ML15112B120
| ML15112B120 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/29/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112B119 | List: |
| References | |
| NUDOCS 8310110569 | |
| Download: ML15112B120 (14) | |
Text
o0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 123 TO FACILITY OPERATING LICENSE NO.
DPR-38 AMENDMENT NO.123 TO FACILITY OPERATING LICENSE NO.
DPR-47 AMENDMENT NO. 120 TO FACILITY OPERATING LICENSE NO. DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 6310110569 830929 PDR ADOCK 05000269 P
1.0 INTRODUCTION
By letter dated March 10, 1983, as supplemented June 24 and 30, 1983, July 14, 1983 and August 8, 1983, 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 (TSs) to allow an increase in the Unit No. 3 spent fuel pool (SFP) storage capacity from 474 to a maximum of 825 fuel assemblies through the use of neutron absorbing "poison" spent fuel storage racks.
The expanded storage capacity would allow the Oconee units to operate until about 1990 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.
2.0 EVALUATION 2.1 CRITICALITY 2.1.1 EVALUATION The Duke Power Company has provided an analysis of the criticality aspects of the proposed spent fuel pool expansion. The analysis was performed using the KENO-IV code, a Monte Carlo program optimized for reactivity calculations.
The code has been benchmarked and verified with a large number of critical experiments which spanned the enrichment range of interest in Oconee. The cross sections used for the analysis were from the ENDF/B-IV. The moderator was assumed to be pure water at a density of 1 g/cm which would yield the largest reactivity within the temperature design limits of the pool.
The comparison of the calculated and measured values yielded a bias (value of bias = 0, method uncertainty =.013 k) which is used in the calculated results. In addition, calculated uncertainties due to mechanical effects were examined.
These include uncertainties due to mechanical design tolerances, particle self shielding in the boron, bowing in the cans, etc.
The uncertainties are for 95% probability at a 95% confidence level.
The total uncertainty is the sum of its constituents (the square root of the sum of the squares).
When the calculational bias and the sum of the uncertainties were included, the effective multiplication factor was fouhd to be.9411.
The effects of accidents on the reactivity of the racks has been analyzed.
Mislocation of an assembly is precludedby design.
All configurations which could result from an accident are estimated to yield effective multiplication factors lower than the design value. The effective multiplication factor is dominated by the large amount of boron in the "boraflex"-which is attached to each cell for which, however, no credit, is taken but credit is taken for the boron dissolved in the pool water.
-2 2.
1.2 CONCLUSION
We have reviewed the submittal and conclude that the rack design is acceptable from the nuclear physics and criticality' point of view. This conclusion is based on the following:
- 1. The Duke Power Company used analysis methods which are state-of-the-art and have been validated with critical experiments of arrangements incorporating the main design features of the racks,
- 2. An evaluated calculational bias (zero) and the sum of the expected uncertainties have been applied to the calculated value of the effective multiplication factor,
- 3. A series of credible accidents have been considered and were shown to have acceptable consequences, and
- 4. The value of the effective multiplication factor meets the acceptance criteria, i.e., less than or equal to.95 with the bias and the uncertainties taken into account.
We conclude that any number of fuel assemblies of the Babcock and Wilcox 15x15 design can be stored in the racks provided that the uranium-in the fuel has an enrichment no.greater than 4.00 w/o in U-235. We also conclude that the proposed revision to the Technical Specifications concerning the expansion of the storage capacity for the Oconee Unit 3 is acceptable.
2.2 MATERIALS 2.
2.1 INTRODUCTION
The spent fuel racks in the proposed expansion would be constructed entirely of type 304 stainless steel, except for the nuclear poison material.
The existing spent fuel pool liner is constructed of stainless steel.
The high density spent fuel storage racks will utilize Boraflex sheets as a neutron absorber. Boraflex consists of 42 weight percent of boron carbide powder in a rubber-like silicone polymeric matrix. The spent fuel storage rack configuration is composed of individual storage cells interconnected to form an integral structure. The major components of the assembly are the fuel assembly cells, the Boraflex material, the wrapper and the upper and lower spacer plates.
The upper end of the cell has a funnel shape flare for easy insertion of the fuel assembly. The wrapper surrounds the Boraflex material, but is open at the top and bottom to provide for venting of any gases that are generated. The Boraflex sheets sit in a square annular cavity formed by the square inner stainless steel tube and the outer wrapper. Each sheet is supported by lower spacer plate.
-3 The pool contains oxygen-saturated demineralized water containing boric acid, controlled to a temperature below 150 0F.
