ML19345H462

Safety Evaluation Supporting Amends 48 & 42 to Licenses DPR-42 & DPR-60,respectively

ML19345H462
Person / Time
Site:	Prairie Island
Issue date:	05/13/1981
From:	Office of Nuclear Reactor Regulation
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ML19345H461	List: ML19345H460 ML19345H462 ML19345H463 ML19345H464
References
NUDOCS 8105200332
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fg UNITED STATES O*8,c(3 NUCLEAR REGULATORY COMMISSION g

E WASHINGTON. O. C. 20666 US SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NOS.48 AND 42 TO FACILITY OPERATIM LICENSE NOS. OPR-42 AND DPR-60 RELATING TO MODIFICATION N THE SPENT FUEL POOL NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 i

1.0 Introduction By letter dated January 31, 1980 the Northern States Power Company (NSP) proposed to change the spent fuel pool storage design for the Prairie Island Nuclear Generating Plant (PINGP) Unit Nos.1 and 2.

The presently approved design was reviewed and approved in Amendment Nos. 22 and 16 to Facility Operating License Nos. OPR-42 and DPR-60 issued August 16, 1977.

The presently installed storage capacity is 687 fuel assemblies in a compartmented double pool which serves both units. The proposed modifi-cations would make available up to 1582 storage locations of which only 1120 will be allowed to contain spent fuel assemblies. In response to our questions NSP submitted supplemental information by letters dated June 10, and November 21, 1980, January 14, February 3, March 10, March 31, and April 20, 1981.

2.0 Background

The Prairie Island Nuclear Generating Plant (PINGP) spent fuel pool (SFP) was originally designed with a storage capacity of about one and two-thirds cores (198 fuel assemblies) felt to be adequate for the storage of tha discharge of the 40 assemblies per reactor year for a period of one year prior to shipment off-site for reprocessing.

By Amendment Nos. 22 and 16 dated August 16, 1977, we approved NSP's request to expand their SFP capacity to a total of 687 fuel assemblies "or both units, through the use of high density spent fuel racks. NSP further realized that an additional increase in SFP capacity would likely be necessary before any reprocessing facility or offsite storage facility is ready. By letter dated January 31, 1980, NSP submitted their request to expand the SFP capacity to 1582 fuel assembly storage locations with high capacity poison racks. The licensee proposes to fill up to 1362 of these location: with spent fuel resulting from nonnal refueling. As discussed in Section 3.3 of this report, current approval is limited to a total storage of 1120 fuel assemblies due to the current status of resolution of the heavy loads handling issues for the PINGP.

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. Our reviews except for the heavy loads handling issue were based on the 4

design proposed by the licensee.

NSP, as the licensee, is responsible for the modification to the SFP.

Nuclear Services Corporation is retained to design the spent fuel racks, contract for fabrication, perform analysis pertinent to the modification, l

and provide technical assistance during installation.

l 3.0 Discussion and Evaluation In reviewing the SFP modification, we considered:

(1) criticality analysis, (2) spent fuel ( Soling, (3) installation of racks and fuel handling (4) structure design, (5) fuel and other heavy loads handling, (6) occupational radiation exposure, (7) radioactive waste treatment, and (8) material ompatibili ty.

The proposed new higher density racks are to be made up of individual double-walled containers which are about fourteen feet long. The inner wall of each of these containers will be made from 0.090 inch thick sheet of 304 stainless steel which will be formed into a square cross section container with an inside dimension of 8.27 inches. The outer, or external wall will be a sheet of 0.036 inch thick stainless stee.. Borated, neutron absorbing plates, which are 8.05 inches wide and 0.125 inches thick will be placed in each of the four spaces between the tw walls. Thus J

each of the four sides of every container will have a bot :ed plate in it which; as NSP states in its January 31, 1980 submittal, will initially contain at least 0.04 grams of boron-ten per square centimeter of plate.

NSP also shows in this submittal that the average center-to-center pitch between fuel assembly storage tubes will be maintained at 9.50 + 0.060 inches

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by the external sheets and by welded spacers.

3.1 Criticality Considerations The Prairie Island fuel pool criticality calculations were performed under the assumption of 39 grams of U-235 per centimeter of assembly length, clean unborated water in the pool, fresh fuel without burnable poison, and an array of infinite extent in the Tateral direction. Calculations were done for the standard Westinghouse 14x14 fuel assembly, the slightly different Exxon replacement design, and a proposed optimized (for uranium utilization) Westinghouse design. Since the latter two designs had smaller pellet diameters than the Westinghouse standard design, it follows that larger enrichments were used for them. Credit was taken for axial leakage but none was taken for rack structure (other than the storage cans) cr for the fuel assembly spacers.

