ML19309H353
| ML19309H353 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/21/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309H348 | List: |
| References | |
| NUDOCS 8005130090 | |
| Download: ML19309H353 (13) | |
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8005180 010 4
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UNITED STATES NUCLEAR REGULATORY COMMISSION y-g WASHINGTON, D. C. 20555 5
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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 74 TO LICENSE N0. DPR-57 AND AMENDMENT N0.15 TO LICENSE N0. NPF-5 GEORGIA POWER COMPANY, ET AL.
EDWIN I. HATCH NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET N0. 50-321 AND 50-366 1.0 Introduction By application dated July 9,1979, and supplements thereto dated July 27, September 21, October 29, November 30, December 31, 1979 and February 18, 1980, the Georgia Power Company (the licensee) proposed to change the spent fuel pool (SFP) storage design for E. I. Hatch Nuclear Plant Units 1 and 2 (Hatch 1/2) from the design which was reviewed and approved in the operating license review and described in the FSAR.
The proposed change consists of increasing tha total spent fuel storage capacity of the SFP from 840 fuel assemblies.to 3171 fuel assemblies for Hatch 1 and from 1120 fuel assemblies to 2755 fuel assemblies for Hatch 2.
The increase will be accomplished within the existing spent fuel pool by the installation of spent fuel storage racks which utilize fixed neutron absorbers to allow higher density storage.
2.0 Discussion The prwosed spent fuel assembly racks are to be made up of alternat-ing stainless steel containers.
Thus, there will be only one con-tainer wall between adjacent spent fuel assemblies.
Each container wall is to have a sheet of Boral sandwiched between stainless steel, 0.0355 inch inside and 0.090 inch outside.
The containers will be about 14 feet long and will have a square cross section with an outer dimension of 6.653 inches and a total wall thickness of 0.2015 inches.
The nominal pitch between fuel assemblies will be 6.563 inches, which when combined with the fuel region dimension of 5.12 inches gives i
an overall fuel region volume fraction of 0.61.
The approximately 6.25 inch wide Boral sheets, which are to be in every container wall, are made up of a central segment of a 0.056 inch thick dispersion of boron carbide in aluminum, which is clad on both sides with 0.010 inches of aluminum. GPC stated in its July 9, 1979 submittal that the minimum, equivalent homogeneous, areal concentration of boron will be 0.013 grams of the boron ten isotope per square centimeter of Boral plate.
. 2.1 Criticality Analyses The General Electric Company (GE) performed the criticality analyses for GPC.
GE made the calculations with the MERIT Mor te Carlo program with cross sections which were processed from ENDF/B-IV data. The accuracy of this calculational method was assessed by using it to calculate representative critical experiments.
From this qualification program, by GE determined that this calculational method underpredicts Keff 0.5 percent A.
GE used these computer programs to calculate the neutron multiplication factor for an infinite array of fuel assemblies in the nominal storag' lattice at 20 C.
The calculational model of the fuel assembly which uE 0
used for these spent fuel pool calculations had discrete, unirradiated fuel pins with varying amounts of uranium-235 in them with a total fuel loading of 15.2 grams of uranium-235 per axial centimeter of fuel assembly.
There was no burnable poision included in any of the fuel pins.
The Boral plates, which were also discretely represented in the calculational model, had the quoted minimum concentration of boren in them, i.e., 0.013 grams of boron - ten per square centimeter of plate.
For the minimum possible as-built pitch (i.e., 6.503 inches)
GE's calculated value for this storage lattice k.
is 0.87.
GE then calculated the k.
's for the following conditions: (1) increasing the temperature to 650C; (2) increasing the lattice pitch; (3) locating every four fuel assemblies as close together as possible; and (4) reducing the density of the water.
GE found that all of these changes resulted in a decrease in k..
Because of the alterna' ting lattice design, wherein there will be only one storage container for every two fuel assemblies, there will be spaces on the periphery of the rack modules which will not have Boral plates.
Thus, it Will' be possible for two rack modules to be put together so that adjacent fuel assemblies will not have a Boral plate between them.
GE calculated the effect of these missing Boral plates for the minimum attainable gap between rack modules and found that it would not increase the maximum k of 0.87.
GE also analyzed the situation where the defective fuel storage space, which are attached externally to some of the storage modules, are all filled with fuel assembl ies.
GE st'ates that these analyses have demonstrated that the keff is <0.95.
