ML19352B015
| ML19352B015 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/15/1981 |
| From: | Office of Nuclear Reactor Regulation |
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| ML19352B014 | List: |
| References | |
| NUDOCS 8106020564 | |
| Download: ML19352B015 (20) | |
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i SAFETY EVALUArION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE MODIFICATION OF THE SPENT FUEL STORAGE POOL FACILITY OPERATING LICENSE NO. OPR-6 CON 5tjMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET NO. 50-155 i
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DATE:
May 15,1981 i
8106 02 0T6Y
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1.0 INTRODUCTION
By letter dated April 23, 1979, as supplemented on June 26, October 1, October 19, October 25, December 28, 1979, January 7, January 16, February 1, June 20, August 11 August 14, and December 5,1980, Consumers Power Company (CPC) (the licensee) requested an amendment to Facility Operating License No. OPR-6 for the Big Rock Point Plant.
The amendment would authorize an increase in the stoiage capacity of the spent fuel pool (SFP) from the present 193 fuel assemblies to A41 fuel assemblies at the Big Rock Point Plant.
This increase in storage capacity will allow the storage of spent fuel until 1990 while retaining tna capability to offload a full core up to that time.
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a 2.0 DISCUSSION The proposed increase in the spent fuel pool capacity would be acco.,lished by installing three new independent fuel assembly storage racks.
The existing failed fuel storage rack would be removed.
The new rack assemblies consist of one 8 x 11 array, an 8 x 13 array, and a 9 x 9 array of storage locations.
Each new rack consists of an array of storage containers approximately 7.5 inches square by about seven feet long.
The containers are fabricated from one quarter (1/4) inch-thick type 304 stainless steel, with a nominal 9.0 inch center-to-center spacing.
The dimensions of the storage containers and the spacing will result in 1.5 inches of water between the containers.
The general arrangement and details of the proposed new spent fuel racks are presented in Figures 2-1, 3-1, a< d 4-1 of the " Spent Fuel Rack Addition Description and Safety Analysis," April 1979, which is attached to the licensee's letter dated April 23, 1979.
The major safety considerations associated with the proposed expansion of the Big Rock Point Plant spent fuel peol storage capacity are addressed below.
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3.0 EVALUATION 3.1 Criticality Considerations The Big Rock Point fuel pool criticality calculations are based on unirradiated fuel assemblies with no burnable poisons which have a fuel enrichment of 3.8 weight percent U-235.
This is the highest anticipated U-235 enrichment in an 11 x 11 fuel assembly and results in a maximum fuel loading of 28.3 grams of U-235 per axial ce.3imeter of fuel assembly.
The NUS Corporation (NUS) performed the criticalf ty analyses for CPC.
For parametric calculations, NUS used their version of the LEOPARD somputer program, which is called NUMICE, to get four group cross sections for P0Q-7 diffusion theory calculations.
NUS checked the accuracy of this method by using it to calculate water-moderated, uranium lattice experiments.
As indicated in Section 4.3 of the licensee's April 23, 1979 submittal, the calculated neutron multiplication factors from NUMICE/PDQ-7 deviated from the experimental values by an average of + 0.009.
In order to ensure that the results of these four group calculations for the storage lattice are accurate, NUS used the KENO Monte Carlo program with 123 group cross sections from the XSDRN program with the GAM-THERMOS library to check selected cases and to verify the neutron multiplication factor of the final design.
This method was checked by using it to calculate critical experiments of shipping cask configurations.
This series of calculations showed that this GAM-THERMOS / KEN 0 method yielded neutron multiplication factors that are within + 0.008 of the experimental values.
However, there is an additional statistical uncertainty of +.009 in these calculations that was also taken into consideration.
NUS's~use of these computer programs gave a neutron multiplication factor of 0.89 for an infinite array of these spent fuel assemblies located in the nominal storage lattice, which is assumed to be at a temperature of 68 F.
NUC calculated that this neutron multiplication factor would increase to 0.90 when the pool outlet water temperature is increased to its maximum possible value of 212 F, and it calculated that, with all tolerances, mislocations, and uncertainties included at a temperature of 212 F, the maximum possible neutron multiplication factor could be 0.946.
CPC states in its submittal that it will not be possible to inadvertently position a fuel assembly close enough to the outside of a filled rack to cause an increase in this maximum neutron multiplication factor of 0.946.
3.1.1 Evaluation A comparison of the above results with the results of parametric calculations made with other methods for similar fuel pool storage lattices shows them to be acceptably accurate.
By assuming new, unirradiated fuel with no burnable
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poison or control rods, these calculations yield the maximum neutron multipli-cation factor that could be obtained throughout the life of the fuel assemblies.
This includes the effect of the plutonium which is generated during the fuel cycle.
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B ine NRC acceptance criteria for the criticality aspects of fuel storage racks is that the neutron multiplication factor in spent fuel pools shall be less tha9 or equal to 0.95, including all uncertainties, under all conditions, throughout the life of the racks.
