ML20073S483
| ML20073S483 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/15/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073S481 | List: |
| References | |
| TAC-48219, NUDOCS 8305090023 | |
| Download: ML20073S483 (15) | |
Text
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c UNITED STATES
[ '$., y[ i NUCLEAR REGULATORY COMMISSION V -
E WASHINGTON, D. C. 20555 0,
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE MODIFICATION OF THE SPENT FUEL STORAGE POOL FACILITY OPERATING LICENSE N0. NPF-2 ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT - UNIT 1 DOCKET NO. 50-348 O
i 0305090023 830415 PDR ADOCK 05000348 P
j CONTENTS 1.0 Introduction
2.0 Background
3.0 Discussion and Evaluation 3.1 Criticality Considerations 3.1.1 Conclusion 3.2 Spent Fuel Cooling System 3.2.1 Introduction 3.2.2 Evaluation 3.2.3 Conclusion 3.3 Installation of Racks and Fuel Handling 3.3.1 Conclusion 3.4 Structural and Seismic Loadings 3.4.1 Introduction 3.4.2 Applicable Codes, Standards and Specifications 3.4.3 Seismic and Impact Loads 3.4.4 Load and Load Combinations 3.4.5 Design,and Analysis Procedures 3.4.6 Structural Accentance Criteria i
3.4.7 Materials, Quality Control, and Special Construction Techniques 4
3.4.8 Conclusion 3.5 Materials Evaluation 3.5.1 Structural Aspects 3.5.2 Corrosive Aspects 3.5.2.1 Introduction i
3.5.2.2 Evaluation
.3.5.2.3 Conclusion 3.6 Occupational Radiation Exposure 3.6.1 Radiation Expo:ure to Workers During Modifications 3.6.2 Conclusion 3.6.3 Onsite Radiation Exposure During Normal Operation 3.6.4 Conclusion 3.7 Radioactive Waste Treatment 3.7.1 Introduction 3.7.2 Evaluation 3.7.3 Conclusion 4.0 Overall Safety Conclusion
1 1.0 Introduction By letter dated March 19, 1982, as supplemented April 21 and September 14, 1982, Alabama Power Company (APCo) (the licensee) _ requested an amendment to Facility Operating License No. NPF-2 for Joseph M. Farley Nuclear Plant
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Unit No. 1.
The request would revise the Technical Specifications to i
allow an increase in the spent fuel pool (SFP) storage capacity from 675 to a maximum of 1407 fuel assemblies through the use of neutron absorbing
" poison" spent fuel storage racks.
The expanded storage would allow Farley Unit 1 to operate until the year 2004 with capability for a full core discharge, assuming annual one-third core reloads of 52 assemblies each reload. Total storage capacity would be expended with the discharge in the year 2007, when 29 assemblies would remain in the reactor vessel.
i The major safety considerations associated with the proposed expansion of SFP storage capacity are addressed below.
A separate Environmental Impact Appraisal has been prepared as part of this licensing action.
2.0 Background
The SFP contains 136 spent fuel assemblies discharged from cycles 1, 2, and 3.
Only 32 assemblies were discharged at the end of cycle 3 since cycle 3 was shortened by several months due to a failure in the main electrical generator which occurred on September 10, 1981. APCo's current i
plans are to off-load the cycle 4 spent fuel assemblies during the cycle 5 refueling outage which started January 15, 1983.
Later these spent fuel assemblies will be located to one end of the fuel pool.
Thus, removal of the i
old used type fuel racks will b.e accomplished at the opposite end of the pool l
to minimize radiation exposure to divers.
The modification to the spent fuel pool is scheduled to start in July 1983 and to complete about end of September 1983 on a phased basis. We will modify the effective date of the license amendment accordingly.
i 3.0 Discussion and Evaluation APCo proposed to replace the existing storage racks in the SFP with high density, stainless steel, fixed poison type, free standing storage racks.
The storage racks will have three basic module configurations with dimensions of 6 x 7, 7 x 7, and 7 x 8 feet, and weights of 6 3/4 tons, 7 9/10 tons, and 9 tons, respectively.
There will be two 6 x 7 modules, nineteen 7 x 7 modules and seven 7'x 8 modules.