2.2.2 EVALUATION The pool liner, rack lattice structure and fuel storage tubes 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 10-5 inches in 100 years, which is negligible relative to the initial thickness.
Dissimilar metal contact corrosion (galvanic attack) between the stainless steel of the pool liner, rack lattice structure, fuel storage tubes, and the Inconel and the Zircaloy in the spent fuel assemblies will not be significant because all of these materials are protected by highly passi vating oxide films and are therefore at similar potentials. The Boraflex is composed of non-metallic materials and therefore will not develop a.
galvanic potential in contact with the metal components. Boraflex has undergone extensive-testing to study the effects of gamma irradiation in various environments, and to verify its structural integrity and suitability as a neutron absorbing material. 2 The evaluation tests have shown that the Boraflex is unaffected by the pool water environment and will not be degraded by corrosion.. Tests were performed at the University of Michigan, exposing Boraflex to 1.03 x 10" rads of gamma radiation with substantial concurrent neutron flux in borated water. These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of borated water and gamma irradiation.
Irradiation will cause some loss of flexibility, but will not lead to breakup of the Boraflex. Long term borated water soak tests at high temperatures were also conducted. 3 The tests show that Boraflex withstands a borated water immersion of 240'F for 260 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide.
The annulus space which contains the Boraflex is vented to the pool at each corner storage tube assembly. Venting of the annulus will allow gas generated by the chemical degradation of the silicone polymer binder during heating and irradiation to escape, and will prevent bulging or swelling of the inner stainless steel tube.
The tests have shown that neither irradiation, environment nor Boraflex composition has a discernible effect on the neutron absorption of the Boraflex material.
The tests also show that Boraflex does not possess leachable halogens that might be released into the pool environment in the presence of radiation.
Similar conclusions are reached regarding the leaching of elemental boron from the Boraflex. Boron carbide of the grade normally in the Boraflex will typically contain 0.1 wt percent of soluble boron., The test results have confirmed the encapsulation function of the silicone polymer matrix in preventing the leaching of soluble specie from the boron carbide.
e,
To provide added assurance that no unexpected corrosion or degradation of the materials will compromise the integrity of the racks, the licensee has committed to conduct a long term fuel storage cell surveillance program.
Surveillance samples are in the form of removable stainless steel clad Boraflex sheets, which are proto-typical of the fuel storage cell walls.
These specimens will be removed and examined periodically.
2.
2.3 CONCLUSION
From our evaluation as discussed above, we conclude that the corrosion that will occur in the.Oconee spent fuel storage pool environment should be of little significance during the 40 year life of the plant.
Components in the spent fuel storage pool are constructed of alloys which have a low differential galvanic potential between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosion. Tests under irradiation and at elevated temperatures in borated water indicate that the Boraflex material will not undergo significant degradation during the expected service life of 40 years.
We further conclude that the environmental compatibility and stability of the materials used in the Oconee expanded spent fuel storage pool are adequate based on the test data cited above and actual service experience in operating reactors.
We have reviewed the surveillance program and we conclude that the monitoring of the materials in the spent fuel storage pool, as proposed by the licensee, provides reasonable assurance that theiBoraflex material will continue to perform its function for the design life of the pool.
The materials surveillance program enacted by the licensee will reveal any instances of deterioration of the Boraflex that might lead to the loss of neutron absorbing power during the life of the new spent fuel racks. We do not anticipate that such deterioration will occur. This monitoring program will ensure that, in the unlikely situation that the Boraflex will deteriorate in this environment, the licensee and the NRC will be aware of it in sufficient time to take corrective action.
We therefore find that the implementation of a monitoring program and the selection of appropriate materials of construction by the licensee meets the requirements of 10 CFR Part 50, Appendix A, Criterion 61 (having a capability to permit appropriate periodic inspection and testing of components) and Criterion 62 (preventing criticality by maintaining structural integrity of components and of the boron poison).
2.3 STORAGE RACKS 2.3.1 EVALUATION The high density spent fuel storage racks are of the fixed poison, free standing modular design, and are designed to seismic Category I requirements.
The two basic storage modules ( 8 x 10 and 8 x 12) weigh from 10 to 12 tons.
Each module is composed of storage cells formed by two concentric stainless steel storage tubes with an inner core of Boraflex. The storage cells are arranged in a 10.60 inch center to center rectangular lattice configuration.