This 0.95 acceptance criterion is based on
'ae overall uncertainties associated with the calcu!ational methods, a A it is our judgment that this provides sufficient margin to preclude critici
=; in fuel pools. Accordingly, there is a technical specification which limi c the neutron multiplication factor, k'he,nt fuel pools is not a quantity which is in spent fuel pools to 0.95.
Since the nuetron multiplication factor in measured with good accuracy, the only available value is a calculated one.
To preclude any unreviewed increase, or increased uncertainty, in the calculated value of the neutron multiplication factor which could raise the actual k ff in the fuel pool above 0.95 without being detected, a limit on the maximufn fuel loading is also required.
Therefore, we find that the storage racks proposed for Big Rock Point will meet the NRC criteria when the fuel loading in the assemblies, described in the CPC submittals, is limited to 28.3 grams of uranium -235 per axial centimeter of fuel assembly or equivalent.
We will require a Technical Specification to limit the fuel loading to this value prior to the use of the new racks.
3.1.2 Conclusion We find that when any number of the fuel assemblies described by CPC in these submittals including the G-1(MOX) fuel, which have not more than 29.3 grams of uranium-235 per axial centimeter of fuel assembly, or equivalent, are loaded into the proposed racks, the k in the fuel pool will be less than the 0.95 litit.
PendingNRCreview,weYllprohibittheuseofthesestorageracks for fuel assemblies that contain more thar. 28.3 grams of uranium-235,_ or equivalent, per axial centimeter of fuel assembly.
On the basis of the information submitted by the licensee, the k not exceeding 0.95, and the fuelloadinglimitstobeaddedtotheTechnTN1Specificationsasstated above, we concluda that the use of the proposed racks is acceptable.
3.2 Spent Fuel Cooling The licensed thermal power for the Big Rock Point Plant is 240 MWth.
CPC is refueling this reactor annually, at which time 22 of the 84 fuel assemblies in the cora are replaced.
To calculate the maximum heat loads in the spent fuel pool, CPC assumed a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time interval between reactor shutdown and the time either the 22 fuel assemblies in the normal refueling or the 84 fuel assemblies in a full core offload are placed in the spent fuel pool.
For this coolingtime,CPCusedthemethodgiveningheNRCStandardReviewPlan9.2.5 to calculate maximum heat loads of 1.4 x 10 BTU /hrforthefifteegthsucces-sive annual refueling after the proposed modification and 3.8 x 10 BTU /hr for the full core offload sixty days after the eleventh refueling after the proposed modification.
The spent fuel pool cooling system at the Big Rock Point Plant consists of two pumpsandgwoheatexchangers.
Each pump is designed to pump 250 gpm (1.25xf0 pounds per hour), and each heat exchanger is designed to transfer 3.0 x 10 BTU /hr from 119 F fuel pool water to 70 F cooling water which is 5
flowing through the shell side of the heat exchanger at a rate of 1.25 x 10 pounds per hour.
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CPC states that this system, with two pumps running, will be able to keep the spent fuel pool outlet temperature below 82 F through the fifteenth annual refueling and 101*F for a full core offload sixty days after the eleventh refueling.
In the event that one of the cooling loops were to fail just after the full core was offloaded into the pool, CPC states that the maximum spent fuel pool outlet temperature that could be obtained is 132 F.
In regard to postulated accidents such as the loss of offsite power or some reactor accident which would prevent personnel from entering the containment and, hence, getting to the spent fuel pool for an extended period of time, CPC stated the following:
(1) Even if all pool cooling systems were lost, acalyses show that less than one-half inch of the assembly height would be in bulk boiling.
(2) Makeup water to the spent fuel pool will be available for an 11 gpm evaporation rate.
The capability to provide makeup water to the pool from outside containment will be available.
(3) Remote indicating spent fuel pool water level instrumentation will be installed prior to the installation of the new spent fuel storage racks.
(4) The containment enclosure spray system can be used to prevent overpressurization of the containment.
The emergency diesel generator serves as a backup power supply to the spray pumps in the event of loss of normal Nxiliary power.
3.2.1 Evaluation By using the method given on pages 9.2.5-8 through 14 of the November 24, 1975 versionoftheNRCStandardReviewPlan,withtgeuncertaintyfactor,K, equal to 0.1 for fuel with decay times longer than 10 seconds and with a total cooling time of forty eight hours on the newly discharged fuel, we find that the maximum refueling heat load in the modified spent fuel pool would be 1.44 x 108 BTU /hr and that the maximum heat load for a full core offload that fills the pool one year after the last refueling would be 4.43 x 108 BTV/hr.
We also find that the maximum incremental heat load that could be added by increasing the number of spent fuel assemblies in the pool from 193 to 441 is 0.17 x 106 BTU /hr. This is a 3% increase in peak heat loads from the present to the modified pool and is the result of an increase in the older spent fuel assemblies that wi!' be present in the modified pool.