The individual poison cans or cannisters of the modules are formed using 0.024 inch thick sheets of stainless steel wrapped around a neutron absorbing material (vented Boraflex). The center-to-center spacing of the cans will be 10.75 inches. A water plenum is provided by supporting the modules a.t their four corners by stainless steel support feet equipped with large leveling screws.
.. The racks are in compliance with the applicable portions of the following:
Regulatory Guides 1.13 and 1.29; and 10 CFR Part 50 Appendix A General Design Criteria 1, 2, 61, 62, and 63.
Based on the above we conclude that the proposed storage rack design and arrangement is adequate and, therefore, it is acceptable.
3.1 Criticality Considerations The criticality aspects of the proposed high density spent fuel racks have been analyzed using the PDQ-7 diffusion theory code for purposes of scoping and design.
The KENO-IV Monte Carlo code with AMPX cross section code has been used to verify the final design.
These codes have been benchmarked against experiment and a calculational bias, as well as calcu-lational and mechanical uncertainties,were obtained.
The effective multiplication factor for the racks was calculated under the assumption of fresh fuel of 4.3 weight percent U-235 enrichment (54.25 grams of U-235 per centimeter of assembly length) fully flooded with unborated water at a pool temperature of 68 degrees Fahrenheit.
No credit is taken for control rods or any noncontained burnable poison in the Westinghouse 17 x 17 fuel assemblies and the fuel racks are assumed to be infinite in extent.
Under these assumptions the nominal effective multiplication factor for the storage racks in their design configuration is 0.92171 0.0044 as detennined by the KENO-IV code.
To this value must be added a calculational bias.of 0.0027 (obtained from benchmark comparisons) and a total uncertainty of 0.0159 (obtained by a statistical combination of the calculational and mechanical uncertainties).
The mechanical uncertainty accounts for variations in center-to-center spacing, B-10 loading in the poison plates, and U-235 enrichment.
After all uncertainties are added, the resulting value of the effective multiplication factor is 0.9403 0.0044.
This meets our acceptance criteria for criticality calculations of 0.95 when flooded with unborated water including all uncertainties. The calcu-lational uncertainty is such that the true multiplication factor will be less than the calculated value with a 95 percent probability at a 95 percent confidence level.
The effect of credible accidents has been calculated and the most consequen-l tial one is the dropping of a single fuel assembly outside the rack between the periphery of the storage racks and the side walls of the pool. The i
effective multiplication factor remains below 0.95 for this accident with all uncertainties and biases included.
The pool water was assumed to con-tain soluble boron for this analysis.
This is permitted by the double contingency principle of ANSI N16.1-1975 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside l
Reactors," which states that two unlikely, independent, concurrent events are required to produce a criticality accident.
The staff has accepted I
this principle in previous safety evaluations.
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3.1.1 Conclusion We conclude that the proposed storage racks meet the requirements of General Design Criterion 62 as regards criticality.
This conclusion is based on the following considerations:
1.
state-of-the-art calculation methods which have been verified by comparison with experiment have been used; 2.
conservative assumptions have been made about the enrichment of the fuel to be stored and the pool conditions; 3.
credible accidents have been considered; 4.
suitable uncertainties have been considered in arriving at the final value of the multiplication factor; and 5.
the final effective multiplication factor value meets our acceptance criterion.
We also conclude that the modifications to the Technical Specification 5.6.1.1 increasing the maximum allowable enrichment in the spent fuel pool to 4.3 weight percent U-235 and reducing the nominal center-to-center distance between fuel assemblies in the storage racks to 10.75 inches are acceptable.
The revised Technical Specification 5.6.3 which allows an increase in the spent fuel storage pool capacity from 675 to 1407 fuel assemblies is also acceptable for the high density storage racks described in the Farley Unit 1 Spent Fuel Pool Modifi-cation Report dated March 1982.
The maximum fuel enrichment presently allowed in the new fuel pit storage racks is 3.5 weight percent U-235 4
(Technical Specification 5.6.1.2).
Therefore, the higher enriched, extended cycle fuel of 4.3 weight. percent enrichment can only be stored in the proposed high density spent fuel storage racks at present.