The racks are also designed in such a manner that will not permit the insertion of fuel assemblies in other than prescribed locations. Further, they can withstand the maximum uplift forces produced by the fuel handling machine, and are.designed such that the accidental dropping of a fuel assembly will
e
-5 not result in a geometry which can result in criticality.
2.
3.2 CONCLUSION
We conclude that the proposed high density storage racks meet the requirements of GDC 2 and 62 with respect to seismic design considerations and prevention of criticality, and the guidelines of Regulatory Guides 1.13 and 1.29 with respect to fuel storage design and seismic design classification, and are, therefore, acceptable.
2.4 RACK HANDLING AND INSTALLATION 2.4.1 EVALUATION The reracking modification will not commence until all fuel in the Unit 3 pool has been transferred to the Unit 1/2 pool, thus eliminating the concerns associated with carrying loads over stored spent fuel.
The spent fuel pool expansion modification will utilize the.100 ton spent fuel pool crane (cask handling crane), a temporary construction crane, a lift bag and several special lifting devices. The handling of spent fuel racks constitutes the handling of heavy loads and thus is subject to the purview of NUREG-0612 - Control of Heavy Loads. The heaviest of the racks to be handled will be the existing 32 ton C-E double rack module. This module will be handled only by the temporary construction crane. The module will then be cut into individual rack assemblies of 16 tons each, prior to their removal from the pool.
The Phase I review of the licensee's conformance to the guidelines of NUREG-0612 (whose analysis included the spent fuel pool crane) was completed and found acceptable as documented in our April 20, 1983 letter to the licensee. We conclude that the use of the 100 ton spent fuel pool crane is acceptable during the reracking modification. However, the temporary construction crane, the lift bag and various special lifting devices were not included in the original heavy loads analysis and thus are discussed below.
The temporary construction (T-C) crane is a gantry type bridge crane which traverses the length of the pool.
The T-C crane has a design load rating of 40 tons to accommodate the heaviest of the existing modules (32 tons).
The crane is designed to meet CMAA-70 and ANSI B30.2 criteria and will be load tested to 40 tons (e.g., 125% of the highest expected load).
The hoist is reeved with stainless steel wire rope and employs a submersible block. We conclude that the use of the temporary construction crane is acceptable during the reracking modification.
The lift bag system is used to move the two interconnected modules located furthest from the cask storage pit to a:point where they can be rerigged to the T-C crane. The lift bag system employs the lift bag itself,.a compressor, an air regulator, and a globe valve. The lift bag is a pneumatic bladder which
-6 when inflated, will lift the rack assembly no higher than six inches off the pool floor. Adverse consequences are minimal in the event the lift bag fails in such a manner which results in a load drop. An analysis performed by the licensee for the NUREG-0612 evaluation determined that there was no equipment below the pool needed for safe shutdown. Further, the worst expected consequence would be the puncture of the pool liner rather than pool floor failure. We conclude that the use of the lift bag system is acceptable during the reracking modification.
The special lifting devices of concern are those associated with the existing C-E racks, and the new Westinghouse racks. The C-E lifting device has a load rating of'32 tons and will be load tested to 125% of rating capacity (i.e., 40 tons).
The lifting device is built with a safety factor of 5 based on ultimate strength. The Westinghouse lifting device is the same as that which was used during the Unit 1/2 reracking. However, since the Unit 3 racks are lighter than those in the Unit 1/2 pool, usingthe same lifting device results in a safety factor of 5 based on yield strength.
Lifting apparatus such as sling, shackles and fittings are sized as necessary to conform with the guidelines of NUREG-0612, Control of Heavy Loads.
We conclude that the lifting devices and other apparatus used for the handling of the storage racks are adequate, and therefore, acceptable.
2.
4.2 CONCLUSION
We conclude that the use of the proposed cranes and load lifting devices meet the requirements of GDC 4 and 61 with respect to protection of systems or components needed for safe shutdown from load drops, and the guidelines of NUREG-0612, Section 5.1.1, with respect to safe load handling practices.
Specific safe load paths for the spent fuel racks will be those already developed for cask handling. The height of the rack(s) over floor areas during the modification will be limited to six inches. Previous analysis by the licensee has shown that a load drop of a 25 ton spent fuel cask will not adversely affect the capability to safely shut down the plant. The imple mentation of safe load paths, operator training and qualification, and proced ures will be consistent with NUREG-0612.
2.5 SPENT FUEL POOL COOLING SYSTEM 2.5.1 EVALUATION The spent fuel pool cooling system (SFPCS) is designed to maintain the spent fuel pool water temperature, water inventory, clarity and chemistry at an acceptable level.
