ML20141M134
| ML20141M134 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/27/1992 |
| From: | HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20141M131 | List: |
| References | |
| NUDOCS 9204010290 | |
| Download: ML20141M134 (352) | |
Text
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HOLTEC 1NTERNATIONAL SPENT FUEL POOL MODIFICATION for INCREASED STORAOE CAPACITY 4
SEQUOYAH U' NIT 1 DOCKET NO. 50-327 SEQUOYAH UN~IT 2 DOCKET NO. 50-328 i
Tennessee Valley Authority Chattanooga, Tennessee JER"1883RES8S8527 p PDR w
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TABLE OF CONTFErS SECTION PAGE l Le INTRODUCTION 11 2.0 MODULE DATA 21 2.1 Synopsis of New Modules 21 f 2.2 Mixed Zone Three Region Storage (MZTR) 22 l 25 23 Matedal Considerations 2.3.1 Introduction 25 1 23.7 Stnictural Matedals 2-5 2.33 Poison Material 25-
, 2.3.4 Com 3atibility with Coolant 28 2.4 Existing Rack Moc ules and Proposed 28 <
Reracking Operation 3.0 CONSTRUCTION OF RACK MODULES !
3.1 Fabrication Objective 31 3.2 Mixed Zone Three Region Storage (MZTR) 32 :
33 Anatomy of Rack Modules 32 3.4 Welding Types and Processes- 3-4 i 3.5 Codes, Standards and Practices for the 6 i
Sequoyah Spent Fuel Pool Racks 3.6 Materials of Construction 3 10 4.0 CRITICALITY SAFEW ANALYSES 4.1 Design Bases .
4-1 4.2 Summary of Criticality Analyses 4-3 4.2.1 Nortnal Operating Conditions 4-3 -
4.2.2 - Abnormal and Accident Conditions 4-7.
43 Reference Fuel Storage Cells 4-8
-43.1 Reference Fuel Assembly- 4-8 .
43.2 - High Density Fuel Storage Cells -.4-8 4.4 Analytical Methodology .
4-9 4.4.1 Reference Design Calculations 4-9 l 4.4.2-- Fuel Burnup Calculations and Uncertainties 4-11 4.4.3 Effect .of Axial _ Burnup Distribution 4 4.5 - Criticality Analyses and Tolerances - 4 13 4.5.1 Nominal Design . .
4 13 4.5.2 Uncertainties Due to Manufacturing Tolerances '4-14 1
4.5.2.1 Boron Loading Tolerances 14 l 4.5.2.2 Boral Width Tolerance 4 14 l 4.5.23 Tolerances in Cell 1.attice Spacing - 4-15 4.5.2.4 Stainless Steel Thickness Tolerances 4-15 4.5.2.5 Fuel Enrichment and Density Tolerances - .4 15 g' '
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OF CONTENTS
..catinued) 4.5.3 Water Gap Spacing Between Modules 4-16 '
4.5.4 Eccentric Fuel Positioning 4 16 1 4.6 Abnormal and Accident Conditions 4-16 4.6.1 Ten;perature and Water Density Effects 4 16 4.6.2 Dropped Fuel Assembly 4 17 4.6.3 Lateral Rack Movement 4 17 :
4.6.4 Abnormal Location of a Fuel Assembly 4-17 4.7 Existing Spent Fuel 4 18 4.8 References 4-19 i Appendix A to Section 4 A1 5.0 TIIERMAL.IIYDRAULIC CONSIDERATIONS f 5.1 Introduction 51 5.2 Spent Fuel Pit Cooling System 52 5.2.1 System Description 53 5.2.2 Component Descriptica 5-4 5.3 Decay _ Heat Load Calculations 57 5.4 Discharge Scenarios 57 ;
5.5 Bulk Pool Temperature _ 5-8 5.6 Local Pool Water and Fuel Cladding Temperatures 5 13 5.6.1 Basis 5 13 5.6.2 Model Description 5-14 5.6.3 Cladding Temperature 5 15 5.6.4 Block:d Cell Analysis 5 18 5.7 References 5-18 6.0 STRUCTURAI/ SEISMIC CONSIDERATIONS 6.1 Introduction 61 6.2 Analysis Outline 6-1 6.3 Artificial Time Histories 66 ,
6.4 Rack Modeling for Dynamic Simulations 69 6.4.1 General Remarks 69 ,
6.4.2 The 3 D 22 DOF Model for Single Rack Module 6 11 6.4.2.1 Assumptions 6 11 6.4.2.2 Model Details _ _ _ - 6 13 6.4.2.3 Fluid Coupling Details - 6 14 6.4.2.4 Stiffness Element Details 6 15 6.4.3 Whole Pool Multi. Rack (WPMR) Model 6-16 6.4.3.1 General Remarks 6 16
- 6.4.3.2 Whole Pool Fluid Coupling - 6 17 6.4.3.3 Coefficients of Friction 6 17 4
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TABLE OF CONTENTS (continued) 6.4.3.4 Modeling Details G18 6.5 Acceptance Criteria, Stress Lhnits, and Material Properties 6 19 6.5.1 Acceptance Criteria G19 6.5.2 Stress Limits for Various Conditions 6-21 6.5.2.1 Normal and Upset Conditions 6 21 (Level A or Level B) 6.5.2.2 Level D Service Limits G23 6.5.2.3 Dimensionless Stress Factors 6-23 6.5.3 Materia! Properties 6 24 6.6 Governing Equations of Motion 6 24 6.7 Results of 3 D Nonlinear Analyses of Single Racks 6-26 6.7.1 Racks in the Fuel Pool 6-26 6.7.1.1 Impar.t Analyses 6 28 6.7.1.2 Weld Stresses 6-28 6.7.2 Racks in the Cask Pit Area 6 29 6.8 Results from Whole Pool Multi Rack Analyses 6 30 6.9 Bearing Pad Analysis 6 31 6.10 References 6 32 7.0 ACCIDENT ANALYSIS AND MISCELLANEOUS STRUCTURAL EVALUATIONS 7.1 Introduction 7-1 7.2 Refueling Accidents 71 7.2.1 Dropped Fuel Assembly 72 7.3 Local Buckhng of Fuel Cell Walls 73 7.4 Analysis of the Impact Shield for Cask Pit 74 7.5 Analysis of Welded Joints in Rack due to Isolated Hot Cell 7-5 7.6 References 7-6 l
8.0 FUEL POOL STRUCIVRE INTEGRITY CONSIDERATIONS 8.1 Introduction 8-1 8.2 General Features of the Model 83 8.3 Loading Conditions 8-6 8.4 Results of Analyses 8-8 8.5 Pool Liner 89 8.6 Conclusions 89 8.7 References S.10 111
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l TABLE OF CONTENTS (continued) 9.0 RADIOLOGICAL EVALUATION 9.1 Fuel Hand!!ng Accident 9-1 9.1.1 Assumptions and Source Term Calculations 91 9.1.2 Results 9-4 9.2 Solid Radwaste 95 9.3 Gaseous Releases 95 9.4 Personnel Exposures 95 9.5 Anticipated Exposure During Re Racking 96 10.0 DORAL SURVEILLANCE 10.1 Purpose 10-1 10.2 Coupon Surveillance Equipment 10 2 10.2.1 Coupon Description 10-2 10.2.2 Surveillance Coupon Location 10-3 10.2.3 Measurements 10 3 10.2.4 Measurement Acceptance Criteria 10 4 10.3 References 10 5 -
11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 11.1 Introduction 11 1 11.2 Imperative for Increasing Spent Fuel Storage 11 1 11.3 Appraisal of Alternative Options 11 2 11.4 Project Cost Estimate 11 3 11.5 Resource Commitment 11 3 11.6 Emironmental Considerations 11-4 11.7 References 11 5 l
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[ 1.0 INTRODf)CTION gy.s p[ The Sequoyah Nuclear Plant (SQNP or Sequoyah) is a twin unit pressurized water nuclear power reactor installation owned and operated by the Tennessee Valley Authority. Sequoysh is located in the township of Soddy-Daisy, which is approximately 18 miles i northeast of the city of Chattanooga, Tennessee. Sequoyah received A its construction permit from the NRC (farmerly AEC) in May, 1970, and its low power License in February, 1980 for Unit 1 and June, 1981 for Unit 2. The two reactors went into commercial operati.on in July, 1981 (Unit 1) and June, 1982 (Unit 2), respectively. The Sequoyah fuel storage system is made up of a fuel pool 474 inches
) l ate and 380.5 inches wide with a separate cask laydown area. The paol yasently contains 1386 spent fuel storage locations. Prior to the October, 199, refueling, a total of 589 fuel bundles were s ured in the Sequoyah pool. Since the full core has 193 fuel assemblies for each of the two Sequoyah reactors, maintaining full core offload capability from one reactor implies that 1193 storage cel's (1386 minus 193) be available for normal offload storage. Table 1.1.1 provides the data on previous and projected fuel assembly discharges in the Sequoyah spent fuel pool. Table 1.1.2, constructed from Table 1.1.1 data, indicates that Sequoyah will lose full core discharge capability (for one reactor) in 1996. This prcjected loss of full core discharge capability promptes the present undertaking to increase spent fuel storage capability in the Sequoyah pool.
The purpose of this submittal is to request the _ euthori=ation to rerack the Sequoyah pool and equip it with new poisoned high density storage racks containing 2091 storage cells.
The owner and operator of Sequoyan, the Tennessee Valley Authority, entered into a contract with Holtec International of Cherry Hill, New Jersey, in March, 1991 to design, procure material, fabricate and deliver high density racks to the Sequoyah
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site. This licensing document has been prepared by TVA and its contractor, Holtec International.
Twelve free-standing poisoned rack modules positioned in the spent fuel pool with a prescribed and geometrically controlled gap
-between them will contain a total of 2091 storage cells. The design and construction of the storage cells is described in Section 3 of this document. As stated in Section 3, the design and construction of all cells are identiecl, although their physical location in a rack gives them some special attributes.
Those storage cells which are located on the periphery of a rack module are referred to as flux-trap cells *, and the interior ones are of the so-called non-flux trap type. A great majority of the
" flux trap" cells and scme non-flux trap cells ~are cuitable for storing fresh fuel (up to 5% enrichment) as depicted in Section 4.0, Figures 4.2.1 and 4.2.2. These fresh fuel tells are surrounded by other non-flux trap cells which have a burnup restriction on the fuel which they can store. Consistent with the concept of multl region storage, the placement of fuel with a given burnup ?.n the allowable location is administrative 1y controlled. No credit is taken for soluble boron in normal refueling and full core offload storage conditions.
In- addition to the twelve modules u the spent fuel pool,.TVA plans to install one 15x15 module (225 cells) in the cask region,-
as further described in Section 2 of this report. This rack will be identical in construction to the aforementioned spent fuel pool racks, except that it will be positioned on pedestal " tables" (in contrast to " bearing pads" for the spent fuel racks) so as to pennit the use of the standard fuel handling tools. This 15x15 A flux trap construction implies that there is a water gap between adjacent storage cells such that the neutrons emanating from a fuel assembly are therralized before reaching an adjacent Diel assembly.
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module, henceforth referred to as the " cask pit rack", will not be installed in the proposed 1994 rerack campaign, but would rather '
be utilized at a ltter date when the increase in the fuel inventory in the pool warrants its deployment.
, It is noted that the proposed raracking eff.rt in 1994 will -
increase the number of licensed storage locations to 2091 and, as indicated in Table 1.1.2, will extend the date of loss of full core discharge capability (193 bundles) through the year 2003 and possibly 2004 because the batch size projections in Table 1.1. 2 were estimated conservatively (80 bundles 0 18 month cycles).
Table 1.1.3 presents key comparison data for existing and proposed maximum density rack modules for Sequoyah. Finally, the cask pit rack with 225 storage cells would further extend Sequoyah's in-pool storage life by approximately 3 cycles (about 2 years). Thus, the proposed capacity expansion would extend the loss-of-full core offload capacity to the yesr 2005/2006.
The new spent fuel storage racks are free-standing and self nupporting. The principal construction materials for the new racks are SA240-Type 304L stainless steel sheet and olate stock,
- and SA564-630 (precipitation hardoned stainless stau.) for the adjustable support spindles. The only non-stainless material s utilized in the rack is the neutron absorber material which is
! boron carbide and alun,inum-composite sandwich available under the patented produci name "Boral".
The new racks are designed and analyzed in accordance with Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B&PV) Code. The mat (rial procurement, analysis, and fabrication of the rack modules conform to 10CFR 50 Appendix-B [
requirements.
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e This Licensing Report documents the design and analyses performed to demonstrate that the new spent fuel racks satisfy all governing requirements of the applical:1c codes and standards, in particular, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", USNRC (1978) and 1979 Addendum thereto.
The safety assessment of the proposed rack modules involved demonstration of their thermal-hydraulic, criticality and structural adequacy. Hydrothermal adequacy requires that fuel cladding will not fail due to excessive thermal stress, and that the steady state pool bulk temperature will remain within the limits prescribed for the spent fuel pool to satisfy the pool structural strength constraints. Demonstration of structural adequacy primarily involves analyses showing that the free-standing rack modules will not impact with each other in the cellular region or with the pool v:lls under the postulated Design Basis Earthquake (DBE) and Operating Basis Earthquake (OBE) events, and that the primary stresses in the rack module structure will remain below the ASME BkPV Code (subsection NF) allowables.
1 The structural qualification also includes analytical demonstration that the subcriticality of the stored fuel will be maintained under accident scenarios such as fuel assembly drop, accidental risplacement of a fuel assembly outJide a rack, drop of a gate, etc. The structural consequences of these postulated accidents are evaluated and presented in Section 7 of chio report.
The criticality safety . analysis shows that the neutron multiplication factor for a stored fuel array is bounded by the USNRC limit of 0.95 (OT Position Paper) under assumptions of 954 probability and 954 confidence. Consequences of the inadvertent placement of a fuel assembly are also evaluated as part of the criticality analysis. The criticality safety analysis synopaized in Section 4 sets the requirements on the length of the Boral panel ad the areal B-10 density, s
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This Licensing Report contains documentation of the analyses performed to demonstrate the large margins of safety with respect to all USNRC specified criteria. This report also contains the results of the analysis performed to demonstrate the integrity of the fuel pool reinforced concrete structure, and an appraisal of radiological considerations. A nummary of the cost / benefit consideration demonstratinr; reeacking as the most cost effective approach to increase the on-site storage capacity of the Sequoyah Nuclear Plant is also included in this report.
All computer programs utilized in performing the analyses documented in this licensing repcrt are ide:itified in the appropriate sections. All computer codes at. e benchmarked and verified in accordance with Holtec International's nuclear Quality Program.
The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety from all four vantage points: thermel-hydraulic, criticality, 77ructural and radiological. The No-Significant-Hazard Considerations submitted to the Commission elong with this Licensing Report is based on the descriptions and analyses synopsized in the subsequent sections of this report.
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Table 1.1.1 PAST AND PROJECTED FUEL DISC'IARGE SCHEDULE ' (ca. 1991)
CUMULATIVE-D DRGE.Noa MI HEE IN THE POOL PAIg-1 1 63 68 9/82 2 2 68 136 . Qg3 3 1 72 208 2/84-4- 2 68 276 g/s4 5 1 72 343 g/g5 6 2 -80 423 1/85
- 7 1 81 %C9 -3/90 8 2 80 589 3/30 9 1 72 661 10/91 10 2 76 737 3/92 11 1 30 gg7 47,3 12 2 30 gg7 -9f93 13 1 30 977 10/94 14- 2 80 1057 3/95 15- 1 80 1137 4/96-16 2 30 1217 3/g6 17 1 80 1297
_ 10/37 18 2 30 1377 379, 19 1 30 1457 4/g3 20 2 30 - 1537 g/93 21 1 30 . 1617 10/00 22 2 30' is37 3 fog 23 1 80 1777 4/02-24 2'- 80 - 1857-- 9/02 25 1 30 1937 10/03 26 2 30 2017- 3/04 6 ,
Table 1.1.2*
PROJECTED AVAILABLE STORAGE IN THE SEQUOYAH-POOL (ca. 1991) wits PRESENT DISCHARGE MONTH / LICENSED CAPACITY AFTER RERACRING EUMBER M f 1346 crils) J10,91 CELLS 1 p 8 9/90 797 (Racks m,n't be 9 10/91 '725 installed until 10 3/92 649 June 1994) 11 4/93 569 12 9/93 489 13 10/94 409 1114 14 3/95- 329 1034 15 4/96 249 954 16 9/96 169" 874 17 10/97 --- 704 18 3/98 ---- 714 19 4/99 ---- 634 20 9/99 --- 554 21 10/Of: -- 474.
22 3/01 --- 394 23 4/02 - - - 314 24 9/02 -- '234 23 10/03 ---- 154**
26 3/04 ----
This table does.not include the. additional storage capacity _
provided by the 15x15 cask pit inodule.
Loss'of reserve (one reactor core) 3
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Table 1.1.3 ,
RACK HODULE DATA, EXISTING AND PROPOSED RACKS ITEh EXISTIbiG _ RACKS EHOPOSED RACKS Number of cells 1386 2091 Number of modules 24 12 Neutron Absorber Boral Boral (Nom.) cell pitch, inch 10.375 8.972 (Nom.) cell opening size, inch 8.75 8.75 s
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2.0 MODULE DATA 2.1 Synoosis of New Modules The Sequoyah spent fuel storage pool consists of a 380.5" x 474 (nom.) rectangular pit and a separate 380.5" x 144" (nom.) pit next to the fuel pool pit for cask handling operations. The pool is connected to the cask handling area and the fuel transfer canal through weir gates on the intermediate wall between the two pits and the west wall. Figure 2.1.1 shows a planar section of the Sequoyah spent fuel pool region.
At the present time, the Sequoyah pool contains medium density racks with a 10.375" nominal assembly center-to-center pitch.
There is a total of 1386 storage cells in the pool. There are eighteen 7x8 nodules (56 cells each) and six 7x9 modules (63 cells each).
Figure. 2.1.2 shows the module layout for the Sequoyah pool after the proposed reracking campaign assuming nominal pool dimensions.
The actual layout, however, is predicated, on the miniuum envelope dimension (374" x 468") of the spent fuel pool as determined by a prismatic envelope - survey of the pool in the summer of 1991.
