ML15112A865

Safety Evaluation Supporting Amends 72,72 & 69 to Licenses DPR-38,DPR-47 & DPR-55,respectively

ML15112A865
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Site:	Oconee
Issue date:	06/19/1979
From:	Office of Nuclear Reactor Regulation
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ML15112A864	List: ML15112A863 ML15112A865 ML15112A868 ML15112A869
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NUDOCS 7907250600
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4o oUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20655 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE MODIFICATION OF THE OCONEE UNITS 1/2 COMMON SPENT FUEL STORAGE POOL FACILITY OPERATING LICENSES NOS. DPR-38, DPR-47 AND DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 Dated:

June l9, 1979 79072 5oI44o

The spent fuel pool serving Units 1 and 2 is sized to accmr.odaLe a full core of irradiated fuel assemblies in addition to :he concur rent storage of the largest quantity of new and spent fuel assemblies predicted by the fuel management program.

Provisions are made in the Unit 1, 2 spent fuel pool to accommodate up to 750 fuel assemblies and in the Unit 3 spent fuel pool up to 474 fuel assemblies.

5.4.2.2 Spent fuel may also be stored i.n storage racks in the fuel transfer canal when the canal is at refueling level.

5.4.3 Except as provided in Specification 5.4.1.4, whenever there is fuel in the pool, the spent fuel pool is filled with water borated to the concentration that is used in the reactor cavity and fuel transfer canal during refueling operations.

5.4.4 The spent fuel pool and fuel transfer canal racks are designed for an earthquake force of O.lg ground motion.

REFERENCES FSAR, Section 9.7 Amendments Nos.

72, 72, 69 5.4 -2Z

INTRODUCTION By letter dated February 2, 1979, as supplemented April 20 and May 2, 1979, Duke Power Company (DPC or the licensee) requested an amendment to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55 for the Oconee Nuclear Station, Units Nos.1, 2 and 3. The request would revise the provisions in the Station's common Technical Specifications (TS) to allow an increase in Units Nos. 1 and 2 common spent fuel pool (SFP) storage capacity from 336 to a maximum of 750 fuel assemblies through the use of high capacity spent fuel storage racks.

The expanded storage capacity would allow the Oconee units to operate until about 1981 while still maintaining the capability for a full core discharge.

The major safety considerations associated with the proposed expansion of the SFP storage capacity for the Oconee Station are addressed below. A separate environmental impact appraisal has been prepared as part of this licensing action.

DISCUSSION AND EVALUATION Criticality Considerations The proposed spent fuel racks are to be made up of individual containers which are approximately nine inches square by 16 feet long. These containers are to be fabricated from 0.250 inch-thick, type 304 stainless steel.

The rack structure is designed to hold these square containers on a 13.75 inch pitch under safe shutdown earthquake accelerations. Thus, there will be over three inches of water between neighboring containers. The 13.75 inch pitch combined with the overall dimension of the fuel assembly, which is 8.52 inches, gives a fuel region volume fraction of 0.38 for the.storage, lattice.

DPC states that the highest anticipated U-235 enrichment is 3.5%.

This value was used in the neutron multiplication factor calculations.

This enrichment in the present fuel assemblies results in a fuel loading of 46.0 grams of U-235 per axial centimeter of fuel assembly.

As stated in DPC's February 2, 1979 submittal, the fuel pool criticality calculations are based on unirradiated fuel assemblies with no burnable poisons which have a fuel enrichment of 3.5 weight percent U-235. This corresponds to a fuel loading of 46.0 grams of U-235 per axial centimeter of these fuel assemblies. For the criticality calculations, it was also assumed that the water in the pool was pure., i.e., unborated.

Combustion Engineering's (CE's) CEPAK computer program was used to get the multi-group cross sections for the criticality analysis. The NUTEST computer program was used to calculate the self-shielding and flux advantage factors for the material heterogeneity, and the DOT-2W discrete ordinates transport program was used for the overall storage lattice cell calculations.