It is designed to withstand the effects of a seismic event, and meets the requirements of Quality Group C classification. The existing system which is designed to accommodate the heat load from 474 fuel assemblies
-7 6
is composed of two trains with a heat removal capability of 15.5 x 10 Btu/hr at 125 0F. The major components of the SFPCS consist of two pumps in parallel with one heat exchanger in series with each pump. These heat exchangers are cooled by the recirculating cooling water system. To supplement decay heat removal for the anticipated added heat load following the pool expansion modification, an additional cooling train with a heat removal capability of 7.75 x 106 Btu/:hr at 125 0F will be installed and operational prior to exceeding the current design inventory of 474 assemblies. The additional train is designed to meet the same seismic requirements and quality group clas sification as that of the origina.1 system.
The refueling cycle for Oconee 3 is an annual one-third core discharge of 59 fuel assemblies. Each assembly is assumed to have experienced a continuous power level of 14.5 MWt prior to discharge. For both the normal refueling and full core discharge, the fuel will be subject to a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> decay period after shutdown prior to its transfer to the spent fuel pool.
The licensee's calculated spent fuel discharge heat load to the pool, which was determined in accordance with the Branch Technical Position ASB 9-2, "Residual Decay Energy for Light Water Reactors for Long Term Cooling," indicates that the expected maximum normal heat load following the last normal refueling is 12.6 x 106 Btu/hr. This heat load results in a maximum bulk pool temperature of 140OF with two of three trains in operation (assumed single failure). The expected maximum abnormal heat load following a fuel core discharge after the last normal refueling discharge is 30.8 x 106 Btu/hr. This abnormal heat load results in a maximum bulk pool temperature of 150OF with all cooling trains in operation or 205*F with the loss of one train. The above maximum normal and abnormal heat load temperatures are within our guidelines. In the event of a loss of the SFPCS under maximum normal heat load calculations, the time to reach bulk pool boiling is approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> respectively, which is sufficient time to provide emergency makeup to the pool.
The required makeup of less than 70 gpm can be provided from alternate sources.
2.
5.2 CONCLUSION
We have reviewed the calculated heat load values and conclude that the heat loads are consistent with the Branch Technical Position ASB 9-2. The spent fuel pool cooling system performance and the available makeup systems have been reviewed and found to be acceptable. We conclude that the spent fuel pool cooling system meets the requirements of GDC 44 with respect to heat removal capability, and the guidelines of Regulatory Guide 1.13 with respect to system design considerations.
2.6 STRUCTURAL ASPECTS 2.
6.1 INTRODUCTION
Oconee Unit 3 is a two-loop B&W PWR. The plant is founded on rock. The spent fuel pool serves Unit 3 exclusively and is located in the auxiliary building. The pool is a concrete box and is rectangular in plan view.
-8 Inside dimensions are approximately 58 ft long by 24 ft wide x 42 ft deep (maximum dimensions).
The bottom of the pool is elevated aboveithe basemat and the inside is at elevation 802 ft.
The basemat is at elevation 758. In general the walls of the pool extend to the basemat. The north end of the pool (cask area) rests on a massive concrete pier (21.5 ft x 31 ft) which extends to basemat.
The floor and the walls of the pool vary in thickness. The bottom of the pool is a minimum of 4.5 ft thick and the walls are a minimum of 3.5 ft thick. The pool is lined with a 1/4 inch thick stainless steel watertight liner plate. A leak-chase channel system is provided.
The existing fuel storage racks are to be replaced by ten new poisoned racks.
These are free standing box-shaped structures with individual rectangular cells to store the fuel.
The 8 cell by 10 cell rack is approximately 14 ft high by 9 ft wide by 7 ft long. The racks are constructed of type 304 stainless steel.
The cells are constructed of cold formed 0.075 inch thick cold formed sheet. The cells are supported by and welded to a top and bottom grid of structural tubing. The bottom grid rests on and is welded to a heavy base plate which is in turn supported by and welded to four corner leveling pads.
2.6.2 EVALUATION The racks were designed in accordance with the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and revised January 18, 1979 (referred to hereafter as the "NRC Position").Section III, Division 1, Subsection NF of the ASME code was the basis of design of the racks.
The existing concrete pool structure was originally designed in accordance with ACI 318-63. The pool and liner were reevaluated in accordance with the original plant design criteria for the new rack loads.
Rack structural materials are in conformance with the requirements of the ASME code.