Based on our calculations, we find that, with two pumps operating, the spent fuei pool cooling system can maintain the fuel pool outlet water temperature below 83 F for the normal refueling offload that fills the modified pool and below 108 F for any full core offload.
In the highly unlikely event that both spent fuel pool cooling loops fail to operate during the peak heating period of the full core offload that fills the modified pool, the maximum possible heat-up rate of the spent fuel pool water would be 5.5 F per hour. Thus, assuming that the average spent fuel pool water temperature is about 100 F at, the time this complete loss of cooling occurs, there would be approximately twenty hours before the water would boil.
We calculate that af ter boiling starts the maximum possible required water 3-3
makeup rate will be 9 gpm.
Because at this time the whole core would be in the spent fuel pool, the reactor vessel would not contain any fuel.
Thus, for this situation there should be nothing that would prevent personnel from entering the containment and getting access to the spent fuel pool and its cooling system.
For this unlikely event, we find that twenty hours will be sufficient time to either repair at least one loop of the spent fuel pool cooling system or establish the required 9 gpm makeup rate to maintain the water level in the spent fuel pool.
For all situations other than the full core offload the maximum possible heat load in the modified Big Rock Point pool will be 1.44 x 10s BTU /hr.
This is at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the core is shutdown.
At this heat load it would take more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the pool to heat up to 212*F.
Aftertheadditionaf72 hours of decay, the heat load in the pool would be reduced to about 1 x 10 BTU /hr.
Thus, if there were an extended loss of offsite power and emergency power could not be hooked up to the spent fuel pool cooling system in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or if both spent fuel pool cooling loops were to fail during some postulated accident which would prevent personnel from entering the containment, the spent fuel pool could go into a boiling condition with a 1 x 106 BTU /hr heat load.
If the resultant steam were all carried away from the pool, a water makeup rate of 2 gpm would be required to keep the spent fuel pool water level from dropping.
In its June 20, 1980 submittal, CPC stated that the capability to actuate the supply of makeup water from outside of containment will be available..
Based on our review, we find that the licensee's commitment to provide the capability to supply 11 gpm of makup water to the spent fuel storage pool, actuated from outside containment, is acceptable.
CPC's letter dated June 20, 1980, included a revision to response number 4 of their January 16, 1980 letter concerning the provisions of makeup water to the spent fuel pool in the event containment cannot be entered.
CPC stated in the revised response that they are considering three types of water for the makeup to the fuel pool:
a) demineralized water, b) treated waste water, c) fire water.
However, in response number 1 in their January 16, 1980 letter, CPC indicated that only the demineralized water system and the fire system have been designed to remain intact following a design basis earthquake.
Thus, to assure that the 11 gpm makeup will be available in the event of a design basis earthquake, the demineralized water system or the fire system should be used as the source of this water. We will require a Technical Specification that specifies that this 11 gpm makeup water must be available.
The containment enclosure spray system is designed to remove 8 x 106 BTU /hr from the containment.
It does this by spraying 100 F water out of nozzles located on top and within the steam drum enclosure.
On its way down to the bottom of the containment the temperature of the spray increases by 40 F.
The heat picked up by the spray is then removed in a heat exchanger located outside containment.
Based on our review, we find that the containment spray system has the capability to remove the heat and prevent containment overpressurization should pool boiling occur as discussed above.
g Thus, we find that it is highly unlikely that the level of water in the spent fuel pool will decrease to a level such that the fuel will no longer continue to be immersed in water even assuming the unlikely sequence of postulated events discussed above. We also find that the incremental increase 3-4 l
in hea't icid due to the modified pool is not sufficient to obviate the containment spray system's capability to control containment pressure.
3.2.2 Conclusion We find that the present cooling capacity for the Big Rock Point spent fuel pool will be sufficient to handle the incremental heat load that will be added by the proposed modification. We, therefore, conclude that the proposed design is acceptable.
- 3. 3 Installation of Racks and Fuel Handling CPC states in their submittal that the following safety precautions will be taken during the installation of the new racks:
(1) Installation paths shall follow routes analyzed in regard to cask-drop hazard with particular attention to precluding possible interaction between the racks (e.g., direct drop or tipping)and safety-related structures, systems, or components.
In no case shall the route chosen present a hazard to safety related structures, systems or components that has not been previously reviewed and/or analyzed.
(2) Crane speeds in both vertical and horizontal directions shall be maintained so as to prevent uncontrolled motion of the racks.
(3) No rack shall be moved in the vicinity of stored spent fuel so that a direct drop or tipping of the rack will result in damage to the spent fuel.
Additional acministrative controls will be developed as required to preclude the dropping or tipping of a rack on stored spent fuel.
(4) All racks shall t,e moved only when empty.
(5) One operation will be conducted at any one time only (e.g., shuffling fuel, movinn racks, etc.).