In regard to an onsite neutron attenuation test to verify the presence of the boron plates, GPC states the following:
"The presence of the neutron absorber material in the fabricated fuel storage module will be verified at the reactor storage pool site by scanning each storage tube in the modules with a neutron source and neutron detectors.
The recorded results provide a comparison between neutron absorption through each Boral sheet and neutron absorption measured through the stainless steel without Boral.
A signi-ficant_ increase in neutron absorption verifies the presence of Boral."
_3 2.1.1 EVALUATION GE's use of a non-uniform distribution of fuel in the criticality calculations for the spent fuel pool results in a lower value of the neutron multiplication factor than would have been obtained with a.
unifom fuel distribution.
If a uniform distribution of fuel had been used, the calculated value of the neutron multiplication factor would have been about 0.91 instead of 0.87.
The maximum possible neutron multiplication factor in these pools is higher than 0.87.
This maximum value is set by the comppsition of fuel assemblies which are placed in the defective fuel storage locations and in the adjacent locations.
By assuming new, unirradiated fuel with no burnable poison or control mds, 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.
Because the neutron multiplication factor is not measured in spent fuel pools, the only available value is a calculated one.
This calculation is complex, ard it has many inputs and possible uncertain-ties. Thus, the NRC staff is required to review these calculations to determine that the uncerta:nties in the calculated neutron multiplication factor are not excessive. Accordingly, we have reviewed the calculations l
which were made for the fuel assemblies which are described in this submittal.
We find that all factors that could offect the neutron multiplication factor of the fuel assemblies, which are described in this submittal, in the spent fuel pools with the proposed racks have been accounted for and that the neutron multiplication factor will not exceed 0.95.
This is NRC's acceptance criterion for the maximum (worst case) calculated neutron multiplication factor in a spent fuel pool.
Accordingly, there is a technical specification which limits the neutron multiplication factor (keff) in the spent fuel pool to 0.95.
GPC may want to store fuel assemblies other than those described in this submittal in the proposed storage racks.
Since the maximum neutron multiplication in the pool could be detemined by the composition of these other fuel assemblies, the NRC required an additional technical specification to preclude any unreviewed increase, or increased uncer-tainty, in the calculated value of the neutron multiplication factor which could raise the actual neutron multiplication factor in the fuel pool above 0.95 without being detected.
An acceptable fom of this additional technical specification is to limit' the fuel loading of any assembly stored in the proposed racks to that of the fuel assembly which was modeled for the calculations in this submittal.
. In regard to G C's onsite neutron attenuation testing of the Boral plates, we f t..d that with the quality assurance program procedures in effect there should be no Boral plates missing from the prescribed locations in the fabricated fuel storage modules.
If GPC finds any Boral plates missing it shall specifically note and document this finding in its test report, and report it to the NRC.
2.
1.2 CONCLUSION
We find that when any number of the fuel assemblies, which GPC described in these submittals, which have no more than 15.2 grams of uranium-235, or equivalent, per axial, centimeter of fuel assembly, are loaded into in the fuel pool will be less than the the proposed racks, the keff 0.95 limit.
We also find that in order to preclude the possibility of in the fuel pool fmm exceeding this 0.95 limit without being the keff detected, it is necessary pending an NRC review, to prohibit the use of these high density storage racks for fuel assemblies that contain more than 15.2 grams of uranium-235 per axial centimeter of fuel assembly.
On the basis of the information submitted, and the keff and fuel loading limits stated above, we conclude that the health and safety of the public will not be endangered by the use of the proposed racks.
2.2 SPENT FUEL COOLING The licensed thermal power for each of the two Hatch reactors is 2436 MWth.
GPC plans to refuel these reactors annually at which times about 140 of the 560 fuel assemblies in each core will be offloaded.
To calculate the maxirr.am heat loads in the spent fuel pool GPC assumed a 150 hour0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> (6.25 day) time interval between reactor shutdown and the time when either the 140 fuel assembiies in the normal refueling or the 560 fuel assemblies in the full core offload are placed in the spent fuel pools.
For this cooling time GPC used the method given in 11.6 x 10gandard Review Plan 9.2.5 to calculate maximum heat loads ofBTV the NRC S x 106 BTU /hr for the full core offload which fills the pool after nineteen annual refuelings.
This calculation was made for Unit I which is to have the larger amount of spent fuel stored in it and consequently the greater heat load.
There are three trains of spent fuel cooling at the E.I. Hatch Nuclear Plant.
There is one train for each of the tv units and another one, called a swing train, which can cool either of the two spent fuel pools.
