ML20054K406
| ML20054K406 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/23/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054K404 | List: |
| References | |
| TAC-47523, NUDOCS 8207020037 | |
| Download: ML20054K406 (15) | |
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i SAFETY EVALUATION BY THE OFFICE OF l
NUCLEAR REACTOR REGULATION 1
RELATING TO THE MODIFICATION OF THE SPENT FUEL STORAGE P0OL FACILITY OPERATING LICENSE NO. NPF-8 ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT - UNIT 2 DOCKET NO. 50-364 l
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8207020037 820623 PDR ADOCK 05000364 P
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h CONTENTS l.0 Introduction
2.0 Background
3.0 Discussion and Evaluation 3.1 Criticality Considerations 3.2 Spent Fuel Cooling 3.2.1 Introduction 3.2.2 Evaluation 3.2.3 Conclusion 3.3 Installation of Racks and Fuel Handling 3.4 Structural and Seismic Loadings l
3.4.1 Introduction 3.4.2 Seismic and Impact Loads 3.4.3 Load and Load Combinations 3.4.4 Design and Analysis Procedures t
3,4,5 Structural Acceptance Criteria 3,4.6 Materials, Quality Control, and Special Construction Techniques 3.4.7 Conclusion 3.5 Materials Evaluation 3,5.1 Structural Aspects 3.5.2 Corrosive Aspects 3.5.2.1 Introduction 3.5.2.2 Evaluation 3._5. 2. 3 Conclusion 3,6 Occupational Radiation Exposure 3.7 Radioactive Waste Treatment 3.7.1 Introduction 3.7,2 Evaluation 3,7.3 Conclusion 4,0 Conclusions i
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I 1.0 Introduction By letter dated December 18, 1981, as supplemented February 1, March 19, April 5, and April 21, 1982 Alabama Power Company (APCo) (the licensee) requested an amendment to Facility Operating License No. NPF-8 for Joseph M. Farley Nuclear Plant Unit No. 2.
The request would revise the radiological Technical Specifications to 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 2 to operate until the year 2006 with capability for a full core discharge, assuming annual one-third core reloads.
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 currently contains racks with a capacity of 675 fuel assemblies. The proposed expansion modification will increase the existing storage capacity to 1407 fuel assemblies.
For simplification of the work involved, APCo has proposed a schedule for reracking that would allow all modifications to be completed before the first scheduled refueling outage (November 1982).
This simplifies the modification because currently the SFP and racks are clean, dry and uncontaminated. Thus, standard construction procedures can be used.
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 sandwiched in between the two sheets of stainless steel. The center-to-center spacing of the cans will be 10.75 inches. A water plenum is provided by supporting the modules at their four corners by stainless steel support feet equipped with large leveling screws. Pool water will flow down along the pool walls where it will enter the water plenum, and then travel laterally where it enters the bottom of the storage cans through the bottom grid.
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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 calculational 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) 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 0.9217 as determined by the KENO-IV code. To this value must be added a calculational bias of 0.0027 (obtained from benchmark comparisons) and a total uncertainity of 0.0159 (obtained by-a statistical combination of the calculational and mechanical uncertainties). The mechanical uncertainity 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.
This meets our acceptance criteria for criticality calculations of 0.95 including all uncertainties. The calculational uncertainty is such that the true multiplication factor will be less than the calculated value with a 95 percent probability of a 95 percent confidence level.
The effect of credible accidents has been calculated and the most consequential 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 effective multiplication factor remains below 0.95 for this accident with all uncertain-ties and biases included. The pool water was assumed to contain soluble boron for this analysis. This is permitted by the double contingency principle l
of ANSI N16.1-1975 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors" which states that two unlikely, independent, concurrent events are required to produce a criticality accident.
The staff has accepted this principle in previous safety evaluations.
We conclude that the proposed storage racks meet the requirements of General Design Criterion 62 regards criticality. This conclusion is based on the following considerations:
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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 proposed modifications to 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 accepta ble. The proposed 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 2 Spent Fuel Pool Modification Report dated December 1981.
The maximum fuel enrichment presently allowed in the new fuel pit storage racks is 3.5 weight percent U-235 (Technical Specification 5.6.1.2).
Therefore, the higher enriched, extended cycle fuel of 4.3 weight percent enrichment can be stored only in the proposed high density spent fuel storage racks at present.
