ML20083M833

Spent Fuel Disposition Plans

ML20083M833
Person / Time
Site:	Millstone
Issue date:	03/31/1984
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20083M821	List: B11039, Forwards Spent Fuel Disposition Plans, Outlining Program Undertaken to Increase Spent Fuel Storage Capacity ML20083M833
References
PROC-840331-01, TAC-54671, NUDOCS 8404180362
Download: ML20083M833 (45)

Text

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s MILLSTONE NUCLEAR POWER STATION 4

UNIT NO. 2 DOCKET NO. 50-335 LICENSE NO. DPR-65 .

i SPENT FUEL DISPOSITION PLANS

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CONTENTS 1.0 Introduction 1.1 The importance of Spent Fuel Storage Capacity 1.2 General Description of Millstone Unit No. 2 Pool 1.3 Basis for the NU Capacity Expansion Program 1.4 General Description of Proposed Pool 1.5 Generic Applications 2.0 Nuclear Considerations 2.1 Criticality Determinations 2.2 Two Region Fuel Storage 2.3 Reactivity Monitoring 3.0 Thermal and Hydraulic Considerations 3.1 Review of Current Spent Fuel Pool Cooling System 3.2 Methodology and Analysis

3. 3 Thermal H)draulic Test Program 4.0 Mechanical Considerations 4.1 Material of Construction 4.2 Seismic Methods Development 4.3 Seismic Analysis 4.4 Pool / Auxiliary Building Analysis 5.0 Radiological Considerations 6.0 Accident Analyses

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7.0 Rerack 7.1 General Description of Procedtre 7.2 Seismic Considerations 7.3 Load Handling A

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1.0 INTRODUCTION

1.1 The Importance of Spent Fuel Storage Capacity Effective utilization of industry's substantial investment in nuclear power generation f acilities requires that they continue to be operated for their full design life. Continuing operation requires adequate provisions for accommodating the spent fuel which must be discharged periodically. Only limited spent fuel storage capacity has been provided in commercial nuclear power station site facilities because it generally has been assumed that longer term fuel disposition provisions such as reprocessing or ultimate disposal in geologic repositories would be avadable relatively early in the design life of such plants.

However, it is not clear that this will be the case. Furthermore, mder current Federal policy, industry is responsible for interim storage of spent fuel until longer term disposition provisions are available. This mandates that f uel owners give priority attention to meeting their own needs f or spent f uel storage capacity.

1.2 General Description of the Millstone Unit No. 2 Spent Fuel Pool Northeast Nuclear Energy Company's (NNECO) Millstone Unit No. 2 facility received its operating license (DPR-65) in August,1975. At that time there was capacity to store 301 spent f uel assemblies or about 1.3 f ull cores in the spent f uel storage pool.

In November 1976, NNECO concluded that a capacity expansion of the spent fuel pool was necessary to support the engineering practice and corporate policy of reserving storage space in the spent f uel pool to receive an entire discharged reactor core (" full-core-of f-loa #)

should it become necessary due to operational considerations.

Additionally, spent fuel reprocessing facilities were not expected to be available in the near-term.

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_4 The spent fuel pool rerack license amendmentl was obtained and the project was completed prior to the first ref ueling in the f all of 1977.

The modified storage pool provided storage locations for 667 f uel assemblies. The increased spent f uel storage capacity provided for the " full-core-off-load" capability needed beyond the 1978 ref ueling, as well as the capacity needed for the spent f uel discharges through 1934. The pool design configuration is composed of nine rack modules, each containing 63 f uel assembly storage locations in a 7 x 9 array and one rack module containing 100 fuel assembly storage locations in a 10 x 10 array. The modules are arranged as shown in Figure 1.0 and store the f uel assemblies with a nominal center-to-center spacing of 12.19 inches.

1.3 Basis for the Millstone Unit No. 2 Capacity Expansion Program in 1985, Millstone Unit No. 2 will lose the reserve capacity necessary to discharge the entire reactor core into the spent f uel pool with the present capacity of 667.

Current circumstances in the back-end of the nuclear f uel cycle make it necessary that f uel owner's establish and implement a plan for " life-of-reactor-storage" of nuclear spent f uel. Numerous utilities are presently planning capacity expansion projects for which interim storage of spent fuel will have to be provided over the next fif teen to twenty years.

Spent fuel consolidation has the potential to meet the needs of a large number of utilities for additional storage capacity in their existing pools at reasonable costs. Spent fuel consolidation refers to a process whereby individual fuel rods are removed from the fuel assembly and arranged in a compact array. Spent f uel consolidation can effectively double the storage capacity of the spent fuel pool by providing a more eificient use of available storage space. Storage of consolidated f uel in spent f uel racks properly designed for the service

O loadings associated with consolidated f uel is the best approach to the maximum utilization of storage space available in the spent f uel pool.

Moreover, when other modes of interim storage are dictated by weight limitation on existing spent f uel pools, consolidation holds the prospect of substantially reducing the cost associated with the transport and/or storage of the spent f uel.

NNECO intends to apply for the license amendments necessary to support the reracking and the storage of consolidated spent fuel in the Millstone Unit No. 2 spent fuel pool. NNECO intends to develop benchmarked analytical methods and related data on consolidated fuel characteristics that will support the licensing of a spent fuel pool with storage racks that have been designed to criteria for consolidated spent f uel. Additionally, NNECO intends to conduct a "hnt demonstration" of spent f uel consolidation with production-scale equipment and processes on approximately five to ten spent fuel assemblies.

1.4 General Description of Proposed Pool Several recent developments within the nuclear industry have produced new design considerations that influence the spent fuel storage options available to utilities currently planning capacity expansion projects.