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. The calculations were performed for Northern States by Nuclear Services Corporation. The CHEETAH and XSDRN computer codes were used to generate cross sections for use in the CITATION diffusion theory program. The results of these calculations were verified against critical experiments which included boral blades, and against a Monte Carlo calculation which used the KENO-IV code with cross sections prepared by the AMPX code package.

The effect of " normal" variations - uncertainties in box dimensions and variations in pool temperature have been considered along with calculational biases and uncertainty in arriving at the total uncertainty in the calculation.

The result for the optimized fuel case is a K-effective value of 0.942 with all uncertainties included.

3.1.1 Evaluation A comparison of the calculation presented in this application with others made for similar configurations shows it to be acceptably accurate. This is confirmed by the results of the comparison of calculated and experimental results referred to above which showed that the calculation tended to overestimate the multiplication factor.

In order to assure that ths appropriate amount of neutron poison is present a strict quality assurance procedure will be followed during the manufacture of the fuel storage boxes. In addition, after the,

racks are manufactured and are on-site a final verification of the presence of boron will be made by using a neutron source and detector combination. Each storage box will be examined. Continued presence of baron in the racks will be verified by removal of sample coupons periodically in order to test for deterioration.

3.1.2 Conclusion We conclude that the design of the proposed spent fuel storage racks is acceptable for storage of fuel for the Prairie Island reactor with respect to potential criticality in normal usage and in credible accident configurations. Our conclusion is based on the following:

1.

Standard state-of-the-art analysis methods are used for the analysis.

2.

The methods were verified by comparison with critical experiment.,

3.

The result for thass storage racks is consistent with that for similar rack designs calculated by other applicants using other methods.

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. 4.

Normal or expected variations in design parameters have been treated, either by assuming worst case values or by performing sensitivity studies.

5.

Credible " abnormal" configurations have been analyzed.

6.

Our criterion of less than or equal to 0.95 for the calculated effective multiplication factor has been met for both nomal and abnomal storage configurations.

7.

Tests will be conducted to assure the presence of the poison in the racks on the site.

Accordingly, we find that the proposed racks are safe with respect to criticality for the storage of fuel assemblies containing up to 39 grams of U-235 per cm. of assembly length.

3.2 Spent Fuel Cooling The shared PINGP spent fue'l storage facility consists of a small (pool #1) and a large (pool #2) storage pool housed within a seismic Category I structure. Transfer slots, with pneumatically sealed gates, pemit the fuel to be moved from the fuel transfer canal into either pool or betweer pools. The elevation of the bottom of the transfer slots is above t!)e top of the stored fuel.

Above the operating floor a seismic Category I concrete reinforced structure encloses the two spent fuel storage pool.s and the new fuel storage pit. The walls and roof of the enclosure serve as a tornado missile barrier. The passage of heavy loads in or out of the enclosure, using the overhead crane, is restricted to a slot in the roof and walls of the enclosure. The slots are located such that all loads being handled by the overhead crane would travel over pool #1, the smaller storage pool.

The new storage racks will be modular assemblies consisting of stainless steel storage tubes held and scoported on a 9.5 pitch by upper and lower grids. The grids are a box like structure fabricated from heavy plates.

A cypical storage rack is shown in Figure 3.2-1 Exhibit-C of the January 31, 1980 submittal.

3.2.1 Evaluation The licensed themal power for each of Units 1 and 2 is 1,650 MWt. During a nomal annual refueling cycle one third of a core's 121 fuel assembif es is discharged to the storage pools. The re*ueling cycles for the two units are scheduled such that a refueling operation takes place about t

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! every six months. The licensee's March 10, 1981 submittal provided the resultant decay heat loads. These heat loads have been calcu-lated in accordance with Branch Position ASB 9-2.

Further, in evaluating the adequacy of the cooling system it has been conservatively assumed that 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> will be required to prepare the facility for refueling and that either a normal core discharge or a full core discharge can be accomplished in 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> following shutdown, i.e.,

50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> are required for transferring fuel between the reactor vestel i

and the storage pools.

The spent fuel pool cooling system consists of two pumps and two heat exchangers. These are cross connected such that the loss of any one pump or heat exchanger will not prevent the operation of the remaining components. The decay heat is removed from the spent fuel pool heat exchangers by either Unit 1 or Unit 2's Component Cooling Water System.

Each unit's Component Cooling Water System consists of two,100%

capacity, normally interconnected parallel loops each comprised of one pump and one heat exchanger having a rating of 29x106 BTU /hR. In the unlikely event that a LOCA should occur in the unit whose Component Cooling Water System is connected to the Spent Fuel Pool Cooling System, the operator would conservatively have more than one hour to transfer the pool cooling system to the unaffected unit's Components Cooling Water System.

