ML20083H818
| ML20083H818 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 12/15/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20083H798 | List: |
| References | |
| NUDOCS 8401050041 | |
| Download: ML20083H818 (14) | |
Text
,
UNITED STATES e
NUCLEAR REGULATORY COMMISSION i.
g waswinorow,o.c.2oses l
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 61 TO FACILITY LICENSE NO. DPR-71 AND AMENDMENT NO. 87 TO FACILITY LICENSE NO. OPR-62 CAROLINA POWER & LIGHT COMPANY SRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 i
- 1. 0 INTRODUCTION By letter dated April 16, 1981 and supplemented by letters dated June 22, and November 23,1981, March 16, April 5, May 20, and September 16, 1982, February 23, March 31, May 5, and September 20, 1983.
Carolina Power & Light Company (CP&L, the licensee) requested amendments to Facility Operating Licenses DPR-71 and DPR-62 for Brunswick Steam Electric Plant, Units 1 and 2.
The request is to authorize increased storage capability for boiling water reactor (BWR) fuel in the spent fuel pools (SFP) for the two nuclear units.
The authorized storage capability for pressurized water reactor (PWR) fuel would be decreased.
The proposed modifications would change the SFP storage spaces from 616 PWR or 1386 BWR spaces per unit to 160 PWR and 1803 BWR licensed spaces for Unit 1 and 144 PWR and 1839 BWR licensed spaces for Unit 2.
This expanded storage capacity will allow the continued operation of the two nuclear units with onsite storage of spent fuel to 1988 for Unit 1 and 1987 for Unit 2.
This expanded capacity would provide full core discharge capability until the stated times.
The licensee's proposal would increase the SFP storage capacity by replacing some of the existing spent fuel storage racks with new high density storage racks.
The new racks will contain neutron absorber material in the rack walls so that spacing between stored assemblies can be reduced while maintaining adequate criticality margin.
1173 new spaces in each fuel pool are contained in high density racks made up of modules, each module being composed of six-inch square cells, each cell accommodating a single BWR fuel assembly.
The cell walls contain a neutron absorber material sandwiched between sheets of stainless steel. The cells making up the module have 6.56-inch center-to-center spacing. The spacing is sufficient to maintain KeH. below 0.95.
The racks are also designed in such a manner that accidental dropping of a fuel assembly will not cause a geometry that could result in criticality, o
The licensee's basic supporting document for this action is a report, Spent Fuel Storage Expansion Report, that was attached to their submittal dated April 16, 1981.
The report contains an overall description of the racks, their design bases, a description of the proposed installation as well as the licensee's analyses and evaluations supporting the proposed spent fuel pool expansion.
The staff evaluation of the safety considerations associated with this proposed action are addressed below.
A separate Environmental Impact Appraisal has 4
been prepared for this action.
1 l
"~ 8401050041 831215 I
PDR ADOCK 05000324 P
PDR J
2.0 DISCUSSION AND EVALUATION 2.1 Structural and Mechanical Design Considerations Description of the Spent Fuel Pool and Racks Both units at Brunswick are Mark I BWRs.
The spent fuel pools are right and left hand and symmetrical with respect to the transverse centerline of the plant.
The pools are elevated with the top of the pools at the fueling floor level, elevation 117'4".
Grade is at elevation 19'6".
The inside dimensions of the pools are:
height = 38.75' length = 46' width = approximately 28' The pool structures are reinforced concrete with floor thickness of about 5.5' and walls of various thickness from 4 to 5'.
The ends of each pool (column lines N and P) are a portion of two prestressed concrete girders which run the length of each reactor building.
Each of these girders is 140' long by 5' wide by 42.33' deep.
The girders support the spent fuel pool, reactor well, steam separator well, and portions of the floor slabs at elevations 80'0",
98'8", and 117'4".
The ends of the girders are supported by the exterior reactor building walls and are independent of the reactor containments.
Each pool is lined with a continuous, welded, watertight, 1/4" thick stainless steel plate which is backed up by leak-chase channels at all seams.