The first development is the Nuclear Waste Policy Act (NWPA) of 1982, which requires fuel owners' to provide on-site spent fuel storage until a government repository is available. The second development is proposed NRC Regulatory Guide 1.13, Revision 2, which permits credit to be taken for reactivity depletion in nuclear spent f uel. In particular, this proposed NRC policy permits a new approach to the design of a spent f uel storage rack which will provide a substantially closer center-to-center spacing for increastd capacity while eliminating the use of poison material and its

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associated engineering complications, complexities and cos ts.

Additionally, the new policy introduces the concept of a " region

strategy" that can be employed to achieve the maximum utilization 4

,' of the storage space available in the spent f uel pool.

The " region strategy", proposed by NNECO for the spent f uel pool design configwation is a two-region dual-pitch pool of both poisoned and non-poisoned spent f Jel racks as shown on Figure 1.1.

l Region I would contain the high-enrichment, core off-load I assemblies. Fuel assemblies would be stored in every location. The I

region consists of poisoned spent f uel racks with a nominal center-to-center cell spacing of 9.8 inches. The five modules of Region I totaling 362 storage locations are designed to accommodate 1.6 reactor cores of high enrichment nuclear spent fuel.

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l The spent f uel rack design for Region 1 is based upon the commonly j accepted physics principle of a " neutron flux trapP with the use of neutron absorber materials. The racks are designed to store u

Millstone Unit No. 214 x 14 fuel with an initial enrichment c ' 4.5 -

! w/o U-235. The poison material to be used is Boroflex.

1 i Region 2 is reserved for f uel that has sustained at least 85% of-its

! design bwn-up. Fuel assemblies are stored in a three-out-of-four logic pattern as shown on Fgure 1.2. The spent fuel rack design is based on the criticality acceptance criteria specified in Revision 2 of Regulatory Guide 1.13 which allows credit for reactivity depletion in i spent fuel. (Previously, the physics criteria for fuel stored in:the -

4 spent fuel pool were defined by the maximum mirradiated initial l enrichment of =the f uel.) The fourth. location' of ' the storage 1

~ configuration remains empty to provide the flux trap for reactivity.

control. Blocking devices .will be used to prevent inadvertent '

l placement of a iuel assembly into the fourth location.

1 Region 2 consists of twelve modules of non-poisoned spent fuel racks i whose nominal . center-to-center cell' spacing is 9.0 inches. The- ,

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I modules consist of 907 cells with storage capacity for 680 fuel l assemblies that have sustained at least 85% of design burn-up.

NNEGO intends to utilize the " state of the art" techniques developed

. for measuring the many properties of irradiated nuclear spent f uel to verify that a f uel assembly complies with the burnup criterion. In particular, a reactivity monitor, an apparatus for measuring directly the subcritical multiplication of individual fuel assemblies, is being developed.

[ The reactivity monitor is an improved, more precise method for determining the characteristics of the discharged spent f uel. Current i practice relles upon administrative procedures which require the detailed power history be made available for each f uel assembly. The power history is then caref ully reviewed with complex and time-l consuming calculations to determine the burn-up of each assembly.

j The reactivity meter places the emphasis on the direct measurement that improves the speed of operation with increased reliability of the i burnup determination and verification.

i j With the reactivity meter, the fuel assembly bwn-up determination

! will be performed in the fuel upender prior to placement of the discharged f uel into the spent f uel pool. Fuel intended for storage in Region 2, once it has passed the reactivity monitor test, could be placed directly into Region 2 of the pool. if the fuel does not pass i

the reactivity test,it would be placed into Region 1 of the pool.

The spent fuel rack modules in both regions of the f uel pool are designed to be f ree-standing and direct bearing onto the spent f uel pool floor liner. The rack modules will be fabricated from 304

stainless steel and will be seismically quallfled without mechanical deper.dence on neighboring modules or the pool walls.

I i The spent fuel rack modules in both regions of the fuel pool will be structwally capable of accommodating the loads generated by the

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j f storage of intact fuel assemblies as well as consolidated spent fuel in 1

i every storage location with a consolidated assembly compaction ratio I of not less than 2:1.

Since consolidated spent fuel with the water gaps removed is less

~ reactive than intact spent fuel assemblies, storage of consolidated spend fuel can be achieved in every location of the pool.

Both regions of the spent fuel pool have been designed to store both l Intact fuel assemblies and consolidated spent fuel in a safe, coolable, j subcritical configuration with Keft essl than 0.95.

l The new spent fuel storage racks and supporting analyses will be provided by Combustion Engineering (C-E), Inc.

1.5 Generic Applications 1

] Aspects of the Spent Fuel Capacity Expansion Project at Millstone i Unit No. 2 are being performed under cooperative agreements with l the Electric Power Research Institute (EPRO. Specifically, one.

{ aspect of the project will be the demonstration of the applicability l and cost-effectiveness of spent fuel consolidation as a means of

! meeting future needs for interim on-site spent fuel storage. NNECO

! Intends to apply for a license amendment _ authorizing spent fuel- t j consolidation at Mllistone Unit No. 2.

I Fuel consolidation can double the effective storage capacity of properly designed spent fuel storage racks. Loading densities for .

some pools will be greater than the original design basis. Pool

[ cooling and purification systems and structural capabilities must be t

examined in detall. Spent fuel racks must be designed to carry the

greater loads associated with consolidated fuel. Methods must be developed to model the consolidated spent fuel envelope for seismic, structural, criticality, and thermal-hydraulic an11yses.