Therefore, the accual rack-to-wall gap at certain locations in the pool will be smaller than the nominal gaps indicated in Figure 2.1.2. The minimum gaps, however, were utilized in the dynamic impact evaluation described in Section 6 of this report. As shown in Figure 2.1.2 and compiled in Table 2.1.1, there are twelve racks containing a total of 2091 storage cells with a 8.972" nominal pitch in the rcracked configuration in'the Sequoyah pool.
The essential cell data for all storage cells is given in Table 2.1.2. The physical size and weight data on the modules may be found in Table 2.1.3. Not included in Tables 2.1.1 and 2 .1'. 3 is:
the 15x15 module which will be installed in the cask pit at a 2-1 l
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later date. If this rack is included in the stcrage capacity, then the present reracking application-will increase the licensed storage capacity at Sequoyah from 1386 to 2316 cells.
2.2 Mixed Zone Three Recion Storace (MZTR):
The high density spent fuel storage racks in tne sequoyah pool and cask pit will provide storage locations for up to 2316 . fuel assemblies and will be designed to maintain the -stored fuel, having an initial enrichment of up to 5 wt% U-235, in a cafe, coolable, and sub-critical configuration during normal discharge and full core offlood storages and postulated accident conditions.
All rack modules for Sequoyah spent fuel pool are of the so-called " free-standing" type inasmuch as the modules are not i attached to the pool floor and they do not require any lateral braces or restraints. Thebe rack modules will be placed in the pool in their designated locations using a specifically designed lifting device, and the support legs will be remotely leveled using a telescopic removable handling tool. 'The leveling operations will be done with the support legs lifted off the floor. Except for the cra'3, no additional lifting equipment will be needed while leveling the rack modules.
As described in detail in Section 3, all modules in the Sequoyah pool an of "non-flux trap" construction. The baseplates on all rack nodules extend out beyond the rack module wall such that the contiguous edges of the plates act to set a geometric separation between the facing cells in the modules. The caseplate projection in the north-south direction is 1" which establishes a 2" (min.)
separation between the modules in the north-south direction.
Providing for a total non-straightness in the facing baseplate edges of 1/8 inches, the minimum separation of 2- inches 'is consistent with 2.125" (nominal) N-S spacing indicated in Figure 2.1.2 for the emplacement of the racks in the pool. Similarly, to 2-2
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ensure a nomipsi module-to-module gap of 1.5 inch, the baseplates are fabricated to project 11/16" in the east-west direction. The baseplates on the module sides facing the pool walls, however, are trimmed off to within 1/4" of the rack wall.
'The geometric separation between the modules created by the baseplate serves to establish a " flux trap" space between the adjacent modules. In othar words, although there is a single paneA of the neutron absorber between two fuel assemblies stored in the same rack, there are two poison panels with a specified water flux trap between them for fuel assemblies located in cells in two facing modules. Out of these flux trap locations and peripheral cell locations (cells adjacent to pool walls), a certain number of storage cells are designated for storing fresh fuel. In this i manner, a sufficient number of locations without any burnup restriction (Regica 1 c611s) arc identified to enable unrestricted full core offload of the Sequoyah reactor in the spent fuel pool.
These so-called Region 1 cells are identified in Figures 4.2.1 and 4.2.2. of this report. The remaining storage cells have errichment/burnup restrictions. Appropriata restrictions on the enrichment /bornup of uhe stored fuel in - Region 2 and Region 3 cells are presented in Section 4.
Most rack modules are supported by four legs which are remotely adjustable. Racks with adjustable podestals can easily be-made co-planar with each other. The rack module pedestals are engineered to accommodate variations in the flatness of the pool floor. To avoid interference with wall mountings, certain racks have pedestals of fixed height so as to reduce the overall height of the rack. The support legs also provide an under rack plenum for natural circulation of water through the storage cellc. The placement of the rack pedestals in the spent. fuel pool has been designed to preclude any support legs from being located over existing obstructions on the pool floor.
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The Sequoyah racks are riubjected to designated seismic loadings. .
A set of three two-dimensional seismic rouponse spectra are provided for the Design Basis Earthquake (DBE) . The Operating Basis Earthquuke (OBE) spectra are specified as the DBE spectra multiplied by a factor of 0.5. Aa described in Section 6 of this
'rroort, synthetic time-histories in three orthogonal directions ,
were generated from the design response spectra. In addition, time-histories were also generated for the non -mandatory site specific spectra. Tho; new Sequoyah racks were quali"ied for both the FSAR commitment spectra (DBE and OBE) and the optional nite specific spectra.
Under the seismic events, the rack modules have four designated locations of potential impact:
(i) Support leg to hearing pad (ii) Storage cell to fuel assembly contact surfaces '
(iii) Baseplate edges (iv) Rack top corners The support leg to pool slab bearing pad impact would occur whenever the rack support foot lifts off the pool floor during a seismic ravent . The " rattling" of the fuel assemblies in the storage cell is a natural phenomenon associated with seismic
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conditions. The baseplate and rack top corners impacts would occur if the rack modules' tend to slide or tilt towards each other d'Aring the postulated DBE or OBE seismic events. Section 6 of this report presents the analysis methodology and lesults for all four locations of impact, and establishes the structural- integrity of the racks under the postulated load combinations.
A bearing pad, made of austenitic stainless steel, is interposed between the rack support foot and the pool liner"such that the loads transmitted to the slab by the. rack module under steady state as well as seismic conditions are diffused into the pool slab, and alloubtle local concrete surfcce pressures are not 2-4 w_ _ _ _ _ - - . _ _ _ _ _ _ _
exceeded. The cask pit rack, due to a slightly lower elevation of the cask pit liner (Figure 8.1.1) requires a support' leg " box" in lieu of a bearing pad. Section 8 of this report presents the details of pool structure analyses performed in support of this e
licensing application.
2.3 M erial Considerations 2.3.1 Introductign s
safe storage of nuclear fuel in the Sequoyah spent fuel pool req tires that the materials utilized in the fabrication of rncks be of proven durability and be compatible with the pool water environment. This section provi. des the necessary information on this subject.
2.3.2 Structural Materinig The following structural materials are utilized in the fabrication of the spent fuel racks: *
- a. ASME SA240-304L for all sheet metal stock.
- b. Internally threaded support legs: ASME SA240-304L.
- c. Externally threaded support spindle: ASME ~ SA564-630 precipitation hardened stainless steel.
d.- Weld material -
per the - following - ASME ~ specification:
SFA 5.9 ER308L.
2.3.3 Poison IQ1;9 rial In addition to the structural and non-structural stainless material, the racks employ Boral, a patented product of_AAR Brooks
& Perkins, as the - thermal neutron absorber material.- A brief description-of Boral, and its fuel pool experience list follows.
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Boral is a thermal neutron aosorbing material consisting of finely divided particles of boron carbide (B4 C) uniformly distributed in type 1100 aluminum and pressed and sintered in a hot rolling process. Boron carbide is a compound having a high boron content in a physically stable e,nd chemically inert form. The 1100 alloy
'alumintim is a light-weigb; metal with high tensile strength which is protected from corrosion by a highly resistant oxide film. The two materials, boron carbide and aluminum, are chemically compatible and ideally suited for long-term use in the radiation, thermal and chemical environment of a spent fuel pool.
The selection of Boral for use in the spent fuel pool as the neutron absorbing material can be attributed to the following reasons:
(i) The contant and placement of boron carbido provid.es a very high removal cross section for thermal neutrons.
J (ii) Baron carbide, in the form of fine particles, it 4
homogenously dispersed throughout the central layer of the Boral.
(iii) The boron carbido and alumintkm aterials in Boral do not degrade as a result of le ig-term exposure to gamma radiation.
(iv) The thermal neutron absorbing central layer of 3 oral is clad with permanently bonded surfaces of aluminum.
(v) Boral is stable, strong, durable, and corrosion resistant.
Boral has garnered an excellent record of application in light water reactor fuel pools. -In fact, the existing racks in the Sequoyah pool also rely on Boral for neutron attenuation.
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Boral is manufactured by AAR Brooks & Perkins under the control and surveillance of a computer-aided Quality Assurance / Quality Control Program that conforms to the requirements of 10CFR50 Appendix B, " Quality Assurance Criteria for Nuclear Power Plants t.nd Fuel Reprocessing Plants". As indicated in Table 2.3.1, Boral
'has been licensed by the USNRC for use in numerous BWR and PWR spent fuel storage racks and has also been extensively used in overseas nuclear instalAations.
Boral Mater _ial.ChqrActeristiqa
, Aluminum: Aluminum is a silvery-white, ductile metallic element that is abundant in the earth's crust. The 1100 alloy aluminum is used extensively in heat exchangers, pressure and storage tanks, chenical equipment, reflectors and sheet metal work.
It has high resistance to corrosion in industrial and mari'ne environments. Aluminum has atomic number of 13, atomic weight of 26.98, specific gravity of 2.69 and valence of 3. The physical /
mechanical properties and chemical composition of the 1100 alloy aluminum are listed in Tables 2.3.2 and 2.3.3.
The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion and the film remains stable between a pH range of 4.5 to 8.5. (See Brooks and Perkins Technical Bulletin #624, Livonia, Michigan.)
Boron Carbide: The boron carbide contained in Boral is a fine granulated powder that conforms to ASTM C-750-80 nuclear grade Type III. The particles range in size between 60 and 200 mesh and the material conforms to the chemical composition and properties '
listed in Table 2.3.4.
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2.3.4 Comnatibi,lity with Coolant All materials used in the construction of the Sequoyah rackn have an established history of in-pool usage. Their physical, chemical and radiological compatibility with the pool environment is an
' established fact at this time. As noted in Table 2.3.1, Boral has been used ir both vented and unvented configurations in fuel pools with equal success. Consistent with the recent practice, the Sequoyah rack construction allows full venting of the Boral space.
Austenitic stainless steel (304L) is widely used in nuclear power plants.
2.4 Existina Rack Modules and Prcoosed Rerackinq Operation The Sequoyah fuel pool currently has medium density rack modules containing a total of 1386 storage cells in twenty-four module's.
At the time of the proposed raracking operation, approximately 65%
of the 1386 locations will be occupied with spent fuel. By temporarily s oring a numbec of fuel assemblies in the cask pit as described later, there is a sufficient number of open (unoccupied) cells in the pool to permit relocation of all fuel such that the existing modules can be emptied and removed from the pool, and new modules installed in a programmed manner.
- Remotely engagable lift rigs, designed to meet the criteria of NUREG-0612 " Control of Heavy Loads of Nuclear Power Plants", will be used to lift the new and old rack modules. Auxiliary Building cranes will be used for this purpose. A module change-out plan and procedure will be developed which ensures that all modules g being handled are empty wnen the module is moving at a height which is more than 12" above the pool floor.
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The Auxiliary Building has one overhead crane which rides on rails that traverse the entire fuel handling area of the building.
The crane has a main hook ratad at 40 tons- (designed at 125 tons) . In addition there is an auxiliary hoist on the overhead crane rated at 10 tons.
Pursuant to the defense-in-depth approach of NUREG-0612, the following additional measures of safety will be undertaken for the rere.cking operation.
(i) The crane and _ hoist will be given a preventive maintenance checkup and inspection within-3 months of the beginning of the reracking operation.
(ii) The crane hook will be used to lift-no more than 20% of its rated capacity of 80 tons at any time during the reracking operation. (The maximum weight of any module and its associated handling tool is less than 15 tons).
(iii) The old fuel racks will be lifted no more'than 6" e above the pool floor and held in-that elevation for approximately 10 minutes before beginning the vertical lift.
(iv) The rate of vertical lift will not exceed 6' per minute.
(v) The rate of horizontal movement'will not exceed 6' per minute.
(vi) Preliminary safe _ load paths have ' been _ developed.
i The "old" or "new" racks will not be carried over any region of the pool containing fuel.
(vii) The rack upending or laying . down will be carried out in an area which does not encroach on any' space -
ascribed to safety related equipment.
(viii) All crew members '-involved in the- reracking operation will'be given training.in the use of the-
' lifting and upending equipment._-The -: training seminar will utilize videotapes of the actual lifting and upending rigs-on-the actual modules;to
-be installed in the pool. Every crew member will be required to: pass _a written examination in the use of lif ting and upending apparatus _ administered by i the rack designer.
a 2 -3
(ix) In addition to the video and in-class training, the-rack installation crew will be given " hands-on" rack handling experience prior to executing any handling operation over the fuel pool. The unloading, rigging, upending, and staging of the racks, upon their arrival at the sequoyah site, will be carried out by the installation crew. As a
, result, the crew members will acquire considerable handling " feel" of the racks before bringing the hardware to the refueling floor level of the Auxiliary Building.
(x) It is noted that the fuel handling bridge crane cannot access some storage cells due to obstructions. Therefore, it will be necessary to load the inaccessible cells with fuel when the rack is staged a certain distance (approximately 20")
from the pool wall. Having loaded these cells, the module will be lifted approximately 4" above the pool liner, and laterally transported to its final designac% location. A fue) shuffling and rack installw.lon sequence will be developed to ensure 5 that all heavy load handling criteria of NUREG-0612 are satisfied. The rack handling rig is designed with consideration of the rack module weight alo'ng with the contained fuel assembly mass.
The fuel racks will be brought directly into the Auxiliary Building through the access door which is at ground level. This direct access to the building greatly facilitates the rack removal and installation effort.
A preliminary fuel reshuffle scheme f or - the -- spent fuel pool has been develope 6 by TVA which is predicated on the' followinc' criteria:
(1) No heavy load (rack or rig) with a potential to drop on a rack has less than 3 feet lateral free zone clearance from active fuel.
(2) All heavy loads are lifted in Luch a manner that the C.G. of the lift point is aligned with the C.G. of the load being lifted.
(3) Turnbuckles are utilized to " fine tune" the verticality of the rack being lifted.
2-10
All phases of the rcracking activity will be conducted in accordance with written procedures which will be reviewed and approved.
. Compliance with the objectives of NUREG-0612 will follow the guidelines contained in section 5 of that document. The guidelines of NUREG-0612 call for measures to " provide an adequate defense-in-depth for handling of heavy loads near spent fuel. . .".
The NUREG-0612 guidelines cite four major causes of load handling accidents, namely
- 1. operator errors
- 11. rigging failure 111. lack of adequate inspection iv. inadequate procedures The sequoyah rerack program will ensure maximum amphasis- to mitigate the potential lead drop accidents by implementing measures to eliminate shortec iings in all aspects of the operation including the four aforemet;cioned areas. A summary of the measures specifically planned to deal with the major causes is '
provided below.
Operator errors: As mentioned above, TVA plans to provide comprehensive training to the installation ctew using videotapes of the actual lift rigs built for the rerack project.
Bj.caino failure: The lifting devices designed for handling and installation of the old and new racks in the Sequoyah fuel pool will have redundancies in the lift legs, and life eyes such that-there are four independent load patns. Failure of any one load bearing - member would not lead to uncontrolled lowering of the load. The rigs will comply with all provisions of ANSI 14.6-1978, including compliance-with e.he primary stress criteria, load testing at 150% o' maximum lift load, and dye examination of critical welds.
The proposed rig for installing the new Sequoyah racks is similar to the rigs used in the rerack of numerous other plants, such as Hope Creek, Millstone Unit 1, Indian Point Unit Two, - Ulchin II, and Laguna Verde.
2-11
__ _-______A
4 Lack of adeauste inspection: The designer of the racks wi)1 develop a set of inspection poir.ts which have proven to eliminate any incidence of rework or erroneous installation in numerous prior rerack projects.
Inadeaut e.e nrocedures: TN A plans approximately twenty operating procedures to cover the entire gamut of operations pertaining to
.the rerack i,ffort,, such as mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. Distinct procedures for old racks and new racks will be devoloped, e
The series of operating procedures planned for Sequoyah rerack are the successor of the procedures implemented successfully in other project.s in the past.
A rack change-out plan consistent with the foregoing commitments is illustrated in Figures 2.4.1 -
2.4.15. The sequential steps to accomplish the reracking are also noted in those figures. It is noted that this proposed sequence of operations presumes that one new rack module (Module A-1 in Figure 2.1.2) will be temporarily placed in the cask pit and preloaded with approximately 150 bundles before ' the beginning of rack change-out. This preparatory operation reduces the amount of fuel shuffles required and increases the " width" of permissible safe load paths.
The reracking operation involves removal of the existing racks and the underlying structural grid, referred to as the " Floor Support Grid" (FSG), as well as removal and modification of other obstructions such as the coolant diffuser /sparger piping at the North end of the storage pool. Th=2 FSG structure will be removed in discrete segments by saw cutting the grid at designated locations. The sequence of operations illust?ated in Figures 2.4.1 - 2.4.15 is intended to demonstrate the feasibility of the proposed 1994 rack change-out in the Sequoyah pool which. complies with all handling commitments outlined in the foregoing. Further optimization of - the rerack sequence which will reduce the overall number of rack handling and fuel handling operations are being considered. The final operational sequence may be a slightly optimized version of the one presented herein.
2-12 i I
. . . . .. . __ a
Pursuant to the defense-in-depth concept, TVA has designed an impact shield which will be placed on top of the deep cask pit to protect the temporarily stored fuel in the pit from damage due to an accidental y drop of a heavy load. during reracking. This shield, shown in Figure (
2.4.16, is designed to withstand a total uniform load of 144 tons, which envelopes all heavy loads involved in the rarack operation by a A large margin. In Section 7.4, a parametric chart correlating the cross-section, drop height and allowable heavy load mass is provided for implementing administrative control on the heavy load mo';ements over the impact shield.
Procedural controls will be established for loads carried over the cask load.ng area of the cask pit when the impact shield is in place and to e. sure that no loads are carried over the cask loading are. if l fuel is present and the impact shield is not in place.
Table 2.4.1 provides a synopsis of the requirements delineated in NUREG-0612 and our intended compliance.
In summary, the measures to be implemented in Sequoyah reracking are identical to the-those utilized in the most recent successful rerack projects (such as Indian Point Unit 2, concluded in October, 1990).
2-13 l 1
)
, x _ _ _ . . _ _ _ . _ _ _ _ _ _ . - . _ _ . --
Table 2.1.1 MODULE DATA
- Module Array Cell Total Cell Count LJ) . Quant h Size for this Module Tyne A 6 13x14 1092 B 2 12x14 C
336 3 13x13 507 D 1 12x13 156 Total 12 2091 3
- Excluding the 15x15 cask pit rack to be installed at a later time.