These computer programs were first used to calculate the neutron multiplication factor for an infinite array of fuel assemblies in the nominal storage.

lattice. The maximal effects of the stainless steel thickness tolerance,

-2 fabrication tolerances, fuel assembly positioning uncertainties, and water temperature were then calculated. The accuracy of these methods was checked by calculating the following sets of experiments:

1. The criticality of five, cold, clean pressurized water reactors (PWR's).

2. Stainless steel clad U02-H20 lattice experiments.

3. The LaCrosse Boiling Water Reactor critical experiments with stainless steel shrouds.

4. The reactivity worths of stainless steel reflectors on a uranyl fluoride solution reactor.

The results of these calculations indicate that the total uncertainty in the storage lattice cell calculations might be as large as 1.8% Ak; so DPC allowed this amount of margin in the design.

The above described results compare conservatively with the results of parametric calculations made with other methods for similar fuel pool storage lattices. By assuming new, unirradiated fuel with no burnable poison or control rods, these calculations yield the maximum neutron multiplication factor that could be obtained throughout the life of the fuel assemblies.

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

We conclude that all factors that could affect the neutron multiplication factor in this pool have been conservatively accounted for and that the maximum neutron multiplication factor in this pool with the proposed racks will not exceed 0.95. This is NRC's acceptance criterion for the maximum (worst case) calculated neutron multiplication factor in a SFP. This 0.95 acceptance criterion is based on the uncertainties associated with the calculational methods and provides sufficient margin to preclude criticality in the fuel.

Accordingly, there is a TS which results in a limitation of the effective neutron multiplication factor in the SFP to 0.95.

Conclusion on Criticality We conclude that when any number of the fuel assemblies, which DPC described in their submittals, having no more than 46.0 grams of uranium-235 per axial centimeter of fuel assembly or equivalent are loaded into the pro posed racks, the keff in the fuel pool will be less than the acceptance criteria of 0.95. We also conclude that in order to preclude the possibility of the Keff in the fuel pool from exceeding this 0.95 limit without being detected, the use of fuel assemblies that contain more than 46.0 grams of uranium-235, or equivalent, per axial centimeter of fuel assembly will be prohibited. On the basis of the information submitted and the Keff and fuel loading limits stated above, we conclude that the criticality calcula tions are acceptable.

-3 Spent Fuel Cooling The licensed thermal power for Oconee Units Nos. 1 and 2 is 2568 MWt each. DPC plans to refuel these reactors every 18 months at which times about 70 of the 177 fuel assemblies in the cores will be replaced.

To calculate the maximum heat loads in the SFP, DPC assumed a 168-hour time interval between reactor shutdown and the time when either the 70 fuel assemblies in the normal refueling or the 177 fuel assemblies in the full core offload are placed in the SFP. For this cooling time, DPC used the method given in NRC Standard Review Plan 9.2.5 to calculate maximum heat loads of 19.6 x 106 BTU/hr for a normal refueling and 31.7 x 106 BTU/hr for a full core offload.

Thespent fuel cooling system presently consists of two pumps and two heat exchangers. Each pump is designed to pump 1000 gpm (5.0 x 105 lbs./hr.),

and each heat exchanger is designed to transfer 7.75 x 106 BTU/hr from 125'F fuel pool water to 90'F Recirculating Cooling Water (RCW), which is flowing through the heat exchanger at a rate of 5.0 x 105 lbs./hr.

DPC states that this system will be sufficient to keep the SFP water temperature below 150 0F, the pool design temperature, until the first quarter of 1980 when an additional SFP cooling pump and heat exchanger of the same capacity will be installed. We find this acceptable.

Using the method given on pages 9.2.5-8 through 14 of the November 24, 1975 version of the NRC Standard Review Plan, with the uncertainty factor, k, equal to 0.1 for decay times longer 'than 107 seconds, "we calculate that the maximum peak heat load during the refueling which would fill the pool could be 20 x 106 BTU/hr and that the maximum peak heat loads for a full core offload that essentially fills the pool could be 34 x 106 BTU/hr. This full core offload was assumed to be a fully irradiated core which was taken out of its reactor vessel 35 days after the other Oconee unit, which shares this SFP, had been refueled. We also find that the maximum incremental heat load that could be added by increasing the 6

number of spent fuel assemblies in the pool from 336 to 750 is 1.9 x 10.