Our evaluation is based on PWR fuel pins and fuel assemblies similar in design to the Westinghouse fuel presently installed in the Farley Unit 1 plant.
Fuel designs differing from this would require a reeval-uation even though the U-235 enrichment and fuel assembly spacing spec 4-fications are not violated.
3.2 Spent Fuel Pool Cooling System 3.2.1 Introduction The SFP Cooling System consists of two pumps and two heat exchangers.
One pump and one heat exchanger is used for normal operation and the second pump and heat exchanger serves as a backup.
The heat exchangers are cooled by the component cooling water system. The SFP cooling connections to the pool are provided with anti-siphon holes or are located in such a manner that protects against inadvertent drainage of the pool to less than 4 feet below the normal level of 23 feet above the fuel.
In event of a loss of the cooling system, makeup is available from the seismic Category I reactor water makeup system.
. The refueling cycles for Farley Unit 1 are twelve month cycles where one-third of the core is removed and stored in the SFP after each cycle.
To limit the decay heat load, the one-third core (52 assemblies) will be removed from the reactor vessel and stored in the SFP no sooner than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown.
In the event of a full-core discharge, the decay heat load is based on a ten day decay time after shutdown before core discharge.
3.2.2 Evaluation To calculate the heat loads for the discharges of spent fuel to the pool, APCo used Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long-Term Cooling." The maximum normal heat load which occurs after the twenty-ninth refueling discharge, was calculated to be 19.755 x 106 BTU /HR.
The normal heat load resulted in a maximum bulk pool temperature of approximately 139 F with one cooling train operating which is in compliance with Standard Review Plan Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System."
The maximum abnormal heat load results from a full-core discharge after the last nonnal refueling discharge was calculated to be 30.384 x 106 BTV/HR.
The abnormal heat load resulted in a maximum bulk pool temperature of approximately 158 F with one train operating and 131 F with two trains operating.
The American National Standard 57.2 " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" indicates that the maximum pool temperature should not exceed 150 F under normal operating conditions with all storage' full.
The design, therefore, meets this standard.
To verify that natural circulation of the pool water for the proposed expanded rack configuration provides adequate cooling of all fuel assemblies in the event of a loss of external cooling, APCo performed a thermal-hydraulic analysis.
For this event, a complete failure of the SFP cooling system, even for the maximum abnormal heat load, at least four hours is available before boiling occurs.
The maximum boiloff rate is between 50 and 60 gpm.
Each of three makeup water sources can be initiated in the required time.
The reactor water makeup tank supply can be provided to the pool by either of two 165 gpm l
reactor water makeup pumps. The reactor water makeup tank, piping', and l
the makeup pumps are seismic Category I.
Sufficient makeup rates are also available from the refueling water storage tank (via two paths) and the demineralized water system; however, neither source is completely seismic Category I.
... 3.2.3 Conclusion We have reviewed the calculated decay heat values and conclude that the heat loads'are consistent with the Branch Technical Position ASB 9-2 and, therefore, are acceptable.
The SFP cooling system performance and the natural circulation assumptions have been reviewed and we conclude that the pool cooling is adequate.
The available makeup systems, the respective makeup rates, and the time required before makeup is needed have been reviewed and found acceptable.
Based on the above, we conclude that the SFP cooling system is acceptable.
3.3 Installation of Racks and Fuel Handling The SFP, a seismic Category I reinforced concrete structure, is housed within the fuel storage area of the auxiliary building.
A 125-ton capacity outdoor overhead unequal leg gantry crane is provided to handle heavy loads such as the spent fuel shipping cask.
The crane, a seismic Category I, single-failure proof crane,was evaluated and found acceptable by the staff as documented in Supplement No. 2 to the Farley Units 1 and 2 Safety Evaluation Report, NUREG-0117, dated October 1976.
The range of travel of this crane is limited by design such that it cannot pass over the SFP, The only crane that can pass over the SFP is the spent fuel bridge crane with a main hoist rated at 4000 pounds.
Therefore, the removal and installation of storage racks will require a temporary traveling bridge and hoist installed on the fuel handling bridge rails for the movement of the storage racks to and from the spent fuel cask area.