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, The NNECO Spent Fuel Capacity Expansion Project will develop j benchmarked analytical methods and related data on consolidated fuel characteristics in support of a license amendnent for the storage of consolidated f uel in the Millstone Unit No. 2 spent f uel pool. A demomtration of fuel consolidation with production-scale equipment and processes is also planned.

Specific outputs of the project that are expected to be of generic I

value include the iollowing:

o Application for appropriate licenses to the NRC for fuel consolidation in the Millstone Unit No. 2 spent f uel pool, o Development of non-proprietary codes (where applicable) for use in licensing activities for iuel consolidation.

j o Design, fabrication, and demonstration of advanced fuel j consolidation equipment.

i o Demonstration of equipment to handle, disposition and package waste f uel assembly skeleton hardware.

4 o Generation of a Thermal / hydraulic data base by performing heated flow experiments which simulate consolidated fuel

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POWER MAX CAPM REGION ll SYSTEMS SPENT PUEL STORED IN 3 0F 4 CAVITIES i.2 CCMSUSTION EN0iNEf*NG ltC

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2. NUCLEAR CONSIDERATIONS 2.1 Criticality Determinations C-E will perform calculations of the criticality of the f uel rack design, using developed methodologies covering the design of closer center-to-center spacing fuel racks. Within the scope of the proposed design, C-E will perform criticality saf ety calculations.

Composition dependent nuclear cross-sections for the spent f uel will be generated using the CEPAK(2) code or other codes developed during this study as well as auxiliary processing codes. Spatial calculations employed in the determination of the multiplication f actor will be perf ormed using the two dimensional transport code, DOT (3). Three dimensional and benchmark cases will be solved using 1

KENO.(4)

The 123 group cross-sections f rom Oak Ridge National Laboratory (ORNL) will be used as a starting point, since this cross-section set apparently will be utilized by the NRC dtring licensing reviews. The DOT code will be utilized to describe the pin geometry in the compacted f uel criticality analyses. In using DOT, the space limitations of DOT will require that the analyses be limited to a single f uel array. Theref ore, the K-infinity will be calculated by DOT and then compared to the KENO analyses for a homogenized fuel array. If this results in an acceptable comparison, KENO will then be run using the critical geometries.

These codes are the same codes used in C-E's design of the Millstone Unit No. 2 reactor core and internals. The codes have broad empirical data bases to support them and have been accepted by the nuclear indtstry. In addition, the NRC has licensed C-E's nuclear steam supply systems and spent fuel racks based on the tse of CEPAK and DOT.

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3 The calculations will be performed for an infinite array, and include both nominal and minimun compacted rack dimensions. The minimum dimensions will include the effects of concwrent adverse dimensional tolerances and the effects of structwal elements within [

the compact f uel racks. Calculations will be performed at several pool temperatures to insure that the multiplication factor changes are considered in the design.

l 2.2 Two Region Fuel Storage Region I i

Region I is being designed as the core of f-load region with a monolith to accommodate Millstone Unit No. 2 fuel in the rack structwe utilizing borated poison. The f uel assemblies are to be stored in j every location because of the use of poison plates and the corresponding space lef t for water flux traps. In addition to the

] Region I storage capacity needed for a f ull core off-load, storage

! space is also provided for storage of a batch of fresh fuel and space j ior some other f uel assemblies.-

1 Region II Region 11 is reserved f or f uel that has sustained at least 85 percent of

its design bwnup. Within Region II, f uel assemblies are stored in 75 percent of the total cavities with a tighter rack center-to center pitch and no poison utilized. The mconsolidated fuel is stored in a 3

, out of 4 storage configuration as illustrated in Figure 1.2. Cell blocking devices are used to preclude inadvertent placement of fuel assemblies into every fourth cavity which will ensure the flux trap ior reactivity control.

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O Both regiors of the pool have been designed to store f uel in a safe, coolable, subcritical (K-effective less than 0.95) configtration.

Region 11 presents a criticality philosophy which is approved for use by the most recent proposed revision to Regulatory Guide 1.13 on Spent Fuel Storage. Before Regulatory Guide 1.13, the physics criteria for fuel placed into the spent f uel pool were defined by a maximum mirradiated initial enrichment. Regulatory Guide 1.13 currently allows for credit to be taken for the reactivity depletion in spent f uel. It is important to note that reactivity depletion is not a imetion of the initial enrichment of the fresh fuel or of its target burnup but is a f metion of the subsequent percentage of target discharge btrnup achieved for that particular f uel assembly.

Since consolidated f u::1 (with water gaps removed) is less reactive than intact spent f uel assemblies, it can be stored in every location in Region II. Preliminary analysis indicates that the consolidated f uel could also be stored in Region I among the fresh f uel.

2.3 Reactivity Monitoring The reactivity monitor is an apparatus and method ior measuring the subcritical multiplication of individual f uel assemblies. The device will allow verification of the predicted credit for the burnup of the f uel when storing consolidated f uel in the spent f uel racks. In actual practice, a f uel assembly with a reactivity that meets the spent f uel rack requirements is used to calibrate the device.

Thereaf ter, each spent fuel assembly is measured ming the same apparatm, and only those assemblies having a lower subcritical muldplication than the calibration f uel assembly are placed in the spent fuel rack. Those assemblics having a higher multiplication are stored separately in racks designed ior mburned iuel. In this way,it is possible to verif y that the consolidated f uel has a reactivity within

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the design limits of the rack design.