It is proposed to remove existing racks at the end of each pool and replace them with high density racks.
Existing racks in the center portions of each pool will remain in place.
The existing racks are fabricated of stainless steel and are supported within an " egg-crate" grid of trusses which rests on the pool floors.
Individual existing racks will be removed from the grid of trusses in order to make room for the new racks and the grid is to remain in place.
The new racks (also stainless steel) are to be installed ever the existing grid on a system of free-standing pedestals which will allow the free-standing racks to bridge the grid.
Leveling of the racks will be accomplished with shims, if necessary.
The new racks are stainless steel " egg-crate" structures.
The 15 by 17 cell rack is about 8.33' wide by about 9.33' long by about 14.5' hign.
The pedes-tais mentioned above are heavy stainless steel plate, the largest of which is about 3.33' long by about 2' wide by about 1.8' high.
The pedestals are placed at the corners of the racks and several of them support adjacent cor-ners of two racks.
The pedestals are constrained by friction.
2 i
t m
Evaluation Applicable Codes, Standards and Specifications Structural material of the racks conforms to the ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Subsection NF.
Computed stresses were com-pared with the ASME B&PV Code,Section III, Subsection NF.
Load combinations and acceptance criteria for the racks are in accordance with the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and amended January 18, 1979 (hereafter referred to as the "NRC Position").
Buckling criteria for the cold formed stainless steel portions of the racks were in accordance with the AISI " Cold Formed Stainless Steel Structural Design Manual", 1979 edition.
The pool structure was evaluated in accordance with the requirements of ACI-318-71 which is found acceptable.
, Load and Load Combinations Loads and load combinations for the racks and the pool structure were reviewed and found to be in conformance with the applicable portions of the NRC Position.
Seismic and Impact Loads Seismic loads are based on the original design floor acceleration response spectra calculated at elevation 80'0" of the plant.
This was based on a 0.16 g. DBE with 4 percent structural damping and 1 percent equipment damping.
Acceleration in the vertical direction was computed as being 2/3 of horizontal acceleration.
Impact loads due to rack / fuel bundle interaction were considered.
Loads due to a fuel bundle drop accident were considered in a separate analysis for such an occurrence.
The postulated loads from such events were found to be acceptable.
Design and Analysis of the Racks Artificial earthquake time Histories were generated from the floor response spectra and used as input to a 2 dimensional stick model of the racks.
The racks were found to be essentially rigid in the vertical direction for pur-poses of structural design.
Effects of added hydrodynamic effects due to motion of the racks in the water were accounted for.
A separate analysis using a simplified model and a constant 1 g acceleration was used to assess the effects of rack / fuel bundle impact on the rack structural design.
An analysis for sliding and tipping of the racks was accomplished.
It was found that the racks are placed sufficiently far apart to preclude the possi-bility of rack-to-rack or rack-to pool structural interaction.
The racks and pedestals were found to be stable for all postulated events.
The rack structural design produced calculated stresses for the rack compo-nents which were well within allowable limits.
The peaestals were designed 3
l
for worst-ctse lo& ding.
It was found that an analysis w:s conduct;d to assess the potential effects of a dropped fuel bundle on the racks and results were considered satisfactory.
Upward loads are bounded by other loading conditions for the racks.
- Also, overload cutoff devices are installed on the fuel handling cranes in order to preclude uplift damage to the racks or fuel.
Thus the design basis foF upward loads was found to be acceptable.
Seismic Analysis of the Pool Structure An analysis of the reinforced concrete pool structure was conducted by the licensee, and it was found that each pool floor is adequate to withstand the effects of added loads due to the new racks under seismic loads.
The pre-stressed concrete girders (previously described) were found to be unaffected i
by the worst-case new rack loads.
We find that with respect to structural and mechanical design the subject modification proposed by the licensee satisfies the applicable requirements of General Design Criteria 2, 4, 61, and 62 of 10 CFR, Part 50, Appendix A and is acceptable.
2.2 Material Considerations Discussion We have reviewed the compatibility and chemical stability of the materials (except the fuel assemblies) wetted by the pool water.
