ML20141M899
| ML20141M899 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/19/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20141H654 | List: |
| References | |
| NUDOCS 8603030027 | |
| Download: ML20141M899 (13) | |
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'o, UNITED STATES d
NUCLEAR REGULATORY COMMISSION o
E WASHINGTON, D. C. 20555 y
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENTS N05.116 AND l20TO FACILITY OPERATING LICENSES NOS. DPR-44 AND DPR-56 PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GA5 COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION, UNITS h05. 2 AND 3 DOCKETS NOS. 50-277 AND 50-278 P
TABLE OF CONTENTS 1.0 Introduction 2.0 Evaluation 2.1 Criticality Considerations 2.2 Spent Fuel Pool Cooling and Makeup 2.3 Installation of Racks and Load Handling 2/4 Structural Design 2.5 Materials 2.6 Spent Fuel Pool Cleanup System 2.7 Occupational Radiation Exposure 2.8 Radioactive Waste Treatment 2.9 Radiological Consequences of Cask Drop and Fuel Handling Accidents 3.0 Summary 4.0 Environmental Considerations 5.0. Conclusions 6.0 References I
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1.0 INTRODUCTION
I By letter dated June 13,1985, Philadelphia Electric Company (the licensee or PECo) made application for approval to install and use new high density spent fuel racks at Peach Bottom Atomic Power Station, Units 2 and 3.
Revision 1 to the application was submitted by letter dated August 1, 1985 in order to include some confirmatory calculations.
Further information in response to staff questions was provided in letters dated October 9, 1985 and January 30, 1986. The proposed action would increase the spent fuel pool storage capacity I
in each unit from 2608 to 3819 storage cells by replacing existing storage racks with higher density storage racks.
l 1.1 Discussion l
There are two spent fuel pools (SFPs) at Peach Bottom; one for each unit.
l The existing racks in each of these pools have 2608 total storage cells.
Amendment Nos 49 and 48 for Units 2 and 3, respectively, dated November 30, 1978, increased the original SFP storage capacity from 1110 fuel assemblies to the present design of 2608 assemblies per pool.
In the 1987-1988 time frame',
these SFP units will lose their full-core discharge reserve storage capacity (764 fuel assemblies); and in the 1991-1992 time frame, they will no longer have the capacity to store additional fuel discharges from the operating units.
The licensee, therefore, is proposing to replace the existing spent fuel storage racks with new spent fuel storage racks whose design will allow for more fuel in the same space as occupied by the current racks. The new rack structures will increase the existing spent fuel storage capacity from 2,608 to 3,819 storage cells for each unit.
The following general description of the proposed action is based upon the licensee's August 1, 1985 submittal.
i The proposed new racks are being designed and fabricated by Westinghouse Electric Company located in Pensacola, Florida.
The new racks, designed to be free standing, will be installed by setting them on the spent fuel pool floor as the old racks are removed.
As in the previous storage rack replacement at Peach Bottom in 1978, some of the pool floor swing bolts (no longer functional) will be removed to within one i
inch of the fuel poul liner to avoid interference with the support feet on the new racks. Also, to avoid rack feet interference with the pool liner seam welds, leak detection trenches, sparger support brackets and support bases of removed swing bolts, some stainless steel plates will be set in place to span these items and provide a surface for the rack feet to rest. Also the end sections j
and diffusers of the spent fuel pool (SFP) cooling discharge piping will be removed.
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2.0 EVALUATION The " Spent Fuel Storage Capacity Modification Safety Analysis Report" provided by the licensee on June 13, 1985 and revised on August 1, 1985, in support of l
this application for approval was the basis for the NRC staff evaluation.
l Supplemental information provided by the licensee is also reflected in i
4 the following Safety Evaluation which summarizes the NRC staff effort.
2.1 Criticality Considerations The rack design consists of square stainless steel cylinders which are fastened together in an egg crate-like structure.
A Boraflex sheet is located on each outer surface and held in place by a stainless steel wrapper which is welded to the cylinder.
The calculations of rack reactivity (K-effective) were performed by the licensee with the KENO-IV Monte Carlo code.
Cross sections were generated with the AMPX system of codes using the EN,)F/B-IV data base.
This code package has been used in numerous fuel rack calculations and the NRC staff finds it acceptable.