There the racks are handled by the single-failure-proof cask handling crane. This is the same seismic Category I crane and lifting-fixture that was used for reracking Farley Unit 2, thereby demonstrating i
i its ability to safely perform the reracking.
Following the installation of the temporary crane and before use in Unit 1, the crane will be tested I
using a load of 117 percent of rated capacity.
l APCo indicated that the movement of all loads into and out of the-auxiliary building, associated with this modification, will be accomp-lished with the single-failure-proof cask crane and double rigging to assure that a single failure will not result in an unanalyzed load-drop event. No heavy loads will be carried over any spent fuel assemblies during the rerack program.
The spent fuel assemblies will be moved and located so that no heavy loads will be carried over them.
APCo also stated in response to NUREG-0512, " Control of heavy loads at Nuclear Power Plants," that all crane operators and signalmen will be trained in accordance with ANSI B30.2-1976, and no exceptions are taken regarding training, qualification or operator conduct.
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.. 3.3.1 Conclusion We have reviewed the described load handling operations and equipment needed for the spent fuel rack modifications and conclude that the stored spent fuel and safety related equip-ment will be adequately protected against a load drop accident.
We, therefore, conclude that the health and safety of the public will not be endangered by the expansion-program of the Farley Unit 1 SFP.
Therefore, the expansion program is acceptable.
3.4 Structural and Seismic Loadings 3.4.1 Introduction The Farley Unit 1 SFP is an existing reinforced concrete box structure.
The walls of the pool vary in thickness from about 3.5 feet to 7.5 feet.
The floor is 5 feet thick and rests on 9.5 foot long columns surrounded by fill concrete, which in turn, are supported by a 5 foot thick base slab, which rests on rock.
The inside dimensions are approximately 40.5 feet deep by 27 feet wide by 45 feet long. The pool is lined with a water-tight, continuous, 1/4 inch thick, stainless steel plate.
The new spent fuel storage racks are to be constructed of 300 series stainless steel with vented "Boraflex" poison material sandwiched between stainless steel sheets.
The racks are vertical
" egg-crate" structures, each of which is free-standing on four pads on the pool floor. A 7 x 8 rack (56 cells) would be approxi-mately 14.9 feet high by 7.2 feet wide by 6.3 feet long. The pitch of all cells will be 10.75 inches, center-to-center. The racks are individually installed with the bottom grids of adjacent racks butting to one another leaving a nominal 5/8 inch gap at the top.
The minimum clearance between a rack and the pool wall is to be approximately 3 inches while the maximum is about 9 inches.
3.4.2 Applicable Codes Standard and Specifications The design, fabrication, installation and quality assurance standards for the new spent fuel racks are compared with the staff's "0T Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications" dated April 1978 including revisions dated January 1979 (to be referred to henceforth as the "0T Position").
The racks are designed in accordance with the requirements of the American Institute of Steel Construction (AISC) Manual which is an acceptable alternative in the OT Position.
. APCo proposes to use austinitic stainless steel conforming to ASTM A666, Grade B, for certain portions of the racks.
This e
material specification is not found in the ASME Code.
The staff's position is that all rack material should conform to all applicable requirements of Section III, Division 1, Subsection NF of the
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ASME Code.
APCo has committed to qualify the rack material in question to ASME Code Subsection NF (material specification SA240) in all respects, and in addition, to obtain valid test results to justify the higher yield stress allowed by ASTM-A666, Grade B.
APCo has also furnished test results and cited experience with this material to satisfy staff concerns.
Complete documentation of material quality will be maintained.
This is acceptable to the staff.
3.4.3 Seismic and Impact Loads The SFP floor response spectra used for the seismic analysis were as provided in the Farley FSAR and approved as part of the license review. A computer program, "SIMQKE," was then used to develop artificial time histories from these spectra.
Damping values of 2 percent for OBE and 5 percent for SSE, which are plant specific values and previously approved in the plant license review were used.
The dynamic model, consisting of springs, masses, gaps, and damping elements for a double rack system includes the potential for rack-to-rack interaction, fuel-to-fuel interaction and floor-to-rack interaction.. The se.ismic time history analysis was conducted using a coefficient of friction between the pool and rack of 0.2 in order to define maximum credible sliding.