In addition, our review has included an evaluation of the Boral neutron absJrber material used in the high density storage locations for environmental stability.
The proposed spent fuel storage racks are fabricated primarily of Type 304 stainless steel, which is used for all structural components, except for a special low friction material (which is 99% graphite) used as a foot pad between the module and the support pad.
Beral plates, used as a neutron absorber, are an integral nonstructural part of the basic fuel storage tubes.
The Boral plates are sandwiched between the inner and outer wall of the storage tubes, and are not subject to dislocation, deterioration, or removal.
The compartments in the storage tubes containing the Boral are exposed to the spent fuel pool environment through small openings formed during fabrication in the top and the bottom of each tube assembly.
The Brunswick Units 1 and 2 spent fuel storage pools contain high purity water.
The chlorides are specified to be less than 0.2 ppm, the pH specified to be in the range 6.0 to 7.5, and the conductivity specified to be less than 1 pS.
The design temperature of the water in the pool is 150 F, maximum.
At most times during normal operation, the spent fuel pool temperature would never reach this design level, reaching a calculated maximum of 145'F immedi-ately after a refueling operation and dropping rapidly thereafter until the next scheduled refueling.
4 4
The Type 304 stainless steel rack modules have been welded and inspected by i
nondestructive examination in accordance with the applicable provisions of ASME Boiler and Pressure Vessel. Code,Section IX.
The licensee will perform a materials compatibility monitoring program consist-ing of two types of specimens:
the first are 8" x 8" coupons of Boral covered with stainless steel, and the second consisting of 6" x 6" samples of Boral without stainless steel cladding.
The stainless clad coupons have two sides open to permit water access to assess if any galvanic attack may be occurring betwe3n the aluminum and the stainless steel.
It is particularly important to evaluate if this potential degradation mechanism might lead to loss of the neutron absorbing capabilities of the Boral.
Sufficient coupons will be included to permit destructive examination of a sample on inspection intervals of 1 to 5 years over the life of the facility.
Evaluation i
The Brunswick Units 1 and 2 spent fuel pools contain neutral, extremely high quality water in which all the materials of fabrication are expected to have good compatibility.
The corrosion rate of Type 304 stainlass steel in water of this quality and temperature is so low as t0 defy our ooility to measure it.
Galvanic effects between stainless steel, aluminum, and graphite are also unlikely in water of this quality as is stress corrosion cracking of weld sensitized stainless steel that may be present where the fuel storage cells are welded together.
No instances of corrosion of these materials in water of this quality has been observed at any spent fuel storage pools in the country, some of which have been in operation for close to 20 years.
No 1
loss of boron carbide from the Boral has been observed in material exposed to an environment similar to that in the Brunswick spent fuel storage pools for periods up to 20 years,
2 The venting of the cavities containing the Boral to the spent fuel pool environment will ensure that no gaseous buildup will occur in these cavities that might lead to distortion of the racks.
Flooding them will not intro-2 duce any significant corrosion problems where the aluminum is in contact with stainless steel in water of this quality.
The low friction foot pads are made of a stable material, graphite, that will not be significantly affected by radiation or water and will not release significant quantities of potentially corrosive materials to the envirnnment.
The codes and standards used in fabricating and inspecting these new spent fuel storage racks should ensure their integrity and minimize the likelihood that any stress corrosion cracking of the racks themselves will occur during service.
The materials surveillance program spelled out by the licensee as outlined above will reveal any instances of corrosion of the Boral that might lead to loss of neutron absorbing power during the life of the new spent fuel racks.
While we do not anticipate that such corrosion processes will occur, this monitoring program will ensure that, in the unlikely situation that corrosion should develop, the licensee and the NRC will be aware of it in sufficient time to take corrective action.
From our evaluation as discussed above, we find that the corrosion that will occur in the Brunswick Steam Electric Station Units 1 and 2 spent fuel storage 5
pools will be of little significance during the remaining life of the plant.
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 pcals at other boiling water reactor sites where the water chemistry is maintained to comparable standards to those in force at Brunswick.