(6) Racks will be transferred to the spent fuel pool area during a plant shutdown unless a transfer route is available which will preclude damage to safety-related equipment as a result of any postulated direct drop or tipping of a rack.
3.3.1 Evaluation We find that strict adherence to the above procedures will provide adequate protection to the spent fuel assemblies stored in the pool during the cnange of the racks.
After the racks are installed in the pool, the fuel handling procedures in and around the pool will be the same as those procedures that were in effect prior to the proposed modifications. The proposed increase in spent fuel pool storage capacity does not change the consequences of fuel handling accidents in the spent fuel pool from those presented in our Safety Evaluation dated February 6, 1976 for the Big Rock Point plant.
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- 3. 3. 2-Conclusion We conclude that the removal of the failed fuel storage rack and the installation and use of the proposed racks can be performed safely and is therefore acceptable.
3.4 Structural and Mechanical Design and Material Considerations The proposed modification to the spent fuel pool storage capacity involves the removal of the existing failed fuel storage rack and the installation of three new independent fuel assembly storage racks.
The new rack assemblies consist of one 8 x 11 array, an 8 x 13 array, and a 9 x 9 array of storage locations.
Each new rack consists of an array of square storage cans, or tubes, fabricated from one-quarter (1/4) inch thick type 304 stainless steel, with a nominal 9.0 inch center-to-center spacing.
The cans are welded to one-half (1/2) inch thick base' plates.
The storage cans have lead-in surfaces at the top to provide guidance for insertion of the fuel assemblies. Openings at the top and in the base plate provide a flow path for convective cooling of the fuel assemblies through natural circulation.
The storage cans are welded to a grided base, and are structurally tied at the top with an over-under stainless steel grid system.
Both the bottom and top grid systems will be comprised of a lattice of stainless steel beams.
Each storage rack is self-supporting, and is supported by adjustable leveling legs.
Four leveling legs are provided for the 8 x 11 and 9 x 9 racks, while the 8 x 13 racks are supported on six leveling legs.
Except at the leveling legs, all welded construction is used in the fabrication of the spent fuel rack assembly.
The racks are not supported laterally by the fuel pool walls.
Load transfer to the pool structure from the fuel racks occurs only at the base of the racks at the pool floor / leveling legs interface.
Horizontal seismic loads are transmitted through the cans and grid systems to the floor of the fuel pool via the leveling legs.
Vertical seismic and dead-weight loads are transmitted through the cans and baseplates to the grid beams, to the gussets and bosses into which the leveling legs are screwed.
The loads are then transmitted to the pool floor via the leveling legs.
Each rack is provided with four lifting lugs which will be welded to the upper grid beam system.
Lifting forces will be transmitted through the lifting lugs, to the upper grid beams, to the cans, and into the lower grid beam system and base plates.
3.4.1 Evaluation 3.4.1.1 Structural and Mechanical The new spent fuel storage rack designs were reviewed in accordance with the applicable parts of Sections 3.7 and 3.8 of the Standard Review Plan and the Branch Technical Position entitled "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (forwarded to all licensees in April 1978) including modifications (forwarded to all licensees in January 1979).
The following structural and mechanical aspects of the new rack design were reviewed:
the supporting arrangements for the racks. including their restraints, design, fabrication, and installation criteria; the structural 3-6
design and analysis procedures for all loadings, including seismic and impact loadings; the load combinations; the structural acceptance criteria; and the quality assurance requirements for design, fabrication, and installation.
The allowable stresses for the stainless steel racks are in accordance with Part 1 of the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings." Yield strengths at appropriate temperatures were obtained from Table I-2.2 of Section III of the ASME Boiler and Pressure Vessel Code.
The spent fuel pool is located in the Reactor Building.
The seismic analysis of the racks conducted by the licensee was a time history analysis, using structural damping values which conform to the positions in Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Pcwer Plants." The 8 x 11 rack was used in the analysis to determine loads, stresses, and deflections, since this rack has the greatest potential for tipping and will develop the greatest internal forces due to seismic and deadweight loadings.
Three models were used in the analysis of the fuel rack.
A nonlinear model is used in a tipping analysis to determine rack forces resulting from horizontal seismic excitation at the rack supports.
A three-dimensional finite element model is used to determine stresses in the rack resulting from seismic, ther-mal, and deadweight loading. And a one-dimensional three mass model is used to determine maximum rack sliding distances during a Design Basis Earthquake (DBE).
The analytical model used to determine rack loads allows tipping, but conservatively restrains sliding, to obtain a conservative value for the s
structural loading of the rack members.
The seismic input used was a displacement time history of the fuel pool floor, obtained from a seismic time history analysis of the plant.
In addition to the weight of the rack and fuel assemblies, hydrodynamic masses were accounted for.
Fuel can interaction was accounted for by explicitly modeling the gaps and the hydrodynamic coupling between a fuel assembly and the can which encloses it.
In the three dimensional analytical model used to determine rack stresses, north-south and east-west equivalent seismic loading was represented by uniform forces applied along the length of the cans.