The licensee's fuel rack designer (Westinghouse) has verified the application of the code by calculating a number of critical experiment configurations and j
comparing calculated results with the experiment.
These comparisons showed essentially zero bias for the calculations with an uncertainty of 0.0032 at the 95 percent level with 95 percent confidence interval. We conclude that the calculation procedure has been suitably qualified.
Calculations of the K-effective value of the racks were performed for the three types of BWR fuel assemblies to be stored in the racks - 7x7, 8x8 and 8x8R.
Calculations were done for an enrichment of 3.5 w/o U-235 for each type.
It was determined that 7x7 assembly was the most reactive.
Uncertainties were treated either by assuming worst case conditions or by performing sensitivity studies and obtaining appropriate values.
Worst case assumptions were made for asymrtetric fuel assembly position and material properties (e.g., boron loading).
Uncertainty values were obtained for material thickness, and spacing and bowing tolerances.
Poison particle self-shielding effects were treated as a bias in the calculations.
This treatment of uncertainties meets our requirements and is acceptable.
l Postulated accidents which were considered include the loss of cooling systems, dropping a fuel assembly on top of the racks and dropping of an assembly outside l
the periphery of the racks. These accidents either do not cause an increase in the K-effective value or the increase is small compared to the margin between the nominal K-effective and the acceptance criterion of 0.95.
We conclude that l
proper analyses of the accident conditions have been performed.
The maximum value of K-effective for normal stora0e or a postulated accident condition is 0.936 including uncertainties at a 95/95 probability / confidence level. This meets our acceptance criterion of 9.95 for this quantity and is acceptable.
i We conclude that the proposed high density spent fuel storage racks are acceptable with respect to criticality.
This conclusion is based on the following:
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Calculations are performed for the fuel having the maximum reactivity.
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The calculation method has been verified against experiment.
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Uncertainties in the calculations have been properly treated.
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Credible accidents have been analyzed.
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The results of the analyses meet NRC acceptance criterion for K-effective.
Finally, the Technical Specifications (TSs) for Peach Bottom limit storage in the pool to fuel having less than 17,3 grams of U-235 per centimeter of assembly length. The licensee has confirmed that this is equivalent to 3.5 w/o U-235 enrichment in the most reactive (7x7) assembly. We conclude that the proposed rack design is acceptable for storage of assemblies meeting the TS requirements.
2.2 Spent Fuel Pool Cooling and Makeup h
The licensee calculated 13.14 MBTU/hr as the maximum " normal" heat load, to j
the pool (all spent fuel storage locations full with fuel from successive cyclic discharge) following the last refueling.
The staff performed an independent calculation for the maxir.um " normal" heat load to the pool in I
accordance with the guidelines of Branch Technical Position ASB 9-2, " Residual i
Decay Energy for Light Water Reactors for Long Term Cooling," and Standard i
Review Plan Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System" which i
resulted in a value of 16.69 MBTV/hr.
The licensee indicated that two of the j
three existing spent fuel pool cooling heat exchanger trains have a combined i
heat removal capabflity of 17.g3 MBTU/hr when maintaining the bulk pool i
temperature at 150 F.
The 150 F pool temperature is the upper limit previously approved by the staff. Thus,theheatloadcalgulatedbythe licensee results in a maximum bulk pool temperature of 135 F.
This value is based on assuming a single failure in the spent fuel pool cooling system which leavestwofuelgoolcoolingsystemheatexchangersinoperation.
This value i
is below the 150 F upper limit for bulk pool temperature,F based on the staf f for normal storage l
conditions.
The pool temperature will also be below 150 i
calculated maximum "nornal" heat load.
Thus, we have verified that the pool temperature is maintained within acceptable limits for the maximum " normal" heat load condition.
In addition, the licensee has concluded based on their analysis that no boiling would occur within the storage racks when the normal l
fuelpoolcoolingsystemiginoperationorwheneverthepooltemperatureis maintained at or below 150 F.
The licensee calculated 23.12 MBTU/hr as the l
maximum " abnormal" heat load following a full core discharge, with the remaining storage spaces full with fuel from successive cyclic discharges.
l Thig ", abnormal" heat load results in a maximum bulk pool temperature of 143 F with all cooling train heat exchangers operating.