Each cooling train consists of a pump and a heat exchanger. The FSAR for Unit I states that the pumps for its cool {ng train and the swing train are designed to pump 610 gpm (3.05 x 10 pounds per hour) and 6
that each of the two heat exchangers is designed to transfer 4.25 x 10 BTU /hr from 1250F fuel pool water to 1050F Reactor Building Component Cooling Water (RBCCW), which is flowing through the shell side of the
5 pounds per houti.
heat exchanger at the rate of 1200 gpm (6.0 x 10 The FSAR for Unit 2 states that its spent fuel pool cooling pumc is 5 pounds per heat exchanger is designed to transfer 4.25 x 10pur) and that its designed to pump 650 gpm (3.25 x 10 BTU /hr from 1250F fuel pool water to 1050F RBCC water which is flowing through the shell side of the heat exchanger at the rate of 1200 gpm.
GPC states that this system, with two trains operating, for either of the two spent fuel pools, will be able to keep the outlet water temperature at or below 1330F through the final annual refueling.
For cooling an offloaded full core GPC states that a single train of the Residual Heat Removal (RHR) system, when aligned to either of the fuel pools, will, by itself, maintain the spent fuel pool outlet water temperature at or below 145 F.
GPC states that, when the reactor vessel 0
head and the spent pool gates are removed, the,RHR system can be aligned to the spent fuel pool by installing two spectacle flanges and operating four isolation valves. The estimated time for this realignment is eight hours.
in regard to emergency m;ake up water for the spent fuel pools, GPC states that the water level in the fuel pools will be maintained by the Seismic Category I Plant Service Water system in each unit.
2.2.1 EVALUATION Using the method given on pages 9.2.5-8 through 14 of the NRC Standard Review Plan, with thq uncertainty factor, K, equal to 0.1 for decay times longer than 103 seconds, we find that GPC's calculated values for the maximum peak heat loads in the modified spent fuel pools are conservatively high.
From our calculations we also find that the maximum incremental heat load that could be added by increasing the number of spent fuel assemblies in the Unit 1 pool from 840 to 3181 will be 2.3 x 106 BTV/hr. This is the difference in peak heat loads for full core offloads that essentially fill the present and the modified pools.
Since more spent fuel is to be stored in the Unit 1 pool, the incremental heat load for the Unit 2 pool will be smaller.
We calculate that with two trains in operation, the spent fuel pool cooling system of Unit 1 can maintain the fuel pool outlet water tem ature below 133 F for a peak annual refueling heat load of 11.6 x 10ger-0 BTU /hr.
Since the Unit 2 pool will have a smaller heat load its outlet water temperature will be less. We find that when the RHR system of Unit 1 is aligned with its spent fuel pool it will, by l
l itself, have sufficient capacity to keep the temperature of the outlet 6
water below 1500F for a peak, full core, heat load of 28.7 x 10 BTU /hr.
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. Since the peak heat load for Unit 2 is somewhat smaller, its RHR system can also keen the outlet water below 1500F.
Because, as stated in GPC's submittal, the Residual Heat Removal System (RHR) and the spent fuel pool cooling system piping that is exposed to RHR flow are designed to Seismic Category I criteria, the cocling of a full core offload in the spent fuel pool would not be precluded due to a seismic event. Thus, the worst credible accident for spent fuel cooling at either Hatch Units 1 or 2 is the complete loss of cooling after a normal refueling and a resumption of power operation. For this situation the heat up rate of the water in either of the spent fuel pools will be less than 50F/hr.
Assuming a maximun fuel pool temperature of 1330F there will be a minimum time period of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> before bulk pool boiling commences. After this the maximun boil off rate will be 24 gpm.
We find that sixteen hours will be sufficient for GPC to establish a 24 gpm make up rate from the Unit's Seismic Category I Plant Service Water system.
We also find that under bulk boiling conditions the temperature of the fuel will not exceed 350 F.
This is an acceptable temperature from the stand-0 point of fuel element integrity and surface corrosion.
2.2.2 Spent Fuel Pool Cleanup System The SFP cooling and cleanup system consists of a filter vessel, a resin trap, a holding pump, a precoat mixing tank and pump and the required piping, valves and instrumentation. The entire fuel pool cooling water flow is processed through the filter /demineralizer system. The normal flowrate will produce approximately four complete water changes per day of the fuel pool. The filter is enclosed in a shielded cell to keep exposures of plant personnel to minimum levels of radiation.