The apparatus and method include placing a neutron flux detector in one control rod guide tube of a spent fuel assembly and a neutron

)- source in another guide tube. The subcritical multiplication is measwed, and compared with the measurement for the standard

assembly. The detector and source are attached to rods on a movable support member, as seen in Figwe 3.0; the arrangement very closely resembles a control element assembly. Since the rods carrying *Se source and detector are maintained in a constant spatial relationship due to the close fit of each rod within a respective rigid guide tube, t

the distance and angle between the detector and source used in the

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measurement of the standard fuel assembly can be accurately l l repeated ior measuring each spent f uel assembly.

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METHOD OF VALIDATION 123 Group X-Section Set lf Geometry to Model Fuel

• Pin Array in DOT-IV 1V X-Section Processing NITAL+XSDRN PM If DOT 123 Group Analysis i k.

SY KENO Analysis 123 Group with Homogenized Fuel k.

1 F KENO Analysis 123 Group of Criticals IP KENO Analysis with Reduced Group Structure i

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3. THERMAL-HYDRAULIC CONSIDERATIONS 3.1 Review of Cwrent SFP Cooling System The process of f uel cor.solidation may impose requirements upon the Spent Fuel Pool Cooling System above those originally comidered in the design basis of the system. These new requirements are:

o The ability to cool and purify an isolated portion of the fuel pool (the fuel cask area) during fuel consolidation operations and insertion into the consolidation storage canisters, and o The ability to provide cooling f or an increased quantity of spent f uel (more than twice the amount prior to consolidation) in the spent f uel pool in a new arrangement af ter the f uel is returned to the storage racks.

The existing system must be evaluated in terms of its ability to provide these new services and appropriate modifications must _ be made to the system should it be necessary.

The demand for additional cooiing, despite the approximate doubling of the density of stored f uel, is not expected to be a major corrern as consolidation is anticipated to occw after .the fuel has been discharged f rom the core for a minimum of five years. Decay heat generation rates at that time are relatively low and therefore the additional cooling requirements will also be low. In contrast, purification during fuel disassembly may require some system modifications due to the potential for additional crud not normally encountered during normal operation of the spent fuel cooling system. A temporary crud filtration. vacuum system . connected directly to the work location will substantially mitigate this phenomenon. ' Regardless of' the' anticipated impact, however, both of the new system design requirements will be evaluated prior to the perf orma ice of a f uel amolidation operatton.

The new demands upon the spent f uel pool cooling system will be quantified to provide a basis for the. evaluation of the existing systems' capability to meet the needs of f uel consolidation. Heat load requirements in the consolidation area during the f uel consolidation operation and in the f uel pool itself after the consolidated fuel is returned for long term storage will be developed assuming finite f uel irradiation and anticipated consolidation sequences (i.e., when af ter removal from the reactor core, the f uel f rom a specific batch is to be consolidated). Filtration system demands will be developed based upon experience in crud release dtring f uel reconstitution processes which are physically similar to the consolidation operation and should provide the basis for a reasonable estimate of the crud release during consolidation.

Ctrrent regulatory standards will be considered in the development of the requirements, particularly with respect to fuel pool temperattres and radiological considerations.

The capacity of heat exchangers, filtration devices, and ion exchangers will be compared with the new demands. The adequacy of flow paths -

particularly in the tiel cask areas where the consolidation will be performed - will be examined. This evaluation will provide a listing of these requirements for the consolidation grocess which are not met by the current systems.

For those areas where the existing system needs modification, system modifications will be developed. Additional components, instrumentation, and controls will be sized and appropriate design information provided along with necessary document changes. If changes to the system must be made, a description of the changes and the impact of the changes in terms of the modified systems' ability to meet regulatory requirements will be prepared. This description will be in the sufficient detail to support additional licensing,1f necessary.

3.2 Methodology and Analysis A thermal-hydraulic analysis is being perf ormed to verif y under normal and accident conditions the Millstone Unit No. 2 spent f uel racks can be cooled adequately for the storage of mconsolidated and consolidated f uel.

The proposed design bases for the analysis are shown in Table 3-1.

The maximum bulk water temperature (1500F)is assumed to be kept constant by the external cooling system dtring normal operation. It is assuned that the rack contains f uel assemblies which are loaded at the rate of about one-third core per calendar year plus the option of loading a f ull core at any time. The shortest decay time for one assembly of a core is three days. The shortest decay time for the entire core is six days, given that it takes at least three days to transf er the f ull core into the rack. It is assumed that part of the rack will be loaded with f uel assemblies and the other part will contain consolidated f uel cells. Fuel assemblies will decay for a specified amount of time in the rack and then fuel rods will be removed and placed into a Comolidated Fuel Storage Box. The nurnber of consolidated cells and the minimum cooling time for an assembly bef ore its consolidated will be determined f rom the thermal hydraulic analysis.

The thermal-hydraulic analysis acceptance criteria are as iollows:

o Bulk boiling of the entire pool must not exist dtring normal operation.

o Maximum fuel clad temperature will not exceed 6500F dtring both normal operation and accident conditions.

The thermal-hydraulic accident analysis meets the double contingency principle of ANSI N 16.1-1975, that is, that two unlikely, independent, concurrent events must be assumed to eccur. The events to be evaluated are:

o dropped iuel assembly in pool o dropped iuel cask or large object in pool o tornado, earthquake, or other severe, external phenomena occurring o loss of coolant f rom pool To meet these criteria, ncrmal operation is defined as any arrangement of the fuel assemblies in the rack for which the conditions of maximum pool fluid temperature of 1500F and minimum pool depth of 23 feet of water above the f uel are maintained.