BTU/hr. This is the difference in peak heat loads for the present and the modified pools.

We conclude that with the three pumps operating, as DPC has committed to provide by the first quarter of 1980, the cooling system can maintain the fuel pool outlet water temperature below 125 0F for-the normal refueling offload that fills the pool and below 1360F for the full core offload that fills the pool.

In the highly unlikely event that all three SFP cooling systems were to fail at the time when there was a peak heat load from a full core in the pool, we calculate that the maximum heatup rate of the SFP water would be 9.0aF/hr. Thus, if the water were initially at an average temperature of 125 0F/hr it would be more than nine hours before boiling would start. We also calculate that after boiling starts the required water makeup rate will be less than 70 gpm. We conclude that nine hours will be sufficient time to establish a 70 gpm makeup rate.

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-4 Conclusion on Spent Fuel Cooling We conclude that the cooling capacity of the three loop system proposed by DPC for the Oconee Nuclear Station Units 1 and 2 SFP cooling system will be sufficient to handle the heat load that will be added by the proposed modifications. We also conclude that the incremental heat load due to this modification will not alter the safety considerations of spent fuel cooling from that which we previously reviewed and found to be acceptable.

Installation of Racks and Fuel Handling In their February 2, 1979 proposal,DPC states that at the time of the installation of the new racks there will be 140 spent fuel assemblies in the pool.

Initially, these will all be placed in existing racks at the south end of the pool.

This will allow the removal of approximately one third of the existing racks, which are at the north end of the pool, and the installation of two new racks without getting close to the spent fuel.

For the installation of the rest of the racks, DPC has developed a detailed procedure for redistributing the 140 fuel assemblies between the south end of the pool and the new racks in the north end of the pool so there will be a minimum of 14 feet of open space between the work area and.racks with fuel in them. Also, the plan is to move the racks in the pool at an elevation which is lower than the top of any stored fuel assemblies, such that there will be no movement of racks over stored fuel.

Conclusion on Fuel and Rack Handling We conclude that DPC's plan will insure than no racks will be moved over the spent fuel assemblies in the pool.

After the racks are installed in the pool, the fuel handling procedures in and around the pool will be the same as those procedures that were in effect prior to the proposed modifi cations. On this basis we conclude that the fuel and rack handling pro cedures are acceptable.

Structural and Mechanical The proposed modification consists of replacing the existing fuel -assembly racks with the CE supplied High Capacity (Hi-Cap) Fuel Assembly Rack, without changing the basic structural geometry of the SFP. Fourteen independent Hi-Cap fuel assembly storage rack modules are to be installed in the pool.

Each fuel assembly storage module is composed of an array of rectangular storage cavities or tubes, fabricated from one-quarter inch thick stain less steel plate, with each tube capable of accepting one fuel assembly. The fuel assembly storage tubes have lead-in surfaces in top castings to pro vide guidance for insertion of fuel assemblies. The tubes are open at the

-5 top and bottom to provide a flow path for convective cooling of the fuel assemblies by natural circulation. The fuel assembly storage tubes are structurally connected to a chevron grid structure to form the modules.

The chevron grid structure, placed at the bottom and upper elevations of the module, limits structural deformations and assures that a nominal center-to-center spacing of 13.75 inches is maintained between adjacent tubes for all design loading conditions, including seismic. Each storage rack module is self-supporting, and is supported by four U-channels, connected along the outer periphery of the base of the module, which in turn rest on bearing pads placed on the pool floor liner. All welded construction is used in the fabrication of the spent fuel rack assembly.

Load transfer to the pool structure from the fuel racks occurs only at the base of the racks, and consists of transmitting the vertical com pression loading and horizontal shear forces due to frictional restraint at both the module/pad and pad/liner interfaces.