2-14
_ _ _ _ J
l Table 2.1.2 COMMON MODULE DATA Storage cell inside dimension: 8.75" i-0.04" Storage cell height 168" i 1/16" (above the baseplate):
Baseplate thickness: 0.75" (nominal)
Support leg height: 5.25" (nominal)
Support leg type: Remotely adjustable legs Number of support legs: 4 (minimum)
Remote lifting and handling provision: '
Yes Poison material: Boral Poison length: 144" (nominal)
Poison width: 7.5" - ( na ninal)
Cell Pitch: 8.972" (nominal) 2-15 ,
_-- e l'
Table 2.1.3 MODULE DATA Dimensions (inch)* Shipping Module _ Weight x 10-3 I.D. North-South East-West (lbs)
A 117 126 26.0 B 108 126 24.0 C 117 117- 24,2 D 108 117 22.3 All dimensions are bounding rectangular envelopes rounded to the nearest one quarter of an inch.
2-16
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Table 2.3.1 BORAL EXPERIENCE LIST (Domestic and Foreign)
Pressurized Water Reactors Vented Construc- Mfg.
. Plant Utility tion Year Bellefonta 1,2 Tennessee Valley Authority No 1981 Donald C. Cook Indiana & Michigan Electric No 1979 Indian Point 3 NY Power Authority Yes 1987 Maine Yankee Maine Yankee Atomic Power Yes 1977 Salem 1, 2 Public Service Elec & Gas No 1980 Sequoyah 1,2 Tennessee Valley Authority No 1979 Yankee Rowe Yankee Atomic Pover Yes 1964/1983 Zion 1,2 Commonwealth Edison Co. Yes 1980 Byron 1,2 Commonwealth Edison Co. Yes 1988 Braidwood 1,2 Commonwealth Edison Co. Yes 1988 Yankee Rowe Yankee Atomic Electric Yes 1988 Three Mile Island I GPU Nuclear Yes 1990 Boiling Water Reactors ,
Browns Ferry 1,2,3 Tennessee Valley Authority Yes 1980 Brunswick 1,2 Carolina Power & Light Yes 1981 Clinton Illinois Power Yes 1981 Cooper Nebraska Public Power Yes 1979 Dresden 2,3 Commonwealth Edison Co. Yes 1981 Duane Arnold Iowa Elec. Light & Power No 1979 J.A. Fitzpatrick Ni Power Authority No 1978 E.I. Hatch 1,2 G6orgia Power Yes 1981 Hope Creek Public Service Elec & Gas Yes 1985 Humboldt Bay Pacific Gas & Electric Yes 1986 Lacrosse Dairyland Power Yes 1976 Limerick 1,2 Philadelphia Electric No 1980 Monticello Northern States Power Yes 1978 Peachbottom 2,3 Philadelphia Electric No 1980 Perry, 1,2 Cleveland Elec. Illuminating No 1979 Pilgrim Boston Edison No 1978 Susquehanna 1,2 Pennsylvania Power & Light Nv 1979 Vermont Yankee Vermont Yankee Atomic Power 3, ns 1978/1986 Hope Creek Public Service Elec'& Gas Yes 1989 D.C. Cook American Electric Power Yes 1991 Three Mile Icland GPU Nucl. ear Yes 1990 Unit 1 Shearon Harris Carolina Power & Light Yes 1991 Pool B 2-17
. . . _ _ . , _ . . _ _ . ~ . . _ . . . _ . _ . . _ _ _ _ . . . . _ . _ . . _ _
L
. l Table 2.3.1 (continued) 1 Foreign Installations Using Boril France 12.PWR Plants Electricite de France ;
South Africa Koeberg 1,2 ESCOM' Switzerland Beznau 1,2 Nordostschweizerische Kraftwerke AG.
Gosgen Kernkraftwerk Gosgen-Daniken AG Taiwan Chin-Shan 1,2 Taiwan Power Company Kuosheng._1,2 Taiwan Power Company Mexico Laguna Verde Comision Federal de Electricidad Units 1 & 2 s
1 -.
? : 2-18
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. Table 2.3.2 1100 ALLOY ALUMINUM PHYSICAL AND MECHANICAL PROPERTIES Density 0.098 lb/cu. in.
2.713 gm/cc l Melting Range 1190-1215 deg. F 643-657 deg. C Thermal Conductivity 128 BTU /hr/sq ft/deg. F/ft (77 deg. F) 0.53 cal /sec/sq cm/deg. C/cm Coef. of Thermal 13.1 x 10-6/deg. F-l Expansion 23.6 x 10-6/deg C '
! (68-212 deg. F)
) Specific heat 0.22 BTU /lb/deg. F
- (221 deg. F) 0.23 cal /gm/ dog. C Modulus of 10x106 psi Elasticity Tensile Strength 13,000 psi annealed (75 deg. F) 18,000 psi as rolled Yield Strength 5,000 psi annealed-(75 deg. F) 17,000 psi as rolled Elongation 35-45% annealed (75 deg. F) 9-20% as rolled Hardness (Brinell) 23 annealed-32 as rolled Annealing Temperature 650 deg. F 343 deg. C l-2-19
Table 2.3.3 CHEMICAL COMPOSITION (by weight) - ALUMINUM (1100 Alloy) 99.00% min. Aluminum 1.00% max. Silicone and Iron 0.05-0.20% max. Copper
.05% max. Manganese
.10% max. Zinc
.15% max. others each l
l l
l 2-20
Table 2.3.4 l
BORON CARBIDE CHEMICAL COMPOSITION. Weicht %
Total boron 70.0 min.
B10 isotopic content in 18.0 natural boron Boric oxide 3.0 max.
Iron 1.0 max.
94.0 min.
Total boron plus i
total carbon BORON CARBIDE PHYSICAL PROPERTIES Chemical formula BC4 Boron content (weight) 78.28%
Carbon content (weight) 21.72%
Crystal Structure rombohedral Density 2. 51 gm. /cc (0. 0907 lb/cu. _ in. )
Melting Point 24500 C (444207)
Boiling Point 35000 C (633207) i 2-21
4 Table 2.4.1 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612)
Criterion comoliance
- 1. Are safe load paths defined for Yes the movement of heavy loads to minimize the potential of impact, if dropped on irradiated fuel?
- 2. Will procedures be developed to Yes cover: identification of required equipment, inspection and acceptance criteria required before movement of load, steps and proper sequence for handling the load, defining the safe load paths, and special ,
precautions?
- 3. Will crane operators be trained Yes and qualified?
- 4. Will special lifting devices meet Yes the guidelines of ANSI 14.6-1978?
! 5. Will non-custom lifting devices Yes l be installed and used in accordance with ANSI B30.9-1971?
- 6. Will the cranes be inspected and Yes tested prior to use in rerack?
- 7. Does the crane meet the intent of Yes.
ANSI B30.2-1976 and CMMA-70?
4 2-22
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3.0 CONSTRUCTION OF RACK MODULES The object of this section is to provide a description of rack module construction for the Sequoyah spent fuel pool to enable an l independent appraisal of the adequacy of the design. Similar rack module designs have recently been used in previous licensing applications for South Texas (Houston Light & Power Company) and i Indian Point 3 (New York Power Authority). A list of applicable j codes and standards is also presented. 3.1 Fabrication Objective The requirements in manufacturing the high density storage racks for the sequoyah fuel pool may be stated in four interrelated points: (1) The rack module will be fabricated in such a manner that there is n2 weld splatter on the storage cell surfaces which would come in contact with the fuel assembly. (2) The storage locations will be constructed so that redundant flow paths for the coolant are available. i (3) The fabrication process involves operational sequences which permit immediate verification by the inspection staff. (4) The storage cells are connected to each other by austenitic stainless steel corner welds which leads to a honeycomb lattice construction. The extent of welding is selected to " harden" the racks from the seismic input l motion of the Operating Basis Earthquake -(OBE) and l Design Basis Earthquake (DBE) for the Sequoyah Nuclear Plant. l 3-1
3.2 Hixed Zone _Three Recion Storagg All rack modules designed and fabricated for the Sequoyah spent fuel pool are of the so-called "non-flux trap" type. In the non- I flux trap modules, a single panel of Boral is interposed between two fuel assemblies. The poison material utilized in this project is Boral, which does not require lateral support to prevent slumping due to the inherent stiffness. However, accurate dimensional control of the poison location is essential for nuclea.- criticality and thermal-hydraulic considerations. The design and fabrication approach to realize this objective is presented in the next sub-section. l 3.3 Anatomy of Rack Hoduler As stated earlier, the storage cell locations have a single poison panel between adjacent austenitic stainless steel surfaces. The major components of the rack module ares- (a) the storage box subassembly, (b) the baseplate, (c) the thermal neutron absorber material, and (d) support legs. A synopsis of the anatomy of the rack module is provided in the following, which explains the physical arrangement of the major constituent parts of a Sequoyah rack module. (a) The rack module manufacturing begins with fabrication of the box. The " boxes" are fabricated from two precision formed _ channels by a eam -- welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. Figure 3.3.1 shows the box. 3-2 l s
The minimum weld penetration will be 80% of the box metal gage which is 0.06" (16 gage). The boxes are manufactured to 8.75" I.D. , (nominal inside dimension). As shown-in Figure 3.3.1, each box has two lateral la diameter holes punched near its bottom edge to provide auxiliary flow holes. A double row of matching flat-f aced round dimples are coined into the walls of all square storage boxes. The height of each of these local coined areas is half the thickness of the poison sheet, thus the space provided by the corresponding raised areas on adjacent box walls is the thickness of the poison sheet. The poison sheets are axially centered on the active I fuel region. These sheets are slightly longer than the active fuel length at each end to provide added assurance that there is always suff:.cient poison
- material. The sheets are scalloped along the two long l edges to provide clearance for the raised, coined, areas of the box walls. With the poison installed, the boxes are welded together by fusing them at the coined areas i
using a proprietary fusion welding process. This process has been used to fabricate racks for more than 25,000 storage cells.The Boral panels are thus contained axially and laterally by these raised areas. Each interior poison sheet is supported axially at the bottom by a stainless steel strip of the same thickness as the poison sheet, which is welded to the wall of one of the two adjacent boxes. On the outside wall of the racks, the poison is mounted under a thin sheet of stainless cladding, and four edges of this stainless cladding are intermittently welded to the box wall. Figures 3.3.2 and 3.3.3 show an assemblage of boxes in isometric and sectional views, respectively. (b) Ru eplate: The baseplate provides a continuous horizontal surface for supporting the fuel assemblics. l The baseplate is attached to the box assemblage by l fillet welds. In the center of each storage location,
- the baseplate has a 5" diameter flow hole. The baseplate is 3/4" thick to withstand accident fuel- assembly drop loads postulated and discussed in Section 7 of this report.
3-3 v
I l (c) The thermal neutron absorber material: As mentioned in the preceding section, Boral is used as the thermal neutron absorber material. (d) Supoort Leag Adjustable support legs are shown in Figure 3.3.4. The top portion is made of austenitic steel material. The bottom part is made of SA564-630 i stainless steel to avoid galling problems. Each support leg is equipped with a readily accessible socket to enable remote leveling of the rack after its < placement in the pool. Lateral holes in tha support leg provide the requisite coolant flow path. An elevation cross-section of the rack module shown in Figure 3.3.5 shows two box cells in elevation. The Boral panels and their location are also indicated in this figure. The boral panels are vertically positioned such that the entire enriched fuel portion of the fuel assembly is enveloped in the longitudinal direction by the thermal-neutron absorber material. It is noted that the top of the boxes are flared prior t.o welding to i provide a smooth lead-in contour for the fuel assembly. l The joint between the composite box arrays and the baseplate in made by single fillet welds which provide a minimum of 7" of connectivity between each cell wall and the baseplate surface. As shown in Figure 3.3.4, the support leg is gusseted to provide an increased section for load transfer between the support. legs and the cellular. structure of interconnected boxes above the baseplate. Use of the gussets also minimizes heat input induced distortions of the support / baseplate contact region. Non-adjustable pedestala Me built up from austenitic stainless steel plate sect mus. 3.4 Weldina Tvoes and Processes t The basic types of welds are Tungsten Inert Gas (GTAW or TIG) and Metal Inert Gas (GMAW or MIG). Both fusion and filler metal added TIG arc welds and MIG welds are used. Thi welds are either 3-4
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automatic or manual, using electronically controlled welding j machines. l Elect;onically controlled TIG arc fusion butt welding is used to fabricate each storage box from two full length channel sections. A Jetline velding machine is used. It positions and clamps the parts and provides automatic feed and the electronic weld control. Up to 100% weld penetration is achieved with a minimum of 504 being structurally adequate for this type of construction. Electronically controlled TIG arc local fusion welds are used to fasten storage box to storage box. A special apparatus is used to provide the necessary clamping and spacing for both box-to-box and I row-to-row fabrication. This is a proprietary process used to provide the honeycomb rack design which results in the smallest possible cell-to-cell pitch dimension. The welding procesa is verified by a test procedure implemented at the start of each , manufacturing shift and any other time following a production shutdown or new set-up. Test specimens, made from box wall material, are welded by the machine and destructively tested. The destructive test assures that the , process settings and welding machine operation achieve complete soundness, size, and penetration of the welds. Manual welding with electronicclly cont:; . led welding machines ir performed for all 6ther fabrication 'r e icements. This welding is either TIC fusiu or filler metal added or HIG are welding. 3-5
-_. ,,, .m_ _ , _ . , _ _ _ , , . . - . _ , . _ _ , _ , , , , _ , . . , ,
__- . - ~ - . . - . - . - .. _ ._ . - . _ - - . . -_ - -.-. ._ . . - l l 3.5 Cnics , Standards. and_Eractice s fnI_.1he Secuoyah SEartt Puel En21_Eacht The fabrication of the rack modules for the Sequoyah spent fuel pool is perforar.d under a strict quality assurance system suitable for manufacturing and complying with the provisions of 10CFR50 Appendix D. The following codes, standards and practices will be used as applicable for the design, construction, and assembly of the spent fuel storage racks. Additional specific references relate.d to detailed analyses are given in each section.
- a. Codes add Standards for Den.(cn and Testing (1) AISC Hanual of Steel Construction, 8th Edition, 1980.
(2) ANSI N210-1976, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations". (3) American Society of Hechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section III, Subsection NF, 1989. (4) ASNT-TC-1A, June, 1980 American Society for Hondestructive Testing (Recommended Practice for Personnel Qualifications). l (5) ASME Section V - Hondestructive Examination (6) ASME Section IX - Welding and Brazing Qualifications (7) DuiMing Code Requirements for Reinforced Concrete, ACI318-63/ACI318-71. l l 3-6 l
I l l (8) Code Requirements for Nuclear Safety Related l Concrete Structures, ACI 349-85 and Commentary ACI 349R-85 (9) Reinforced Concrete Design for Thermal Effects on l Nuclear Power Plant Structures, ACI 349.1R-80 . (10) ACI Detailing Manual - 1980 (11) ASME HQA-2, Part 2.7 " Quality Assurance Requirements of Computer Software for Nuclear Facility Applications (draft). (12) ANSI /ASME, Qualification and Duties of Personnel Pngaged in ASME Boller and Pressure Vessel- Code Section III, Div.1, certifying Actinties, H626 1977.
- b. Baterial Codes (1) American Society for Testing and Haterials (ASTM)
Standards - A-240. (2) American Society of Hechanical ::..gineers (ASME), Boiler and Pressure Vessel Code, Section II - Parts A and C, 1989.
- c. Weldina Codes ASME Boiler and Pressure Vessel Code, Section IX-Welding and Brazing Qualifications (1986) or later issue accepted by USHRC.
- d. Qualitv Assurance, Cleanliness, Packacina, Shineina, 82ceivina. Storace, and Handlina Reauirements ~
I (1) ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Har.dling of Items for Nuclear Power Plants. (2) ANSI 45.2.1 <> cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants. I 3-7
,, . , . . . - - . _ . , . . . , , , _ , . . . , . . - - . , _ - . . , , ~
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_ - _ . . _ _. . .-_ . - - . - . _ _ _ _ . . _ - . ~ . . - .- - .___- -.- _ . (3) ASME Boiler and Pressure Vessel, Section V, Nondestructive Examination, 1993 Edition, including Summer and Winter Addenda, 1983. (4) ANSI - N16.1-75 Nuclear Criticality Safety Operations with Fissionable Materials Outside Reactors. (5) ANSI - N16.9-75 Validation of Calculation Methods for Nuclear Criticality Safety. (6) ANSI - N45.2.11, 1974 Quality Assurance Requirwents for the Design of Nuclear Power Plants. (7) ANSI 14.6-1978, "Special Lifting Devices for Shipping Centainers wnighing 10,000 lbs. or more for Nuclear Materials". (8) ANSI N45.2.6, Qualification of Inspection and Testing Personnel. (9) ANSI N41.2.U, Installation, Inspection. (30) ANSI N45.2.9, Records. (11) UlSI N45.2.10, Definitions. (12) ANSI N45.2.12, QA Audits. (13) ANSI N45.2.13, Procurement. l (14) ANSI 45.2.23, QA Audit Personnel. I c. Other References (In the references below, RG is NRC Regulatory Guide) (1) RG 1.13 - Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed). (2) RG 1.123 - (endorses ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants. (3) RG 1.124 - Service Limits and Loading Combinations for Class 1 Linear Type Component Supports, Rev. 1. 3-8
(4) RG 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Puel Handling Accident in the Puel Handling and Storage Pacility of Boiling and Pressurized Water Reactors. (5) RG 1.28 - (endoreer MISI N45.2) - Quality Assurance Program Requirements, June, 1972. (6) RG 1.29 - Seismic Design Classification, Rev. 3. (7) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Metal, Rev. 3. (8) RG 1.38 - (endorses ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants, March, 1973. (9) RG 1.44 - Control of the Use of Sensitized Stainless Steel. (10) RG 1.58 - (endorses ANSI N45.2.2) Qualificationand of Nuclear Power Plant- Inspection, Examination, Testing Personnel, Rev. 1, September, 1980. (11) RG 1.64 - Lendorses ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear Power Plants, October, 1973. (12) PG 1.71 - Welder Qualifications for Areas of Limited Accessibility. (13) RG 1.74 - (endorses ANSI N45.2.10) Quality Assurance Terms and Definitions, February, 1974. (14) RG 1.85 - Materials Code case Acceptability ASME Section III, Division 1. (15) RG 1.88 - (eno rses ANSI N45.2.9) Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records, Rev. 2, October, 1976. (16) RG 1.92 - Combining Modal Responses and Spatial components in Seismic Response Analysis. 3-9
. - . . . , . - - . ,. ,-..-_,r__ _ , . . . . .- . - - . . . .