Because we expect only a small increase in radioactivity released to the pool water as a result of the proposed modification as discussed in Section 4.1, we conclude that the SFP purification system will keep concentrations of radioactivity in the pool to levels which have existed prior to the modification.
2.2.3 Conclusion We find that the present cooling capacities in the spent fuel pools of the Edwin I Hatch Nuclear Plant will be sufficient to handle the incremental heat loads that will be added by the proposed modifications. We also find that these incremental heat loads will not alter the safety considerations of spent fuel pool cooling from that which we previously reviewed and found to be ecceptable. We conclude that there is reasonable assurance that the health and safety of the public wi11 not be endangered by the use of the proposed design.
i
o 7-2.3 Installation of Racks and Fuel Handling There are presently no fuel assemblies in the E.I. Hatch Unit 2 pool.
It is GPC's intention to install the new racks without having fuel assemblies in the Unit 2 pool, There are about 260 spent fuel assemblies in the Unit 1 pool, but the two pools at Hatch are connected by a fuel transfer canal.
Thus, af ter the new racks are installed in the Unit 2 pool the spent fuel assemblies in the Unit 'l pool can be transferred to the new racks in the Unit 2 pool prior to installing the new racks in the Unit 1 pool. Thus, both of these rack changes can be done without fuel assemblies in the pool.
2.3.1 Evaluation Since there will be no fuel assemblies in the fuel pool during the modifica-tion, it will not be possible for an accident to result in any increased j
neutron multiplication factor After the racks are installed in the pools, s
the fuel handling procedures that will be implemented in and around the pool will be the same as those procedures that were in effect prior to the modifi-c a ti.ons. These were previously reviewed and found acceptable by the NRC.
2.3.2 Spent Fuel Handling The NPC staff has under way a generic review of load bandlino operations in the vicinity of spent fuel pools to deternine the likelihood of a beavy load impactino fuel in the pool and, if necessary, the radiological consecuences of such an event.
Because the rain crane meets the reovirenents of NUPEG-0554 we have concluded that the likelihood of a heavy load handlino accident is sufficiently small that the proposed rodification is acceptable and no additional resprictions on load handling operations in the vicir'ty of th'e SFP are necessary while our review is under way.
The consecuences of fuel handlino accidents (i.e., rupture of all the fuel pins 'in the'eauivalent of one fuel assembly and the subseouent release of the radioactive inventory within the pap) in the spent fuel pool are not chanced from those presented in the Safety Evaluation * (SE) dated June 13, 1978.
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2.3.2 Conclusion t
9 We conclude that there is reasonable assurance that the health and safety of the public will not be endangered by the installation and use of the proposed racks.
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e 2.4 Structural and Mechanical Design Georgia Power Company proposes to replace existing spent fuel storage racks at its Edwin I. Hatch Nuclear Plant Units 1 and 2 with racks of increased capacity.
The present Hatch 1 and 2 spent fuel pools contain racks that can hold 840 and 1120 fuel assemblies, respectively. The replacement of high density spent fuel storage racks will provide 3171 storage spaces in Hatch 1 and 2755 in Hatch 2.
The proposed modifica-tion will provide storage capacity up to the year 1997 with a full core reserve, assuming annual quarter reloads. The replacement spent fuel storage racks are to be fabricated primarily from type 304 stain-less steel. The individual fuel assemblies will be stored in square fuel storage cells formed from an inner shroud of stainless steel, a center sheet of boral and an outer shroud of stainless steel.
The cells act as a storage space and provide neutron absorption from the boral sheet to allow spacing of fuel in a 6.5 inch by 6.5 inch array.
The outer and inner tubes are welded together after being sized to the required dimensional tolerances. The completed storage tubes are fastened together by angles welded along the corners and attached to a base plate to form storage modules.
The base plate of each module is supported on all four corners by 2-inch thick foot pads. The foot pads rest on 6-inch thick corner-support pads which in turn rest on the fuel pool floor liner. This raises the base plate of the module a minimum of 8 inches above the floor of the fuel pool, allowing sufficient clear area to permit natural circulation of cooling water to the modules without taking credit for sources of forced cooling.
2.4.1 Evaluation The proposed modification for the spent fuel storage capacity ' xpansion e
program has been reviewed in accordance with the NRC report "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applica-tions," April 1978. The structural and mechanical review consisted of an examination of the following areas:
the proposed design criteria, the design loads and load combinations, methods of analysis, the dropped fuel assembly accident, the material properties and allowable stresses of type 304, stainless steel, the hydro-dynamic effects, the fabrication and installatio'n provisions, and the effect of increased loads on the
. floor slab and liner.