The analysis was also performed using a coefficient of friction of 0.8 in order to define a worst case loading condition.
The spacing of the racks is such that rack-to-rack impacts may occur in some modes; however, in all cases, stresses are maintained within allowable limits.
Fuel casks cannot be transported over the pool due to built-in physical constraints as described above in Section 3.3.
Technical Specification 3.9.7.1 prohibits transporting loads greater than 3000 pounds over the SFP; therefore, the heaviest load that will be carried over the pool is a fuel bundle.
Impact loading on the racks from a fuel bundle drop was considered for the required conditions and combined with dead loads and live loads at suitable thermal levels.
Results were satisfactory.
.. 3.4.4 Load and Load Combinations 4
Loads, load combination were compared with the criteria outlined in SRP Section 3.8.4 and found to be acceptable.
3.4.5 Design and Analysis Procedures As described above, dynamic analyses of the rack and pool were conducted using lumped masses, spring elements, gap elements and damping elements to model the systems.
Hydrodynamic effects were considered. Various loading configurations of fuel in the racks were considered in order to define worst-case conditions.
In addition, a finite element analysis of the racks, using forces developed from the dynamic analysis, was accomplished.
The racks are not attached to the pool walls and the pool itself is founded on bedrock, therefore, any motion of the pool walls will not directly amplify the rack seismic motions.
Seismic loads were imposed simultaneously in three orthogonal directions on the computer models in the dynamic analyses.
APCo's analysis includes consideration of the loads, acting upward, of stuck fuel assembly as it is being lifted out of the rack.
For this case, no permanent deformation of the rack is allowed.
3.4.6 Structural Acceptance Criteria The structural acceptance criteria outlined in APCo's submittal was compared to that outlined in SRP Section 3.8.4 II.5 and was found to be in conformance.
3.4.7 Materials, Quality Control, and Special Construction Techniques With the exception noted previously in Section 3.4.2, all materials are in accordance with the ASME Code, as are fabrication, and inspection procedures.
3.4.8 Conclusion We find that the subject modification with respect to structural and seismic loadings, proposed by APCo is acceptable and satisfies l
the applicable requirements of the General Design Criteria 2, 4, 61, and 62 of 10 CFR, Part 50, Appendix A, regarding such structures.
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., 3.5 Materials Evaluation 3.5.1 Structural Aspects APCo proposes to use austinitic stainless steel conforming to ASTM A666, Grade B, for certain portions of the racks as discussed in Section 3.4.2 above.
3.5.2 Corrosive Aspects 3.5.2.1 Introduction We have reviewed the compatability and chemical stability of the materials, except the fuel assemblies, wetted by the pool water.
The proposed SFP storage racks are fabricated primarily of Type 304 stainless steel, which is used for all structural com-ponents, except for part of the bottom grid where Type 17-4 PH given the H-1100 heat treatment and a cast stainless steel CF8 are used in selected components.
The neutron absorber material is boraflex, which is held firmly between a stainless steel structural can and a stainless steel inner wrapper.
The compart-ments in the storage racks containing the boraflex are exposed to the SFP environment through small openings formed during fabri-cation in the top and bottom of each tube assembly.
The water chemistry in the SFP has been reviewed (Section 3.7) and found to meet NRC specifications.
Type 304 stainless steel rack modules have been welded-and inspected by nondestructive examinations performed in accordance with the applicable provisions of ASME Boiler and Pressure Vessel Code, Section 9.
APCo will perform a materials compatability monitoring program consisting of 10 coupons which duplicate the condition of boraflex which is encased in the poison canisters. These coupons are to be hung alongside the high density fuel racks and will be subjected to the maximum neutron, gamma, and heat fluxes.
Sufficient coupons are included to permit destructive examination of a sample on inspection inter-vals of 1 -to 5 years over the life of the facility.
3.5.2.2 Evaluation The SFP is fabricated of materials that will have good compati-bility with the borated water chemistry of the pool.
The corrosion rate of Type 304 stainless steel in this water is sufficiently low to defy our ability to measure it.