Loads and load combinations for the design of the racks were found to be in accordance with the NRC Position. Loads and load combinations for the analysis of the pool structure were found to be in accordance with the original plant FSAR commitment and are acceptable.
The seismic loads are based on the original design floor acceleration response spectra calculated for the plant at the licensing stage. This was based on a 0.1g SSE and a 0.05g OBE. Damping values for the racks were taken as 2 percent for OBE and 4 percent of SSE. Impact effects due to fuel bundle/rack interaction as well as rack/pool floor interaction were included in the analysis.
-9 A separate fuel assembly drop accident analysis was performed.
A 3000 pound object was postulated to impact the top of the rack from a height of 6 feet.
The same object was postulated to drop 234 inches through a cell and impact the bottom of.the rack.
2.6.2.1 DESIGN AND ANALYSIS PROCEDURES
- a. Racks First a seismic time history analysis of a non-linear 2-dimen sional model was conducted. The model consisted of spring, mass, damping friction, and gap elements to simulate a fuel bundle in a simplified model of a rack. The fuel assembly-to-cell impact loads, support pad lift-off values, rack sliding, and overall rack response were determined,from the non-linear analysis.
Coefficients of friction were varied between minimum and maximum possible values in order to determine worst case conditions of sliding and tipping respectively. Rack-to-rack impacts were precluded by spacing the racks beyond maximum possible excursion. In order to account for 3-dimensional effects, the results of independent orthogonal loadings were combined by the SRSS method.
Next, a seismic response spectrum analysis of a 3-dimensional finite element model of the racks, using inputs from the results of the non-linear analysis, and superimposed with other applicable loads, was conducted. Design stresses and safety margins for appropriate components in the racks were tabulated and found to be acceptable.
- b. Pool The spent fuel pool was reanalyzed for new rack loads.
Results of key structural calculations comparing actual with allowable stresses for the analysis of the pool with the new rack loads with superimposed thermal and dead loads were provided.
The factors-of-safety for the pool and the liner were found to be acceptable.
2.
6.3 CONCLUSION
We conclude that the proposed rack installation will satisfy the requirements of 10 CFR 50 Appendix A, GDC 2, 4, 61 and 62, as applicable to structures, and is, therefore, acceptable.
-10 2.7 SPENT FUEL CASK MOVEMENT AND FUEL HANDLING ACCIDENTS 2.7.1' INTRODUCTION In our Safety Evaluation (SE) dated September 1976, accidents involving the movement of the spent fuel cask were evaluated on the basis of the cask being handled in the spent fuel pool and with fuel present in the pool.
This SE assumed 76 spent fuel assemblies could be damaged and that they had undergone a minimum aging time of 43 days.
Since the issuance of the September 1976 SE, there have been Technical Specifications covering both the Unit 1/2 common spent fuel pool and also the Unit 3 spent fuel pool with regard to the number of rows of stored spent fuel potentially impacted by a cask movement accident and also the aging time of the fuel in these rows.
2.7.2 EVALUATION The licensee has indicated that for this proposed reracking, the modifications to the pool will be accomplished with both the spent fuel cask and presently stored fuel removed from the pool altogether. The presently stored fuel will have been transferred to and stored in the Unit 1/2 common spent fuel pool prior to any work commencing on the presently existing racks.
Thus, the possibility of potential accidents involving the spent fuel cask and stored fuel during the reracking has been precluded. With regard to the potential for an accident involving the stored fuel and the spent fuel cask once the re racking has been accomplished, the licensee has proposed increasing the present limit of 20 rows of fuel aged a minimum of 43 days to a new limit of 31 rows of fuel aged a minimum of 70 days.
This requirement is incorporated in proposed TS 3.8.13. In the licensee's worst case accident analysis, a hoist cable failure would potentially cause the cask to be deflected into the pool wall and the yoke and load block could be deflected into the spent fuel.
There are 128 cans under the projected'cask, yoke, and block impact area.
In considering the cans adjacent to the impact area, a total of 486 cans can potentially suffer a loss of integrity during a cask drop accident. The licensee's analysis for such an accident indicates that the worst radiological consequences experienced would result from 100% of the activity contained in the fission gases trapped in gaps in the fuel stored in the locations being released into the pool water. The exclusion area boundary dose, taking no.
credit for ventilation system filtration, would be 0.1 rem whole body and 55 rem to the thyroid. These doses are well below 10 CFR 100 limits and the licensee has demonstrated compliance with Standard Review Plan Section 15.7.4.
2.