The free volume of pool #2 is slightly less than 2.5 times the free volume of pool #1. The volume of water, storage racks and spent fuel in the respective pools are essentially in the same ratio. The water volumes in pool #1 and pool #2 are 12,692 ft3 and 29,593 ft3 respectively.

Each pool has a high and low water level sensor which provides input to each control room's control board. The spent fuel pool liner seam leakage is directed to a common open sight drain trough for monitoring and then to the waste disposal system.

Temperature detectors are installed at both pools. A high temperature alarm for each pool, nominally set at 130*F, is located on each control room alarm panel. The auxiliary building operator, as a routine shift responsibility, monitors the spent fuel pool water level, temperature, radiation and the leak detection system. The control room operators, as part of their routine shift responsibility, monitor the spent fuel radiation levels.

l In the course of reconfirming the cooling system's heat removal capability NSP found that the hydraulic flow resistance of the two heat exchangers was unequal. Therefore the flow distribution to the two heat exchangers was revised and the maximum pool water temperature was recalculated

. using the revised heat loads, and flows through the two heat exchangers.

The assumed conditions and resultant maximum pool water temperatures are as follows:

(a) 1362 normally discharged fuel assemblies (11.97x106 BTU /HR peak heat load ) are stored in pools 1 and 2.

The pools are cooled by either the main heat exchanger and one of the two pumps or the backup heat exchanger and both of the pumps. The maximum pool water temperature will not exceed 137'F.

(b) 1362 normally discharged fuel assemblies plus a freshly off loaded core consisting of 121 fuel assemblies (25.09x100 BTU /HR peak heat load) are stored in pools 1 and 2.

The pools are cooled by both the main and backup heat exchangers and both pumps.

The maximum pool water temperature will not exceed 145'F.

(c) 1362 normally discharged fuel assemblies plus a freshly loaded core consisting of 121 fuel assemblies (25.09x106 BTU /HR peak heat load) are stored in pools 1 and 2.

The analyses assumed the failure of either one pump or one heat exchanger. The maximum pool water temperature will not exceed 183*F.

Based on the above results we conclude the spent fuel pool cooling system is adequate and therefore acceptable.

NSP investigated the elapsed time before pool boiling under two sets of assumptions following the loss of all external pool cooling.

The two postulated conditions and results are as follows:

l (a) Pools 1 and 2 contain 1362 nomally discharged fuel assemblies plus a full core discharge, and it was assumed that complete mixing of the water in pools 1 and 2 would occur. Calculations indicate that boiling would occur in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Further, since this is the maximum heat load the maximum boil off rate would be 51.4 gpm.

(b) Pool #1 contains one full core discharge plus 266 recently discharged fuel assemblies. It was further assume

• that there would be no mixing of pool 1 water with pool 2 water. Calculations indicate that boiling would occur in 2.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

We concur that it is reasonable to assume that 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is an adequate l

l amount of time to perfom minor maintenance on the cooling system and restore it to an operable condition and therefore the time established in the first postulated condition is acceptable. In order to avoid the

. short boil off time in the second postulated case, we believe that it would be prudent to distribute the 121 freshly off loaded core fuel assemblies in pools 1 and 2.

By distributing these fuel assemblics in a ratio approximating the water volumes of pools 1 and 2 the time to boil would approach 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Further, by adopting this requirement the assumptien made in the first case regarding complete mixing of water in poof s 1 and 2 becomes less crucial in establishing the time to boiling. Therefore we will require that no more than 45 uf the assemblies be placed in pool #1 and the remaining fuel usemblies be placed in pool #2.

Exhibit C of the January 31, 1980 submittal identified six available sources of makeup water in the unlikely event that all spent fuel pool cooling is lost and boiling occurs. Attachment 1 to the June 10, 1980 letter indicates the available makeup rat' from each source and states that ten minutes or less is re@ ired to line up the valves or to cam out the steps necessary in order to make the water available. The six sources of makeup water and their makeup rates are as follows:

(a)

Chemical and Volume Control System - 300 gpn, (b) Chemical and Volume Control System Blender - 100 gpm, (c) Refueling Water Storage Tank -

80 gpm, (d) Reactor Makeup Storage Tanks - 80 pm, (e) four demineralized water hose stations, each station ested at 20 gpm, and (f) the fire protection system - t.'.are are two fire hose stations near the spen +

• l fuel pool each rated at 95 gpm.

Regarding the maximum required makeup rate of 51.4 gpm and the number of makeup sources and their respective makeup rates, we find the makeup sources to be adequate and therefore acceptable.

3.2.2 Conclusion We have reviewed the calculated decay heat values and find them to be consistent with Branch Technical Position ASB 9-2 and therefore accep-table. The described spent fuel pool cooling system performance has been reviewed and found to be adequate and therefore acceptable. The available time required before the water will be available has been reviewed and found to be acceptable.