_...-...-y , _ _ _ _ _ _ __ _ _._ ___._____.___ _ _____ I E I (17) RG 3.41 - Validation of Calculation Methods for Nuclear Criticality Safety. ; (18) General Design Criteria for Nuclear Power Plants, I Code of Federal Regulations, Title 10, Part 50, Appendix A (GDC Nos. 1, 2, 61, 62, and 63). i (19) NUREG-0800, Standard Review Plan, Sections 3.2.1, 3.2.2, 3.7.1, 3.7.2, 3.7.3, 3.8.4. (20) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated-April 14, 1978, and the modifications to this document of January 18, 1979. (Note: OT stands for office of Technology). (21) NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants". j (22) Regulatory Guide 8.8, "Information Relative to
; Ensuring that Occupational Radiation Exposure at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA).-
(23) 10CFR50 Appendix 8,-Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants (241 10CFR21 - Reporting of Defects and Non-Compliance 3.6 Materials of const-uetion Storage Cell: ASME SA240-304L Baseplate: ASME SA240-304L Support Leg (female): ASME SA240-304L Support Leg (male): Ferritic stainless steel (anti-
- galling material) ASME SA564-i 630 Poison: BoralTM i
'3-10 5
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HDLE HDLE FIGURE 3.3.5 ELEVATION VIET OF TWO CONTIGUOUS STOPJGE CELLS 3-1S
- , - - ,,,-e a,. ,- -
4.0 CRITICALITY SAFETY ANALYSES 4.1 Desian Bases The high density spent fuel storage racks for the Sequoyah Nuclear Plant are designed to assure that the effective neutron multiplication factor (k g) is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with unborated water at the temperature within the operating range corresponding to the highest reacti-vity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerances. All uncertainties are statistically combined, such that the final k m will be equal to or less than 0.95 with a 95% probability at a 95% confidence level. Applicable codes, standards, and regulations or pertinent . sections thereof, include the following: o General Design Criteria 62, Prevention of Criticality in Fuel Storage and Handling, o USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3 - July 1981 o USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979, o USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981. l o ANSI ANS-8.17-1984, Criticality Safety Criteria for the l Handling, Storage and Transportation of LWR Fuel Outside Reactors. 4-1 2
, ...r. ., , .,,----w-_3 , ~ ,,yy,
- - -. ~. -. - . . _ _ - - - - - -- . - __. . .
I l To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were madet o Moderator is assumed to be unborated water of the maximum density (1.000 g/cc) at a temperature of 4'C. , o The effective multiplication factor of an infinite radial array of fuel assemblies was used (see section 4.4.1) except for the boundary storage cells where leakage is inherent. o Neutron absorption in minor structural members is neglected, i.e. , spacer grids are analytically replaced by water. The design basis fuel assembly is a 17 x 17 Westinghouse Vantage-5H assembly containing UO 2 at a maximum initial enrichment of 4.95 i O.05 wtt U-235. (If fuel assemblies with natural UO 2 . blankets are used, the design basis enrichment is that of the central enriched zone.) Three separate storage regions are I provided in the spent fuel storage pool, with independent criteria defining the highest potential reactivity in each of the three regions as follows: o Region 1 is designed to accommodate new fuel with a maximum enrichment of 4.95 i O.05 wt% U-235, or spent i fuel regardless of the discharge fuel burnup. l o Region 2 is designed to accommodate fuel of 4.95% initial enrichment burned to at least 50 MWD /KgU (assembly average), or fuel of other enrichments with a burnup yielding an equivalent reactivity. o Region 3 is designed to accommodate fuel of 4.95% initial enrichment burned to at least 41 MWD /KgU l (assembly average), or fuel of other enrichments with l a burnup yielding an equivalent reactivity. 1 I 4-2
The water in the spent fuel storage pool normally contains soluble boron which would result in large subcriticality margins under actual operating conditions. Ilowever, the liRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting kg of 0.95 for normal storage be evaluated in the absence of soluble boron. The double contingency principle of AliSI 11-16.1-1975 and of the April 1978 liRC letter allows credit for soluble boron under other abnormal or accident conditions since only a single independent accident need be considered at one time. Consequences of abnormal and accident conditions have also been evaluated, where
" abnormal" refers to conditior's which may reasonably be expected to occur during the lifetime of the plant and " accident" refers to conditions which are not expected to occur but nevertheless must be protected against.
4.2 Summary of Criticalit? Analyses 4.2.1 fiormal Oneratina Conditions The design basis layout of storage cells for the three regions is shown in Figure 4.2.1. In this configuration, the fresh fuel cells (Region 1) are located alternately along the periphery of the storage rack (where neutron leakage reduces reactivity) or along the boundary between two storage modules (where the water gap provides a flux-trap which reducits reactivity) . Iligh burnup fuel in Region 2 affords a low-reactivity barrier between fresh fuel assemblies and Region 3 fuel of intermediate burnup. At the time of installation of the new racks, there should be an l adequate number of spent fuel assemblies to nearly fill and [ i " block off" the Region 2 barrier locations (see section 4.7) . Under these conditions, the administrative controls required are comparable to those of a conventional two-region storage rack design. 4-3
- .~. . - - . . - - . - , - - . , . - . - , . . . . . - - . . -
i The principles involved in the design and specification of an acceptable loading arrangement for fuel of different burnups are as follows: o Fresh fuel assemblies located along the outer periphery of the storage modules (along the pool wall) must be isolated from each other and from the inner Region 3 cells by at least one (1) Region 2 spent fuel-assembly (i.e., fuel of 50 MWD /KgU burnup or equivalent), o Fresh fuel assemblies located along the wide water-gap between storage modules must be isolated from each other and from the inner Region 3 cells by at least one (1) Region 2 spent fuel assembly (i.e., fuel of 50 MWD /KgU burnup or equivalent), o Fresh fuel assemblies located along the narrow water-gap between storage modules must be isolated from each other by at least two (2) Region 2 spent fuel assem-blies and from the inner Region 3 cells by at least one (1) Region 2 spent fuel assembly (i.e., fuel of 50 MWD /KgU burnup or equivalent), - o A checkerboard loading pattern of fresh fuel intermixed with empty cells may be used internal to or throughout any storage module. In these criteria, the term " fresh fuel" includes any fuel with reactivity greater than that of Region 3 fuel, such as may be encountered with partially-burned fuel in a full core off-load. Similarly, an empty cell is less reactive than any cell contain-ing fuel, and therefore may be used as a Region 2 isolation cell. l Prior to approaching the reactor end-of-life, not'all storage cells are needed for spent fuel. Therefore, an alternative (interim) configuration may be used in which the cells of selected modules may be loaded in a checkerboard pattern of fresh fuel (or fuel of any burnup) sith empty cells. Figure 4.2.2 j illustrates the concept of using a checkerboard loading pattern l 4-4 l 1
~ . , _ _ . _ _ . . __ _ _ _ __ __ _ . _ . ,-
within tho interior of one or more storage modules. A checker-board loading pattern throughout an entire module is also acceptable, with an even lower reactivity than that of Figure 4.2.2. The checkerboard configurations are intended primarily to facilitato a full core unload when needed, prior to the time the racks are beginning to fill up. l l l Figure 4.2.3 defines the acceptable burnup domains for spont fuoi and illustratos the limiting burnup for fuel of various initial enrichments for both Region 2 (upper curve) or Region 3 (lower curvo), both of which assume that the fresh fuel (Region 1) is enriched to 4.95% U-235. Criticality analyses for the design basis storago arrangement (Figuro 4.2.1) showed that the limiting (controlling) configuration is the boundary between modulos where the vator gap
- constitutes a neutron flux trap. There are two different water-gap spacings between modulos which necessitates -
different requirements for the isolation (barrier) cells. Along the poriphery of t'..e modules facing the concrete wall of the pool, the reactivity is substantially lower due to neutron leakago. The bounding criticality analyses are summarized in Table 4.2.1 for the design basis storago condition-(which assumes the single accident condition of the loss of all soluble boron) and in Table 4.2.2 for the most reactive chockerboard loading arrangement (Figure 4.2.2) . The calculated maximum reactivity of 0.938 for the checkerboard arrangemont is within the regulatory limit of a k,, of 0.95. This maximum reactivity includes calculational uncertaintios and manufacturing tolerances (95% probability at the 95% confidence level), an allowance for uncertainty in depletion calculations and the evaluated effect of The thick base-plates on the rack modulos extend beyond the_ storage calls and provido assurance that the noces-sary water-gap between modules is maintained. 4-5
the axial distribution in burnup. Fresh fuel of less than 4.95% enrichment would result in lower reactivities. As cooling time increases in long-term storage, decay of pu-241 (and growth of Am-241) results in a continuous decrease in reactivity, which providos an increasing suberiticality margin with time. No credit is taken for this decrease in reactivity other than to l indicato conservatism in the calculations. The burnup critoria identified above (Figure 4.2.3) for accep-table storage in Region 2 and Region 3 can bo implemented in appropriate administrative procedures to assure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2. Soluble poison is present in the pool water during fuel handling operations, and this serves as a further margin of safety and as a precautic' in the event of fuel misplacement during fuel handling operations. For convenience, the minimum (limiting) burnup data in Figure 4.2.3 for unrestricted storage may be described as a function of the initial enrichment, E, in weight percent U-235 by fitted polynomial expressions as follows: For'Recion 2 Storace Minimum Burnup in MWD /KgU =
-23.761 + 22.075E ~2.0165E2 + 0.1152E 3 Ecr Realon 3 Storaag.
Minimum Burnup in MWD /KgU =
-2 5. 74 2 5 + 18. 7 6 E -1. 393 3 E2+ 0. 0666E' 4-6 h
4.2.2 Abnormal and. Accident conditjj; gig Although credit for the soluble poison normally present in the spent fuel pool water is permitted under abnormal or accident conditions, most abnormal or accident conditions will not result in exceeding the limiting reactivity (k,, of 0.95) even in the absence of soluble poison. The effects on reactivity of credible abnormal and accident conditions are discussed in Section 4.6 and summarized in Table 4.2.3. Of these abnormal or accident condi-tions, only one has the potential for a more than negligible positive reactivity effect. Tho inadvertent misplacement of a fresh fuel assembly har, the potential for exceeding the limiting reactivity, should there be ; a concurrent and independent accident condition resulting in the loss of all soluble poison. Assuring the presence of soluble poison during fuel handling operations will preclude the - possibility of the simultaneous occurrence of the two independent accident conditions. The largest reactivity increase would occur if a now fuel assembly of 4.95% enrichment were to be positioned in a Region 2 location with the remainder of the rack fully loaded with fuel of the highest permissible reactivity. Under this accident condition, credit for the presence of soluble poison is permitted by NRC guidelines", and calculations indicato that approximately 690 ppm would be sufficient to assure that the limiting k,, of 0.95 is not exceeded. ,
" Double contingency principio of ANSI N16.1-1975, au specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Reg. Guide 1.13 (Section 1.4,
! Appendix A). 4-7
, - , - . - , . , , . -.--.-m..p, , , - . , , - . - - - - ..r.,,,,y . , . ,....,,,_..,m.. ,,,,m..,, ,,..r-.- we. ,, ..c.-.,,,6 ._,. , ,, . . --
4.3 Reference Puel Storaue__ Calla P 4.3.1 Reference Fuel Assembiv i The design basis fuel assembly, described in Figure 4.3.1, is a 17 x 17 array of fuel rods (Westinghouse Vantago-5H design) with 25' rods replaced by 24 control rod guide tubes and 1 instrument thimb3e. Table 4.3.1 summarizes the fuel assembly design specifications and the expected range of significant manufac-turing tolerancus. Since the spacer grida are conservatively noglected, the Westinghouse standard fuel design would exhibit the sam 2 reactivity. 4.3.2 Hiqh Density Fuel Stor,Dae Cells The nominal spent fuel storage cell used for the criticality analyses of the Soquoyah spent fuel storage cells is shown in Figure 4.3.1. tach eterage cell is composed of single Boral absorber panels positioned between two 8.75-inch I.D. , 0.060-inch thick stainless stool boxes. Poriphoral cells use a 0.060" stainless steel sheathing on the outsido supporting the Boral panol. The fuel assemblies are normally located in the conter of each storage cell on a nominal lattico spacing of 8.97 i o.04 inches. The Boral absorber han a thickness of 0.102 i o.005 inch l and a nominal B-10 areal donalty of 0.0324 g/cm3 (0.030 g B-10/cm2 minimum). 4.4 Analytical Methodoloav 4.4.1 Reforonge Dog _lgn Calculations In the fuel rack analysos, the primary criticality analyses of the high density spent fuel storage racks were performed with the 4-8
i r KENO-Sa computar code packaga(4.4.1), using the 27-group S CALE*** creso-section library and the NITAWL subroutino for U- ; ! 238 resonance shielding effects (Nordheim integral treatment) . Doplation analycos and deteratination of equivalent enrichments were made with thG two-dimansional transport theery code, CASMO-3(4.4.2). Donchmark ctsiculations, presented in Appendix A, indicato a bias of 0.0000 with an uncortainty of i O.0024 for CASMO-3 and 0.0113 i O.0017 (95%/95%) for HITAWL-KENO-Sa. ORNL, j in benchmsrk calculations with the 27-group SCALE _ library (4.4.3), has reported comparable results. i KENO-Sa Monto Carlo calculations inherently include a statistical { uncertainty due to the random nature of noucron tracking. To . minimize the statistical uncertainty of the KENO-calculated roactivity and to assure convergence of the calculation, a minimum of 1,000,000 neutron histories in 2000 generations of 500 neutrons each, wore accunulated in each calculation. - Figuro 4.4.1 represents the basic geometric model used in the KENO-Sa calculations. With reflecting boundary conditions, this model offectively describes an entire storage module representing both the narrow and the wide water channels. This model was also used to confirm the reactivity calculation for the checkerboard arrangement with fresh fual and empty cells and in the investiga-l tion of the consequences of potential accident conditions with a misplaced fresh fuel assembly. In addition, the axial variation in burnup was explicitly modeled (6 axial zones in Regions 2 and l
- 3) and resulted in osuentially the same reactivity as the referenco design calculation.
i I
***" S CALE" -is an acronym for Etandardized Computer Analysis for Licensing Evaluation, a standard cross-section sot, developed by ORNL for the USNRC, based upon ENDF/B-IV data.
4-9
1 A smaller modol, illustrated in Figure 4.4.2, was used to investigate the effect of the water channels between modulos and to confirm the larger model. This model describes an infinite array of 10 storage cells in the Y-direction (soparated by a watter gGp betwoon modul#4s) and an infinito array of calls in the X-direction. In the axial (Z) direction, the full langth 144- ! inch fuel assembly was described easuming 30-cm water reflector, top and bottom. A similar model was also used for calculations of the rack poripheral cells with both water and concrete reflectors. Because NITAWL-KENO-Sa doos not have burnup capability, burned fuel was represented by fuel of equivalent enrichment as determined by CASMO-3 calculations in the storage cell (i.e. an onrichmont which yiolds the same reactivity in the storace cell as the burned fuel). Figure 4.4.3 shows this squivalent enrichmont for fuel of 4.95% initial enrichment at various - dischargo burnups, evaluated in the storage cell. In tracking long-term (30-year) reactivity offects of spent fuel stored in Region 2 of the fuel storage rack, previous CASMO calculations have demonstrated a continuous reduction in reactivity with time (af ter Xe decay) due primarily to Pu-241 decay and Am-241 growth. No credit is taken for this decrease in reactivity other than to indicato conservatism in the calculations. I i 4.4.2 Fuel Byrnun Calculatiotly and Uncertainties CASMO-3 was used for burnup calculations in the hot operating condition. CASMO-3 has been extensively benchmarked (Appendix A and Ref. (4.4.4)) against critical experiments (including plutonium-bearing fuel) . In addition to burnup calculations, CASMO-3 was used for evaluating the small reactivity increments (by differential calculations) associated with most manutacturing i 4-10
_ ._._ .w. _ . = . - _ . . _ _- - 1 tolerances and for determining temperature ef fects (including the consequence of the library inadequacy in NITAWL) ."" Because the tolorance in boron-10 loading directly af fects a l coupling between cells of differing reactivities, KENO-5a i j differential calculations were used to determine this reactivity 7 uncertainty. l In the CASMO-3 geometric model (cell), each fuel rod and its cladding were described explicitly and reflecting boundary l co'iditj.ons (zero neutron current) were used at the centerline of the Boral and steel plates between storage cells. (CASMO is a two dimensional model.) These boundary conditions have the - . effect of creating an infinite array of storage cells in the X-Y j plane and provide a conservative estimate of the uncertainties in reactivity attributed to manufacturing tolerances. Since there are rio critical experiment data with spent fuel-for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity *"" was assigned based upon the assumption of 5% uncertainty in burnup. At the design basis burnups of 41 and 50 MWD /KgU, the uncertainties in burnup are i 2.05 and-i 2.5 MWD /KgU respective-l ly. To evaluate the reactivity consequences of the uncertainties in burnup, independant calculations were mada' with fuel of 39 and l 47.5 MWD /KgU burnup in Regions ; and 3, and the incremental , change from the referenca burnups assumed to represent the net l uncertainties in reactivity attributable to -burnup tolerance. These calculations resulted in an incremental reactivity The cross-section libraries for_NITAWL at the present time limit the NITAWL-KENO-Sa calculations.to 20 'C. >
"*"Only '.h a t pc.rtion'- of -- the ' uncertainty due to burnup.