The high density spent fuel module has been analyzed for both OBE and SSE conditions.
A damping ratio of 2 percent of the critical was used in the analysis for the SEE condition and 1 percent for the OBE condition i
without any added damping from fluid effects. Seismic spectra are based on Hatch 2 which bound the spectra for Hatch 1.
Combination of the modal response and the effect of the three components of an earth-quake has been performed in accordance with the applicable provisions of USNRC Regulatory Guide 1.92.
The total mass of the module including l
l
9-the hydrodynamic effect, which represents the inertial properties of the fluid surrounding the submerged modules, was calculated and then used to perform the seismic analysis.
Stress analyses were done for both OBE and SSE conditions based upon the shears and moments developed in the finite-element dynamic analysis of the seismic response. These values were then compared and found to be within the allowable stresses referenced in ASME Section III, Subsection NF.
Seismic loads obtained from the response spectra analysis were increased by impact factors which account for the effects of the clearance gap between the storage cells and the fuel assemblies contained therein.
The sliding analysis was done using the two-dimensional, non-linear DRAIN-2D and SEISM computer codes.
DRAIN-2D was originally developed at the Ur,1versity of California at Berkely, SEISM was developed by GE.
Both computer codes meet NRC-QA requirements.
Sliding and overturning of the module were studied for the SSE and OBE conditions. All of the modules were found to be stable under the worst postulated seismic loading conditions, and the minimum 2-inch clearance between modules precluded contact during a seismic event.
The spent fuel pool structure was re-evaluated based on the increased loads caused by the new high density spent fuel storage racks. The ACI code 349-76 " Code Requirements of Nuclear Safety Related Concrete Structure" as supplemented by USNRC Regulatory Guides and positions was used as the design basis for the structural re-evaluation.
A three-dimensional methematical model was developed for each spent fuel pool structure.
Each mathematical model is composef of plate shell elements, beam elements, truss elements, and boundary elements to idealize the existing structure.
Structural properties for the elements were selected based on in-situ conditions.
The re-evaluation for the Unit 1 spent fuel pool structure showed that the existing structure and liner plate would have adequate capaci+.y to carry the additional loads imposed by the high density spent fuei racks.
Fuel assembly drop accidents were analyzed using analytical methods in accordance with the " Operating Technical Position for Review and Accept-ance of Spent Fuel Storage and Handling Applications".
In estimating local damages in the nodule, the maximum strain energy resulting from plastic deformation was equated to the maximum potential energy of the fuel.
Energy dissipation attributable to the viscosity of the water and plastic deformation of the fuel bundle was ignored for conservative results. The stainless steel for the module was assumed to exhioit a bi-linear hystresis relationship, with yield stress and ultimate stresses as the two control points, t
. A free fell of a fuel assembly into the fuel pool il=r was evaluated to serve as a basis for concluding that the leak tightness of the fuel pool liner plate is maintained. The evaluation demonstrated that the energy developed by a freely falling fuel assembly from a height extending 27 inches above a module would not cause liner plate perforation.
Specifications which impose quality control requirements during the design, procurement, fabrication, installation, and testing of the storage system were developed specifically for the high density spent fuel.
Periodic audits of the various facilities and practices are performed by certified quality assurance personnel to ensure that these Quali ty Assurance / Quality Control requirements are being met. All welding and nondestructive examinations will be done in accordance with the applicable provisions of the ASME Boiler & Pressure Vessel Code,Section IX, and the American Society for Nondestructive Testing.
2.4.2 Material Considerations The Type 304 stainless steel used in the new spent fuel storage racks is compatible with the storage pool environment, which is demineralized water.
Based on our review of previous operating experience with similar materials approved in use, we have concluded that there is reasonable assurance that no siW': cant corrosion of the racks, the fuel cladding, or the pool liner will occur over the lifetime of the plant. The aluminum in the Boral neutron absorber plates will experience some galvcnic corrosion with the stainless steel tubes encapsulating the Boral being vented to the pool water environment, although in the high resistivity pure water environment any galvanic action will be mini-mized. The more noble stainless steel will not be affected by any galvanic attack when contacted with aluminum. Although slight galvanic corrosion may occur in the aluminum of the Boral plates, it should not have a.y significance on the neutron absorption capability of the Boral and certainly no effect on storage rack structural integrity for a period in excess of 40 years.
2.4.3 Conclusion The structural, mechanical and material aspects of the spent fuel storage racks have been evaluated based upon NRC guidance provided in the report entitled, "0T Position for Review and Acceptance of Spent Fuel Storage and llandling Applications," April 1978.