Since all materials in the pools are stainless steel, no galvanic corrosion effects are anticipated.
No instances of corrosion of stainless steel in spent fuel pools (l) containing boric acid has been observed throughout the country.
(1 )
J. R. Weeks, " Corrosion of Materials in Spent Fuel Storage Pools,"
BNL-NUREG-23021, July 1977.
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Boraflex has been shown to be resistant to radiation doses in excess of any anticipated in the SFP.
The venting of the-cavities containing the boraflex to the SFP environment will ensure that no gaseous buildup will occur in these cavities that might lead to distortion of the racks.
The type 17-4 PH stainless steel in the threaded feet of the racks has been given an H-1100 heat treatment, in which condition it is resistant to stress corrosion cracking in SFP environments. The Codes and Standards used in fabricating and inspecting these new fuel storage racks should ensure their integrity and minimize the likelihood that any. stress corrosion cracking will occur during service.
The materials surveillance program proposed by APCo 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.
l This monitoring program will ensure that, in the unlikely situation j
that the boraflex will deteriorate in this environment, APCo and the NRC i
will be aware of it in sufficient time to take corrective action.
3.5.2.3 Conclusion From our evaluation as discussed above, we conclude that the corrosion that will occur in the Unit 1 SFP will be of little significance during the remaining life of the unit.
Components of the spent fuel storage pool are constructed of alloys which are known to have a low differential galvanic potential between them, and that have performed well in spent fuel storage pools at other pressurized water reactor sites where the water chemistry is maintained to comparable standards to those in force at Farley. The proposed materials surveillance program is adequate to provide warning in the unlikely event that deterioration of the neutron adsorbing properties of the boraflex will develop during the design life of the racks.
Therefore, with the selection of the materials we believe i
that no significant corrosion should occur in the spent fuel storage racks at Farley Unit 1 for a period well in excess of the 40 years design life of the unit.
l Therefore, we conclude that the compatability of the materials and I
coolant used in the spent fuel storage pool is adequate based on tests, data, and actual service experience in operating reactors.
We find that the selection of appropriate materials by the licensee meets the require-ments of 10 CFR Part 50, Appendix A, Criterion 61, by having a capability to permit appropriate periodic inspection and testing of components, and Criterion 62, by preventing criticality by maintaining structural integ-rity of components, and is, therefore, acceptable.
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.. 3.6 Occupational Radiation Exposure 3.6.1 Radiation Exposure to Workers During Modifications We have reviewed APCo's plan for the re.noval 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 APCo to range from 6.1 to 8.1 person-rems. This estimate is based on a detailed breakdown of occupational exposure for each phase of the modification.
APCo has 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 is being performed.
Throughout the SFP modification operation, personnel exposure controls will be administered in accordance with APCo's radiological control procedures to assure exposures as low as is reasonable achievable (ALARA) to workers. These procedures include pre-job planning and worker briefings, checking water clarity for visibility, extensive surveys of the work area, physical barriers to prevent divers entering prohibited areas, and use of local filtered ventilation when necessary.
In addition, APCo has developed specific operating procedures for divers to assure that their doses are ALARA.
APCo has presented two alternative plans for the removal and disposal of the old racks.
These are (1) transfer of the old racks to another utility for use as spent fuel racks or (2) decontamination of the old racks prior to disposal. 'APCo will follow ALARA guidelines for workers regardless of which disposal method chosen.
3.6.2 Conclusion Based on the manner in which APCo proposes to perform the modifications, and relevant experience from other operating reactors that have performed similar SFP modifications, we conclude that the SFP modifications can be performed in a manner that will ensure as low as is reasonably achievable (ALARA) exposures to workers.
3.6.3 Onsite Radiation Exposure During Normal Operations We have estimated the increment in onsite occupational dose during normal operations after the SFP modifications which will result in an increase in stored spent fuel assemblies at Unit 1.
This estimate is based on information supplied by APCo for occupancy times and for dose rates in the SFP area from radionuclide concentrations in the SFP water. The spent fuel assemblies contribute a neglible amount to dose rates in the pool area because of the depth of water shielding the fuel.