The magnitude of this equivalent loading was determined by equating the maximum bending moment at the base of the can (from the peak response of the time history tipping analysis) to the bending moment at the base in a model subjected to a uniform horizontal load.
In the vertical direction, the seismic analysis consisted of a static analysis using the maximum vertical floor acceleration.
This is accep-table based on a frequency analysis of the rack which showed that its vertical fundamental frequency was greater than 33Hz, indicating a rigid response in the vertical direction.
The response of the rack in the three component directions were combined using the square root of the sum of the squares (SRSS) procedure, in accordance with Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analyses." Thermal stress resulting from the differential heating effect between a full and an empty can were added to the stresses resulting from the seismic and deadweight loadings, using an absolute sum method.
A time-history sliding analysis was performed to determine any potential impacting between adjacent fuel racks, and between the racks and the spent fuel 3-7 l
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pool structure.
The coefficient of friction between the stainless steel liner and the rack leveling legs used in the analysis was conservatively chosen to be 0.2, based on information provided in a report by E. Rabinowicz to Boston Edison Comptny, " Friction Coefficients of Water-Lubricated Stainless Steels,"
November 5, 1976.
The results of the time history analysis indicate that during a DBE, the racks slide less than 0.5 inch, based en the absolute sum of the displacements of two adjacent racks.
The minimum asailable gaps between adjacent racks, between racks and pool walls, and between racks and other pool equipment are approxmately 1.75, 3.0, and 2.0 inches, respectively.
The loads and load combinations considered in the analysis of ".he spent fuel storage racks are in accordance with Section 3.8.4 of the Stanurd Review.
Plan.
Results of the seismic analysis show that the racks are capable of witnstanding the loads associated with all the design loading conditions without exceeding allowable stresses.
Seismic floor time histories used in the analysis were taken from a report by D'Appolonia Consulting Engineers, Inc., entitled, " Derivation of Floor Responses Reactor Building," dated June 1978.
These floor responses were determined using ground acceleration time histories compatible with smooth design resporse spectra which confor.n to the positions in Regulatory Guide 1.60, " Design Response Spectra for Seismic Design of Nuclear Power Plants," and scaled to peak zero period horizontal and vertical accelerations of 0.12g and 0.08g, respectively.
This document is currently under staff review, and has been sAmitted in support of the Systematic Evaluation Program review of seismic issues.
Two postulated fuel assembly drops were considered in the analysis of the racks.
Energy methods were used to determine the effects of the impact of a fuel assembly dropped from a maximum height of 43'-3" above the bottom of the fuel pool, based on the upper limit of travel of the auxiliary hook.
The worst case drop, where the fuel assembly is postulated to drop into an empty fuel can and impacts the baseplate at the bottom of the rack, results in a kinetic energy at impact of 195,000 inch pound.
The second case postulates the drop of the fuel assembly onto the top of the fuel rack lead-in guides.
In both cases, the impact energy 13 dissipated by local yielding or crushing; however, gross stresses in the rack remain below allowable limits and the overall structural integrity of the rack is maintained.
The licensee has st?ted that the heaviest load that will be transported over the stored spent fuel is the 24-ton spent fuel transfer cask.
The 24-ton spent fuel transfer cask has a redundant support (safety sling) assembly that was designed to catch the cask in the event of a main hoist / hook failure.
By letter dated August 14, 1980, CPC proposed a Technical Specification that would require functional testing of the trip mechanism of the spent fuel transfer cask redundant support assembly prior to commencing refueling activities.
Analyses performed by the licensee for casks smaller than the 24-ton fuel transfer cask indicate that even if they drop on the fuel pool liner, it will not result in a loss of pool water in excess of the 200 gpm makeup capability.
In addition, the licensee has stated that administrative controls, for casks other than the fuel transfer cask, will be established to ensure that:
(a) no I
cask is moved over or near stored spent fuel, (b) all cask handling operations are limited to the southwest corner of the spent fuel pool and (c) no spent fuel is stored in the two existing "A" racks adjacent to the cask har.dling arca during cask handling operations.
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The NRC staff has completed its generic review of load handling operations in the vicinity of spent fuel pools. A report describing the results of this review has been issued as NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants - Resolution of TAP A-36."
This report contains several recommendations to be implemented by all licensees to ensure the safe handling of heavy loads.
By letter dated December 22, 1980, which was sent to all licensees, CPC was requested to review the controls for the handling of heavy loads to determine the extent to which the guidelines in NUREG-0612 are presently satisfied at the Big Rock Point Plant, and to identify the changes and modifications that would be required in order to fully satisfy these guidelir.et. Our December 22, 1980 letter also identified interim actions that the licensee should implement within 90 days (by letter dated February 3,1981, this was extended to May 15, 1981).