The supporting arrangements of the modules, including their restraint, design, fabrication, and installation procedures; the structural design and analysis procedures for all loadings, including seismic and impact loadings; the load combinations; the structural acceptance criteria; the quality assurance requirements for design, fabrication, and installation; and applicable industry codes were all reviewed in accordance with the applicable portions of the current Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications, April 1978, including errata, January 1979.

The SFP is located in the Auxiliary Building. Seismic analysis was per formed using pool floor response time histories which conform to those approved in the original plant design. The pool floor response time histories were determined in the seismic analysis of the Auxiliary Building using a base acceleration time-history compatible with smoothed response spectra which.conform to the positions in Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants," and structural damping values which conform to the positions in Regulatory Guide 1.61, "Damping Values for Seismic Design of Nuclear Power Plants."

The pool floor horizontal time histories were then used as input to perform non-linear time history analyses of the lateral motion of the fuel racks.

The pool floor vertical time history was converted to a response spectrum for use in a vertical linear response spectrum analysis. The use of non-linear time history analyses in the horizontal directions was neces sitated by the non-linear characteristics of the fuel racks in the lateral directions. The methods of analyzing and combining responses for the racks in the three component directions are in accordance with Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic Response Analyses."

-6 In the spent fuel rack horizontal dynamic analysis, the effects of a gap between a storage cavity and a fuel assembly, and the effects of submer gence in water on the motion of the fuel racks were accounted for. The analysis was performed in two steps: In the first step, a modal extraction analysis of a detailed finite element model of the empty rack module in air was performed to determine its dynamic characteristics (e.g., natural frequency and mode shapes). In the second step, the modal parameters of the rack module were used to derive a dynamically equivalent spring-mass model of the module which was then incorporated into a lateral non-linear model which included the rack and contained fuel assemblies, and the water surrounding and contained within the cavities. This model considered the rack module, the fuel assemblies, the effect of impacting between the two; the hydrodynamic mass and coupling among the fuel, racks, and pool walls, friction between the fuel racks and pool floor; and rocking of the modules on their supports. The analysis was then performed to determine the dynamic response due to the effects of fuel impacting, hydrodynamic action, and the acceleration time history of the pool floor.

Non-linear time-history sliding base analyses were also performed to deter mine any potential impacting between adjacent fuel racks and between the racks and the SFP structure. Using the dynamic model discussed above, the motion of the racks relative to the pool floor was determined. The co efficients of friction used in the analysis, between the racks and the d pool floor, were based on test data for stainless steel in water provided in a report by P. Hoffman, "Wear Behavior of Friction Materials and Pro tective Layers with Regard to their Application Possibilities in Water Cooled Nuclear Reactors," ForderungsYorhaben BMFT-Inv. Reakt. 72/711, Kraftwerk Union, August 1973. This analysis resulted in conservative values for the rack sliding, and indicated that the ratios of horizontal displacement to the minimum available gaps between adjacent racks and between the racks and the nearest SFP structure are less than 0.11 and 0.22, respectively, and that the actual sliding distance will not exceed 0.133 inch. An additional analysis was made using an infinite friction coefficient to oltain a conservative value for the peak structural loading of the rack members and pool interfaces. These seismic loadings on the racks and the embedments, along with the maximum deflections, the maximum fuel assembly impact loadings, and the normal and thermal loads were con sidered in the design of the fuel racks.

Rack material properties used in the analysis of the spent fuel racks are in accordance with the requirements of Subsection NF and Appendix I of Section III of the ASME Boiler and Pressure Vessel Code.

Results of the seismic analysis show that the racks are capable of with standing the loads associated with all the design loading conditions with out exceeding allowable stresses.

-7 An analysis was performed to calculate the consequences of a fuel cask drop accident. The worst case was considered to be an eccentric drop onto the fuel pool wall from the design height of six feet. In this case the cask, yoke, and load block could be deflected onto the spent fuel.