The proposed Materials Surveillance Program is adequate to provide warning in the unlikely event that deterioration of the neutron absorbing properties of the Boral will develop during the design life of the racks.
Therefore, with the selec+ ion of the materials and water chemistry, we believe that no significant corresion should occur in the spent fuel storage racks at Brunswick Units 1 and 2 for a period well in excess of the 40 years design life of the unit.
2.3 Installation and Heavy Load Handling Consideration The results of the generic review of " Control of Heavy Loads at Nuclear Power Plants" (NUREG-0612) will not be completed until after the spent fuel storage modification has commenced.
Thus, our review and evaluation of the heavy load handling operation has been limited to those activities associated with the spent fuel storage modification.
The installation of the spent fuel racks will be accomplished with the reactor building crane.
The main hook of this crane is rated at 125 tons, while the heaviest load identified with the modification is the 15 x 17 storage rack weighing approximately 16 tons.
Based on modifications to the crane and
" upgrade" commitments by the licensee, the Staff previously concluded that the integrated design of the crane and controls with respect to the single failure criterion was acceptable.
Based on the above, we conclude that the reactor building crane is acceptable for use for the modification.
A single failure proof lifting device for the handling of the new storage racks has been provided Dy the manufacturer.
The lifting device has a rating of 39,000 lbs. and has been tested to 125% of capacity, while the heaviest storage rack is approximately 32,000 lbs.
The removal of the existing storage racks will use the existing lifting device which has a safety factor of 5 on yield stress for cesign load of 9400 lbs. (maximum empty rack weight).
The lifting apparatus such as slings, shackles and fittings are sized to maintain a minimum safety factor of 5 (based on ultimate strength static load only).
We conclude that the lifting devices and other apparatus used for the handling of the storage racks are adequate, and therefore, acceptable.
Specific load path instructio7s precluding the travel of heavy loads over stored spent fuel will be developed and implemented prior to the modifica-tions.
Also, handling procedures will be such that the storage racks which contain fuel will not be immediately adjacent to the rack being moved. We conclude that the proposed procedures are adequate, and therefore, acceptable.
Regarding operator training, qualifications and conduct, the licensee has stated that all operators are trained in accordance with the requirements of ANSI / 830. 2-1976.
The crane inspection, testing and maintenance program is also in conformance with the above referenced industry standard.
We conclude that these are acceptable.
6 O
M
l 2.4 Spent Fuel Pool Cooling Considerations System Descriotion Each BSEP Unit has an independent spent fuel pool and spent fuel pool cooling system (SFPCS).
The major components of the SFPCS consists of two pumps in parallel, with one heat exchanger in series with each pump.
These heat exchangers are cooled by the reactor building closed cooling water system.
The heat removal capability of the SFPCS is 6.53 x 108 Btu /hr at 125*F and 12.0 x 106 Btu /hr at 150 F.
The residual heat removal (RHR) system can be crosstied with the SFPCS in the event supplemental heat removal capability is required.
The refueling cycle for BSEP is an annual quarter core discharge of 140 fuel assemblies.
Each assembly is assumed to have experienced a continuous power level of 4.35 MW, prior to discharge.
For both the normal refueling and full core discharge, the fuel will be subject to a 24-hour decay period after shutdown prior to its transfer to the spent fuel pool.
The licensee's calculated spent fuel discharge heat load to the pool, which was determined in accordance with the Branch Technical Position ASB 9-2 "Resi-dual Decay Energy for Light Water Reactors for Long Term Cooling," indicates that the expected maximum normal heat load following the last refueling is 14.1 x 106 Btu /hr.
This heat load results in a maximum bulk pool temperature of 145.2 F.
The expected maximum abnormal heat load following a full core discharge after the last normal refueling discharge is 29.2 x 108 Btu /hr.
This abnormal heat load results in a maximum bulk pool temperature of 124.6 F if the RHR system supplements the SFPCS.
A maximum pool temperature of 197.2 F is expected if only the spent fuel pool cooling system is used.