These interim actions are:
(1) Safe load paths should be defined per the guidelines of Section 5.1.1(1) of NUREG-0612, (2) Procedures should be developed and implemented per the guidelines of Section 5.1.1(2) of NUREG-0612, (3) Crane operators should be trained, qualified and conduct themselves per the guidelines of Section 5.1.1(3) of NUREG-0612, (4) Cranes should be inspected, tested, and maintained in accordance with the guidelines of Section 5.1.1(6) of NUREG-0612, and (5) In addition to the above, spe.cial attention should be given to procedures, equipment, and personnel for the handling of heavy loads over the core, such as vessel internals or vessel inspection tools.
This special review should include the following for these loads:
(1) review of procedures for installation of rigging or lifting devices and movement of the load to assure that sufficient detail is provided and that instructions are clear and concise; (2) visual inspections of load bear-ing components of cranes, slings, and special lifting devices to identify flaws or deficiencies that could lead to failure of the component; (3) appropriate repair and replacement of defective components; and (4) verify that the crane operators have been properly trained and are familiar with specific procedures used in handling these loads, e.g.,
hand signals, conduct of operations, and content of procedures.
In addition to these actions, the licensee has stated that the 71/2 ton,15 ton, and 24 ton casks will not be moved during reactor operation.
Based on our review, we conclude that implementation of the above measures will provide reasonable assurance tha'. handling of heavy loads will be performed in a safe manner until fina: implementation of the guidelines of Section 5.1 of NUREG-0612.
The effects of a postulated stuck fuel assembly have been considered and accounted for by examining the case where an upward force of 2000 pounds is exerted on a rack due to attempted assembly withdrawal.
A mechanical trip switch is set to actuate below this load.
The spent fuel pool is constructed of concrete walls and floor, lined with a 3/16 inch thick stainless steel plate.
The fuel pool concrete, reinforcing 3-9
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steel', and liner were analyzed to account for the additional loadings imposed by the new and existing racks.
The more conservative load combinations of Section 9.3 of ACI 349-76 or Standard Review Plan Section 3.8.4 were used in the analysis. The allowable stress / load limits were taken from ACI 318-77 and the American Iron and Steel Institute " Stainless Steel Cold-Formed Structural Design Manual."
By letter dated June 20, 1980, the licensee provided an analysis performed by NUS regaiding the effects of pool boiling on the concrete structure, racks, and liner of the spent fuel pool in the unlikely event that both pumps in the spent fuel pool cooling system should fail.
The analyses show that the stresses that develop under these conditions are within design limits.
Results of the analysis for the most severe loading conditions indicate that the maximum stresses are within the allowables, and that the structural members of the fuel pool are adequate to withstand the additional loads imposed by the new racks and additional fuel.
Detailed written installation procedures will be prepared by the installation contractor and approved by the licensee prior to commencement of rack installa-tion. We have reviewed the licensee's acceptance criteria for the detailed written procedures, and find that they provide an acceptable means for ensuring that any deleterious impact on safety related structures, systems and components or additional risk to the health and safety of the public is precluded.
In addition, the licensee has agreed to provide the staff with a copy of the detailed written installation sequence and procedures for our review prio? to the commencement of any movement of racks in the spent fuel pool.
3.4.1.2 Materials The spent fuel storage racks, their associated hardware and support frame components, and the pool liner are constructed almost entirely of Type 304 stainless steel.
The only exception is the leveling legs which are ASTM A276 UNSS21-800 stainless steel.
The storage pool environment, is demineralized unborated water controlled to a temperature less than the design value of 93 F.
The fuel pool water chemistry is maintained at a pH of 6.9 and the conductivity typically of 0.3 umho.
This corresponds to extremely pure neutral water.
The pool liner, rack lattice structure and fuel storage tubes are stainless steels which are compatible with the storage pool environment.
In this environment of oxygen saturated demineralized neutral (pH of 6.9) water, the corrosion rate of the stainless steel is so small as to be unmeasurable.
Corrosion rate measurements for this material in water of tais quality and temperature are not available, and any estimate of corrosion rates must be extrapolated down from measurements at higher temperature.l'2 Calculated corrosion degradation, using extrapolated estimated rates, of type 304 stainless steel would not exceed a depth of 6.00 x 10 5 inch 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 ;tructure, fuel storage tubes, and the Inconel and Zir-caloy in the spent fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are therefore at similar potentials.
Though localized corrosion of various types such as pitting corrosion cannot be completely excluded, any corrosion will 3-10
be minor and insignificant relative to the initial thickness for periods well in excess of 40 years.
Because of the lack of neutron poisons and the extensive experience with the use of type 304 stainless steel in spent fuel storage pools with demineralized neutral water, we conclude that it is not necessary to have a particular surveillance program for this storage pool.
3.4.2 Conclusion 3.4.2.1 Structural and Mechanical The analyses, design, fabrication, and criteria for establishing installation procedures of the proposed new spent fuel racks are in conformance with accepted codes, standards and criteria.
