The analysis and design are in compliance with the current licensing basis given in the VEGP UFSAR and applicable provisions of the SRP and, therefore, , ~ are acceptable. 4.0 TS CHANGES The licensee has proposed changes to the VEGP Units'1 and 2 TS to reflect the proposed increase in the VEGP Unit 1 SFP storage capacity and the revised criticality analysic, described ~ in Section 3.1, herein. The following revisions to the TS are proposed: .(1) TS 3.7.18 would be changed to reflect that separate criticality requirements apply to the Units 1 and 2 SFPs. Currently, TS 3.7.18 references the VEGP Udis 1 and 2 criticality requirements in TS 4.3.1. The proposed TS 3.7.18 references TS 4.3.1.1 for criticality - requirements in the VEGP Unit 1 SFP and TS 4.3.1.2 for criticality requirements in the VEGP Unit 2 SFP. '(2)- TS Figure 3.7.18-1, "Vogtle Unit 1 Bumup Cr' edit Requirements for All Cell Storage," . would be replaced with a revised figure based on the criticality analyses for the VEGP Unit i racks containing Boral as previously evaluated. (3) TS 4.3.1, " Criticality," would be separated into two sections,4.3.1.1 and 4.3.1.2, to ] address the design features and the criticality requirements for the VEGP Units 1 and 2 l SFPs, respectively. The criticality requirements for the VEGP Unit 2 SFP would not i change. (4) TS 4.3.?, ' Capacity," would be revised to increase the VEGP Unit 1 storage capacity 3 from 288 to 1476 assemblies. ) (5) TS Figure 4.3.1-4, "Vogtle Units 1 and 2 Empty Cell Checkerboard Storage ' Configurations," TS Figure 4.3.1-6, "Vogtle Units 1 and 2 Interface Requirements (All Cell to Checkerboard Storage)," and TS Figure 4.3.1-7, "Vogtle Units 1 and 2 Interface Requirements (Checkerboard Storage Interface)," titles would be revised to reflect the elimination of a 2-out-of-4 storage configuration for VEGP Unit 1. l (6) Administrative changes to the TSs are preposed to change the Table of Contents and I renumber the sections of TS 4.3, " Fuel Storage," to accommodate the separation of . TS 4.3.1 into proposed TS 4.3.1.1 and TS 4.3.1.2. The TS changes proposed as a result of the revised criticality analysis are consistent with l NRC-approved methodology. On the basis of this consistency with the approved methoc' ology ~ and on the preceding evaluation, the staff finds these TS changes acceptable. The proposed i

associated Bases changes adequately describe these TS changes and are also acceptable.

j

or ..... 5.0 LICENSEE COMMITMENTS RELIED UPON BY THE NRC STAFF in letters dated September 4,1997, May 19,1998, and June 12,1998, the licensee committed to the following: (1) The SFP heat loads will be managed by administrative controls. These controls will be placed in applicable procedures before transferring irradiated fuel into the VEGP Unit 1 SFP. (2) The UFSAR will be updated to include the heat load ' hat will ensure the temperature limit of 170 'F will not be exceeded, as well as the requirement to perform a heat load evaluation before transferring irradiated fuel to either pool. This will be included in the next appropriate UFSAR update following the installation of the VEGP Unit 1 spent fuel racks. (3) A temporary gantry crane, with a hoist rated for 20 tons, will be erected on the existing fuel handling bridge rails to move the racks within the SFP area. This commitment will be implemented before commencing reracking operations. (4) The licensee will implement all applicable crane, load path and height, rigging and load test guidelines of NUREG-0612 and ANSI Standard B30.2 before and during reracking operations, as appropriate. In the May 19 and June 12,1998, supplements, the licensee proposed that these commitments be incorporated in Appendix D of the VEGP Unit 1 Facility Operating License. In addition, Commitment 2 should be incorporated in Appendix D of the VEGP Unit 2 Facility Operating License. The NRC staff agrees that these commitments should be incorporated in Appendix D of the VEGP Units 1 and 2 Facility Operating Licenses in that fulfilment of the preceding commitments is necessary to maintain the integrity of the analyses associated with installation and use of the MYAPP spent fuel racks in the VEGP Unit 1 SFP.

6.0 STATE CONSULTATION

in accordance with the Commiscion's regulations, the Georgia State odicial was notified of the proposed issuance of the amendments. The State official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21,51.32, and 51.35, an Environmental Assessment and Finding of No S!gnificant impact was published in the Federal Reaister on June 24,1998 (63 FR 34491). Accc'dingly, based on the Environmental Assessment, the Commission has determined that issuance of the amendments will not have a significant effect on the quality of the human l environment. j

. _ ~ =

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by i operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. PrincipalContributors: Y. Kim L. Kopp D. Jaffe V. Ordaz B. Thomas i Date: :3une 29,1998 i L __}}