Assuming the loss of all cooling, boiling would occur after 82 hours9.490741e-4 days <br />0.0228 hours <br />1.35582e-4 weeks <br />3.1201e-5 months <br /> for the maximum " abnormal" heat load condition.
This is a substantial time period for actions to be i
taken such as initiating makeup to the spent fuel pool.
No upper limit for the maximum " abnormal" storage condition is established in the staff criteria, and therefore, the above temperatures are acceptable.
i The spent fuel pool cooling system is normally cooled by the service water
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l system. The licensee proposed no modifications to this system as part of l
this spent fuel pool expansion project.
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_. _. Under emergency cenditions, the reactor building cooling water heat exchangers can be manually connected to provide cooling to the spent fuel pool cooling system.
It is in turn cooled by the emergency service water system. The residual heat removal system can also be utilized to supplement the spent fuel pool cooling system under abnormal heat load conditions.
The licensee has also analyzed the effects of spent fuel pool boiling on the outside environment.
The licensee utilized a ucdel similar to that previously employed for a comparable analysis on the Limerick Station to determine the offsite radiological consequences of pool boiling.
The results indicated that the resulting offsite dose was a very small fraction of 10 CFR Part 100 limits and was a negligible offsite contribution. We find this analysis and it conclusion to be acceptable.
2.3 Installation of Racks and Load Handling Currently, there is spent fuel in the Peach Bottom Units 2 and 3 spent fuel l
pools.
However, the licensee has stated that at no time will the cask handling crane carry a spent fuel storage rack : nodule over stored spent fuel. The licensee has committed to employ heavy load handling procedures, safe load paths and installation procedures as part of the administrative controls to precluce the potential for the mishandling of rack modules and miscellaneous heavy load items during the rerack operation over the spent fuel pool.
The licensee has also performed a load drop analysis for the rack module.
The results of that analysis indicate that the proposed spent fuel pool modifications will not result Jn fuel damage and that the resulting radiological consequences will not be in excess of the fuel handling accident previously evaluated in the upcated Peach Bottom FSAR. The postulated rack drop also would not change tn't minitc.uS separation distance between the stored fuel assemblies or the concentratics of fixed neutron absorbing material between the adjacent fuel assemblies.
O,erefore, the margin of safety to criticality will also not be affacted by the postulated rack drop accident.
The licensee has committed to use the main book of the reactor building cr.ane for lifting the existing spent fuel storage racks and the new storage racks.
The main hook and its associated load lifting system on the reactor building t
crane are of a single failure proof design, such that a single failure will not result in dropping the load.
The auxiliary nook on the reactor building crane will be used only for lifting small miscellaneous items whose weights are less than that of a fuel assembly.
This will ensure that the consequences of their being dropped is bounded by the existing FSAR fuel handling accident analysis.
The refueling bridge crane will be used for lifting fuel assemblies and transferring them within the pool in accordance with the existing station approved procedures.
The licensee has stated that the cask handling crane meets the desigr. and operational criteria of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" and NUREG-0554, " Single Failure Proof Cranes for Nelear Power Plants." We have verified that the stoff's previous safety e nluation report i
for NUREG-0612 has found this crane in ccepliance with the applic6ble guidelines for the control of heavy loads.
Therefore, we conclude that the handling of heavy loads during the spent fuel pool expansion modification will be in conformance with staff criteria and is acceptable.
-S-Based upor. the discussions in Sections 2.2 and 2.3 above, we conclude that the proposed SFP modifications for each SFP with respect to the developed heat loads, pool water temperatures, and load handling practices are in accordance with applicable criteria and are, therefore, acceptable.
2.4 Structural Design Our evaluation of the structural aspects of the proposed modifications are based on a review performed by the staff's consultant, Franklin Research Center (FRC). The FRC Technical Evaluation Report (TER) is appended to this safety analysis and provides additional details relating to the structural evaluation.
The SFPs are reinforced concrete structures located inside the Reactor Building l
in an elevated position adjacent to the North (Unit 2) and South (Unit 3) sides j
of the drywell shield walls.
The walls and floor of the SFP are lined with a stainless steel liner.
This liner serves only as a water tight boundary, and it is not a structural member.
The new high density racks are stainless steel " egg-crate" structures.