Accident conditiors are defined by two conctrrent events; namely a dropped f uel assembly and a loss of coolant event. A loss of coolant event assumes that coolant is evaporated becatric of loss of external cooling. The decay heat of the f uel is removed by bolling the pool coolant. The loss of coolant event is bounded by assuming a minimum pool depth of 10 feet of water above the racks sind a maximum fluid temperature in the racx region of 2330F.

3,3 Thermal Hydraulle Test Program An experiment, using a heated test section, will be performed to confirm the results of the thermal /hy&aulic analyses. In the experiment, electrically heated ros will simulote the heat generation of the f uel rods.

The test section will contain 19 f ull length f uel rod simulators which are arranged in a hexagonal lattice. The test section will be placed in a vessel so that the pressure and temperature condtlors in the

actual spent f uel rack may be simulated, instrunentation will be installed to meastre the coolant and cladding temperattre distribution, flow rate and pressure drop in the test section u1 der nattral circulation conditions.

The benefits of the test would be:

1. Experimental verification of the thermal-hydraulic computer codes utilized.

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2. Confirmation of the adequacy of the correlatlors chosen for heat transfer coefficients and presstre drop in the triangular flow channel over the range of interest for the spent f uel pool.

3. A verification of predicted flow rates produced by natural convection in a triangular flow channel.

11 cornidered necessary, NNECO will irstall a test rig in the spent f uel pool to verif y the pedicted heat generation rates in spent f uel l assemblies and to itrther determine whether a f uel assetnbly is ready for iuel consolidation. In this test rig, a f uct assembly is placed in a closed canister filled with pool coolant. Therinocouples are placedin the canister to menstre the coolant temperattre rise within the canister. The heat generation rate f or the f uel asseinbly is then calculated using the temperattre rise data. The assembly will be cornolidated if the heat generation determined frorn the thermocouple incastrevnents is lower than the minimten valte required Ior f uel corsolidation. The required value will be based upon analytical enetimh and the results of the trian,gtdar flow channel test.

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4. MECHANICAL CONSIDERATIONS r

4.1 Materials of Corstruction I i

The new spent f uel racks will be designed similar to C-E standard j Super Hi-CAP and MAX-CAP f uel storage racks which are f abricated from 304 stainless steel with a maximtm carbon content of 0.065  ;

percent. The racks are monolithic honeycomb structtres. Each j storage location is formed by welding stainless steel sectiors along  ;

the intersecting seams, permitting the assembled cavities to become [

the load bearing structure, as well as framing the storage cell ..

enclostres.

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Stainless steel bars, which are inserted horizontally tirough the rectangular slots in the lower reglon of the module, support the f uel' i assemblies. These support bars when welded in place support an entire row of f uel assemblies.  !

The relevant standarda are listed below: I r

American Society ior Tenting Materials Doctanents .

l A. ASTM - A240 - SpecliIcation ior Corrosion Resisting Chromium Nickel Steel Plate, Sheet & 5 trip for Fusion welded Unfired i l

Pressure Vessels.

B. ASTM - A479 - 5pecl!! cation for Stainless and Heat Resisting

! Bara and Shapes l Other codes and standards may be invoked as necessary by applicab%

C-E specifications, dawing procedtres, or other documents.

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-2 0-4.2 Seismic Methods Development

1. Seismic

a. Te;t Program in Support of Seismic Analysis of the Consolidated Spent Fuel Rack System CFC Lateral Load Deflection Tests A single Consolidated Fuel' Storage Box, filled with depleted UO2 fuel rods, will be supported in a vertical orientation at both ends. The end conditions will be of the pinned-pinned type. A horizontal load will be applied externally to the f uel canister at it= nidpoint and the static deflection behavior of the - structure will be monitored at up to 12 elevations by means of LVDT type.

displacement transducers. The hysteresis ef fects of the .

canister will be determined by a series of push-pull cycles with increasing force amplitudes. The test results will be condensed into representative stiffness properties and deflection curves. All the testing will be. performed in the elastic deflection range which will be monitored by a series of strain gauges attached to the f uel storage box.

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Consolidated Fuel Storage Box Forced Vibration Testing : .%

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The _ f uel ; stc, rage box, filled -lwith Lf . rods,' will' . b,e f ,

supported ' vertically at its ~upperQlocation.' Ti }g ,

attachment to a horizontal sliding fixture will be of the g , ,, x "pinnect' L type. A large water tank ~ enclosure will Ipg. .

3 placed arotmd the Fuel Storage Box, which' allows for both - 'jb

' air and still1 water testing. Closed-loop controlled

x sinusoidal ~ excitations 1will' be introduced: to the test-

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' specimen by a: hydraulic ~ actuator connected directlhto-

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6 J The f uel storage box behavior will be monitored along its

) length by a series of LVDT type displacement transducers and strain gauges. Initially, low level sine sweep tests will be conducted over a frequency range from 1 to 33 Hertz. Frequency response plots will be developed for all y monitoring locations and the test intensity will be h increased incrementally up to a maximum acceptable

, level. From the frequency response plots, the nattral frequencies and associated model deflection patterns,' as well as the model damping properties, will be determined.

2 The non-linear response characteristics will then be presented as a f unction of excitation level for use in s

j model correlation efforts. The hydrodynamic effects of a' water meditm on the dynamic behavior of the' f uel i

canister will be determined by comparison.of . air and .

., water environment test results.

4 Box Section Stiffness Properties 4

, During a seismic event, the fuel storage- box can impa'ct the storage rack walls. LIn the model, the stiffness .

. properties of the impacting" components have ~ to be

simulated by springs and gap-elements. These' spring .

properties are a combination of storage box and " stand-of f" flexibilities.