7.3 CONCLUSION
We conclude that the consequences of a cask drop accident and fuel handling accident in the Unit 3 spent fuel pool are not changed from those previously evaluated.
With the implementation of the limits prescribed in TS 3.8.13, the staff finds the proposed reracking acceptable.
-11 2.8 OCCUPATIONAL RADIATION EXPOSURE 2.
8.1 INTRODUCTION
We have reviewed the licensee's plan for the removal and disposal of the low density racks, and 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 22 person-rems.
This estimate is based on the licensee's detailed breakdown of occupational exposure for each phase of the modification.
2.8.2 EVALUATION Throughout the SFP modification operation, the licensee's personnel exposure controls will be administered in accordance with the licensee's health physics control procedures to assure as low as is reasonably achievable exposures (ALARA) to workers. The procedures include pre-job planning and worker briefings, checking water clarity, extensive surveys of the work areas, vacu uming the pool floor, walls and fuel rack surfaces around diver working areas and removing all spent fuel assemblies stored in the pool prior to diving operations. In addition, the licensee has developed specific operating proced ures for divers to assure that their doses are ALARA.
The licensee has presented two alternative plans for the removal and disposal of the old racks. These are:
(1) decontamination of the old racks prior to disposal as non-radioactive waste or (2)' transfer of the old racks to an authorized burial site. The licensee will follow ALARA guidelines for workers regardless of which disposal method is chosen.
Based on the manner in which the licensee will perform the modification, and relevant experience from other operating reactors that have performed similar SFP modifications, we conclude that the Oconee Nuclear Station, Unit No. 3 SFP modification can be performed in a manner that will ensure ALARA exposure to workers.
We have estimated the increment in onsite occupational dose for normal operations after the pool modification which can result from the proposed increase in stored fuel assemblies at Unit 3. The spent fuel assemblies themselves contrib ute a neglible amount to dose rates in the pool area because of the depth of water shielding the fuel.
The proposed increase of the storage capacity of the SFP would not create any significant additional radiological effects to the population. The additional
-12 total body dose that might be received by an individual at the site boundary, and by the population within a 50-mile radius, is estimated to be less than 0.10 mrem/yr and 0.02 person-rem/yr, respectively. These doses are extremely small compared to the fluctuations in the annual dose this population receives from background radiation. The populati6n dose represents an increase of less than 0.01 percent of the dose previously evaluated in the FES for the Oconee Nuclear Station, Unit No. 3. We find this to be an insignificant increase in dose to the population resulting from the proposed action.
'Similarly, the proposed increase in storage capacity of the SFP would not affect radiological impact to the work force significantly. The dose to plant workers at Oconee, over the years 1974 to 1981, has averaged about 900 person-rems/year. The total projected worker dose is 22 person-rems, which is about one-fourteenth of the normal annual rate.
2'.
8.3 CONCLUSION
The small increase in radiation exposure should not affect the licensee's ability to maintain individual occupational doses to as low as is reason ably achievable levels and within the limits of 10 CFR Part 20. Thus, we conclude that storing additional fuel in the pool will not result in any significant increase in doses received by workers.
3.0 CONCLUSION
S Based on our review, we conclude that the proposed modified fuel storage design of the Oconee Unit No. 3 spent fuel pool meets our requirements. The proposed increase in the spent fuel pool storage capacity to a maximum of 825 fuel assemblies (maximum allowed enrichment of 4.0 weight percent U-235) through the use of neutron absorbing (poison) spent fuel racks meets the requirements of the General Design Criteria, as discussed above, of Appendix A to 10 CFR Part 50 and is, therefore, acceptable. Based on our review, we have also determined that the proposed TS changes for the Oconee Nuclear Station's Common TS's are acceptable.
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 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: September 29, 1983 The following NRC personnel have contributed to this Safety Evaluation:
L. Lois, B. Turovlin, T. Chan, 0. Rothberg, M. Lamastra, J. Nehemias, J. Suermann.
-13
4.0 REFERENCES
- 1. J. S. Anderson, "Boraflex Neutron Shielding Material --
Project Performance Data," Brand Industries, Inc., Report 748-30-1, (August 1979).
- 2. J. S. Anderson, "Irradiation Study of Boraflex Neutron Shielding Materials," Brand Industries, Inc., Report 748-10-1, (July 1979).
- 3. J. S. Anderson, "A Final Report on the Effects of High Temperature Borated Water Exposure on BISCO Boraflex Neutron Absorbing Materials,"
Brand Industries, Inc., Report 748-21-1, (August 1978).
dated July 14, 19893.