Cther uncertainties are accounted for elsewhere. ; 4 . 4s--==-,-,w ,- e -e,c ww w . m m ,a w-e-e w, r ew -s %w+-r e + rw ,- v vy + r -- y - w -wwv3,,c+,mer-,w-i-y--,we, 4 y --d-- y w- dn v , w m -e --w -- --w , r er a t er 1 -w- r- M '" y N *v e my '+ W m'V-
**T*F9'- T V T weveever, y y wg-, gg s t urreet-,g e g"
. - - - - _ . . _ - . . . . . - - - - ~ - .
r uncertainty in k m of 0.00 70 4k in Region 2 (isolation barrier at 50 MWD /KgU burnup) and t 0.0024 4k for Region 3 (at 41 MWD /KgU , burnup). In the racks, the fresh unburned fuel in Region 1 strongly dominate the reactivity which tends to minimizo the l reactivity consequoness of uncertainties in burnup. The allowance for uncertainty in burnup calculations is a conser-vative estimate in view of the substantial reactivity decrease with timo as spent fuoi ages. , 4.4.3 Effect of Axial Burnuo Distributiom i Initially, fuel loaded into the reactor will burn with a slightly skewed cosino power distribution. As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the contral regions than in the upper and lower ends, as may be seen in the curves compiled in Ref. [4.4.5) . At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to noutron leakago. Consequently, it would be expected that over most of the burnup history, fuel assemblies with distributed burnups would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup pro-gresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of largo regions of significantly l rottcod burnup. Among others, Turnor(4.4.6) has provided generic l analytic results of the axial burnup effect based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially 1 i distributed burnup. l l 4-12
1 l Calculations were made with KEllo-Sa in three dimensions, based I upon the typical axial burnup distribution of spent fuel (that observed at the Surrey plant was taken as representative) . In these calculations, the axial height of the burned fuel was divided into a number of axial zones (6-inch intervals near the more significant top of the fuel), each with an enrichment equivalent to the burnup of that zone. These calculations resulted in a negligible reactivity increment for the reference l design burnups. Fuel of lower initial enrichments (and lower burnup) would have a more r.ogative reactivity effect as a result ! of the axial variation in burnup. These estimates are conserva-tive since smaller axial increments in the calculations have been shown to result in lower incremental reactivities (4.4.6). 4.5 Criticality Analyses and Tolerances 4.5.1 ffominal Desian For the nominal storage cell design, the liITAWL-KElio Sa calcula-tion resulted in a bias-corrected k,* * * * *
- of 0.9331 i O.0012 (95%/95%), which, when combined with all known uncertainties, results in a k,g of 0.9331 i O.010 or a maximum k,g of 0.943 with a 95% probability at the 95% confidence level (4.4.7).
For the interim loading pattern of fresh fuel checkerboarded with empty cells in Region 3, calculations resulted in nearly the same reactivity (maximum k,, of 0.938) as the reference design (maximum k,, of 0.943) within the normal KEllo-Sc statistics. The alternate checkerboard loading pattern can also be used through-out any module (maximum k g of 0.043).or in the internal region (Region 3) of any module. l * **"*A s used in this report, k, refers to the reactivity of an infinite radial array of finite length, with a top and bottom water reflector. i 4-13
.. . - - . - . - _ - . _ - . _ _ - ~ . . - _ . . - . - - . - - . - - - . _ - - . - i 4.5.2 Uncertainties Dge to Manufacturina Tolerances I The uncertainties due to manufacturing tolerances are summarized in Table 4.5.1 and discussed below. . 4.5.2.1 Boron Loadina ToleranG21 The Boral absorber panels used in the storage cells are nominally 0.102 inch thick, 7.50-inch wide and 144-inch long, with a nominal B-10 areal density of 0.0324 g/cm 2. The vendors manuf ac-turing tolerance limit is i O.0024 g/cm2 in B-10 content which assures that at any point, the minimum B-10 areal density will
- not be less than 0.030 g /cm'. Differential KENO-Sa calculations for thm reference design with the minimum to'.orance B-lo loading results in an incremental reactivity of i 0.0045.6k uncertainty.
4.5.2.2 Boral Width Toleranc.g
- i The reference storage cell design uses a Boral panel with an initial width of 7.50 t 0.06 inches. For the tolerance of 0.06 inch, the dif ferential CASMO-3 calculated reactivity uncertainty is i O.0010 6k.
4.5.2.3 Tolerances in cell Lattice Spacinct The manufacturing tolerance on the inner box dimension, which dir.ctly affects the storage cell lattico spacing between fuel assemblies, is i o.04 inches. This corresponds to an uncertainty in reactivity of i O.0016 Sk determined by differential CASMO-3 calculations. 4-14
._ _ _ . . , _ . . _ , _ . . _ . , , . . . _ _ . = . . _ . . _ ,-
4 4.5.2.4 Stainless Steel Jhickness Tolerances i The nominal stainless steel thickness is o.060 t 0.005 inch for the stainless steel box. The maximum positive reactivity effect-of the expected stainless steel thickness tolerances was calculated (CASMO-3) to be t 0.0004 dk. 4.5.2.5 Fuel Enrichment and Density Tolerancos The design maximum enrichment is 4.95 i 0.05 wt% U-235. Separate CASMo-3 burnup calculations were made for fuel of the maximum enrichment (5.0%) and for the maximum 002 density (10.61 g/cc). Reactivities in the storage coll were then calculated using the restart capability in CASMO-3. For fresh fuel, the incremontal reactivity uncertainties were i 0.0021 for the enrichment tolerance and i 0.0013 for the tolerance in fuel density. Using equivalent enrichments determined for the reference fuel burnups of 41 and 50 MWD /KgU, 3-dimensional KENO-Sa calculations were made to confirm the CASMc estimated uncertainties. The small incremontal reactivities datormined with KENO-Sa were consistent with those of the CASMo calculations within the normal statisti-cal variation in KENO results. For the tolerance on U-235 enrichment, the uncertainty in k2 is 0.0021 6k and for fuel density.is 0.0021 6k. 4.5.3 Water-can Snacina Between Modulgg The wator-gap between modulos constitutes a neutron flux-trap for the outer (periphoral) row of storage cells. Calculations with KENO-5a wore made for the tolerances in water-gap spacings. From these data, it was dotermined that the incremental reactivity consequence (uncertainty) - for the water-gap tolerance- of i 1/8 inch is t o.0022 6k (narrow gap) and i 0.0020 6k (wide gap) .
'4-15 - + , - - , y. --m,.,,..y-,..,-#--y..y ,,y y-, . , , . + . , - y w ,.v- ,ym..,- p- -,-,e.aw ,.,-yy,-.,,--,,,,...y--..w-- e. ..,,,,,.-_.--,,+r. ,,, .,, + yg
The racks are constructed with the base plate extending beyond the edge of the cells which assures the minimum spacing between storage modules is maintained under all credible conditions. 4.5.4 Ep.2RDtric Puol Positionina The fuel assembly is assumed to be normally centered in the storage rack cell. Calculations were made using KENO-Sa, a0suming the fuel assemblies were located in the corner of the storage rack cell (four-assembly clusters at the closest possible approach). Thesc. calculations indicated that the reactivity ; increment due to eccentricity of assembly locations was slightly l negative. Therefore, the reference case with the fuel assemblies centered is controlling and no uncertainty for eccentricity is necessary. 4.6 Abnormal and Accident Conditions
- 4.6.1 Temnerature and Water Density Effecta The moderator temperature coefficient of reactivity is negative; a moderator temperature of 4'C (39'F) was assumed"""* for the reference designs, which assures that'the true reactivity will always be lower over the expected range of water temperatures.
Temperature effects on reactivity have been calculated (CASMo-3) and the results are shown in Table 4.6.1. With soluble poison present, the temperature coefficients of reactivity would differ from those inferred from the data in Table 4.6.1. However, the reactivities would also be substantially lower at all tempera
~~
"""*Th e library inadequacy in NITAWL results in underpredic-tion of the reactivity at temperatures above 20*C. It is expected that the reactivity would be overpredicted (conserva-tive) at 4'C and this expectation is consistent with CASMO-3 calculations.
4-16
tures with soluble boron present, and the data in Table 4.6.1 is partinent to the higher-roactivity unborated case. 4.6.2 Drooned Puel Assembly I i For a drop on top of the
- ack, the fuel assembly will come to !
rest horizontally on top of the rack with a minimum Saparation distance from the fuel in the rack of more than 12 inches, including the potential deformation under poismic or accident conditions. At this separation distance, the effect on reac-tivity is ins igni f icant . '"*"" Furthermore, soluble boron in the pool water would substantially reduce the reactivity and assure that the true reactivity is always less than the limiting I value for any conceivable dropped fuel accident. 4.6.3 Lateral Rack Movement Lateral motion of the rack aodules undur seismic conditions could potentially alter the spacing between rack modules. However, the thick base plates prevent closure of the water-gaps beyond the I minimum spacing assumed for the analysis. Furthermore, soluble poison would assure that a reactivity loss than the design l limitation is maintained under all credible accident or abnormal l conditions. 4.6.4 Abnormal Locatiqn,_gtf a Fuel Assembly i The abnormal location of a fresh unirradiated fuel assextbly of 4.95 vtt enrichment could, in the absence of soluble poison, result in exceeding the design reactivity limitation (k m of 0.95). Calculations (KENO-Sa) confirmed that the highest l-
""""A - separation distance of 12 inches or more has lon been accepted as effectively being complete neutronic isolation.g 4-17 l
l
~
y.,my 74~. - -,-.,-y-,-e- , = - --__-m.,n,--.m--- ,,-w,- ---...r m.-. ,.m.- ,m., ...y--rc., mrg,-%-.*.3,.w,,e.e-3 -,-,e--my-.c .o r.,--* .
l l reactivity, including uncertainties, for the worst case postu-lated accident condition (fresh fuel assembly in Region 2) would exceed the limit en reactivity in 'che absence of soluble boron, soluble boron in the spent fuel pool water, for which credit is ! permitted under these accident conditions,-_would assure that the reactivity is maintained substantially less than the design limitation. . Calculations indicate that a soluble poison concentration of 690 ppm boron would be required to limit the maximum reactivity to a k, of 0.95 including uncertainties, under the maximum postulated accident condition. To maintain km at the reference design value' of 0.f 43 would require- 765 ' ppm boron. It is not physically possible to install a tuel assembly outside and adjacent to a storage module in the spent fuel storage pool. Nevertheless, a storage module will be installed ir the cask i loading area of the~ cask pit where there would be sufficient room for such an extraneous assembly, ~ However, this module is limited to spent fuel only (a Region 3 storage facility)--and calculations-l show that the maximum k , remains well below the 0.95 limit under the postulated accident condition,even in absence of soluble l boron. 4.7 Existinc Soent Fuel Following the October 1991 refueling, there were 661 spent-fuel assemblies in storage at the - Sequoyah plant. Figure 4.7.1; superimposes the. enrichment-burnup combination of these fuel
- assemblies on the curves defining the acceptable burnup domains.
As may be seen in this figure, most of the spent . fuel- now in-storage falls well into the acceptable domain for-the_ barrier-fuel (Region _2). There are now 562 fuel. assemblies' meeting'the'-
- enrichment-burnup criteria for storage -in Region . 2. By the time-4 l l
.r,,,-3 ..,y-4,,-%& *- 4,-,--- emr e w_v ..w_ ,,1-, m.,_o .m_- ,y._3 , m m -+,ew-, ,i,,y. ,,,.-,cy-
-~,-rm.r-,y, . <, - - - - -,
l the racks are installed, there should be a sufficient number of
- assemblies available to load into Region 2 to fill (or nearly fill) the 826 Region 2 cells. Fourteen fuel assemblies (dis-charged prematurely for various reasons; will need to be kept in a Region 1 storage location, and the remaining assemblies may be
- stored in Region 3 locations. Future discharge batches may reasonably be expected to have a preponderance of highly burned fuel capable of serving as barrier fuel in Region 2, if needed.
An appreciable number of spent fuel assemblies that will be stored ' Regicsn 3 have enrichment-burnup combinations well in excess of the minimum design basis and this provides further conservatism in the criticality safety of the spent fuel storage rack design. 4.8 Esferences 4.4.1 Green, Lucious, Petrie, Ford, White, and Wright, "PSR-63/NITAWL-1 (code package) NITAWL Modular Code l System For Generating Coupled Multigroup Heutron-Gamma l Libraries from ENDF/B", ORNL-TM-3706, Oak Ridge l National Laboratory, November 1975. R.M. Westfall et. al., " SCALE: A Modular System for Performing Standardized Computer Analysis for Licens- ! ing Evaluation", NUREG/CR-02OO, 1979. Volume 2, Section F11, " KENO-Sa An Improved Monte Carlo Critical-ity Program with Supergrouping". 4.4.2 A. Ahlin, M. Edenius, and H. Haggblom, "CASMO - A Fuel Assembly Burnup Program", AE-RF-76-4158, Studsvik report. A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis", ANS.Transac-tions, Vol. 26, p. 604, 1977.
"CASMO-3 A Fuel Assembly Burnup Program, Users Ma7ual", Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986 l 4.4.3 A.M. Hathout, et. 'a l . , " Validation of Three Cross-l section Libraries Used with the SCALE System for Criticality Analysis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.
4.4.4 M. Edenius and A. Ahlin, "CASMO-3: New Features, Benchmarking, and Advanced Applications", Nuclear
-Science and Encineerin_g, 100, 342-351, (1988) 4-19
4.4.5 H. Richings, Some Notes on PWR (H) Power Distribution Probabilities for LOCA Probabilistic Analyses, NRC Memorandum to P.S. Check, dated July 5, 1977. 4.4.6 S. E. Turner, " Uncertainty Analysis - Burnup Distribu-tions", presented at the DOE /SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ ENS Conference, Washington, D.C., November 2, 1988 4.4.7 M.G. Natrella, Experimental Statistics, National Bureau of Standards Handbook 91, August 1963. l l l 1 l 4-20
Table 4.2.1
SUMMARY
OF CRITICALITY SAFETY ANALYSES NORMAL STORAGE CONFIGURATION Design Basis burnups at 4.95% 0 in Region 1 i O.05% initial enrichment 50 in Region 2 41 in Region 3 Temperature for analysis 4*C ( 39 ' F) Reference k m (KENO-Sa) 0.9118 Calculational bias, Skm 0.0113 Uncertainties Bias statistics (95%/95%) t 0.0017 KENO-Sa statistics (95%/95%) t 0.0012 Manufacturing Tolerances i O.0057 Water-gap (wide) i O.0020 Water-gap (narrow) 0.0022 Burnup (41 MWD /KgU) i O.0024 Burnup (50 MWD /KgU) i O.0070 Eccentricity in position Negative Statistical combination .0100 of uncertajnties* Axial Burnup Effect Negligible Total O.9331 i .0100 Maximum Reactivity (k g) 0.943 l m See Appendix A
- Square root of sum of squares.
4-21 rr+ + n
Table 4.2.2
SUMMARY
OF CRITICALITY SAFETY ANALYSES (MOST REACTIVE CHECKERBOARD LOADING) Design Basis burnups at 4.95% 0 in Region 1 0.05% initial enrichment 50 in Region 2 Region 3 -CHECKERBOARD (FRESH FUEL AND EMPTY) - Temperature for analysis 4*C ( 3 9 ' F) Reference k g (KENO-Sa) 0.9191 Calculational bias, 6km 0.0113 Uncertainties (Assumed same as the reference case) Bias statistics (95%/95%) i O.0017 , KENO-5a statictics (95%/95%) i O.0012 Manufacturing Tolerances i 0.OO57 Water-gap (wide) i O.0020 l Water-gap (narrow) 0.0022 Burnup (41 MWD /KgU) NA Burnup (50 MWD /KgU) i O.0024 Eccentricity in position Negativa Statistical combination i O.0072 of uncertaintiesm Axial Burnup Effect Negligible Total O.9304 i O.0072 l Maximum Reactivity (k g) 0.9376 l l l I W See Appendix A A Square root of sum of squares. 4-22
- . _ - . ..~. .- - - . . ._ . - - _ .-. . . . . . - . - . .
Table 4.2.3 REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS Accident / Abnormal Conditions Reactivity Effect Temperature increase (above 4*C) Negative (Table 4.6.1) Void (boiling) Negative (Table 4.6.1) Assembly dropped on top of rack Negligible Late'ral rack module movement Negligible (See Section 4.6.3.) Misplacement of a fuel assembly Positive - controlled by soluble poison i l l l 4-23
_ _ _ _ _ . _ . _ . _ . . _ _ . . ~ . _ . . _ . _ _ - = _ _ _ . _ . _ _ _ _ . . _ . . . _ _ . . _ - _ . _ _ _ . , _ . _ _ _ _ Table 4.3.1 DESIGN BASIS FUEL ASSEMBLY SPECIFICATIONS 4 FUEL ROD DATA Outside diameter, in. O.374 Cladding thickness, in. 0.0225-Cladding innide' diameter, in. O.329 Cladding material Zr-4 Pellet density, % T.D.- 95.0 Stack density, g UO 2 /cc 10.41 i O.20 , Pellet-diameter, in. O.3225 Maximum enrichment, wt % U-235 4.95 i O.05 FUEL ASSEMBLY DATA . Fuel rod array 17 x Number of fuel rods 264-Fuel rod pitch, in. O.496 > l Number of control' rod guide 25 and instrument thimbles-Thimble-0.D., in. (nominal) 0.474 Thimble I.D., in. (nominal) 0.442 4-24 _. , ,__.2 . _ . . .,a... __.2 _ _.- , . _ .. _ _ . , _ _. . . .. ,__-._._..,_.._.._.-._.;_.___.-___....-
l l Table 4.5.1 Reactivity Effects of Manufacturing Tolerances l Tolerance Incremental Reactivity, 6k Boron-10 loading (i O.0024 g/cm 2) i O.0045 Boral Width ( 0.06 inch) i O.0010 Lattice spacing (i O.04 inch) i O.0016 Stainless Thickness (i O.005 inch) 0.0004 Fuel enrichment 0.0021 Puel density 0.0021 Total (statistical sum) i O.0057 t l I l l 4-25 l - __ ., ._. .._ . _ . . _ _
- .. , - . . . - - . . . .~. - . - _ . . . . _. - .. - .. . - .- .- ..-. _ . . . -. --__ ...-. .
E Table 4.6.1 EFFECT OF TEMPERATURE-AND VOID ON CALCULATED REACTIVITY OF-STORAGE RACK Incremental Reactivity Change, 6%- Case Region 1 Region 2 - Region 3 (Fresh)- 50 MWD /KgU
. 41 MWD /KgU-4?C (39'F) Reference Reference Reference 20*C (68'F) -0.0020 -0.0012 -0.0017 60'C (140*F) -0.0096 -0.0059 -0.0073 120'C (248'F) -0.0257 -0.0163 -0.0193 -
120'C (248'F) -0.0518 -0.0384 -0.0425
+ 10% Void
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/NITAWL-1 (code package) NITAWL Modular Code System For Generating Coupled Multigroup Neutron-CAmma Libraries from ENDF/B", ORNL-TM-3706, Oak Ridge National Laboratory, November 1975.