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. In regard to the time before boiling occurs, following a full core discharge, we find the time interval of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> acceptable provided the stored fuel assemblies are distributed in both pools as specified ateve. This distribution of fuel assemblies has been addressed by Technical Specification 3.8.B.5.

3.3 Installation of the New Storage Racks and Lead Handling In order to change the storage racks within the reinforced concrete enclosure, a 15 ton temporary crane will be erecte inside the enclosure.

In addition, a temporary laydown are; ill be provided above pool #1 in order to permit the transfer of loads from one crane to the other one. The temporary laydown area will consist of covers placed above pool !1 in order to support the storage racks while the load transfer is being accomplished.

To accomplish the modification the following operations will be performed:

With all fuel assemblies (442 assemblies) stored in the larger pool (pool #2), the 15 ton temporary crane is erected over the smaller pool (pool #1) and placed on the fuel handling bridge rails. The temporary crane and 25 ton capacity overhead auxiliary hoist will be used in removing the existing empty storage racks in pool #1 and installing the new storage racks. All of the stored spent fuel will then be moved from the old storage racks in pool #2 and placed in the new storage racks in pool #1. The protective covers will be pieced over pool #1 in order to protect the stored fuel and to provide a temporary laydown area. Using the 15 ton temporary crane, the existing storage racks in pool #2 will be individually lifted, transported and placed on the laydown area provided by the protective covers where the 25 ton overhead auxiliary hoist will pick them up and remove them for disposal.

Similarly the new racks will be individually moved, by the 25 ton overhead hoist, into the enclosure and placed on the protective covers where the temporary crane will lift, transport, and locate them in pool #2. When all the new racks have been installed the l

15 ton temporary crane will be disassembled and removed by the overhead hoist. The covers will be removed from pool #1 and the stored fuel will be moved to pool #2 using the fuel handling bridge.

The weights of the old and new storage racks are equal to or less than 12.4 tons. The above procedures effectively limit the possibility of dropped load type damage to the stored fuel to the time that loads are being handled above pool #1 protective covers.

. The 25 ton auxiliary hoist, which is mounted on the Auxiliary Building Crane, will be employed for the movement of loads outside the enclosure and within the enclosure slot. Outside of the enclosure, the travel paths of the loads are such that no equip-ment essential in the safe shutdown of the reactor is located beneath, adjacent to or otherwise within the area of influence of a dropped load. The FSAR indicates this crane has been designed, fabricated and qualified in accordance with the Electric Overhead Crane Institute (EOCI) Standard #61 and the American Standard Institute Standard B30.2-1967. Considering that the heaviest rack is less than one half of the 25 ton hoist's capacity, the safety margin for these operations t 'll be twice what would exist when the hoist is handling its roi.2d load of 25 tons.

While it is not anticipated to move loads, other than those associated with the modifications, through the enclosure and over pool #1, NSP has indicated that should that become necessary these loads will not be moved without first removing all fuel from storage pool #1.

The 15 ton temporary crane is a double leg gantiy type unit having motorized drives for vertical and north / south motions. East / west motions will be powgred manually.

Aside from one rack that will require slings, lifting rigs will be-employed in handling the old and new storage racks. The lifting l

rigs will have four hooks to engage the four corners of the rack.

The vertical dimensions of the rigging will be such that the carrying height of heavy loads above the temporary laydown area will not i

exceed 6 inches, i.e., the hook will be essentially at its upper limit of travel. All rigging will have an overall factor of safety of 10.

NSP states only trained and experienced NSP plant personnel will be permitted to operate the cranes during this modification.

l Since neither of the above cranes are single-failure-proof cranes, the potential exists for dropping their load. A load drop onto the covers above pool #1, while the pool contains all 442 spent fuel assemblies, would potentially be the most severe accident. The i

l adequacy of the covers to withstand load drops is discussed in section 3.4.

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The Prairie Island Nuclear Generating Plant Units 1 and 2 Technical Specification 3.8.B.1 states "No heavy loads will be transported over or placed in either part of the spent fuel pool when irradiated fuel is stored in that part". This limitation makes it impossible l

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to remove and insert the storage racks as described in this tacdification as well as utilizing pool #1 for the long tem storage of spent fuel. As in the previous pool expansion program, NSP is requesting that a temporsry waiver be granted i

to Technical Specification 3.8.B.1 during the load handling operations associated with the pool modification program. In this regard considering the limited number of such operations, the dascribed precautionary measures and safety margins that will exist when carrying out these operations we find the request acceptable. In regards to the proposal contained in the present submittal to have spent fuel stored in pool No.1 when heavy loads such as the four temporary storage racks or the spent fuel shipping cask are inserted or withdrawn from the pool, the following stipulation was agreed upon by NSP, the NRC staff and the State of Minnesota on September 23, 1980:

"Until such time as NRC issues to N3P a license amendment authorizing insertion and withdrawat of a spent fuel shipping cask into and from pool #1 when spent fuel is stored in that pool, NSP shall store in the spent fuel pools no more than 1120 spent fuel assemblies discharged as a result of nomal refuelings. This limitation shall not apply to storage of any fuel which is to be returned to the reactor. NS? may store spent fuel in pool #1 so -

long as there are storage locations in pool #2 into which all spent fuel in pool #1 can be placed prior to insertion of a spent fuel shipping cask".