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Our evaluation is based on PWR fuel pins and fuel assemblies similar in design to the Westinghouse fuel presently installed in Farley Unit 2.
Fuel designs differing from this would require a reevaluation even though the U-235 enrichment and fuel assenbly spacing specifications are not I
violated.
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! 3.2 Spent Fuel Cooling 3.2.1 Introduction The Spent Fuel Pool Cooling System consists of two pumps and two heat exchangers. One pump and one heat exchanger is used for normal operation tnd 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 located in such a manner than protects against inadvertent drainage of the pool to less than 4 feet below the normal level of 24 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 future refueling cycle for Farley Unit 2 will be a twelve month period and one-third of the core will be removed and stored in the SFP after each cycle.
To limit the decay heat load, the 1/3 core will be removed from the reactor vessel and stored in the SFP,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 will be limited by requiring a 10 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-seventh refueling discharge, was calculated to be 19.71 x 10b BTU /HR. The normal heat load resulted in a rnaximum 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 abnonnal heat load results from a full-core discharge after the last normal refueling discharge was calculated to be 30.33 x 106 BTU /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 l
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. The adequacy of natrual circulation was verified using the computer code BP00L. BP00L is a proprietary program of Nuclear Associates Incorporated (NAI). The code is based on the assumption that boiling takes place near the top of the fuel channel. BP00L evaluates the saturation properties of the coolant on the basis of the static pressure at the top of the storage racks.
These properties include water density, temperature, and steam density.
The steam is assumed to separate and flow out of the pool. The water at the saturation temperature corresponding to the pressure at the top of the racks flows downward to the inlet of the storage racks.
The static pressure at this location is higher than the pressure at the top of the storage racks and as a result the fluid is subcooled as it enters the fuel assembly.
The fluid becomes less dense as it passes up the fuel channel. Near the top of the fuel channel the fluid reaches saturation conditions and net boiling occurs. Thus natural circulation will be maintained and the flow is adequate as verified by the computer code BP00L.
Under normal cooling conditions (external cooling available) natural circulation cooling of the spent fuel was verified using the NAI computer code HP00L. HP00L calculates the pressure loss through a fuel assembly for a given flow rate.
This pressure loss is compared with the buoyant head resulting from the difference between the average density of the fluid in the fuel channel and the average density of the fluid in the downcomer (space between the pool wall and the racks).
If the density difference results in a buoyant head greater than the pressure loss, the flow rate through the fuel assembly is increased and a new average density of the fluid is determined. This iterative process is continued until the buoyant head and pressure loss in the fuel assembly are equal.
Using this flow rate, HP00L determines the fuel temperature.
In the event of a complete failure of the SFP cooling system, even for the maximum abnormal heat load there is at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> available before makeup water to the pool is required. The maximum required makeup rate is between 50 and 60 gpm.
Each of four 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 reactor water makeup pumps. The reactor water makeup tank, piping, and 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, their respective makeup rates and the time required before makeup is needed has been reviewed and found acceptable.
Based on the above, we conclude that the SFP cooling system is accepta ble.
3.3 Installation of Racks and Fuel Handling During the expansion program, a temporary crane will be used to remove all of the present racks and insert the new racks. Since there is no fuel in the present -racks, which are uncontaminated, the installation will be done dry. There is also no equipment, essential in the safe shutdown of the reactor or essential to mitigate the consequence of an accident, which is beneath, adjacent to or otherwise within the area of influence or Ony loads that will be handled during the expansion modification. Based 01 the above, we conclude that handling of the present racks and new racks is aweptable.
3.4 Structural and Seismic Loadings 3.4.1 Introduction The Farley Unit 2 SFP is an existing reinforced concrete box structure.
The walls of the pool very 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 j
fl oor. A 7 x 8 rack (56 cells) would be approximately 14.9 feet high i
by 72 feet wide by 6.3 feet long. The pitch of all cells will be i
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. 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.
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 and 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.
3.4.2 Seismic and Impact Loads The SFP floor response spectra used for the seismic analysis were as provided in the Farley Unit 2 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
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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-rack interaction and floor-to-rack interaction.
The seismic 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. The old racks will be removed and the new racks installed before any radioactive material is placed in the pool.
Existing Technical Specification 3.9.7.1 prohibits transporting loads i
greater than 3000 pounds over the spent fuel pool; 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.