Each cell would contain a spent fuel assembly. Weight of the rack and fuel is transmitted to the floor of the pool through supporting legs.
The racks are each free-standing on the pool floor and a gap is provided between the racks and l
between racks to pool wall so as to preclude impact during an earthquake.
j Load combinations and acceptance criteria were compared with those found in the " Staff Position for Review and Acceptance of Spent Fuel 5torage and viandling Applications" dated April 14, 1978 and amended January 18, 1979.
The existing concrete pool structure was evaluated for the new loads in accordance with the i
requirements of the applicable portions of NRC Regulatory Guides 1.12, 1.142, and Standard Review Plan 3.8.4.
The pool structure re-analysis also uses ACI 318-83, ACI 349-80, and AISC Standards.
i Loads and load combinations for the racks and the pool structure were reviewed
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and found to be in agreement with the applicable portions of the staff position.
Additional details are provided in the appended TER.
Seismic loads for the rack design are based on the original design floor acceleration response spectra calculated for the plant at the licensing stage.
The seismic loads were applied to the model in three orthogonal directions.
Loads due to a fuel bundle drop accident were considered in a separate analysis.
The postulated loads from these events were found to be acceptable.
The dynamic response and internal stresses and loads are obtained from a seismic I
analysis which is performed in two phases.
The first phase is a time history analysis on a nonlinear finite element model.
The second phase is a response spectrum analysis of a detailed linear three dimensional finite element model of the rack assembly.
Further details on the methedology may be found in the appended TER.
Calculated stresses for the rack components were found to be within allowable limits. The racks were found to have adequate margins against sliding and
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tipping.
An analysis was conducted to assess the potential effects of a dropped fuel j
assembly on the racks and results were considered satisfactory.
An analysis was conducted to assess the potential effects of a stuck fuel assembly causing an uplift load on the racks and a corresponding downward load on the lifting device as well as a tension load in the fuel assembly.
Resulting stresses were found to be within acceptable limits by the staff.
The existing structures were analyzed for the modified fuel rack loads using a finite element computer program. Original plant respcese spectra and damping values were used in consideration of the seismic loadings, and the existing SFPs are determined to safely support the loads generated by the new fuel racks.
We, therefore, conclude that the proposed rack installation will satisfy the requirements for 10 CFR Part 50, Appendix A (General Design Criteria 2, 4, 61, and 62), as applicable to structures.
2.5 Materials The safety function of the SFPs and storage ra:k system is to maintain the spent fuel assemblies in a sub-critical array during all credible storage conditions. We have reviewed the compatibility and chemical stability of the materials, except the fuel assemblies, wetted by the pool water.
The spent fuel racks in the proposed expansion would be constructed entirely of Type 304 LN stainless steel, except for leveling screws which are Type 17-4 PH stainless steel and the neutron absorter inaterial. The high density spent fuel storage racks will utilize Eoraflex sheets as a neutron absorcer.
Boraflex consists of boron carbide powder in a rubber-like silicone polymeric matrix.
Tre spent fuel storage rack configuration is composed of individual storage cells interconnected to form an integral structure.
The space which contains the Boraflex is vented to the pool.
Venting 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 stainless steel tube.
The pool liner, rack lattice structure and fuel storage tubes are stainless steel which is cespatible with the storage pool environment.
In this environment of oxygen-saturated borated water, the corrosive deterjp, ration of the Type 304 stainless steel should not exceed a depth of 6.00 x 10 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 storate tubes, and the Zircaloy in the spent fuel assemblies will not be significant because the materials are either similar or the materials are protected by highly pessivating oxide films and are J
therefore at similar potentials.
The Boraflex is composed of non-metallic i
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.
The evaluation tests have shown that the Boraflex is unaffected by the pool water environment and will not be degraded by corrosion (1).
Tests g re performed at the University of Michigan (2), exposing Boraflex to 1 x 10 rads of gamma radiation with substantial concurrent neutron flux in deionized water.
Irradiation will cause some loss of flexibility, but will not lead to break up of the Boraflex.
The annulus space in each cell assembly which contains the Boraflex is vented to the pool.
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 and swelling of the inner stainless steel tube.
The tests (1) have shown that neither irradiation, environment nor Boraflex composition has a discernible effect on the neutron transmission 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 present in the Boraflex will typically contain 0.1 wt percent of soluble boron.