- 2. R.M. Westf all et. al. , " SCALE: A Modular System for Performing Standardized- Computer Analysis for Licensing Evaluation",
NUREG/CR-02OO, 1979.
- 3. A. Ahlin, M. Edenius, and H. Haggblom, "CASMO -
A Fuel Assembly Burnup Program", AE-RF-76-4158, Studsvik report.- A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis", ANS Transactions, Vol. 26,
- p. 604, 1977.
I "CASMO-3 A Fuel Assembly Burnup Program, Users Manual", Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986
- 4. M.N. Baldwin et al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel", BAW-1484-7, The Babcock & Wilcox Co., July 1979.
- 5. M. Edenius and A. Ahlin, "CASMO-3 : New Features, Benchmarking, ,
and Advanced Applications", Nuclear Science and Encineerinc, 100, 342-351, (1988)
- 6. M.G. Natrella, Experimentg1 Statistics, National Bureau of Standards, Handbcok 91,-August 1963. '
- 7. R.W. Westfall and-J.-H. Knight, " SCALE System Cross-section Validation with Shipping-cask Critical Experiments", AHE Transactions, vol. 33,.p. 368, November 1979 l 8. S.E. Turner and M.K. Gurley, " Evaluation of - NITAWL-KENO Benchmark Calculations for High Density Spent Fuel Storage R&cks", -Nuclear Science and Encineerino, 80(2):230-237, February 1982.
A-5 - - . - - ,m, ,. - - ,, - yy -
- 9. J.C. Manaranche, et. al. , " Dissolution and Storage Experiment with 4.75% U-235 Enriched 002 Rods", Huclear Technoloav, Vol.
50, pp 148, September 1980
- 10. A . M .~ Hathout, et. al., " Validation of Three Cross -section Libraries Used with the SCALE System for Criticality Analy-sis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.
- 11. S.R. Bierman, et. al. , "Criti Separation between Sub-critical Clusters of 4.29 Wt. % gl U Enriched UO Rods in Water 2
with Fixed Neutron Poisons", Batalle Pacific Northwest Labora-tories, NUREG/CR/OO73, May 1978 (with August 1979 errata). . 4 I A-6 i y - m --gst'wv -->-Tl1 m'n'* Mp-4' -- M'" *' +'Tt' W Table 1 RESULTS OF 27-GROUP (SCALE) NITAWL-KENO-Sa CALCULATIONS OF B&W CRITICAL EXPERIMENTS Experiment Calculated a Number k,, , l I O.9932 0.0016 II O.9915 i O.0015 III O.9916 i O.0013 IX O.9918 i O.0014 X O.9923 i O.0015 XI O.9919 i O.0014 XII O.9961 0.00115 XIII O.9960 t 0.0015 XIV O.9817 1 O.0015 XV O.9843 i O.0014 l XVI O.9912 i O.0015 XVII O.9866 t 0.0013 XVIII O.9904 t 0.0014 l- XIX O.9861 i O.0013 XX O.9934 i O.0013 XXI O.9874 i O.0014 l Mean 0.9910 t 0.0014"3 Bias 0.0090 t 0.0033(2) Blas - (95%/95%) 0.0090 t 0.0021 "3 (*3 Calculated from individual standard deviations. Calculated from k,,, values and used as reference. A-7 Table 2 RESULTS OF 27-CROUP (SCALE) NITAWL-KENO-Sa CALCULATIONS OF FRENCH and BNWL CRITICAL EXPERIMENTS Prench Experiments Separation critical calculated Distance, cm Height, cm k,,, O 23.8 1.0231 i O.0036 2.5 24.48 1.0252 1 0.0043 5.0 31.47 1.0073 i O.0013 10.0 64.34 0.9944 i O.0014 BNWL Experiments Calculated Case Expt. No. k,,, No Absorber 004/032 0.9964 t 0.0034 SS Plates (1.05 B) 009 0.9988 t 0.0038 SS Plates (1.62 B) 011 1.0032 t 0.0033 SS Plates (1.62 B) 012 0.9986 i O.0036 SS Plates 013 0.9980 i O.0038 SS Plates 014 0.9936 i O.0036 Zr Plates 030 1.0044 i O.0035 Mean 0.9990 0 0037 A-8 . _ . - ,_- . .~. - - - - - _ . - . . . . - . . - - . . - _ . . . - . . - - ~~ . ...... .-. Table 3 RESULTS OF CASMO-3 AND NITAWL-KENO-Sa BENCMMARK (INTERCOMPARISON)- CALCUIATIONS Enrichment'U k '" Wt. % U-235 NITAWL-KENO-Sa'33 CASMO-3 ldkl 2.5 0,8408 0.0016 O.8379- 0.0029 3.0 0.8831 i'O.0016' O.8776 0.0055 3.5 0.9097 i O.0016, 0.9090 0.0007-4.0 0.9334 0.0016 0.9346- 0.0012 4.5 0.9569 1.0.0018 0.9559 0.0010-5.0 0.9766 i O.0018- 0.9741 0.0025 ~ Mean -0.0023 Infinite array of assemblies typical of high-density spent fuel storage racks. l (2)
- k. from NITAWL-KENO-Sa corrected - for bias of-- 0.0113 6k.
t -- A-- 9 I , ,-.,,,--,,ve-- ---,e- - ~ , --ra< ,-,m ,-,,w,--e,.we <c-w-, er, ,v- ,s ,--,:g.- ,*.-e-etw,w..,,es s , --,---,--n-, e- w .. . . . . . . _ - - -. . - - - - - - . = - . . - - . . - . _ _ - . _ . - - . . . . . - - - . . . . . I i-l 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction This section provides a summary of the methods, models, analyses and numerical results to demonstrate the compliance of the l 'reracked Sequoyah spent fuel pool with the provisions of Section _ l III of the USNRC "OT Position Paper for Review and Acceptanca of Spent Fuel Storage and Handling Applications", (April 14, 1978). Similar methoda of thermal-hydraulic analysis have been used in + previous licensing ef forts on high density spent fuel racks for Fermi 2 (Docket 50-341), Quad Cities 1'and 2 (Dockets 50-254 and 50-265), Rancho Seco (Docket 50-312), Grand Gulf-Unit 1 (Docket 50-416), Oyster Creek-(Docket 50-219), Virgil C. Summer (Docket 50-395), Diablo Canyon 1 and 2 (Docket Nos. 50-275 and 50-323), Byron Units 1 and 2 (Docket 50-454, 455), St. Lucie Unit One (Docket 50-335), Millstone Point I (50-245), Vogtle Unit 2 (5'0-425), Kuosheng Units 1 & 2-(Taiwan Power Company), Ulchin Unit 2 (Korea Electric Power Company), J.A. FitzPatrick (New York Power Authority), TMI Unit-1 (GPU Nuclear),: Donald C. Cook Units 1 & 2 (Indiana & Michigan _ Electric Company), and Zion Units 1 & 2 !. (Commonwealth Edison Company). The analyses to be carried out for the thermal-hydraulic qualification of the rack array may be broken down into the following categories:- (i) Pool decay _ heat evaluation and pool-' bulk temperature variation with time. . , (ii) Determination;of the maximum pool local temperature at the instant when the bulk temperature reaches its maximum value. (iii) Evaluation of-the maximum fuel cladding temperature-to establish that - bulk . nucleate . boiling at any . location in the vicinity of the fuel assembly does not occur. l l 5-1 l L _ _ _ . . , - - _ . _ . _ . . _ _ . _ . . ~ - _ . . . . . . _ . . . _ _ _ . _ . _ . . _ . , . _ . _ . . _ . . . - _ . . I (iv) Evaluation of the time-to-boil if all heat rejection paths through the cooling and cleanup are lost. , (v) Compute the effect of a blocked fuel cell opening on the local water and maximum cladding temperature. The following seccions present a brief outline of the cooling ! System, a synopsis of the methods employed to perform the thermal-i hydraulic analyses, and a final summary of the resu)ts. 5.2 i}ggpi Puel Pit Coolina Syst.c3B ' l h e Opent Fuel Pit Cooling System is designed to remove from the spent fuel pit water the decay heat generated by stored spent fuel assemb1Eac. Additional functions of the Spent Fue? #it 'Jooling System (SFPCS) are to clarify and purify the water i n t. a s p e h t funi pit, transfer canal, and refueling water stern N te ds. This SFNS is described in Section 9.1.3 of the UFSAR. Syste.n piping la arranged so that failure of any pipeline cannot drain the spont fuel pit below the water level required for radiat:on shielding. A water level of ten feet or more above the top of the stored spent fuel asse.mblies is maintained to limit direct gamma dose to 2.5 mr/hr or '.ess. l *Se system's demineralizer and filter are designed to provide l udequate puriflcation to permit unrestrictec access to the spent l fuel storage area for plant personnel and maintain _ optical clarity
- l. et the spent fuel pit water. The optical clarity of the spent fuel pit water surface is maintained by use of the system's skimmera, strainer, and skimmer filter.
I 5-2 1 l l 1 5.2.1 System Description The Spent Fuel Pit Cooling system consists of two cooling trains (plus a backup pung capable of operation in eIther train), a purification loop, and a service skimmer loop. When the Spont Fuel Pit Cooling System is in operation, water flows from the spent fuel pit to both spent fuel pit pump suctions, is pumped through the tubeside of the heat exchangers, and is returned to the pit. Each pump's auction line, which is protected by a strainer, is located at an elevation four feet below the normal spent fuel pit water level, while the return line contains an anti-siphon hole near the surface of the water to provent gravity drainage of the pit. While the heat removal operation is in process, a portion of the spent fuel pit water may be diverted through a domineralizer and'a filter to maintain spent fuel pit water clarity and purity. This purification loop is sufficient for removing fission products and other contaminants which may be introduced if a fuel assembly with defective cladding is transferred to the spent fuel pit. The spent fuel pit demineralizer may be isolated, by manual valves, from the heat removal portion of the Spent Fuel Pit Cooling Syntum. By so doing, the isolated domineralizer may be used in conjunction with a refueling water purification pump and l filter to clean and purify the refueling water while spent fuel ! pit heat removal operations proceed. Connections are provided such that the refueling water may be pumped from either the refueling water storage tank or the refueling cavity of either l unit, through the domineralizer and filter, and discharged to the refueling cavity or refueling water ctorage tank of either unit. Connections are also provided to allow clean-up - of the water in the transfer canals. Water can be drawn from either the top or the boztom of the canal and is pumped by a refueling water I 5-3 purification pump through the spent fuel pit domineralizar and a refueling water purification filter beforu being returned to the transfer canal. To further assist in maintaining spent fuel pit water clarity, the ' water surface is cleaned by a skimmer loop. Water is removed from the surface by the skimmers, pumped through a strainer and filter, and returned to the pit surface at three locations remote from the skimmers. The spent fuel pit is initially filled with water that is at the same boron concentration as that in the refueling water storage tank. Borated water may be supplied from the refueling water storage tank via the refueling water purification pump connection, t or by running a temporary line from the boric acJd blender, located in the chemical and Volume control System directly into the pit. Demineralized water can also be added for make'up purposes (i.e., to replace evaporative losses) through a connection in the recirculation return line. i The spent fuel pit water may be separated from the water in the transfer canal by a gate. The gate is installed so that the transfer canal may be drained to allow maintenance of the fuel transfer equipment. The water in the transfer canal is pumped, via a refueling water purification pump, into a holdup tank in the chemical and volume control System. When maintenance on the fuel transfer equipment is completed, the water is returned to the transfer canal by the holdup tank recirculation pump. 5.2.2 comnonent Descrintion Spent Fuel Pit Pump.g, The pumps are horizontal, centrifugal units. They circulate spent fuel pit water through the heat exchangers, domineralizer, and filter. The pumps are controlled manually from a local station. 1 5-4 i A third pump is available to serve as a backup to either of the two pumps normally used for cooling the spent fuel pit water. f Spent F,.q1 m Jit Skimmer Pump This hot.1::ontal, centrifugal pump circulaten surf ace water through 'a strainer and a filter and returns it to the pit. Refuelina Water Purification PuttDE These horizontal, centrifugal pumps are used to circulate water from the transfer canal, the refueling cavity and the refueling water storage tank through the spent fuel pit domineralizer and a i refueling water purification filter. The pumps are operated manually from a local station. Spent Fuel Pit Heat Exchancer.s. The spent fuel pit heat exchangers are of the shall and U-tube typo with the tubes welded to the tubasheet. Component cooli'ng water circulatos through the shell, and spent fuel pit water circulatos through the tubes. Snent Fuel Pit Demineralizer This flushable, mixed-bed domineralizer is designed to provide adequate fuel pit water purity for unrestricted access by plant personnel to the pit working area while maintaining visual clarity. Spent Fuel Pit Filter The spent fuel pit filter.is designed to improve the pit water clarity by removing particles which obscuro visibility. Spent Fuel Pit Skimmer Filter The spent fuel pit skimmer filter is used to remove particles which are not removed by the strainer. l 5-5 . I Refuel Jg Water Purification Filterg The refueling water purification filters are designed to improve the clarity of the refueling water in the refueling canal or in the refueling water storage tank by removing particles which obscure visibility. Spent Puol Pit Strainer A strainer is located in each spent fuel pit pump suction line for removal of relatively large particles which might otherwise clog the spent fuel pit domineralizar or damage the spent fuel pit pumps. Snent Fuel plt Skimmer Straingr The spent fuel pit skimmer strainer is designed to remove debris from the skimmer process flow. Spent Fuel Pit Skimmers , Two spent fuel pit skimmers are provided to remove water from the spent fuel pit water surface in order to remove floating debris. l valves Manual stop valves are used to isolate equipment and manual throttle valves provide flow control. Valves in contact with spent fuel pit water are austenitic stainless steel or equivalent corrosion resistant material. Pipine All piping in contact with spent fuel pit water is austenitic stainless steel. The piping is wolded except where flanged connections are used to facilitate maintenance, l 5-6 . . . _ - - . ~ . . _ _ . . _ - . . _ - - - _ - _ 5.3 Decay Heat Load Calculatigag ; The decay heat load calculation is performed in accordance with ,. the provisions of USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981. For purposes of this licensing application, it is assumed that the pool containa an inventory of 1773 fuel assemblies accumulated through twenty-three previous discharges to the pool (Table 5.3.1). Further, since the decay heat load increases monotonically with reactor exposure time, an upper beund of 1260 full power operation days (nominally 3.5 years) is assumed for all stored fuel. The cumulative decay heat load is computed for the instant corresponding to the beginning of fuel transfer for discharge no. 24 in Table 5.3.1 in the year 2002. The ratio of the cumulative decay heat load due to the inventory of previously stored fuel to the average assembly operating power, # is calculated to be 0.2117466 (Table 5.4.1). In the interest of conservatism, this decay heat load is assumed to remain conste.nt for the duration of the pool temperature evaluations performed in the wake of normal and full core offloads discussed below. I l 5.4 Discharae Scenario.g The following discharge scenarios are examined l Case 1: Normal Full Core Discharge l As shown in Figure 5.4.1, the entire core (193 fuel assemblies) from one reactor unit is transferred to the pool after twelve days of decay in the reactor. The total fuel transfer time is assumed to be 36 hours for i 193 bundles. 113 assemblies of the core are reloaded l into the reactor 30 days after completion of download to the pool. The total reload time is assumed to be 21.1 hours. The total duration of the outage is assumed to l be 60 days. l 5-7 Two discrete analyses have been performed for this case assuming two cooling trains in operation and one cooling train in operation, respectively. We denote these two evaluations as Case la and Case Ib. . Case 2: Back-to-Back Normal Full core Discharge As shown in Figure 5.4.2, 18 days after the end of the outage described in case 1, the other unit has a scheduled outage. Case 3: Unplanned Full Core Offload sixty days after the end of the back-to-back normal refueling outage (Case 2 above), the first unit has an , unplanned shutdown (see Figure 5.4.3). The full core transfer to the pool begins 12 days after the shutdown. All fuel assemblies in the discharged core are conservatively assumed to have 1260 full power days of operation. Detailed data for the three foregoing discharge scenarios are given in Table 5.4.2 and 5.4.3. . 5.5 Enik Pool Temocrature In order to perform the analysis conservatively, the heat exchangers are assumed to be fouled to their design maximum and to have had 5% of their tubes plugged. Thus, the temperature effectiveness, p, for the heat exchanger utilized in the analysis la the lowest postulated value calculated from heat exchanger thermal hydraulic codes. p is assumed to remain constant in the calculation, even though the thermal effectiveness of the fuel pool cooler increases with the rise in the pool water temperature. l The mathematical formulation can be explained with reference to the simplified heat exchanger alignment of Figure 5.5.1. Referring to the spent fuel pool / cooling system, the governing differential equation can be written by utilizing conservation of energy: 5-8 dT C = QL - Qux (5-1) dr OL = Pcons + Q (t) - Ogy (T, ta) where: C: Thermal capt. city of the pool (net water volume times water density and times heat capacity), Btu /'F. QL Heat load to the heat exchanger, Btu /hr. Q(t): Decay boat generation rate from recently discharged fuel, which is a specified function of time, r, Btu /hr. Pcons = Pot Heat generation rate from "old" fuel, Btu /hr. (Po = average assembly operating powor, Btu /hr., see Table 5.4.1) QHX: Heat removal rate by the heat exchange', r Btu /hr. i QEV (T,ta): Heat loss to the surroundings, which is a function of pool temper.ature T and the building ambient air temperature tas Btu /hr. QHX is a non-linear function of time. If we assume the temperature effectiveness p to be constant (which is a reasonable assumption if the flow rates through the exchanger , are unchanged), then Qax can be written in terms of effectiveness p for subsequent analysis QHX " Wt Ct p (T - ti) (5-2) I to - ti
- p= '