Based upon our review of_the analyses and the design done by the licensee, we conclude that the rack structure itself, the supporting pool liner a'nd slab, are capable of supporting _ the applied loads without exceeding relevant stresses of Subst.ution NF or the FSAP. Design Criteria. As previously stated, we find the material fabrication, installation, and examination criteria acceptable. We Conclude that the proposed modifications to-the Edwin-Hatch spent fuel storage are in conformance with NRC requirements.
. 2.5 Occupational Radiation Exposure We have reviewed the licensee's plan for the removal and disposal of the low density racks and the installation of the high density racks with respect to occupational radiation exposure.
The occupational exposure for this operation is estimated by the licensee to be about 14 to 28 man-rem as explained below.
We consider this to be a reasonable estimate because it 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 perfonned.
The modi. cation will first be performed in Hatch 2, which is at the present time dry, but would proceed in the same manner if wet.
After completion of the reracking of Hatch 2, the spent fuel presently stored in Hatch 1, will preferentially be transferred to Hatch 2 through the transfer canal, or may be concentrated away from rerack work locations.
In either case, the work to be performed for Hatch 1, will be performed in a manner consistent with as low as is The licensee reasonably achievable (ALARA) occupational exposure.
will remove unnecessary radioactive equipment and material from the Hatch 1 fuel pool prior to the installation work, and will pre-plan procedures necessary for the removal of the old racks and installa-tion of the new ones. The licensee does not anticioate the need fnr divers at this time.
However if they are reouired their occupa-tional exposure is expected tb be insignificant (i.e., less than 0.1 na n-rem). The existing low density racks will be decontaminated upon removal from the pool, packaged and shipped intact to a Althouoh this disposal site as low concentration radioactive waste.
is the present disposal plan of the licensee, because of restrictive limitation of radioactive waste burial, he may choose to consider an alternative option of cutting the racks into smaller sections in order to reduce the volume for burial. Pased on relevant experience from other licensees who have performed this operation, the licensee has learned that volume reduction technioues for disposal of spent fuel racks has required an additional collective dose of from 0.1 to 3 man-rer depending upon the reduction method used.
Therefore, if he does reconsider his disposal method, the licensee will take into account the additional occupational exposure reouired for volume reduction (i.e., cutting contaminated racks into smaller sections) takina into account occupational exposure and economic and environ-mental impact. The occupational exposure expected for the modification of Hatch 1, based on disposal of intact racks, is approxinately 14 ren-rem.
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. Very little additional exposure is anticiDated for modification of Hatch 2 if the pool remains dry prior to the modification. However, if the pool becomes contaminated befare reracking similar tech-nioues will be performed as stated above for Hatci.1, and an additional 14 man-rem would be received. Thus there is some possibility that 28 man-rem could be received for the SFP modification.
Ve have estinated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies on i
the basis of information supplied by the licensee for dose rates in the spent fuel area from radionuclide concentrations in the SFP i
water and deposited on the SFP walls.
The spent fuel assemblies i
themselves will contribute a negligible amount to dose rates in the Dool 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 spent fuel pool area, we estimate that the proposed modification should add less than or.c ;;ercent to the total annual occupational radiation exposure burden at this
)
facility.
The small increase in additional exposp-e will not affect the licensee's ability to maintain individual oce pational doses to as low as is reasonably achievable and within the limits of 10 CFR Part 20 Thus, we conclude that storing additional fuel in the SFP l
will not result in any significant increase in doses received by occupational workers.
2.6 Radioactive Waste Treatment The plant contains waste treatment systems designed to collect and process the gaseous, liquid and solid wastes that might contain radioactive material. The waste treatment systems were evaluated in the Safety Evaluation dated June 1978.
There will be no change in the waste treatment system or in the conclusions given in Section 11.2 of the evaluation of thir system because of the proposed modification.
l 3.0 Summa rv Our evaluation supports the conclusion that the proposed modification to I
the Hatch 1/2 SFP is acceptable because:
(1) The increase in occupationai radiation exposure to individuals due to the storage of additional fuel in the SFP would be negligible.
(2) The potential consecuences of the postulated design basis fuel bandling accident for the SFP are acceptable.
(3) The likelihood of an accident involving heavy loads in the vicinity of the spent fuel pool is sufficiently small that no additional restrictions on load movement are necessary while our generic review of the issues is under way.
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4.0 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 4
Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
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