Based on present and projected operations in the SFP area, we estimate that the proposed modification should add less than one percent
.. to the total annual occupational radiation exposure at Unit 1.
The small increase in radiation exposure should not affect the licensee's
'3 ability to maintain individual occuperional doses to ALARA levels and within the limits of 10 CFR Part 20.
3.6.4 Conclusion Thus, based on these considerations, we conclude that storing the additional fuel in the modified SFP during normal operations of Farley Unit 1 will not result in any significant increase in doses received by workers.
3.7 Radioactive Waste Treatment 3.7.1 Introduction The SFP cleanup system is designed to remove corrosion products, fission products and impurities from the pool water with mixed bed deminera-lizers and filters.
Pool water purity is monitored weekly by chemical and radiochemical analysis. Demineralizer resin will be replaced when pool water samples show demineralizer reduced decontamination effective-ness.
The SFP filters will be exchanged when AP exceeds 20 psid. APCo indicated that no change or equipment addition to the SFP cleanup system is necessary to maintain pool water quality and optical clarity for high density fuel storage.
3.7.2 Evaluation Past experience showed that the greatest increase in radioactivity and impurities in SFP water occurs during refueling and spent fuel handling.
The refueling frequency, the amount of core to be replaced for each fuel cycle, and frequency of operating the SFP cleanup tystem are not expected to increase as a result of high density fuel storage.
The chemical and radionuclide composition of the SFP water is not expected to change as a result of the proposed high density fuel storage.
Past experience also shows that no significant leakage of fission products from spent fuel stored in pools occurs after the fuel has cooled for several months.
To maintain water quality, APCo has established the frequency of chemical l
and radionucli.de analysis that will be performed to monitor the water quality and the need for SFP cleanup system demineralizer resin and filter replacement.
In addition, APCo established chemical and radio-chemical limits to be used in monitoring the SFP water quality and initiating corrective action.
On the basis of the above, we determined that the proposed expansion of the SFP will not appreciably affect the capability and capacity of the SFP cleanup system. More frequent replacement of filters or demineralizer resin, required when the differential pressure exceeds 20 psid or decon-tamination effectiveness is reduced to less than 10 (decontamination i
i factor), can offset any potential increase in radioactivity and impurities in the pool water as a result of the expansion of stored spent fuel.
Thus we have determined that the existing SFP cleanup system with the proposed high density spent fuel storage (1) provides the capability and capacity of removing radioactive materials, corrosion products, and impurities from the pool and thus meets the requirements of General Design Criterion 61 in Appendix A of 10 CFR Part 50 as it relates to appropriate systems to spent fuel storage; (2) is capable of reducing occupational exposures to radiation by removing radioactive products from the pool water, and thus meet the requirements of Section 20.1(c) 4 4
of 10 CFR Part-20, as it relates to maintaining radiation exposures as i
low as is, reasonably achievable; (3) confines radioactive materials in the pool water with the filters and demineralizers, and thus meets Regulatory Position C.2.f(2) of Regulatory Guide 8.8, as it relates to reducing the spread of; containments from the sources; and (4) removes suspended impurities from the pool water by filters; and thus meets j
Regulatory Position C.2.f(3) of Regulatory Guide 8.8, as it relates to removing crud frcm fluids through physical action.
3.7.3 Conclusion On the basis of the above evaluation, we conclude that the existing spentsfuel pool cleanup system meets GDC 61, Section 20.1(c) of 10 CFR Part 20 and the appropriate sections of Regulatory Guide 8.8 and, therefore, is acceptable for the proposed high density spent fuel storage. s,
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4.0 Overall ' Safety Conclusi_o'n On the basis of the foregoing analysis, we conclude that there will be no significant environmental impact attributable to the proposed action.
Having made this ~ conclusion, the Comission has further concluded that no
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environmental impact. statement for the proposed action need be prepared and i
l that a negative declaration to this effect is appropriate.
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 Comission'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.
Date:
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e Principal Contributors:
.L.
Kopp B. LaFace i
M. Lamastra E. Reeves u
- 0. Rothberg
^
B. Turov11n F. Witt M. Wohl i
t s
l L
-