A series' of ' static load deflection tests .are proposed whereby the load is . applied in different axies .directly to ,

the ' stand-offs. The deflection . behavior 1 of . the box' I

4 section,' as well as the stand-offs, will be meastred by a j combination of displacement transducers. The load versus -

L deflection results will be reduced into stiffness properties

! representative of - the respective components. Testing

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again will be confined to the elastic range and be j monitored by strain gauge applications.

4 b) Consolidated Fuel Storage Box / Fuel Rack / Pool Modeling

and Interfaces l

Analytic methods and models will be developed for performing the seismic antaysis of the Consolidated Fuel Storage Racks. the effort will be focused on two arens of model development: 1) Formulation of an analytic model of the Consolidated Fuel Storage Box for use in a nonlinear time history seismic anlaysis; and -2) i development of a model of the spent f uel rack modules..

1 Modeling of the fuel storage box will be based upon full- .

scale dynamic testing. Because . the dynamic characteristics of the f uel storage box are unknown and not easily determined - by analytic means, - tests are ,

. performed to determine nattral frequency, damping and local canister stifinesses. Model simulations of the actual tests will. be performed to. develop a storage box model I

' consisting of a simple vertical' array of Itmped masses interconnected by massless springs. Separate models will be developed - to represent different degrces- ' of compaction and in cases where complete compaction 'is i

not achieved, the models will accotmt f or possible f uel rod .

. impacting. The hydrodynamic effects of fuel storage box natural frequency and damping determined f rom tests will also -be incorporated into the models. For less than complete compaction, the hydrodynamic effects'of fuel .

~ rods for 'various degrees of compaction'will be modeled. -

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Model development of the spent fuel rack module including effects of interaction with' the fuel storage box will also be performed. Because fuel consolidation leads to additional seismic ~ loadings on the coupled fuel rack / pool structures, new rack supporting schemes will be investigated. Since the method of fuel rack support cliects the dynamic characteristics of the modules and their . seismic response, - methods . will be - developed to include the effects of rack support in the dynamic models.

Previous seismic analysis experience with monolith type designs indicated that the gap' and amount of structural coupling between the rack and the fuel container has a significant effect on peak seismic loads. A model of the interface between the fuel storage box and fuel rack will be developed to minimize relative motion between the components and consequently, the seismic loads. Because submergence of the racks affects both the dynamic

~

characteristics of the storage box and fuel-rack and their motion . relative to the pool, methods for modeling this hydrodynamic action between the rack and pool will be-investigated.

Studies will be made to determine the sensitivity of seismic loading to both the rack design and the~ rack-to ~-

pool hydrodynamic representation. z The results of these '

studies will be used to develop fuel storage box and rack designs that are compatible,with the seismic loading associated with consolidated fuel.

b 4 -

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.1 4.3 Seismic Analysis

A site specific seismic analysis of Millstone Unit No. 2 spent fuel racks with consolidated fuel will be performed using the analytic methods and models described in Section 4.2.1. The seismic analysis will consist .of nonlinear- time . history analyses in the two horizontal 4 directions and a response spectrum analysis in the vertical direction.

The first step in the seismic analysis is the determination i of the dynamic characteristics of the basic module as supported in the proposed design. These characteristics are obtained from'a frequency analysis of a linear three-

dimensional model of. a module -using SAP 4, a general l . purpose finite element computer code. The results of the

, frequency analysis are then incorporated into a nonlinear j representation of the fuel rack structures based upon the i methods and models used or. developed in' this program.

l This nonlinear model includes a mathematical model of.

j the Consolidated Fuel Storage Box, based upon both the.

~

box testing and model development. Because of the close proximity of the fuel storage box and the~ rack structure i to the spent fuel pool walls, the hydrodynamic effects will .

i .be accentuated and are, therefore, considered explicitly -

in the nonlinear model. . The results of the engineering

! development effort addressing'the sensitivity of seismic i loadings .to these : hydrodynamic representations - will be used in formulating the hydrodynamic portion of the .

1-model. The nonlinear model will also consider friction forces between the Consolidated Fuel Storage Box and the

. rack module. :

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The program used for the nonlinear time history analysis

is called CESHOCK, which is a C-E proprietary computer code. CESHOCK generates equations of motion for the dynamic system and solves them by direct numerical integration to obtain the system response to an earthquake time history. The model used consists of mass nodes being interconnected not only by linear spring

elements but also by elements that duplicate the action of j friction, nonlinear springs, gaps between structural elements and hydrodynamics. A series of CESHOCK runs s are required to determine the effects of dynamic J

parameters on fuel storage box, rack and pool loadings.

l The seismic excitation to be used for the nonlinear time -

history analyses consists of the acceleration time history of the fuel pool floor derived from a seismic analysis of the auxiliary building which includes the effects of the added weight associated with consolidated fuel. The stress contributions from the seismic analyses of the two horizontal directions and the vertical direction are

combined in accordance with Regulatory Guide' l.92 and j the results are used to determine the structural adequacy of the proposed design.

4.4 Pool / Auxiliary Building Analysis The Millstone' Unit No. 2 fuel' consolidation program requires an analysis of the spent fuel pool and auxiliary building for the increased loads caused by the addition of additional spent fuel and ~ consolidated fuel. . This -

, document addresses the guidelines ' And acceptance..

f criteria that NNECO plans to follow to qualify the spent -

fuel pool and auxiliary building for the new loadings.

i

, l

i The Millstone Unit No. 2 auxiliary building is a multistory.

concrete struct tre. Spent fuel storage is provided between coltann lines H.4 and L.5 at elevation (-)2' -0".