The tests results have confirmed the encapsulation function of the silicone polymer matrix in preventing the leaching of soluble species from the boron carbide.
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 inservice surveillance program.
Surveillance samples are in the form of removable stainless steel clad Boraficx sheets, which are proto-typical of the fuel storage cell walls.
These specimens will be removed and examined periodically over the expected service life.
From our evaluation as discussed above, we conclude that the corrosion that will occur in the spent fuel storage pool environment should be of little significance during the 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 i
temperatures in deionized water indicate that the Boraflex material will not undergo significant degradation during the expected service life.
We further conclude that the environmental compatibility and stability of the materials used in the expanded spent fuel storage pool is adequate based on
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the test data cited above, and the actual service experience in operating reactors. We have reviewed the surveillance program and we conclude that the monitoring of the materials in the SFPs, as proposed by the licensee, will I
provide reasonable assurance that the Boraflex material will continue to perform its function for the design life of the pool.
The materials surveillance program delineated 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.
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.
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_- We, therefore, find that the implementation of an inservice surveillance program and the selection of appropriate materials of construction by the licensee meets the requirements of 10 CFR 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 neutron absorber and is, therefore, acceptable.
2.6 Spent Fuel Pool Cleanup System i
The SFP cleanup system is part of the pool cooling system.
It consists of a full flow (550 gpm) filter-demineralizer composed of a filter precoat powdered ion exchange resin.
This cleanup system is similar to such systems at other nuclear plants which maintain concentrations of radioactivity in the pool water at acceptably low levels. The staff expects only a small increase in radioactivity 3
released to the pool water as a result of the proposed modification. We, therefore, conclude that the spent fuel pool cleanup system is adequate for the proposed modification and will keep the concentrations of radioactivity in the pool water to acceptably low levels.
2.7 Occupational Radiation Exposure The staff has reviewed the licensee's plan for the modification of the Peach Bottom SFP racks with respect to occupational radiation exposure.
The licensee estimates that the exposure for this operation will be approximately 36 man-rems.
This estimate is based on the licensee's breakt'own of occupational J
exposure for each phase of the modification.
The licensee considered the number of individuals performing a specific job, their occupancy time while performing this job, and the average dose rate in the area where the jcb is being performed.
The spent fuel assemblies themselves contribute a negligible i
amount to dose rates in the pool area because of the depth of water shielding j
the fuel.
One potential source of radiation is radioactive activation of corrosion products, termed " crud".
Crud nay be released to the pool water because of fuel movement during the proposed SFP rack modifications. This could increase radiation levels in the vicinity of the pool.
During refuelings, when the spent fuel is first moved into the fuel pool, the addition of crud to the pool water from the fuel assembly and from the introduction of primary coolant to the pool water is greatest.
However, the licensee, based upon previous experience from performing similar modifications, does not expect to have significant releases of crud to the pool water during modification of the 7
i SFP racks.
In addition, the purification system for the pool (SFP Cleanup l
System), which has maintained radiation levels in the vicinity of the pool at low levels during normal operations, will be operating during the modification of the SFP racks.
The staff has evaluated the licensee's proposed crud reduction program in the SFP and finds it acceptable.
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- The presently installed racks will be individually lifted from the SFP and will be rinsed either with low or high pressure water to remove any loose radioactivity.
The racks will then be moved to a receiving area for appropriate disposal.
Currently, the licensee has proposed decontaminating the racks and then disposing of the clean material as industrial waste.
Material that cannot be decontaminated will be disposed of as normal radioactive waste.
Either disposal method used will follow ALARA (as-low-as-reasonably-achievable) guidelines.
Divers will be used during the SFP rack modification.
The licensee has developed specific procedures using the recommendations of Regulatory Guide 8.8 to ensure that doses to the divers will be within the requirements of 10 CFR Part 20 and ALARA guidelines. The ALARA procedures for divers include:
reshuffling of the spent fuel; radiation surveys after the fuel is reshuffled to map radiation zones; instruction to divers on their travel limits within the pool; and constant monitoring of divers' radiation dose.
The staff's evaluation of the Peach Bottom's proposed SFP rack modification includes a review of the manner in which the licensee will perform the modification, the radiation protection program, including the use of area and airborne radioactivity monitoring, and the use of relevant experience from other operating reactors that have performed similar SFP rack modifications.