The structural design and analysis procedures for all loadings, including seismic, thermal, and impact loading; the acceptance criteria for the approp.iate loading conditions and combination; and the applicable industry codes are in accordance with appropriate portions of the "NRC OT Postion for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications."
Allowable stress limits for the combined loading conditions are in accordance with AISC specifications. Yield stress values at the appropriate temperature were obtained from Section III of the ASME Boiler and Pressure Vessel Code.
The quality assurance codes and criteria for the materials, fabrication and installation of the new racks are in accordance with the accepted requirements of an CFR 50 Appendix B, and the provisions of ANSI standards.
The effects of the additional loads on the existing pool structure due to the new fuel racks, existing fuel racks, and equipment have been examined.
The pool structural integrity is assured by conformance with the Standard Review Plan criteria and with all Big Rock Point Final Hazards Summary Report structural acceptaace criteria.
Results of the seismic and structural analyses indicate that the racks are capable of withstanding the loads associated with all design loading conditions.
Also, impact due to fuel assembly / cell interaction has been considered, and will result in no damage to the racks or fuel assemblies themselves.
The findings and conclusions in this Safe.ty Evaluation are based on. specific response spectra, as specified in the D'Appolonia Consulting Engineers Report previously referenced herein.
Results of the dropped fuel assemoly analyses show that local rack deformation will occur, but indicate that gross stresses meet the applicable allowable stresses and that the integrity of the racks is maintained.
In addition, the licensee has stated that a drop of the heaviest load planned for transport over the spent fuel racks (i.e., the 24 ton spent fuel transfer cask) is precluded by the use of safety slings, and that the drop of smaller casks will not result in a loss of pool water in excess of the 200 gpm makeup capability.
R'sults of the stuck fuel assembly analysis show that the stresses are below those allowed for the applicable loading combination.
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O WefiIdthatthesubjectmodificationproposedbythelicenseesatisfiesthe applicable requirements of the General Design Criteria 2, 4, 61, and 62 of 10 CFR, Part 50, Appendix A, and is, therefore, acceptable.
3.4.2.2 Materials Based on our evaluation as discussed above, we conclude that any corrosion that occurs in che Big Rock Point Nuclear Plant spent fuel storage pool environment would be of little significance during the next forty years.
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 corrositn, and galvanic corrosion.
We therefore find that the selection of appropriate materials of construction, with the long term experience of use of the materials in similar service meets the requirements of 10 CFR Part 50, Appendix A, General Design Criterion 62, preventing criticality by maintaining structural integrity of components, and therefore is acceptable.
2.5 Radiological Considerations Our review addressed the following major considerations:
fuel handling, occupational radiation exposure, and radioactive waste treatment.
3.5.1 Evaluation 3.5.1.1 Fuel Handling A fuel transfer cask is used to transfer spent fuel from the core to the spent fuel pool. A postulated drop of the fuel transfer cask.s the worst fuel handling accident.
It could potentially damage up to one third of the fuel assemblies in the core. This accident was evaluated in the NRC staff's Safety Evaluation dated February 6,1976. The potential consequences of this accident are within tne guidelines of 10 CFR Part 100. The potential consequence of damaging a single fuel assembly in a postulated fuel handling accident are 8.8 Rem thryoid and 0.09 Rem whole body.
The potential consequences of fuel handling accidents are not changed because of the increase in the storage capacity of the spent fuel pool; therefore we find that they are acceptable.
A; discussed in Section 3.4.1.1 of this Safety Evaluation, the NRC staff has completed its generic review of load handling operations in the vicinity of pent fuel pools.
Based on the discussion in Section 3.4.1.1, we cenclude that the likelihood of any other heavy load handling accident is suff'ciently small that the proptsed modification is acceptable and no additional restrictions on load handling operations in the vicinity of the spent fuel pool, other than those discussed in this Safety Evaluation, are necessary until final implementation of the guidelines of NUREG-0612.
- 3. 5.1. 2 Occupational Radiation Exposure We have reviewed the licensee's plans for the removal and storage of the failed fuel racks and the installation of the new racks with respect to occupational 3-12
radiation exposure.
The occupational radiation exposure for this operation is estimated by the licensee to be about 23 man rem.
This is based on the licensee's detailed breckdown of occupational exposure for each phase of the pool 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 will be performed.
This estimate represents about 10% of the total man rem burden from occupational exposure at the plant and is expected to be a o'e time exposure to individuals n
performing the operation.
To keep occupational exposure as low as is reasonably achievable, the licensee will vacuum the pool walls, floor and existing racks as needed.
The licensee will decontaminate all areas around the pool, cycle the pool water through the radwaste demineralizer to reduce pool water radioactivity concentrations and will provide surveillance and monitoring of the work area by health physics personnel.
TSc estimated occupational radiation exposure to modify the pool is based on dos rates in the vicinity of the pool which are higher than typical for nuclear power plants.
The licensee has controlled occupational exposure to workers by controlling occupancy in the vicinity of the pool.
The plant annual occupational exposure is below the average value for nuclear power plants.