The storage area consists of a reinforced concrete pool lined with a one-quarter inch thick stainless steel liner to elevation 38'-6". Normal water level is to elevation 36'-

6". A leak chase system consisting of channels embedded

behind the liner at all seams and connected to a collector system is used to monitor and control any possible leak-from the pool. Construction materials used in the corstruction of the pool include ASTM A-240, Type 304 stainless steel, ASTM A-615 Grade 60 deformed bar reinf orcing steel, and 3,000 psi 28-day strength concrete.

The spent fuel rack support system that presently exists in the spent fuel pool consists of several built-up plate

members framing into a set of tees that run in the north-south direction across the pool floor with their flange welded to bearing plates which are welded to the pool liner

plate. These built-up beams frame into the tees with a i

complicated hanger system that is designed to allow the beams to tmdergo free thermal expansion yet not impose any horizontal load on the support . tees. This is

. accomplished through a system of plates, connected by pins in such a manner that the ends of the beams hang freely from the tces. The vertical support 1oads Irom the present f uel racks are supported on this system of built-up members. The horizontal seismic load is carried to the -

spent fuel pool walls and resisted by a series of snubbers - .

at two elevations. The' present fuel rack system was shown to have been designed with no additional capacity to carry. higher loads such as loads associated with consolidated f uel ' Based on this information, the decision was made to remove the present rack support

system to accommodate the installation of f ree standing f uel racks.

The analysis of the spent fuel pool and associated components of the auxiliary building to accommodate the f

loadings associated with corsolidated fuel will be accomplished with the use of a large finite element model. The finite element model will include the entire i

spent fuel pool, fuel transfer canal, and cask laydown ar ea. The foundation of the pool will be included and terminated at points where it is considered the local ef fects are negligible with regard to the overall pool response. The effect of the auxiliary building floors and walls that frame into the pool will be included so that advantage can be taken of the stiffening action and load resisting capacity that this structural system provides.

I The spent fuel pool analysis will proceed with the formation of composite load cases. The composite loads I

consist of a combination of basic load cases which are grouped together for the purpose of application of - the Standard Review Plan load factors. The composite load cases include dead load, live load, operating thermal and accident thermal, and SSE and OBE earthquake loads.

Dead load will consist of a combination of the dead weight of concrete, hydrostatic presswe, and the weight of the f uel rack modules without their f uel components.

Live load consists entirely of the submerged weight of the consolidated fuel and its storage box. Normal operating thermal. loads place the pool water at 1500F with the temperature outside the pool at 550F. The 2120F pool / wall interface temperatwe for accident thermal was-deemed applicable when determining the gross structural effects on the pool walls. OBE earthquake loads involve four composite load cases. These load cases result from

\

l the specification in the Standard Review Plan that the three directions of earthquake m ust be applied simultaneously and permutation of signs must be included.

For this reason four load cases are specified with appropriate permutations of signs of the three orthogonal accelerations. An additional four load cases are developed by multiplying each of the originals by (-)l.0. Load combinations involving SSE earthquake loads utilize these OBE earthquake loads with a coef ficient applied.

The Standard Review Plan rpecifies service load and factored load combinations for Category 1 concrete structtres. Upon examination of these load cases, it can be shown that eight of the composite loads comprising them are not under consideration when analyzing the spent fuel pool. Other accicent loads such as flood loads, tornado loads, or any piping loads will be included, consistent with their definition and in the manner in which they affect the spent f uel pool structtre.

4 Upon examinations of the Standard Review Plan load combinations, it is readily apparent that some load cases are duplicates or envelop other load cases. These combinations will be reviewed and controlling loart combinations will be chosen from this group. Following the load combinations, the concrete sections will be checked against criteria set forth in the latest revision of the American Concrete Institute Code Requirements for Nuclear Saf ety Related Concrete Structures-ACI 349-80.

l Cask drop loads have been addressed in the Millstone Unit l No. 2 ' FSAR. Guidelines set forth for the areas of the pool where the cask may be safely handled will remain as stated.

v 0

4 The above guidelines and acceptance criteria will be used by NNECO. By following these guidelines and acceptance criteria,' NNECO plans to demonstrate the adequacy of j the Millstone Unit No. 2 spent fuel pool and auxiliary building to accommodate the loads associated with

, consolidated f uel.

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5. RADIOLOGICAL CONSIDERATIONS The rod consolidation project will be evaluated to ensure that- personnel expostres are limited in kee,ning with the principles of ALARA and in conformance with NRC Regulatory Guide 8.8. Special attention will be paid to the radiological considerations created by the fuel bundle crud that may be f reed in the pool and by the handling and packaging of waste bundle hardware.

The results of detailed radiological evaluations of both the spent fuel pool-l modifications and the storage of consolidated fuel in the pool will- be -

provided to the NRC.

l Calculations will be performed to determine the incremental' dose at l

certain areas-around the spent fuel pool due to the proposed increase _in spent f uel pool capacity. The dose will be calculated with the pool filled to j its proposed capacities with both unconsolidated and consolidated spent i fuel.  ;

Special attention will be given to the radiological considerations created by j the handling and packaging of consolidated spent fuel . assemblies and l associated waste hardware, to ensure that personnel exposures are in-i- keeping with the principles of ALARA and in conformance with NRC Regulatory Guide 8.8.