3.3.1 Evaluation The described sequence of steps by which the modifications are to be accomplished is such that to the extent possible the stored spent fuel assemblies will be removed from the areas where heavy load handling operations take place. To protect the stored spent fuel assemblies during the removal and installation of the storage racks l

protective covers will be placed over pool #1. These covers have been analyzed assuming the stc. age racks are drcpped from a height of six inches. To preclude greater drop heights the vertical distance of the rigging will be so arranged that the respective hoists will be at their upper limit of travel when the bottom of the rack is six inches above the protective covers. The rated load capacity of both the 25 ton auxiliary hoist and the temporary 15 ton gantry exceeds the weights of both the old and new storage racks. The rigging will have a safety

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. margin of 10. Further, only trained and experienced personnel familiar with the equipment will be pemitted to operate the cranes during this modification.

In the past we have limited our reviews of spent fuel pool load drop accidents to loads equal to or greater than the weight of one fuel assembly and or its associated handling tool. Subsequently, it has become apparent that lighter loads, i.e., other normally handled loads, weighing less than one fuel assembly, and/or its associated handling tool, could potentially, if dropped, impact on stored spent fuel with a greater amount of kinetic energy and hence possibly cause greater damage to the stored spent fuel.

NSP's March 31, 1981 submittal elaborated on their March 10, 1981 response on It states that there are three nomally used handling this subject.

tools which, if dropped from their maximum drop height, could potentially possess more kinetic energy than one fuel assembly and its associated handling tool when dropped from its maximum carrying height above stored spent fuel. These tools are:

the Burnable Poisor Rod Assembly Handling Tool, the Spent Fuel Handling Tool, and the Thimble Plug Handling Tool.

To prevent the potential energy of these three long handled tools from exceeding 3,250 foot-pounds (i.e. the potential energy of one fuel assembly and its handling tool,1,625 pounds, when dropped from its maximum carrying height, 2 feet, above the stored spent fuel) NSP will administrative 1y limit the carrying height of the tools such that the products of their weight times the drop height will not To assist the operator in accomplishing exceed 3,250 foot-pounds.

this, identification marks will be located on the shank of the tool at In addition, the pool's water level when the tool is at this elevation.

only trained and experienced NSP personnel will handle the tools.

Both of the hoists employed in handling the tools (i.e. the 6,000 pound capacity spent fuel pool bridge crane hoist and the 25 ton capacity Auxiliary Building Crane hoist) have keepers on the hook in order to prevent the tool's lifting bail from inadvertently becoming disengaged from the hook. The bails on the handling tools (in the 'oaded condition) have a factor of safety ranging i

from 4.8 to 8.5.

We believe that the identification marks placed on the shank of the handling tools will assist the operator to administratively limit the carrying height of the tools above stored spent fuel.

This, in conjunction with keepers on the hoist hooks and the reserve load capacity of the hoists, leads us to conclude that reasonable assurance has been provided to prevent a light load drop on stored spent fuel which would exceed

- the damage potential of a dropped fuel assembly and is, tnerefore, acceptable.

3.3.2 Conclusion In conjunction with our findings set forth in Section 3.4 of this report regarding the structural adequacy of the pool #1 cover, we conclude that the request for a temporary exemption to Technical Specification 3.8.B.1 using the described handling equipment and handling procedures is acceptable in order to complete the proposed spent fuel pool modifications.

The above procedures and conditions will reduce the possibility of dropping a rack or a fuel cask onto stored fuel assemblies to an acceptable level. From this and the protection provided by the protective covers, we conclude that the health and safety of the public will not be endangered by reracking the spent fuel pools and is therefore acceptable.

The information supplied regarding the handling of light loads above stored spent fuel includes a description of the identification marks on the shank of the handling tools that will assist in limiting the carrying height of tools above the stored spent fuel. This, in conjunction with keepers on the hoist hooks and the reserve load capacity of the hoists, leads us to conclude that reasonable assurance has been provided to prevent a light load drop on stored spent fuel which could exceed the damage potential of a dropped fuel assembly and is therefore acceptable.