The licensee has ccemitted to do additional work to provide additional assurance that the occupational and radiation exposure for this modification is as low as is reasonably achievable.
To reduce occupational exposure the licensee has committed to evaluate doing the following:
(1) clean the pool liner, (2) upgrade the pool purification system, and (3) use the radwaste demineralizer more frequently to clean the pool water.
By letter dated August 11, 1980, the licensee stated that these commitments have been accomplished.
Based on the above, we conclude that the licensee will perform the pool modification in a manner to assure that the occupational exposure will be as low as is reasonably achievable.
There is a small relatively thin region of the walls around the spent fuel pool.
The south wall of the pool tapers down from 6 feet to 3 feet near one corner.
The licensee has calculated a dose rate of 38 mrem / hour near this small region with one year old fuel stored in its pool on the other side of the wall and the modified pool full. This dose rate does not exist now because of the additional water shielding available in the present unmodified pool.
The area along the south wall outside the pool is not a working area.
Because this area is under administrative control, it is chain locked and has radiation signs posted.
The spent fuel pool filter is located in this area but people do not have to be in this area to replace the filte:.
If the radiation levels are too high, tne licensee can control access to this area as he has in the past and place the oldest fuel near the thinnest part of the south wall.
Based on the above, we conclude that the licensee can operate the spent fuel pool in a manner to assure that the occupational exposure near the spent f uel pool will be within the requirements of 10 CFR Part 20 and will be as low as is reasc7 ably achievable.
We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies on the basis of information supplied by the licensee for dose rates in the spent fuel pool area from radionuclide concentrations in the pool water and the spent fuel assemblies.
The spant fuel assemblies themselves will contribute a negligible fraction of 3-13
the dose rates in the pool area because of the depth of water shielding the fuel.
Consequently, the occupational radiation exposure resulting from the additional spent fuel in the pool represents a negligible burden. Based on present and projected operations in the spent fuel pool area, we estimate that the proposed modification should add only a small fraction to the total annual occupational radiation exposure burden at this facility.
The small increase in radiation exposure will not affect the licensee's ability to maintain individual occupational' doses to as low as is reasonably achievable and within the limits of 10 CFR Part 20.
Thus, we conclude that storing additional fuel in the SFP will not result in any significant increase in doses received by onsite personnel.
3.5.1.3 Radioactive Waste Treatment The plant contains wasta treatment systems designed to collect and process the gaseous, liquid, and solid wastes that might contain radioactive material.
The waste treatment systems are evaluated with respect to the requirements of Appendix I to 10 CFR Part 50 in the NRC staff's evaluation dated May 1981~.'
There will be no change in the waste treatment systems or in the conclusions of the evaluation of these systems as described in the above cited evaluation because of the proposed modification.
3.5.2 Conclusions Our evaluation of the radiological considerations supports the conclusion that the proposed modification tc the Big Rock Point spent fuel pool is acceptable because:
(1) 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, other than those aiscussed in this Safety Evaluation until the final implementation of the guidelines of NUREG-0612.
(2) The installation and use of the new fuel racks does not alter the potential consequences of the design basis accident for the spent fuel pool, i.e.,
the rupture of all the fuel pins in the equivalent of a single fuel assembly and the subsequent release of the radioactive inventory within the gap of each fuel pin.
(3) The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the spent fuel pool would be negligible.
(4) The conclusions of the evaluation of the waste treatment systems are unchanged by the modification of the spent fuel pool.
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4.0 CUMMARY The results of our evaluation indicate that the proposed modification of the Big Rock Point plant spent fuel pool is acceptable because:
(1) The physical design of the new storage racks will preclude criticality for any credible fuel arrangement.
(2) The cooling system for the spent fuel pool has adequute cooling capacity.
(3) The installation and use of the proposed fuel storage racks can be accomplished in a manner that provides reiisonable assurance that there will be no undue risk to the health and safety of the public.
(4) The structural design and the materials of construction are adequate.
(5) -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, other than those discussed in this Safety Evaluation, until the final implementation of the guidelines of NUREG-0612.
(6) The installation and use of the new fuel racks does not alter the potential consequences of the design basis accident for the spent fuel pool, i.e.,
the rupture of all the fuel pins in the equivalent of a single fuel assembly and the subsequent release of the radioactive inventory within-the gap of each fuel pin.
(7) The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the spent fuel pool would be negligible.
(8) The cenclusions of the evaluation of the waste treatment systems are unchanged by the nodification of the spent fuel pool.
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5.0 CONCLUSION
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 proposed license amendment will not be inimical to the common defense and security or to the health and safety of the public.
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REFERENCES 1.
J.R. Weeks, " Corrosion of Materials in Spent Fuel Storage Pools",
j BNL-NUWEG 23021, July 1977.
L.
2.
A.B. Johnson, Jr., " Behavior of Spent Nuclear Fuel in Water Pool Storage", BNWL-2256, 1977.
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