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6. ACCIDENT ANALYSES The following accidents will be analyzed and the results provided to the NRC for review:

o dropped f uel assembly in pool o dropped f uel cask or large object in pool o external phenomena o loss of spent f uel pool coolant For thermal-hydraulic analyses, normal operation is defined as any arrangement of the fuel assemblies in the rack for which the conditions of maximum pool bulk temperature of 1500F and minimum pool depth of 23 feet of water above the fuel are maintained. Accident conditions are defined by two concurrent events; namely a dropped f uel assembly and a loss of coolant event. A loss of coolant event assumes that coolant is evaporated because of loss of external cooling. The decay heat of the fuel is removed by boiling the pool coolant. The loss of coolant event is bounded by assuming a minimum pool depth of 10 feet water above the racks and a maximum fluid temperattre in the rack region of 2350F.

7.0 PLANT MODIFICATION 7.1 General Description of Procedure NNECO intends to begin reracking modifications at Millstone Unit No. 2 during 1985 with a spent fuel pool inventory of approximately 446 spent fuel assemblies. As stated previously, the existing pool configuration consists of ten spent fuel rack modules with a total capacity of 667 storage locations.

The rcracking modifications require that a series of activities be sequenced to permit removal of the existing fuel racks together with the installation of the proposed racks.

The spent fuel population will be concentrated into one section of the pool to vacate the maximura number of fuel rack modules.

Additionally, a temporary spent fuel rack will be placed in the cask laydown area for storage of fuel assemblies in order to facilitate the rerack installation. Engineering analyses will be performed to support the temporary placement of one module in the cask laydown area.

The reracking transition between removal of the old racks and installation of the new racks requires that several pool modifications be implemented. The existing fuel rack support steel on the floor of the spent fuel pool must be removed to permit installation of the new racks.

Additionally, NNECO will evaluate the need for temporary fuel pool modifications during transition periods to maintain reasonably achievable seismic integrity of the spent fuel pool configuration. -

1

O 4

j The pool layout portrayed on Figure 7.0 represents an arrangement whereby an optimum number of fuel racks can be vacated and j removed from the pool and a large area of the flooring steel exposed

for access, support, and removal.

i 7.2 Seismic Considerations The Millstone Unit No. 2 spent fuel pool design configuration consists l of ten spent fuel rack modules that are horizontally braced to the T

j pool walls and vertically supported by a built-up flooring system of structural steel beams (Figures 7.1 and 7.2). By design, the spent fuel  !

. e

', rack modules and the floor support system are intended to' function  !

i together as a composite unit.

Horizontal seismic loadings are transferred between rack modules in bearing to hydraulic " Lock-up" snubbers then to the pool walls. The hydraulle snubbers are mounted between the outer rack periphery and the pool walls to maintain continuous contact between the racks and

! the walls.

Vertical deadweight and seismic loadings are transferred through the j floor support structure to the pool floor in he'aring.

i The spent fuel storage racks, peripheral equipment, and all associated l structures are designed to the qualifications of a seismic Category. I structure.

The rerack transition from a braced pool design to a free-standing pool design requires the removal of the floor support structures to ,

I permit installation of the free-standing racks. Analyses conducted by NU Indicate that the structural design of the floor support system is

! not capable of seismic qualification under a higher vertical service

! loading criterion than presently specified by design. Since the floor '

support system is built up above the pool liner and transfers all

vertical loads to the pool floor in bearing, temporary supports will be installed between the structure and the pool floor to provide alternate load paths and maintain vertical stability.

NNECO will evaluate the need for temporary fuel pool modifications during the period when the primary horizontal load path of the racks is interrupted during the reracking transition.

7.3 Load Handling The reracking operation at Millstone Unit No. 2 will be conducted in accordance with strict procedures to prevent inadvertent dropping of heavy objects into the pool during the reracking operations. Strict -

procedural controls, technical specifications,' and mechanical controls prohibit the movement of heavy objects over spent fuel stored in the -

pool.

During the reracking no objects routinely supporting the activities in the spent fuel pool will have a handling path that brings the objects above or in the immediate vicinity of the stored spent fuel unless it is absolutely necessary. Additionally, during the reracking transition, no fuel racks old or new will have a handling path directly over stored spent fuel.

e I

4

Table 11 I

Proposed Thermal Hydraulic Design Bases Assemblies 4

Number Assemblies Normal Reload 1/3 core Minimum Decay Time One Assembly 3 days Minimum Decay Time Full Core 6 days Minimum Cooling Time of a Fuel Assembly To be determined

j. before the Rods are Consolidated Water Height Above Fuel 23 feet l Minimum Water Height Above Rack (Accident) 10 feet

Maximum Bulk Water Temperature (Normal) 1500F Maximum Water Temperature In the Rack Region (Accident) 2350F.

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LIST OF FOOTNOTES (1)

G. Lear letter to D. C. Switzer, date November 22, 1976, transmitting license amendment Nos. 39 and 30 to DPR-21 and DPR-65, respectively.

(2) CEPAK is a synthesis of the following computer codes:

FORM - A Fourier Transform Fast Spectrum Code for the IBM-709, McGoff, D. J., NAA-SR-Memor 5766 (September 1960)

THERMOS - A thermalization Transport Theory Code for Reactor Lattice Calculations, Honeck, H., BNL-5816 (July 1961)

CINDER - A One Point Depletion and Fission Product Program, England, T.

R., WAPD-TM-334 (Revised June 1964)

(3) DOT is a multigroup Two-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering R. G. Sottesy, R. K. Disney, A. Collier " User's Manual for the DOT-11W Discrete Ordinates Transport Computer Code", WANL-TME-1982, December 1969 (4) KENO is a multigroup Three-Dimensional Monte Carlo Code L. M. Petrie and N. F. Cross, " KENO IV An Improved Monte Carlo Criticality Program", ORNL-4938, November 1975

&