Based on this review, the staff concludes that the Peach Bottom SFP rack modification can be performed in a manner that will ensure ALARA exposures to workers.
In addition, the staff has estimated the increment in onsite occupational dose during normal operations after the pool modifications resulting from the proposed increase in stored fuel assemblies.
This estimate is based upon information supplied by the licensee for occupancy times and for dose rates in i
the SPF area from radionuclides concentrations 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, the staff estimates that the l
proposed modification should add less than one (1) percent to the total annual occupational radiation exposure at the plant.
The small increase in radiation exposure should not affect the licensee's ability to maintain individual occupational dose to ALARA levels and within the limits of 10 CFR Part 20.
l Thus, the staff concludes that storing additional fuel in the SFP will not result in any significant increase in dose received by workers.
l 2.8 Radioactive Waste Treatment The plant contains waste treatment systems designed to collect and process the gaseous, liquid and solid waste that might contain radioactive material.
The waste treatment systems were evaluated in the Final Environment Statement (FES) dated April 1973.
There will be no change in the waste treatment systems described in Section III.2 of the FES because of the proposed modifications.
There will be an expected modest increase in the loadings on the Spent Fuel l
Cleanup system ( refer to Section 2.6-Spent Fuel Pool Cleanup System).
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. 2.9 Radiological Consequences of Cask Drop ar;d Fuel Handling Accidents This portion of the staff's review was conducted in accordance with the guidance in NUREG-0800, " Standard Review Plan", Sections 15.7.4 and 15.7.5, Regulatory Guide 1.25 and NUREG-0612 with respect to accident assumptions.
The licensee has committed to follow existing technical specifications regarding allowable loads carried over stored spent fuel during the reracking procedure, and during normal operation after its completion.
The staff agrees with the licensee that the change in radiological conditions which can influence accident conditions in the SFP after the increase in l
storage capacity will be negligible compared with that prior to the modifications.
The fuel burn-up (assumed to be 40,000 mwd /MTU), pool water level, and iodine decontamination factor will remain unchanged.
The Peach Bottom Safety Evaluation Report, dated August 1972, was evaluated for a less tightly packed pool.
However, even though more assemblies could possibly be impacted in a dropped assembly accident with more dense arrangement, the radiological consequences of this accident will not significantly increase.
Therefore, the radiological analysis of the cask drop, fuel assembly, and heavy load accident is unchanged from that previously analyzed for the existing spent fuel pool configuration (Safety Evaluation Report for Peach Bottom Units 2 and 3, August 1972).
In addition, the staff has performed an independent bounding analysis based upon this modification which shows that the doses at the Exclusion Boundary and Low Population Zone will be well within the SRP 15.7.4 dose guidelines.
Therefore, the staff concludes that proposed modification is acceptable.
3.0 Summary Our evaluation supports the conclusion that the proposed modification to the Peach Bottom SFP is acceptable because:
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(1) The physical design of the new storage racks will preclude criticality for any credible moderating condition.
(2) The SFP cooling system has adequate cooling capacity.
(3) The installation and use of the proposed fuel handling racks can be accomplished safely with the limit that ne rack modules will be moved over any spent fuel assemblies.
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(4) The installation and use of the new spent fuel racks can be done I
safely and will not alter the consequences of the design basis l
accident for the SFP, i.e., the dropping and rupture of a fuel assembly and subsequent release of the assembly's radioactive inventory within the gap.
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- (5) Tht. likelihood of an accident involving heavy loads in the vicinity of the SFP is negligible.
(6) The structural design and materials of construction are adequate to function normally for the duration of the plant lifetime and to withstand the seismic loading of the design basis earthquake.
(7) The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the SFP would be negligible.
4.0 Environmental Considerations A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51.
5.0 Conclusions We have concluded, based on the consideration 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 activity 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:
February 19, 1986 The following NRC personnel have contributed to this Safety Evaluation: W. Brooks, J. Raval, S.Kim, R. Fell, M. Lamastra, H.Gilpin, F. Witt and G. Gears 6.0 References 1.
J.S. Anderson, "Boraflex Neutron Shielding--Product 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, August 1981.