The licensee has stated that the results of this accident would be that a maximum of 205 fuel cans could potentially suffer a total loss of integrity before the total energy of the falling cask is absorbed.

The radiological consequences of the cask drop are mitigated by limiting the age of fuel stored in the first 28 rows of the pool closest to the spent fuel cask handling area. Therefore, the proposed TS revision requires that no cask movement will be allowed until fuel in these locations has decayed a minimum of 55 days. Also, the licensee has indicated that the maximum possible drop height will be physically limited to four feet.

In addition, the modified TS, Section 3.8.14, prohibits the transport of loads greater than a fuel assembly with a control rod and the associated fuel handling tool(s).

The SFP is constructed of 'concrete walls and floor lined with one-half inch stainless steel clad plate. The fuel pool concrete reinforcing steel, liner plate, and welds are analyzed to account for any additional loads resulting frop the proposed increase in pool storage capacity. The de sign criteria were in compliance with Oconee Final Safety Analysis Report (FSAR) Appendix 5A for Class I structures. Results of an analysis for the most severe loading conditions indicate that the maximum loads are within the allowables, and that the fuel pool floor is adequate to with stand the effects of the new racks and additional fue.l.

Installation procedures for the new racks have also been reviewed.

Based on handling procedures described to prevent damage to the stored fuel and to prevent interaction between old and new racks, the installation pro cedures have been found to be acceptable to the NRC staff.

Materials The Type 304 stainless steel (ASTM Specification A-240) used in the new spent fuel storage racks is compatible with the storage pool environment, which is demineralized borated water controlled to a maximum 150'F tempera ture. Based on our review of previous operating experience with similar materials approved and in use, we have concluded that there is reasonable assurance that no significant corrosion of the racks, the fuel cladding, or the pool liner will occur over the lifetime of the units.

Conclusion on Structural, Mechanical and Materials The analysis, design, fabrication, and installation of the proposed new spent fuel rack storage system are in conformance with accepted codes and criteria. The analysis of the structural loads imposed by dynamic, static, seismic and thermal loadings; and the acceptance criteria for the appro priate loading conditions are in accordance with the appropriate portions of the NRC Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications, April 1978, including errata, January 1979.

-8 The mechanical properties for the materials used in the rack design are consistent with the normal and accident pool conditions. The quality assurance procedures for the materials, fabrication, installation, and examination of the new racks are in accordance with the accepted require ments of ASME Code,Section III, Subsection HF, Articles NF-2000, NF-4000, and NF-5000.

In addition, the design, procurement, and fabrication of the spent fuel racks comply with the pertinent requirements of Appendix 8 to 10 CFR 50,

'and delineated in Regulatory Guide 1.29, "Seismic Design Classification."

The effects of the additional loads on the existing pool structure due to the high capacity storage racks have been examined. The pool structure integrity is assured by conformance with the original FSAR acceptance criteria.

There is no evidence at this time to indicate that corrosion of the fuel assemblies, the stainless steel rack structures, or the fuel pool liner will occur over the lifetime of the plant, at the temperatures and quality of the denineralized borated water to be maintained in the pool.

We conclude that the subject modification proposed by the licensee is accept able and satisfies the applicable requirements of the General Design Criteria 2, 4, 61, and 62 of 10 CFR, Part 50, Appendix A.

Spent Fuel Cask Movement and Fuel Handling Accidents By letter dated April 20, 1979, the licensee proposed changes to Section 3.8 of the TS for Oconee Nuclear Station Units Nos. 1, 2 and 3. The licensee proposed specifications which restrict (1) the age of spent fuel stored near the cask handling area prior to spent fuel cask movement in the SFP and (2) the weight of loads that can be carried over spent fuel.