Evaluation The American National Standard ANS 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 oper-ating conditions with all storage full.
Thus the RHR system will be operational and crosstied with the SFPCS prior to the discharge of a full core inventory into the pool.
The design of the storage pool is such that the fuel will always be covered with water.
The top of the stored fuel is at an elevation lower than the bottom of the pool gate which separates the reactor well from the storage pool.
Also, all piping which enter the storage pool are equipped with check valves and syphon breakers above the pool elevation to prevent inadvertent drainage.
Normal makeup to the spent fuel pool is the demineralized water system via the SFPCS.
In the event of a loss of the SFPCS, the heatup time to commence boiling is approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
This is sufficient time to provide 28 gpm emergency makeup to the pool by the fire system.
We have reviewed the calculated heat values and conclude that the heat loads are consistent with the Branch Technical Position ASB 9-2.
The spent fuel pool cooling system performance and the available makeup syr,tems have been reviewed and found to be acceptable.
7
,i
/
)
2.5 Criticality Considerations The spent fuel storage modules consist of a checkerboard array of double-walled stainless steel cans which are held together by stainless sterl angles which are welded over the entire length of the cans.
The space between the two walls of the can centains a neutron poison in the form of Boral sheets having a minimum areal concentration of 0.013 grams per square centimeter of boron-10.
The spacing between storage locations is 6.563 inches.
Calculations were performed with fuel having an infinite multiplication value of 1.35 in the standard core configuration at a temperature of 20 C (68 F).
This is equiva-lent to a standard 8 x 8 fuel assembly having a uniform enrichment of about 3.2 weight percent U-235.
This is a higher value than is expected for the reload fuels at their point of maximum reactivity.
The analyses are performed with the MERIT program a three-dimensional Monte-Carlo code that uses the ENDF/B-IV cross section set.
This code has been benchmarked against a wide range of critical assemblies including BWR fuel criticals and criticals with flat plate boron absorbers.
Comparison of calculation and experiment show that MERIT underpredicts the multiplication factor by about 0.5 percert with a standard deviation (1 o) of 0.2 percent.
The calculated effective multiplication factor for the nominal rack design is 0.867 0.009 including the MERIT calculational bias and uncertainty.
Sensitivity calculations were performed in order to assess the effect of stainless steel thickness variations, temperature variations, center-to-center pitch, and locations of the assembly in the storage tube.
The results of these studies showed that either the nominal case had the largest reactivity l
or that the uncertainty due to the variation was within the statistical uncer-tainty of the Monte-Carlo calculation.
The effect of droppir.g a fuel assembly and a loss of fuel pool cooling have been investigated.
For the former event the multiplication factor remains below 0.90 and for the latter case the 20 C water temperature case represents the maximum r.eactivity.
The interaction of the new storage racks with the existing system was investigated.
The minimum separation between the new BWR storage racks and the existing PWR racks will be six inches.
At this distance there is no significant neutron communication between the two systems.
The maximum multi-plication factor for the combined system is therefore that of the existing PWR racks - 0. 95.
Based on our review we find that the reactivity aspects of the proposed BWR rack design are acceptable.
Our bases were as follows:
1.
The calculations were performed with a state-of-the-art code which was benchmarked by comparisons with critical experiments, 2.
Conservative assumptions were made with respect to the input parameters to the code, 3.
The effect of biases and uncertainties have been treated, 4.
Interactions with storage racks existing in the pool have been treated, 8
5.
The resultant calculation meets our acceptance criterion of less than or equal to 0.95 for the multiplication factor, and 6.
Sufficient margin exists between the calculated multiplication factor and the acceptance criterion to account for any uncertainty due to the varia-tion in the way the infinite multiplication factor of the stored assemblies is obtained (for example, a uniform enrichment design as opposed to a poisoned design partially burned up but having the same infinite multi-plication factor as the uniform enrichment design).
We find that BWR fuel having an infinite multiplication factor less than or equal to 1.35 (at its most reactive point in life) in standard core geometry at 20 degrees Centigrade may be safely stored in the proposed racks.