l T - ti l where: Wt Coolant flow rate, Ib./hr. Ct Coolant specific heat, Btu /lb.*F. 5-9 i p Temperature effectiveness of heat exchanger. T: Pool water inlet temperature, 'F ti Coolant inlet temperature, 'F to: Coolant outict temperature, 'F p is obtained by rating the heat exchanger on Holtec's proprietary thermal / hydraulic computer code STER. Q(r) is specified according to the provisions of "USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981. Q(r) is the total heat generation rato from the newly discharged fuel assemblj es in the pool. Q(t) increases as additional avsemblies are transferred to the pool and reaches its maximum value -at the instant when the last bundle is transferred. After that, Q(r) drops monotonically with time. QEV is a non-linear function of pool-temperature and ambient air temperature. QEV containe-the heat evaporation loss through the pool surface, natural convection from the pool surf ace and heat conduction through the pool walls and slab. Experiments show that-the heat conduction takes-- only about' 44 of the total heat loss-(5.5.1), and therefore, can be conservatively neglected.- The i evaporation heat and natural convection heat loss can be expressed as QEV = m P As + he Am 6 (5-3) wheret m Mass evaporation rate, Ib./hr. ft.2 o P Latent heat of pool water, Btu /lb. l A: Pool surface: area, ft.2 l het Convection heat transfer-- coefficient at pool surface, btu /ft.2 .hr. 'F-0 = T-ta: The temperature difference between pool water and ambient air, 'F i 5-10 The mass evaporation rate, m, can be obtained as a non-linear function of G. We, therefore, have m = hp (0) (Wp - Was) (5-4) where: Wps: Humidity ratio of saturated moist air at pool water surface temperature T. Was: Humidity ratio of saturated moist air at ambient temperature ta ho(0): niffusioncoeffi:lentatpoolwatersurface. ho is a non-linear function of 0, Ib./hr. ft. *F The non-linear single order differential equation (5-1) is solved using Holtec's Q.A. validated numerical integration code ONEPOOL. The initial temperature of the pool water is assumed to correspond to the equilibrium bulk temperature which will exist in the pool in the absence of the newly discharged b :,tch and neglecting evaporation. ..s is obvious from heuristia reasoning, numerical computations show that the calculated maximum pool water temperature is rather insensitive to the initial pool water ! temperature value utilized. Therefore, a rigorously accurate value of the initial pool water temperature is not necessary for this analysis. Figures 5.5.2 through 5.5.4 provide the bulk pool temperature profiles for the discharge scenarios described in Section 5.4. Figure 5.5.5 through 5.5.7 show the transient heat load of case 1 to Case 3. Table 5.5.1 gives the peak water- temperature, coincident time, and coincident heat load to the cooler and coincident heat loss to the ambient for three cases. If a 15x15 rack is permanently installed in the cask loading area, the- , maximum bulk pool temperature will - be increased by 2.8'F maximum for the increased accommodation of fuel bundles. 5-11 . - - - . - - _ _ . _ _ _ ._ - - - - - - ._- - - _ = _ _ _ . - . _ - . . _ _ . _ . The next step in the analysis is to determine the temperature rise profile of the pool water if all forced indirect cooling modes are suddenly lost and make-up water is provided with a fire hose. Clearly, the most critical instant of loss-of-cooling is when pool voter temperature has reached its maximum value. It is assumed that cooling water is added through a fire hose at the rate of G lb./hr. The cooling water is at temperature, tcool. The governing enthalpy balance equation for this condition can be written as dT (C + G(C )(T - To)) t = Pcons + Q(T+Iins)+G(Ct)(teool-T)-QEV dr where water is assumed to have specific heat of unity, and the time coordinate r is measured from the instant maximum pool water temperature is reached. to is the time coordinate when the direct addition (fire hose) cooling water application is begun. rins is the time coordinate measured from the instant of reactor shutdown to when maximum pool water temperature is reached. T is the dependent variable (pool water temperature). For conservatism, QEV is assumed to remain constant after pool water temperature reaches and rises above 170'F. A numerical quadrature code TBOIL is used to integrate the foregoing equation. The pool water heat up rate, time-to-boil, and subsequent water evaporation-time profile are generated and compiled for safety evaluation. Assuming no make-up water (G = 0), the time-to-boil output results are presented in Table 5.5.2. Figures 5.5.8 through 5.5.11 show the plot of the inventory of water in the pool after loss-of-coolant-to-the-pool condition begins. Figures 5.5.12 through 5.5.14 show the pool water inventory status after loss of spent fuel pool cooling with make-up water added for Cases la, 2, and 3, respectively. Make-up water at 100'F is assumed to be added to the 5-12 pool beginning 10 hours after loss of cooling at a rate of 55 gallons per hour. It is seen from Table 5.5.2 that sufficient time to introduce manual cooling measures exists and the available time is consistent with other PWR reactor installations. 5.6 Local Pool Water and Fuel Qlpsidina Temperatures In this section, a summary of the basis and calculational methodology for local pool water and fuel cladding temperatures is presented. 5.6.1 Basis - In order to determine an upper bound on the maximum local pool water and fuel cladding temperatures, a series of conservative assumptions are made. The most important assumptions are listed , I below: O The fuel pool will contain spent fuel with varying time-l after-shutdown (ts). Since the heat emission falls off rapidly with increasing I thatallfuelassemblies$n,itisconservativetoassume the spent fuel pool are from the latest controlling discharge batch specified for the bulk pool temperature calculations, and they all have had the maximum postulated years of operating time (1260 full power days) in the reactor. The heat emission rate of each fuel assembly is assumed to be equal and maximum. O As shown in the pool layout (Figure 2.1.1), the modules occupy an irregular floor space in the pool. For the l hydrothermal analysis, a circle circumscribing the actual rack floor space is drawn (Fig. 5.6.1). It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal layout pitch. 5-13 o The actual downcomer space around the rack module group varies. The smaller,t ncminal downcomer gap available in the pool is assumed to be the total gap available around the idealized cylindrical rack; thus, the maximum resistance to downward flow is incorporated into the analysis (Figs. 5.6.2 and 5.6.3). o tio downcomer flow is assumed to exist between the rack modules. O The Westinghouse 17x17 fuel assembly has been used in the analysis which, from the thermal-hydraulic standpoint, bounds the case of the vantage 5 Hybrid fuel bundles utilized in the Sequoyah reactor. O lio heat transfer is assumed to occur between pool water and the surroundings (wall, etc.) 5.6.2 Model Descrintion In this manner, a conservative idealized model for the rack module assemblage is obtained. The water flow is axisymmetric about the vertical axis of the circular rack assemblage, and thus, the flow is two-dimensional (axisymmetric three-dimensional). Fig. 5.6.2 shows a typical " flow chimney" rendering of the thermal-hydraulic model. The governing equation to charteterize the flow field in the pool can now be written. The resulting integral equation can be solved for the lower plenum velocity field (in the radial direction) and axial velocity (inscell velocity field), by using the method of collocation. The hy drodynamic leur coefficients l which enter into the formulation of the integral equation are also taken from well-recognized sources (Ref. 5.6.1) and wherever discrepancies in reported values ext.st, the conservative values l are consistently used. Reference 3.6.2 gives the details of mathematical analysis used in this solution process. l After the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature. The knowledge of the overall flow field enables pinpointing of the storage location with the minimum a>:ial flow (i.e, maximum water 5-14 -. - , - , , , . - - ,, ,y outlet temperatures). This is called the most " choked" location. In order to find an upper bound on the temperature in a typical cell, it is assumud that it is located at the most choked location. Knowing the global planum velocity field, the revised axial flow through this choked cell can be calculated by solving 'the Bernoulli's equation for the flow circuit through this cell. Thus, an absolute upper bound on the water exit temperatura and maximum fuel cladding temporature is obtained. In view of the aforementioned assumptions, the temperntures calculated in this manner overestimato t-he temperature rise that will actually occur in the pool. Holtec's computer code THERPOOL*, based on the theory of Ref. 5 . t' . 2 , automates this calculation. The analysis procedure embodied in THERPOoL has been accepted by the 11uclear Regulatory Commission on numerous dockets. The Code THERPOOL for local temperature analyses includes the calculation of void generation due to locall?ed boiling. The offect of the crud layer on the fuel cladding and the clad stress field, when a vo'id exists, are all incorporated in THERPOOL. The peaking factors used in the analysis are given in Table 5.6.1. 5.6.3 Claddina Tegnerature The maximum specific power of a fuel array qa can be given by: 9A = q Fxy (1) [ where:
- Fxy = radial peaking factor q = average fuel assembly specific power l
l l THERPOOL has been used in qualifying the spent fuel pools for Enrico Fermi Unit 2 (1980), Quad Cities I and II (1981), Oyster Creek (1984), V.C. Summer (1984), Rancho Seco (1983), Grand Gulf I (1985), 'Diablo Canyon I and II (1986), among others. 5-15 l i The maximum temperature rise of pool water in the most ; disadvantageously placed fuel asrembly is computed for all loading cases. Having detirmined the maximum local water temperature in the pool, it is now possible to determine the maximum fuel .uladding temperature. A fuel rod can produce Fg times the average heat emission rate over a small length, where Fz is the axial rod peaking factor. The axial heat distribution in a rod is generally a maximum in the central region, and tapers off at its two extremities. The power distribution corresponding to the chopped cosine power emission rate can be written in the standard form as, l n (a + x) q(x) = qA sin h + 2a where h: active fuel length as chopped length at both extremities in the power curve x: axial coordinate with origin at the bottom of the active fuel region The value of a is given by bz a= 1 - 2z where: _ 1/2 1 1 1 2 z= - - + n Fz n Fz n Fz n2 where F, 7 is the axial peaking factor. 5-16 l The cladding temperature Te is governed by a third order differential equatton which has the form of , d3 T d2 T dT ! + at - a2 ~"- " f (X) dx3 dx2 dx j where al, a2 and f(x) are functions of x, and fuel assembly geomotric properties. The above equation is integrated in the
- program TilERPOOL with appropriate boundary conditions. The solution of this differential equation provides the fuel cladding temperaturo and local water temperature profile.
In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit which results in .005 0F -sq.f t.-hr/ Btu of crud resistance, which , covers the entire surface. Table 5.6.2 provides the kay input data for local temperature analysis. The results of maximum local pool water temperature and maximum local fuel cladding temporc.ture are presented in Table 5.6.3. l The local boiling temperature of water is approximately 242'F at 26' below the free water surface and higher at lower elevations. Tho location where the local water temperature reaches its maximum value is dooper than 26' below the free water surface, and thorofore the coincidet.t boiling temperature of water is greater than 242'F. It is soon from Table 5.6.3 that the local pool water temperature is lower than the coincident local boiling point and therefore, localized boiling of pool water will not occur. If a 15x15 rack is installed in the cask loading area, the maximum local pool water temperature will bo increased by 2.6'F and the maximum fuel l 5-17 . . _ = - - - - _. - - - - - - - _ - _ _ _ . _ - _ _ . _ _ _ . - _ - - l 4 cladding temperature will be increased by 2.7'F. Therefore, it is also concluded that the addition of a 15x15 rack in the cask area does not lead to localized boiling of pool water. Finally, it is noted that the fuel cladding temperature is ' considerably lower than the temperatures to which the cladding is subjected inside the reactor. Therefore, it is concluded that there is sufficient margin against fuel cladding failure in the spent fuel pool. 5.6.4 B19. eked cell Analysis Calculations are also performed assuming that 50% of the top opening in the thermally limiting storage cell is blocked due to a horizontally placed (misplacud) fuel assembly. The corresponding maximum local pool water temperature and local fuel cladding temperature data are also presented in Table 5.6.3. It is se'en while both the local pool water and fuel cladding temperature sustain a slight increase, considerable margin against localized boiling remains. Finally, the maximum local water temperature will be increaoed by 2.8'F and the maximum fuel cladding temperature ~ will be increased by 2.7'F when 225 more fuel bundles are l installed in the cask loading area at a future date. These minor increases do not alter the foregoing conclusion. The condition of localized nucinate boiling of the pool water or potential for fuel cladding damage are, therefore, ruled out for the Sequoyah pool, l 5.7 References for Section 5 5.5.1 Wang, Yu, " Heat Loss to the Ambient from Spent Fuel Pools: Correlation of Theory with Experiment", Holtec Report HI-90477, Rev. O, April 3, 1990. 5.6.1 General Electric Corporation, R&D Data Books, " Heat Transfer and Fluid Flow", 1974 and updates. l l l l 5-18 i 5.6.2 Singh, K.P. et al., 'Hothod for Computing the Maximum Water Temperature in a Fuel Pool Containing Spent Nuclear Fuel", Heat Transfer Engineering, Vol. 7, No. 1-2, pp. 72-82 (1986). W l i i 5-19 l Table 5.3.1 . DISCllARGE SCHEDULE FOR THERMAL-HYDRAULIC AllALYSIS (All fuel assemblies are assumed to-have 1260 day full power operation in the reactor.) DISCHARGE CUHVLATIVE , tit!MBER M M 1}( THE POOL. M 1 1 68 68 9/82 2 2 68 136 7/83 ; 3 1 72 200 2/84 4 2 68 27(i 9/84 5 1 72 3it 8/85 6 2 80 428 1/89 7 1 81 509 3/90 8 2 80 589 9/90 I 9 1 72 661 10/91 ' 10 2 72 733 3/92 11 1 80 813 4/93 12 2 80 893 9/93 13 1 80 973 10/94 14 2 80 1053 3/95 15 1 80 1133 4/96 16 2 80 1213 9/96 17 1 80 1293 10/97 l 18 2 80 1373 3/98 19 1 80 1453 4/99 20 2 80 1533 9/99 21 1 80 1613 10/00 22 2 80 1693 3/01 23 1 80 1773 4/02 24 2 30 1853 9/02 25 1 80 1933 11/02 26 2 193 2126 3/03 l t l l 5-20 w l Table 5.4.1 FUEL SPECIFIC POWER A!!D POOL CAPACITY DATA Total not water volume of Pool 35453 ft3 Averaga Operating Power of a Fuel Assembly, Pos 60.343E6 Btu /hr. Dimensionless decay power of "old" discharges, pt 0.2117466 4 4 5-21 .._ _- _ . _ .- . . _ ~ , _ . . ._. - i i l Table 5.4.2 DATA FOR SCENARIOS 1 through 3 cA_st nuMarn PARAMETERS la lh 2 2 Fool ther:nal capacity, c 2.17 2.17 2.17 2.17 106 stu/'r No. of cooling Trains 2 1 2 2 coolant Inlet Temp., t1 95 95 95 95 *r , coolant Flow Rate / 1.49 1.49 1.49 1.49 cooler, Ws, 106 lb./hr. ruel Pool Water 1.14 1.14 1.14 1,14 Flow Rate / cooler, 106 lb./ht. Temperature 0.3068 0.3068 0.3068 0.3068 Effectiveness / cooler, p 5-22 i Table 5.4.3 DATA FOR SCENARIOS 1 THROUGH 3 i i Tizne Af ter shutdown when ruel ruel case Discharge No. of Transfer Regins, Transfer Exposure No. ID Assemblies hrs. Titne, hrs. TiJne, hrs. la offload 193 288 36 30240-er ib Reload 113 1044 21.1 30240 2 offload 193 288 36 30240 Reload 113 1044 21.1 30240 i 3 offload 193 288 36 30240 l 5-23 i f Table 5.5.1 POOL BULK TEMPERATURE AND HEAT LOAD DATA Time coincident Taax to Tsax, hrs. Coincident Coincident Max. Pool (after Evaporation Case Cooler Duty, Bulk Temp., reactor Heat Loss, No. 106 Btu /hr. 'r shutdown) 106 stu/hr. la 39.326 138.02 332 0.378 lb 26.522 174.91 336 3.061 2 43.501 142.59 332 0.521 3 47.230 146.65 331 0.674 l l 5-24 i Table 5.5.2 TIME-TO-BOIL FOR VARIOUS DISCHARGE SCEllARIOS [1] (WITH NO MAKE-UP WATFR) Time-to-Boil Case liumber (hours 1 r* IHours1 f21 la 5.50 34 - Ib 3.42 56 2 4.71 33 3 4.12 30 1 1 [1] Time coordinate starts from the instant of loss of cooling. [2] r
- is the time clapsed subsequent to the loss-of-cooling when the pool water level drops' to within 10' of the top of the active fuel stored in the fuel racks.
l 5-25 P t t Table 5.6.1 PEAKING FACTOR DATA Radial Bundle Peaking Factor 1.70 Total peaking factor 2.70-D 4 1 5-26 Table 5.6.2 DATA FOR LOCAL TEMPERATURE Type of Fuel Assembly PWR Fuel Cladding Outer Diameter, inches 0.374 Fuel Cladding Inside Diameter, inches 0.329 ; Storage Cell inalde Dimension, inches 8.75 Active fuel length, inches 144 , No. of Fuel Rods / Assembly 264 OperatingPowerperFuelAssembly 60.343 ' Po x 10 , Btu /hr cell pitch, inches 8.97 Cell height, inches 168 Plenum radius, feet 25.33 Min. Bottom height, inches 4.75 Min. gap between pool wall 4.25 and outer rack periphery, inchec y. I i 5-27 t m --n----.'-q . m.eret--- e e w. s.w-4 .--r -.,,-r- - -w- .-- c---r -, e Table 5.6.3 LOCAL AllD CLADDIllG TEMPERATURE OUTPUT DATA FOR Ti!E MAXIMUM BULK POOL TEMPERATURE CollDITIO!1 Maximum Maximum Local Fuel Cladding Condition Water Temo., 'F Temo., 'F llo blockage 227.0 249.4 50% blockage 230.0 251.7 l l 5-28 i s t 30 DAYS - = FULL CORE OFFLOAD -a o RELOAD o REACTOR 113 N h 12 DAYS } 193 ASSEMBUES is1 x tt O k , k
- e w
s 3 a
- TIME 80 DAYOUTAGE _
SHUTDOWN FIGURE 5.4.1 NORMAL FULL CORE DISCIlhRGE SCENARIO - CASE 1 O I a CASE 1 DISCHARGE CASE 2 DISCHARGE o O FULL CORE O! ROAD MW MM REACTOR OFFLOAD M CTOR f 3 '
- 1 3G DAYE N DAYS rw -
e _ . _ _ & TIME 60 DAY OUTAGE OF ONE Ur#T 60 DAY OUTAGE OF THE OTHER UNIT - 1. _18 DAYS _ _ SIMJTDOWN SHUTDOWN FIGURE 5.4.2. BACK-TO-BACK NORMAL FULL CORE DISCIIARGE - CASE 2 J 4 h 4 i ., I . ? 1 , .i i e
- - (
) .