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.. 3.4.3 Load and Load Combinations Loads, load combination were compared with the criteria outlined in SRP Section 3.8.4 and found to be acceptable.
3.4.4 Design and Analysis Procedures As described above, dynamic analyses of the rack and pool were conducted using lumpsed 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 j
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.
3.4.5 Structural Acceptance Criteria The structual acceptance criteria outlined in the applicant's submittal was compared to that outlined in SRP Section 3.8.4 II.5 and was found to be in conformance.
3.4.6 Materials, Quality Control, and Special Construction Techniques With the exception noted previously, all materials are in accordance with the ASME Code, as are fabrication, and inspection procedures.
3.4.7 Conclusion We find that the subject modification with respect to Structural and Seismic Loadings, proposed by the licensee is acceptable and satisfies I
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 r
3.5.1 Structural Aspects APCo proposes to use austinitic stainless steel conforming to ASTM A666, Grade B, for certain portions of the racks. This 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 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 alos 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.
3.5.2 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 i
of Type 304 stainless steel, which is used for all structural components, 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 compartments in the storage racks containing the boraflex are exposed to the spent fuel pool environment through small openings fomed during fabrication in the top and bottom of each tube assembly. The water chemistry in the SFP has been reviewed elsewhere 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 intervals of 1 to 5 years over the life of the facility.
. 3.5.2.2 Evaluation The Unit 2 SFP is fabricated of materials that will have good combatibility with the borated water chemistry of the spent fuel 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 c9ntaining boric acid has been observed throughout the countryll). 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 ins:tances 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 to 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.
3,5.2.3 Concluston From our evaluation as discussed above, we conclude that the corrosion that will occur in Unit 2 SFP will be of little signi-ficance 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 perfonned 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 (1)
J. R. Weeks, " Corrosion of Materials in Spent Fuel Storage Pools," BNL-NUREG-23021, July,1977.
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 that no significant corrosion should occur in the spent fuel storage racks at Farley Unit 2 for a period well in excess of the 40 years design life of the unit.
Therefore, we conclude that the compatability of the materials and 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 integrity of components, and is therefore acceptable.
3.6 Occupational Radiation Exposure 1
We have reviewed APCo's plan for the renoval and disposal of the low density racks and the installation of the high density racks with respect to occupa-tional radiation exposure. Since the SFP for Farely Unit 2 has never had spent fuel stored in it and is currently dry, clean and uncontaminated, there will be no additional radiation exposure to worker due to the SFP modification.
Thus, the staff concludes that SFP modification exposure to workers is as low as is reasonably achievable (ALARA) and acceptable.
We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies at Farley Unit 2 on the basis of information supplied by the licensee and by utlizing relevant assumptions for occupancy times and for dose rates in the SFP area from radionuclide concentra-tions in the SFP water. The spent fuel assemblies themselves contribute a negligible 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 dose at the unit. This small increase in radiation dose in the SFP area should not affect the licensee's ability to maintain individual occupational doses to ALARA levels and within the limits of 10 CFR Part 20. Therefore, we conclude that storing additional fuel in the Unit 2 SFP 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 demineralizers 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 effectiveness.
The SFP filters will be exchanged when AP exceeds 20 psia. The licensee 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 l
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 system 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, the licensee has established the frequency of chemical and radionuclide analysis that will be performed to monitor the water quality and the need for SFP cleanup system demineralizer resin and filter replacement.
In addition, the licensee has also set the chemical and radiochemical 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 decontamination effectiveness is reduced to less than 10 (decontamination 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 fuel pool 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.l(c) of 10 CFR Part 20, as it relates to maintaining radiation exposures as 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 Regulatory Position C.2.f(3) of Regulatory Guide 8.8, as it relates to removing crud from fluids through physical action.
3.7.3 Conclusion On the basis of the above evaluation, we conclude that the existing i
spent fuel 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.
4.0 Conclusions On the basis of the foregoing analysis, it is concluded that there will be i
no significant environmental impact attributable to the proposed action.
Having made this conclusion, the Commission has further concluded that no environmental impact statement for the proposed action need be prepared 'and 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 Commission's regulations and the issuance of this amendment will not be inimical to the common defense and 1
security or to the health and safety of the public.
Date: June 23,1982 Principal Contributors:
M. Fecteau
- 0. Rothberg l
L. Kopp B. Turovlin l
B. Lafave F. Witt E. Reeves M. Wohl i
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