The j
1.35 value shall be calculated for the axial segment of the assembly having j
the highest reactivity.
The licensee proposes to verify the presence of the Boral in the racks by scanning each storage location with a neutron source and detector.
This is an acceptable procedure for verifying the presence of the Boral in the racks.
2.6 Spent Fuel Pool Water Cleanuo Considerations Description The spent fuel pool cleanup system is incorporated as a part of the spent fuel l
pool cooling system.
The spent fuel cooling system for each plant consists of l
two pumps, two heat exchangers, two filter demineralizers, two skimmer surge tanks, associated piping, valves and instrumentation.
The skimmer surge tanks are designed to remove debris from the pool water and provide pump suction.
The filter demineralizers (mixed bed resin) are designed to remove corrosion products, fission products, and impurities from the pool water.
The deminer-alizers, like the pumps, are connected in parallel for operational flexibility.
Pool water purity is monitored by a continuous conductivity meter installed on the inlet to the fuel pool demineralizers, and by periodic grab samples for laboratory analysis.
Once a week, samples are taken foY chemical and radio-chemical analysis.
Demineralizer resin will be replaced when either:
(1) effluent conductivity equals influent conductivity at values above 1 pmho/cm, (2) effluent conductivity exceeds 1 pmho/cm by a significant margin, or (3) differential pressure reaches 25 psi.
The licensee indicated that no change or equipment addition to the spent fuel pool cleanup system is necessary to maintain pool water quality for the increase in fuel storage capacity.
E_ valuation Past experience showed that the greatest increase in radioactivity and impurities in spent fuel pool water occurs during refueling and spent fuel handling.
The refueling frequency and the amount of core to be replaced for each fuel cycle, and frequency of operating the spent fuel pool cleanup system are not expected to increase as a result of high density fuel storage.
The chemical and radionuclide composition of the spent fuel pool water is not expected to change as a result of the proposed high density fuel storage.
9
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 radiochemical analysis that will be performed to monitor the water quality and the need for spent fuel pool cleanup system demineralizer resin and filter replacemcat.
In addition, the licenste has also set the chemical and radiochemical guidelines to be used in monitoring the spent fuel pool water quality and initiating corrective action.
These guidelines are consistent with the reactor coolant Technical Specification water quality requirements.
The facility contains waste treatment systems designed to collect and process the gaseous, liquid, and solid wastes that might contain radioactive naterial.
The waste treatment systems were evaluated in the Safety Evaluation, dated November 1973.
There will be no change in the waste treatment system or in the conclusions given in Sections 9.0 and 11.0 of the evaluation of these systems because of the proposed modification.
On the basis of the above, we determined that the proposed expansion of the spent fuel pool will not appreciably effect the capability and capacity of the spent fuel pool cleanup system.
More frequent replacements of filters and demineralizer resin, if necessary, could offset any potential increase in the i
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 fuel storage (1) provides the capability and capacity of removing radioactive materials, corrosion products, and impurities from the prol and thus meets the requirements of GDC 61 in Appendix A of 10 CFR Part 50 as it relates to appropriate systems to 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) of 10 CFR Part 20, as it relates to maintaining radiation exposures as low as is reason-ably achievable; (3) confines radioactive materials in the pool water within 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 contaminants from the sources; and (4) remov,es 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.
Therefore, no change to the spent fuel pool cleanup system is required.
2.7 Occuoational Radiation Exposure We have reviewed the licensee's plan for the removal and disposal of the low density racks and the installation of the high density racks with respect to occupational radiation exposure.
The occupational exposure for this operation is estimated by the licensee to be approximately 81 man-rem.
This estimate is based on the licensee's detailed breakdown of occupational exposure for each phase of the modification.
The licensee considered the number of individuals performing a specific job, their occupancy time while performing this job, and the average dose rate in the area where the job was being performed.
In several instances he is conservative in his ' estimation of dose-rate and man-hours to perform a specific operation.
Crud may be released to the pool water because of fuel movements during the proposed SFP modification.