3.4 Structural Design The design for the racks, fabrication, and installation criteria; the structural design and analysis procedures for all loading, including seismic and impact loadings; the load combinations; the structural acceptance criteria; the quality assurance requirements for design, and applicable industry codes were all reviewed in accordance with the applicable portions of the current "0T Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications",

dated April 1978, including revisions, dated January 1979.

1 The design of the spent fuel storage modules utilized the AISC Code

" Specification for the Design, Fabrication and Erection of Structural Steel for Buildings". The basic material allowables were taken from the ASME Code,Section III, Division and Standard Review Plan Section 3.8.4 for the applicable loao combinations. The fabrication and insta11stion of the modules are in accordance with the ASME Code Section III, Division 1, Subsection NF with the following exceptions:

(1) material traceability is preserved for each module and not for each individual piece of the module, (2) the neutron absorber

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• material was purchased in accordance with ANSI documents since the ASME code does not address this material, and (3) tre welds ssed to install the Boraflex into the fuel racks were made by the tungsten inert gas welding process. These welds are not strucNral welds and a random failure of 50% of these welds would not result in dislocation of the Boraflex.

3.4.1 Evaluation The seismic analysis of the racks utilized a time history analysis including structural damping consistent with Regulatory Guide 1.61,

" Damping Values for Seismic Design of Nuclear Power Plants" and Regulatorf Guide 1.92.

The structural evaluation of the proposed racks was based on the results from previous seismic analyses contained in the report entitled

" Revised Earthquake Analysis for Prairie Island Nuclear Generating Plant" by John A. Blume and Associate Engineers, February 16, 1971.

The spectrum corresponding to the spent fuel pool floor was used in i

the analysis. Althuugh the racks have no floor attachment nor lateral l

supports, it was assumed in the analysis that the coefficient of L

friction was such that the racks would not slide thus maximizing the internal stresses produced by seismic forces.

The 7X8 rack was used in the ana' lysis to determine loads, stresses and deflections, sint.e this rack had the greatest potential for tipping I

and would develq the greatest internal forces due to seismic and deadweight loadings. Two models were used in the analysis of the fuel rack. A three dimensional finite element model was used to determine stresses in the rack resulting from seismic, tMrmal, grapple, buoyancy and dead weight and a one-dimensional model wa. used to determine maximum rack sliding distance duriag an SSE. All the stress analyses on the finite element model were perfonned using the computer program STARDYNE and the stresses resulting from dead weight and buoyancy loads l

were evaluated simultaneously.

A nonlinear sliding analysis was performed to determine the maximum displacement and velocity of the rack relative to the pool floor under the action of SSE vibratory motion. 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 the information provided in a report by E. Rabinowicz of the Massachusetts Institute of Technology entitled " Friction Coefficients of Water Lubrication Stainless Steel for a Spent Fuel Rack Facility" dated November 5,1976.

The result of this analysis indicates that during an SSE, the proposed racks which are free-standing may slide towards each other and impact in a random fashion. The methodology used by the licensee to predict l

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. the stresses generated by the impact has been found to be acceptable.

The postulated fuel assembly drpp was considered in the analysis of the racks. The energy balance method was used to determine the effects of the impact of a fuel assembly dropped from a maximum height of 18 inches at the most c-1tical location on the rack. Three postulated drops were analyzed, namely, the straight drop of a fuel assembly through an individual cell, an inclined drop on the rack and a vertical impact on the rack where a point load was assumed instead of finite impact area. In all cases, the impact energy is dissipated by local yieldinu or crushing, however gross stresses in the rack remain below allowable and the overall structurai integrity is maintained.

The effects Jf a postulated stuck fuel assembly due to the attempted assembly withdrawal was considered and the stresses due to this postulated accident were computed by an elastic analysis.

In order to protect fuel assemblies in the small spent fuel pool against the accidental drop of a heavy load, a protective cover over the pool is provided. The pool cover is made of 3/16 inch stainless steel plate welded to a grid of structural tees and built-up wide-flange beams which are made of structural steel ASTM A588 Grade A.

Underneath each end of the beams, one pad made of one-inch thick compressible material is used between the cover and the concrete floor.

The licensee had evaluated the protective cover when subjected to a postulated drop of 24,800 pounds at a height of 6 inch clearance above the cover. The results of the evaluation show that although local plastic deformation may occur, the overall structural integrity of the cover will be maintained. Thus, the effect of the postulated drop of this heavy load is considered to be within the acceptable Ifmit.

The loads and load combinations considered in the analysis of the spent fuel storage racks are in accordance with SRP Section 3.8.4.

Results of the analysis show that the racks are capable of withstanding the loads associated with all the design loading conditions without exceeding allowable stresses.

The spent fuel pool is constructed of concrete walls and floor, lined with a stainless steel liner and reinforced in both vertical and horizontal f

directions. The fuel pool concrete, reinforcing steel and liner were analyzed to account for the additional loadings imposed by the new racks. The structural adequacy was verified using conventional concrete building codes ( ACI 318).