The current restrictions on the age of spent fuel stored near the cask laydown area in the Oconee Units 1 and 2 and Oconee 3 SFP's result from the Safety Evaluation (SE) dated September 1976. The licensee's proposal would (1) restrict the age of significantly more spent fuel than required in the 1976 analysis SE (even accountin for the increased density of a modified Oconee Units 1 and 2 SFP) and (2) specify a minimum age for spent fuel in the modified Oconee 1 and 2 SFP which maintains constant the potential consequences of a spent fuel shipping cask falling into the Oconee 1 and 2 SFP over the values given in the 1976 SE.,

9 In our SE dated September 1976, we assumed that 76 spent fuel assemblies may be damaged if a spent fuel shipping cask fell into the Oconee Unit 3 SFP and the minimum age for this damaged fuel was 43 days. The proposed specification 3.8.13.b requires that more than 76 spent fuel assemblies have a minimum of 43 days decay before spent fuel cask movement in the Oconee Unit 3 SFP. Based on this, and in that the potential consequences for the postulated accident are well within the exposure guidelines of 10 CFR Part 100, we conclude that the proposed Specification 3.8.13.b is acceptable.

In our SE dated September 1976, we assumed that less than 76 spent fuel assemblies may be damaged if a spent fuel shipping cask fell into the Oconee Units 1 and 2 SFP and the minimum age for this damaged fuel was 43 days. The licensee has determined that up to 205 assemblies may be damaged in the modified SFP. This is more assemblies than were assumed to be damaged based on the evaluation given in the SE for the increased capacity of the modified Oconee Units 1 and 2 SFP. Based on 205 assem blies being damaged, we would calculate a minimum age of 55 days for these damaged assemblies for the potential consequences of a postulated cask falling into the Oconee Units 1 and 2 SFP to not be greater than the values given in the SE dated September 1976. We have asked the licensee to specify 55 days as the minimum age of spent fuel stored near the cask handling area in Oconee Units 1 and 2 SFP in proposed Specification 3.8.13.a.

The licensee has agreed to this change. Based on this and on the potential consequences for the postulated accident of a cask falling into the Oconee Units 1 and 2 SFP being within the exposure guidelines of 10 CFR Part 100, we conclude that the proposed Specification 3.8.13.a is acceptable as modified by the NRC staff and agreed'to by the licensee.

The licensee has proposed Specification 3.8.14 to prohibit the transport of loads greater than a fuel assembly with control rod and associated handling tool over spent fuel in either the Oconee Units 1 and 2 SFP or Oconee Unit 3 SFP. This restriction on loads allowed over spent fuel will ensure that in the event the load is dropped, the activity release will be limited to'that contained in the equivalent of a single fuel assembly. We concluded, therefore, that the proposed Specification 3.8.14 of the Oconee TS is acceptable as written.

The NRC staff has under way a generic review of load handling operations in the vicinity of SFP's to determine the likelihood of a heavy load impacting fuel in the pool and, if necessary, the radiological consequences of such an event. Because Oconee Units 1 and 2 will be required topro hibit loads greater than 3000 pounds (the nominal weight of a fuel assembly, control rod and handling tool) to be transported over spent fuel in the SFP, we have concluded that the likelihood of any other heavy load handling accident is sufficiently small that the proposed modification is acceptable and no additional restrictions on load handling operations in the vicinity of the SFP are necessary while our review is under way.

-10 The consequences of fuel handling accidents in the SFP are not changed from those presented in the SE dated June 1973 for the SFP at Oconee Units 1 and 2 and are acceptable.

Occupational Radiation Exposure We have reviewed the licensee's plan for the removal and disposal o f the---_

low density racks and the installation of the high density racks with respect to occupational radiation exposure. The occupational exposure for this operation is estimated by the licensee to be about 75 man-rem.

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 average dose rate in the area where the job was being performed. In several instances, the licensee is conservative in his estimation of dose-rate and man-hours to perform a specific operation.