This could increase radiation levels in the vicinity of the pool and decrease the 10
1 l
clarity of the water.
There will be a number of fuel movements in each pool during the modification.
The plants have not experienced significant releases
'of crud to the pool water during refuelings when the spent fuel is first moved a
into the pools and the addition of crud to the pool water, from the fuel assembly and from the introduction of primary coolant water to the pool water, is the greatest.
The licensee does not expect to have significant releases of crud to the pool water during the modification of the pools.
The purification system for each pool, which has kept radiation levels in the vicinity of the 1
pool to low levels, includes a filter to remove crud and will be operating during the modification of the pools.
The pool floor will be vacuumed during l
the modification to remove particles which fall to the floor.
The staff finds that the SFP modification can be performed in a manner that will ensure as low as is reasonably achievable (ALARA) exposures to occupational workers.
The licensee has proposed decontaminating the spent fuel racks which are removed and storing them on site.
The dose estimated for this activity is 6 man rem.
We have estimated that the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies at both units on the basis of information supplied by the licensee and by utilizing relevant assumptions for occupancy times and for dose rates in the spent fuel area from radionuclide i
concentrations in the SFP water.
The spent fuel assemblies themselves contri-bute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel.
The occupational radiation exposure resulting l
from the proposed action represents a negligible burden.
Based on present and projected operations in the spent fuel pool area, we estimate that the proposed modification should add less than one percent to the total annual occupational radiation exposure burden at both units.
The small increase in radiation l
exposure should not affect the licensee's ability to maintain individual occupational doses to as low as is reasonably achievable levels and within the limits of 10 CFR Part 29.
Thus, we find that storing additional fuel in the two pools will not result in any significant increase in doses received by occupational workers.
i i
11
l
3.0 CONCLUSION
We have performed an evaluation of the licensee's proposed modifications based primarily on information provided to us in the licensee's basic supporting document.
This document has been revised and supplemented during the course of our review in response to staff questions, and from meetings and discussions with the licensee, and to address new or more refined information regarding the proposed modification.
Our evaluation concludes that the proposed modification of the Brunswick Station Units 1 and 2 spent fuel storage is acceptable because:
(1)
The structural and mechanical design for the proposed modification satisfies the applicable requirements of General Design Criteria 2, 4, 61, and 62 of 10 CFR Part 50, Appendix A and is acceptable.
(2) The compatibility of the materials and coolant used in the spent fuel storage pool is adequate based on tests, data, and actual service exper'i-ence in operating reactors.
The selection of appropriate materials of construction by the licensee meets the requirements 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.
(3) The installation of the proposed' fuel handling racks can be accomplished safely.
(4) The likelihood of an accident involving heavy loads in the vicinity of the spent fuel pool is sufficiently small that no additional restrictions on load movement are necessary while our generic review of the issues is underway.
(5) The cooling system for each of the spent fuel pools has acceptable cooling capacity for normal fuel off loading.
The spent fuel ccoling system, as supplemented by the RHR system has acceptable cooling capacity for discharge of full core inventory into the pool.
(6) The physical design of the new storage racks will preclude criticality for any credible moderating condition.
(7) The existing SFP cleanup system is adequate for the proposed modification.
(8) The conclusions of the evaluation of the waste treatment systems are l
unchanged by the modification of the spent fuel pool.
(9) The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the spent fuel pool would be negligible.
12 l
Wm conclude, th:n, 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 proposed license amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
R. A. Hermann B. Turovlin
- 5. MacKay P. Wu T. Cain T. Chan R. J. Serbu J. S. Boegli W. Brooks
- 0. Rothberg Dated:
December 15, 1983 i
- ~
4 a
l i
}
1 13 i
I References 1.
J. R. Weeks, " Corrosion of Materials in Spent Fuel Storage Pools,"
BNL-NUREG-23021, July 1977.
2.
J. R. Weeks, " Corrosion Considerations in the use of Boral in Spent Fuel Storage Pools," BNL-NUREG-25582, January 1979.
s s
s N
k d'
+
4 A
14
/
i e
s
%