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Results of the analysis for the most severe loading conditions indicate that the maximum stresses are within the allowables, and that the structural meters of the fuel poci are adequate to withstand the additional loads imposed by the new racks and additional fuel.

3.4.2 Conclusion The analysis, 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 leadings, including seismic, thermal, and igact loading; the acceptance criteria for the appropriate loading conditions and combinations; and the applicable industry codes are in accordance with appropriate sections of the NRC Staff "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications".

Allowable stress limits for the combined loading conditions are in accordance 41th the AISC specifications. Yield stress values at the appropriate tegeratures were obtained from Section III of the ASME B&PV 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 the ASW. Code.

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 Standard Review Plan Section 3.8.4.

Results of the seismic structural analyses indicate that the racks are capable of withstanding the loads associated with all design loading conditions. Also, igact due to fuel assembly / cell interaction has been l

considered, and will result in no damage to the racks or fuel asse211es.

The methodology used by the licensee to predict the stresses generated by the impact during the SSE has been found acceptable.

Results of the dropped fuel assembly analyses show that local rack deformation will occur, but indicate that gross stresses meet the applicable allowables and +.bt the integrity of the racks is maintained.

Results of the dropped heavy loads over the protective pool cover indicate that although local damage and plastic deformation may occur, the overall structural integrity of the cover is maintained and is within the acceptable licits.

Results of the stuck fuel assembly analyses show that the stresses are below l

those allcwed for the applicable loading combinations.

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. We find that the subject modification proposed by the ifcensee is acceptable and satisfies the applicable requirements of the General Design Criteria 2,4,61, and 62 of 10 CFR, Part 50, Appendix A.

3.5 Occupational Radiation Exposure We have reviewed the licensee's plan for the removal and disposal of the existing racks that were installed during a previous modification in 1977/1978 and the installation of the new racks with respect to occupational radiation exposure. The occupational exposure for this operation is conservatively estimated by the lic :isee to be about 40 man-rems. We consider this to be a reasonable estimate because it is based on the licensee's detailed breakdown of occupational exposure for each phase of the modification. The licensee considered the number of individuals perfoming a specific job, their occupancy time while perfoming this job, and the average dose rate in the area where the job is being performed.

The modification will be perfomed by reracking pool 1, transferring spent fuel elements from pool 2 into pool 1 and then reracking pool 2.

The spent fuel will then be returned to pool 2.

The licensee has indicated that alternative plans are being evaluated for the disposal of the present racks which include removing and crating the racks intact for shipment offsite versus removing, cleaning by electropolishing and subsequently disposing of the racks. The licensee has estimated that either alternative will result in an occupational exposure of four man-rems or less. This contribution to occupational exposure has been included in the over all estimate of 40 man-re:as discussed above. Selection of a disposal method has not been finalized. The

~icensee will estimate the exposures associated with the different ways to dispose of the present racks from measurements of the activity levels on them when they are removed from the pool and are ready for disposal. At that time taking into account alternative disposal costs and exposures, the licensee will select the method of disposal so that exposures will be kept to levels that are as low as is reasonably achievable. All work will be effected in accordance with a radiation permit to fientify all protection requirements. Health physics personnel will be available to assure that ALARA radiation exposures prevail.

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 rpent fuel

. area from radionuclide concentrations in the SFP water and deposited on the SFP walls. The spent fuel assemblies themselves will contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel. The occupational radiation exposure resulting from the additional spent fuel in the pool represents a negligible impact. 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 this facility. The small increase in additional 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 occupational workers.

3.6 Radioactive Waste Treatment The plant contains waste treatment systems designed to collect and process the gaseous, liquid, and solid wastes that might contain radioactive material. The waste treatment systems were evaluated in the Safety Evaluation dated September 1972. There will be no change in the waste treatment system or in the conclusions given in Section 11.0 of the evaluation of these systems because of the proposed modification. Our evaluation of the SFP cleanup system, in light of the proposed modification, has concluded that any resultant additional burden on the system is minimal and therefore the existing SFP cleanup system is adequate for the proposed modift-cation and will keep the concentrations of radioactivity in the pool water within acceptably low levels.

Our evaluation of the radiological considerations supports the conclusion that the proposed modification to the Prairie Island 1 and 2 spent fuel pools is acceptable because:

(1) The conclusions of the evaluation of the waste treatment systems, as found in the Prairie Island Safety Evaluation Report of 1972, are unchanged by the modification of the SFP.

l (2) The existing SFP cleanup system is adequate for the proposed modification.

3.7 Material The proposed spent fuel storage racks are fabricated of Type 304 stainless steel with the exception of the adjusting bolts of the rack feet. These bolts are made from Type 17-4 PH stainless steel.