For example, although dose rates used to establish the collective (man-rem) exposure to many work groups is based on measurements that average 10 to 15 mrem/hr, the licensee is planning on reducing, or has already reduced, these dose rates by the following methods: (1) by adding two feet of water to the SFP to shield the crud "ring" around the pool; (2) by cleaning the walls of the pool near the pool water surface to remove the buildup of this crud "ring;" and (3) by using a skimmner and filter system to remove insoluble activity that is on the surface of the pool water. Based on the above, the staff concludes that the SFP modifications will be performed in A manner that will ensure as low as is reasonably achievable (ALARA) exposures to occupational workers.

The licensee is considering two methods of disposal of the old racks:

(1) cutting the old racks into small sections to significantly reduce the vol'ume to be shipped to the burial site or (2) crating the racks whole which will reduce the man-rem exposure involved with disposing of these racks. -Cutting the old racks into small sections will permit more efficient packaging in the shipping containers.

This will result in a smaller volume of radio active waste to be disposed of, with resulting economic-and environmental benefits, e.g., fewer waste shipments and conservation of low level waste burial site space. This will also require that the licensee expend effort to cut the old racks and will result in an increase in occupational'exposure.

The licensee has estimated that the occupational exposure to decontaminate the old racks and dispose of them whole would be 0.5 man-rem, while to decontaminate and cut the old rack into small'sections would be two man-rem.

the licensee has estimated that the burial costs for the old racks would be

$50,700 if they are cratedwhole (13,950 cubic feet, 28 boxes) and $3,500 if they.are cut into small sections (720 cubic feet). Therefore, in burial costs alone and not considering additional savings in shipping costs, cutting the racks into small sections represents a savings of over $47,000 for an estimated additional exposure of 1.5 man-rem. The licensee has stated that he will estimate the exposures associated with the different ways to dispose of the old racks from measurements of the dose rates from the old racks whe6 he has the racks outside the SFP, decontaminated and ready for disposal.

At this time, taking into account alternative disposal costs and exposures, the licensee will make the final decision as to the choice of method of disassembly and disposal of the old racks so that exposures will be kept to levels that are as low as is reasonably achievable (ALARA).

We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies on the basis of information supplied by the licensee for dose rates in the spent fuel area from radio nuclide concentrations in the SFP water and deposited on the SFP walls. The spent fuel assemblies themselves will contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel.

The occupational radiation exposure resulting from the additional spent fuel-in the pool represents a negligible burden. Based on present and projected operations in the SFP area, we estimate that the proposed modification should add less than 1% to the total annual occupational radiation exposure burden at this station. The small increase in additional exposure will not affect the licensee's ability to ma*intain individual occupational doses to as low as in reasonably achievable and within the limits of 10 CFR Part 20.

Thus, we conclude that storing additional fuel in the SFP will not result in any significant increase in doses received by occupational workers.

Radioactive Waste Treatment The station contains waste treatment systems designed to collect and process the gaseous, liquid and solid wastes that might contain radio active material.

The waste treatment systems were evaluated in the SE dated December 1970 for Oconee Unit 1 and in the SE dated July 1973 for Oconee Unit 2. There will be no change in the waste treatment systems or in the conclusions of the evaluations of these systems because of the proposed modification.

Conclusion on Cask Movement Fuel Handling, Occupational Exposure and Radioactive Waste Treatment Our evaluation supports the conclusion that the proposed modification to the Oconee Units 1 and 2 SFP is acceptable because:

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1. The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the SFP would be negligible.

2. The potential consequences of the postulated design basis accident for the SFP, i.e., the rupture of the fuel pins in the equivalent of one fuel assembly and the subsequent release of the radioactive inventory within the gap, are acceptable.

3. The likelihood of an accident involving heavy loads in the vicinity of the SFP is sufficiently small that no additional restrictions on load movement are necessary while our generic review of the issues is under way.

Based on the above, we conclude that the proposed Specifications 3.8.13 and 3.8.14 are acceptable with the minimum age of spent fuel near the cask handling area in the Oconee Units 1 and 2 SFP being 55 days.

Based on the above, we also conclude that the proposed modification of the Oconee Units 1 and 2 SFP is acceptable.

CONCLUSION We have 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 operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and that the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

Dated: June 19, 1979