The existing spent fuel pool liner is stainless steel.

The high density spent fuel storage racks will util':e Boraflex sheets i

as a neutron absorber, within an annulus formed by ncentr c

-,...,,._,,--._-_.,-.._-,.-m,

,-w m.

..-~~,._,_,--,,y_.--

..w-.r-

-.,-,-r--. ~. - -..

, -- -, _ ~ --

_ inner and outer, square stainless steel tubes. The Boraflex is t

composed of boron carbide powder in a rubber-like silicone polymeric matrix. The Boraflex sheets will have a minimum boron-ten content of 0.04 gm/cm2 of sheet surface area. The spent fuel storage rack is composed of individual storage cells interconnected to form an integral grid structure. The annulus region between the concentric tubes that contains the Boraflex is vented at both the top and bottom. The Boraflex is not attached to either of the stainless steel tubes. It is captured in the annulus and is supported on a stainless steel strip at the bottom J

of the annulus.

The pool contains oxygen-saturated demineralized water containing boric acid, generally controlled to a temperature below 130*F.

3.7.1 Evaluation The pool liner, rack lattice structure and fuel storage tubes are l

stainless steel which is compatible with the storage pool environment.

In this environment of oxygen-saturated borated water, the corrosive deterioration of the type 304 stainless steel should not exceed a depth of 6.00x10-5 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 structure, fuel storage tubes, and the Inconel and the Zircaloy 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. The Boraflex is composed of non-metallic materials and therefore will not develop a galvanic potential in contact with the metal i

comoonents. Baraflex has undergone extensive testing to study the effects of gamma irradiation in various environments, and to verify its structural I

integrity and suitability as a neutron absorbing material. The evaluation tests have shown that the Boraflex is unaffected by the pool water environ-ment and will not be degraded by corrosion. Tests wgre performed at the University

• Michigan, exposing Boraflex to 1.03x10'l rads of gamma radiation with substantial concurrent neutron flux in borated water.

These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of borated water and gamma irradiation. Irradiation will cause some loss of flexibility, but will not lead to break up of the Boraflex.. l.ong term borated water soak tests at high temperatures were also conducted. The test showed that Boraflex withstands a borated water immersion of 240*F for 260 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide.

(

, The annulus space which contains the Boraflex is vented to the pool at each corner storage tube assembly. 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 or swelling of the inner stainless steel tube.

The tests have shown that neither irradiation, environent 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 thac 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 in the Boraflex will typically cor.tain 0.1 wt percent of soluble boron. The test results have confirmed the encap-sulation function of the silicone polymer matrix in preventing the leaching of soluble specie from the boron carbide.

To provide added assurance that no unexpected corrosion or degradation of the materials will comnromise the integrity of the racks, the licensee has committed to conduct a long tenn fuel storage cell surveillance program.

Surveillance samples are in the form of ranovable stainless steel clad Boraflex sheets, which are proto-typical of the fuel storage cell's walls.

l These specimens will be removed and examined periodically.

1 3.7.2 Conclusion From our evaluation as discussed above, we conclude that the corrosion that will occur in the Prairie Island spent fuel storage pool environment should be of little significance during the 40-year 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 i

corrosion. Tests under irradiation and at elevated temperatures in borated water indicate that the Boraflex material will not undergo significant degradation during the expected service life of 40 years.

We further conclude that the environmental compatibility and stability of the materials used in the Prairie Island expanded spent fuel storage pool is adequate, based on the test data cited above and actual service experience in operating reactors.

We have reviewed the surveillance program and we conclude that the monitoring of the materials in the spent fuel storage pool, as proposed by the licensee, will provide reasonable assurance that the Boraflex material will continue to perfonn its function for the design life of the pool. We therefore find that the implementation of a monitoring l

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. program and the selection of appropriate materials of construction by the licensee meets the requirements of 10 CFR Part 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 poison.

4.0 Technical Specification As indicated in the criticality analysis of this Safety Evaluation and in the licensee's referenced submittals the maximum Uranium-235 content is specified in Technical Specification 5.3 and 5.6 to be. 39 grams per axial centimeter of fuel assembly. Therefore fuel assemblies that are bound by the fuel assembly designs described in the licensee's referenced submittals may be stored in the spent fuel pool.

Technical Specification 3.8.B.1 is modified to pennit its suspension during the fuel pool reracking operation so that the old racks may be removed and the new racks installed.

Technical Specification 3.8.B.5 is added to implement the specification associated with a discharge of a complete core of fuel assemblies as discussed in Section 3.2.1 of this report.

Technical Sp'ecification 5.6.D has been added to implement the stipulation on allowable spent fuel storage discussed in Section 3.3 of this report.

5.0 Safety 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 issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Dated: May 13, 1981 l

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