- t I O BACK-TOaACK NORMAL OUTAGES FASE 1 & CASE 2) M FLOAD
- o ;
? A i
- tu ,
- ,3 80AsseMeuEs (( ,
8 80 ASSEMBUES l II 'f .- !E 1 5 W ~ 4 12 DAYS 4 s i W l- ta. ! 80 DAY OUTAGE 1E'3AY 80 DAY OUTAGE 8C DAYS ; = = = = = : : : r_ i !, SHUTDOWN SHUTDOWN UNSCFEDULED M1HE 1st UNIT &TtE 2nd UNIT SHUTDOWN s WTHE 1st UNIT f l
- 3. .
i ! r i ! 1 l PIGURE 5.4.3 ' UNPLANNED' FULL CORE OFFLOAD- - CASE 3-i ; l l l p. t } , 4 i j !- J -* e---,---__e.a. -4 A e.-44 4_a 4 m:: .,a3,.i. ,a .s 6 a4 .# , so -- - --- - - - - - - - - - h 1 4 ) EVAPOMTION HEATLOSS SPENT FUEL POOL = 9 HEATEXCHANGER T, P -g3 COOLANT tg % L t FIGURE 5.5.l' PCOL BULK TEMPERATURE MODEL 32 _.._, ,--s.,, c-,.. -y ..,ec.. w. v a wmN *
- e '"' d
- W I -*-*P --B-p
i 190 ; l $ SEQUOYAII SFP DISCIIARGE CASE 1 & 2, ONE COOLING TRAIN ; .180 l - . 170--
- u. 160 -
*n o' _ i u W _- t a 150 - o _ o .
- n. !
3 _ t a 140 - m 2 , _ a i ! 130 - l ~ r ~ ] t 120 - t : _ t 110 -- i i i s. i i i i i e i i ;e i i i , i i i . . ., iiie i ;ii,,,ii,,iii 0 250 500 750' 1000 1250 1500 1750 '2000 2250 i TIME AFTER FIRST SHUTDOWN OF THE REACTOR, HR FIGURE.5.5.2 ; i I t 145 . 140 SEQUOYAH SFP DISCHARGE CASL 1 & 2, 'IVO COOLItG TRAINS . 135_ - 130 _ .c u_ _ Y a.- _~ %- 3 125 - L - a , o - o -
- n. 120 _ i 3
o 115 2 .J t 110 - l 105 - I _- i 100 , ,,,,,,,,,,,,,,,, ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, 0 250 500 750 1000 1250 1500 1750 2000 2250 TIME AFTER FIRST SHUTDOWN OF THE REACTOR. HR FIGURE 5.5.3 I r 150 l ._ \ 145 I SEQUOYAll SFP DISQlARGE CASE 2 & 3, MD COOLItB TRAltG 1 140 _ t 135_
- u. -
n.- _- T 3 130 - w g _ sn _ _s > o ~ o -
- n. 125 - j
~ 3 O t 120 : i 115 - ~ .i i - 110_ : ~_ 105 i,,,gsiiigsii gein,iiiiigsini;iii ;iiiigisiigiii ;iisigi ,,; .;,j,, 0 250 500 750 1000 1250 1500 .1750 2000 2250 2500 2750 3000 3250 TIME AFTER FIRST SHUTDOWN OF THE REACTOR. HR ; FIGURE 5.5.4 ( b 4.50E+007 .. 1 SEQUOYNI SFP DISCIIARGP, CASC 1 & 2, ONE COOLItG TRAIN 4.00E+007 - 3.50E F007 ~ 3.00E+007 _ I _ N T' 3 m 2.50E+007 [ w yi [ HEAT LQADS TO THE COOLERS Q - i g 2.00E+007 - - w ^ b ~ 1.50E+007 _ _ l 1.00E+007 - 1 5.00E+006 I _ EVAPORATION HEAT LOSS 4 _ ~ 0.00E+000 iii,,,,,,,,,,,,,,,,,),,,,,,,,,,,,,,,,,,,,,,,,,, 0 250 500 750 1000 1250 1500 1750 2000 2250 TIME AFTER FIRST SHUTDOWN OF THE REACTOR. HR
- FIGURE 5.5.5
i 1 i 4.50E+007 .- SEQUOYAll SFP DISOIARGE CASE I a 2, HD COOLIfC TRAlts i - , 3.50E+007 - 3.00E+007 ( , .x - I - , .N m a 4 g 2.50E+007.f w '~ vi HEAT LOADS TO RiE COOLERS . 'Q 2 o ' 2.00E+007 i W , 1.50E+007 -- . _~ l i k - 1.00E+007 _ , i 1 , 5.00E+006 -~ i i . EVAPORATION HEAT LOSS l 0.00E+000 ' ,,,,,,,,,j,,,,,,,,,,,,, , , , , , , , , , , , , , , , , , , , , , , , , 0' 250 500 750 1000 1250 1500 1750 2000 2250 , TIME AFTER FIRST SHUTDOWN OF THE REACTOR. HR l d FIGURE 5.5.6 d 5.00E+007 - SEQUOYAll SFP DISOfARGE CASE 2 & 3, 'IID COOLIIC 'IIGIIG 4.50E+007 h ~ 4.00E+007 - , 3.50E+007 ( tr _- Q 3.00E+007 - a ~ . g HEAT LOADS TO THE COOLERS d, vi 2.50E+007 - 9 : :
- 2.00E+007. - !
~ I _ t 1.50E+007 - i 1.00E+007h ~ 5.00E+006 EVAPORADON HEAT LOSS em e l l l l ll l l l lI l 5 5 l5 I O $ l5 0 l 5 l5 5 5 Ill I I I lI I I I l5 5 5 0 l5 5 I E lE I I I lI O I 5l4 $ l 0 250 500 750 1000 1250 1500 1750 2000 2250 2500 2750 3000 3250 TIME AFTER FIRST SHUTDOWN OF THE' REACTOR, HR j FIGURE 5.5.7 l 2.75E+005 _ . 2.50E+005 SEQUOYAH SFP LOSS OF COOUNG SCENARIO - CASE la 2.25E+005 { O 2.00E+005 ( si C .\ l@ 1.75E+005 -~ o _ E- - E 1.50E+005 ( y u. e r m
- o 3 1.25E+005 .--
o ~ 1.00E+005 - e - E a o 7.50E+004 -- o - 1 5.00E+004 - , . F- . ~ 2.50E+004 _ _ ~ 0.00E+000 iii,,,,,iii,,,iiiii i iiiiii:,i, . i,i ,,,,iiii:ii >>>>>g 0- 10 20 30 40 50 60 Time After Max. temp. has beer) reached, hr
- FIGURE 5.5.8
1 2.75E+005 -- . ~ 2.50E+005 ~
- SEQUOYAH SFP LOSS OF COOLING SCENARIO - CASE lb
~ 2.25E+005 O 2.00E+005 -[ s : .E - ,g 1.75E+005 - o _ E - e tr 1.50E+005-( T.. -b - o o 1.25E+005 - 3: _ + - O .- e 1.00E+005 _ E a 3 7.50E+004 - - ~ o q 5.00E+004 -- q H. _ 2.50E+004 ( a . _ 0.00E+000 ii.. . .ii.... iiii .. .iii,,, ,, .iii,i,iiii i .. iiei
- O 10 20 30 40 50 60 Time After Max. temp. has been, reached, hr t
a FIGURE 5.5.9 ] 2.75E+005 - 2.50E+005
- N '
SEQUOYAH SFP LOSS OF COOUNG SCENARIO - CASE 2 2.25E+005 ( O 2.00E+005 ( ~ ch .E - ,g 1.75E+005 - O _ E - , E 1.50E+005 5 . 11 - - .s : . o 1.2SE+005 - 3 _ o I 1.00E+005 - e) - E a o 7.50E+004 - : O .- 5 5.00E+004 -- H 2.50E+004 - 4 f. ~ 0.00E+000 iii iiii iiiiii ii . . 2 i . iiiiiiiei -iiii .a ii...iq 0' 10 20 30 40 50 60 Time After Max. temp. has been reached, hr FIGURE 5.5.10 1 2.75E+005 - 2.50E+005 N SEQUOYAH SFP LOSS OF COOLING SCENARIO - CASE 3 2.25E+005 - O 2.00E+005 -~ e, C @ 1.75E+005 - O _ -E - Y w cE 1.50E+005 -~ w u- - .3 : O 1.25E+005 - 3 _ g - 1.00E+005 C e - E : ~ a - o _7.50E+004 - , > ~ -o .: 5 5.00E+004 - l- _ '~ L 2.50E+004 - ~ 0.00E+000 i., , , s ,,,,,,,,,,,,,,,,,si_ii, ,,: ,,,,,,,,,,,,,,,,,,,i 0 10 20 30 40 50 Time After Max. temp. has _ been reached, hr FIGURE 5.5.11 4 2.75E+005 - . SEQUOYAH SFP LOSS OF COOLING SCENARIO - CASE la 2.50E+005 - (Add makeup water of 100 F at 55 gpm s 10 hours ofter loss of cooling) ~ 2.25E+005 O 2.00E+005 5 cn _ C _ ]E 1.75E+005 - o _ E - y E 1.50E+005 ( ' c u o c 3 1.25E+005 - ~ o ~ 1.00E+005 - 0 E-s a - o 7.50E+004 '- . o .- 5 5.00E+004 - I- _ 2.50E+004 ( ~ 0.00E+000 ....,. .., ,,ii,iii iiiiiiiii ... .. i.. .....i,ii ...i 0 50 100 150 200 250 300 Time- After Max. temp. has been . reached, hr FIGURE 5.5.12 r 2.75E+005 - . ~ SEQUOYAH SFP LOSS OF COOUNG SCENARIO - CASE 2 2.50E+005 ( (Add makeup water of 100 F at 55 gpm 10 hours 'after loss of cooling) - 2.25E+005 2 ~ 0 2.00E+005 5 ch ,~ C _ ] 1.75E+005 - O _ . E - l tE1.50E+0055 y b : 0 3: 1.25E+005 ( - i E ~ 1.00E+005 ~ 0 - E - a _ o 7.50E+004 - O - 5 5.00E+004 - F- _ 2.50E+004 ( 1 ~ 0.00E+000 . . . .. . ,,......... . ....... . .. .......i O 40 80 120 160 4 Time' After Max. temp. has been reached, hr FIGURE 5.5.13 . !:l ;- 0 2 .1 3 . E . S . A . C . 0 i0 .1 r m . h Op I Rg ) . A g . d N5 n e E 5il . C t o . h Sa oc i O c .S a i GFf e r N I o . L0 O0s s n e O1 C f o l . e 4 1 F oe r . b 5 i Ort s0 5 ef 6 s Sa t a . a E R Sws O r . h U G Lp uo . p . I u . F Pe Fk h . i m S a0 i e 0 t H m1 i 4 A . . Y d . x Od . a U (A . M Q . E . r S . . e i0t 2f . A e i i m . T -(: : - ~ : :- _ 0 - _: - _ - - : _: _ - ___- __~ 5 5 5 5 5 5 5 5 4 4 4 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 + + + + + + + + + t + + E E E E E E E E E E E E 5 0 5 0 5 0 5 0 0 0 0 0 7 5 2 0 7. ' 5 2 0 5 0 5 0 2 2 2 2 1 1 1 1 7 5 2 0 O eE.c6EE u3o% ' o cEa5> oyI TO J .* , . .p. a na ,. 2 -. \\ +% / g, v l g\ \ v s :! x-5-46 ..e,m.a -.a,4 _m..,,mm..a.m wa_,.;e J.>mA4m.,.A. r.s, J_ 2gs_,a gy m;g-.m,pr.. p 4-_m ._ . , 4 a A,,44-..am,u,- A_ 4 p @g9 44 -4 I / I / /I"7 l r/u l m IIN g ,F. W 9 C 2 H C d., 6 E - C C I 1 - C HE.AT" AC CITICN t z 3ll I C y A i I ) . 1 j ,, / / (w gi / /1 / j vn vt / FIGURE 5.6.2 THERMAL . CIIIMNEY - FLOW MODEL-5-47 ,. .e ri .> - .i. .. ..J . c. . 4 - s. 5 3 9 . *g. . . . ' , , 4 s
- s. <(
- s- .
w ..e 1-( *.%f . 5 3 .... n>ex .s-9 .J. p i . ,- . T \'.: - Tii n m v \s P P P P v 4.-.I ;4 i i i i i i i
- 1. .
1 . . * -\.
- y. .
1 J .h. 1.J .+7 ,.A (9 ,. , p.. . .t \ .* .-' .h c j,*cc'nN e, r.tsa 4 * .g . . , .' . 5 .. g .\ ). , . 9 . . J. ,..t
- i. Y. W % W **
i . .., f, g.* , .g . 7. - .3 - ..y, . ,. g .T 3 . ., ., . _ a. . . .A . . Q, . g. .. - 1 * . . v . , , .. .
- a. . .. 4 . ,A
- 7 t
.7. . 7 *J s.* . 7 . Pt.ENUM FIGURE-5.6.3 -CONVECTION CURRENTS IN THE POOL 5-48 . _ - - _ . . . . . _ . - _ . . . - . . - . ~ , - . . . , . . . _ . _ _ . . . , . _ _ _ . _ - , 6.O' STRUCTURAL /SIISMIC CONSIDERATIONS 6.1 Introduct19n This section contains analyses to demonstrate structural adequacy of the high density spent fuel rack design under-seismic loadings postulated for the plant spent- fuel pool. Analyses _ and subsequent evaluations are in compliance with the requirements of the - OT Position Paper,_Section IV [6.1.1),'and follow the USNRC Standard Review Plan (SRP) [6.1.2). The dynamic analyses employ a-time-history simulation code used in previous licensing offorts listed in Table 6.1.1. This section provides details of the method of analysis, modeling assumptions, numerical convergence studies and parametric evaluations performed to establish the required-margins of safety. Results reported herein show that- the high density spent fuel racks are structurally and kinematically adequate to meet requirements defined in ref erences - [6.1.1), 16.1. 2 ) , . and- [ 6.1.' 3 ) - with large margins of safety.- l 6.2 Analysis Outline A spent fuel rack is a seismic. category I structure [6.2.1). Furthermore, it is a free -standing structure consisting of discrete storage-cells which -are loaded with-free standing - fuel assemblies. As a result, the response of a. rack module to seismic inputs is highly nonlinear . involving . a complex. - combination of motions - (sliding, rocking, twisting,' 'and turning), 'resulting in-impacts _and friction effects. ' Linear methods such Las- modal analysis and response- spectrum techniques cannot accurately simulate the structural: response of such a_ highly nonlinear l structure to seismic excitation. A correct simulation-is obtained only - by direct - integration 1of the nonlinear equations of motion-using actual pool slab acceleration-time-histories to provide the loading. Therefore,- the initial step _ in. spent- fuel rack 6 . , - - . - - , ., ..o., n ., , . - , , - . . m. .,,.-e-- ,. -,,y--. . , - , , . .r-w..s --*.,2-e .,w ..s, w- ..-w-, .-e,--, , qualification is to develop synthetic time-histories for three orthogonal directions which comply with the guidelines of USNRC SRP 3.8.4 (6.1.2]. In particular, the synthetic time-histories must meet the criteria of statistical independence and enveloping of the design response spectra. As stated above, a free standing spent fuel rack, subject to a seismic loading, executes non-linear motions - even when isolated. The motion of an array of closely spaced racks in the spent fuel pool involves additional interactions due to fluid coupling between adjacent racks and between racks and adjacent valls. Further mechanical interactions between racks occur if rack-to-rack impacts take place during the event. To demonstrate structural qualification, it is required to show that stresses are within allowable limits and that displacements remain within the constraints of the contemplated design layout for the pool. This implies that impacts between rack modules, if they occur, must be confined to locations engineered for this purpose, such as I the baseplate edge for these fuel racks. Similarly, rack-to-pool
- wall impacts, if engineered into the rack design (not contemplated I
for these racks) must be within stipulated limits. Accurate and reliable assessment of the stress field and kinematic behavior of the rack modules calls for a ecmprehensive and conservative dynamic model which incorporates all key attributes of the actual structure. This means that the model must feature the ability to execute concurrent sliding, rocking, bending, twisting and other motion forms available to the rack modules. Furthermore, it must possess the capability to effect the momentum transfers which occur due to rattling of the fuel assemblies inside the. storage l cells and impacts of support pedestals on the bearing pads. Finally, the contribution of the water mass in the interstitial spaces around the rack modules and within storage cells must be modeled in an accurate manner because erring in the quantification of fluid coupling on either side of the actual value is no 6-2 y- - - - - - y- - guarantee of conservatism. Similarly, the coulomb friction coefficient at the pedestal-to-pool liner (or bearing pad) interface may lie in a rather wide range and a conservative value of friction cannot be prescribed a' priori. In fact, a perusal of results of rack dynamic analyses in numerous dockets (Table 6.1.1) indicate that an upper bound value of the.coef -: lent of friction, y, often maximizes the computed rack displacements as well as the equivalent elastostatic stresses. Further, the analysir must consider that a rack module may be fully or partially loaded with fuel _ assemblies - or entirely empty. The pattern of loading in a partially loaded rack may also have innumerable combinations. In short, there are a large number of parameters with potential influence on the rack motion. A comprehensive structural evaluation should deal with all of these without sacrificing conservatism. The 3-D single rack dynamic model introduced by Holtec International in the Enrico Fermi Unit Two rack project (ca. 1980) and used in some twenty rarack projects since that time (Table 6.1.1) tackles the'above mentioned array of parameters in a mos t appropriate manner. The details of this classical methodology are published in the permanent literature (6.2.2) and have been widely replic.ced by other industry groups in recent years. Briefly I speaking, the single rack 3-D model handles the array of variables as follows: l-Interface Coefficient of Friction Parametric runs are made with' upper bound and lower _ bound values i of the coefficient of friction. l The limiting values are based on exper: .atal data. i l Imoact Phenomena i ! Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface. l l 6-3 c - - ms .-- , , . , - . , , --rv.u, I l l Fuel Loadino Scenarios The fuel assemblies are conservatively assumed to rattle in unison which obviously exaggerates the contribution of impact against the cell wall. The different patterns of possible fuel assembly leadings in the rsck are simulated . by orienting the center of gravity column of the assemblage of fuel assemblies with respect of the module geometric centerline in an appropriate manner. Fluid Coucling The contribution of fluid coupling forces is ascertained by prescribing the motion of the racks (adjacent to the one being analyzed). The most commonly used assumption is that the adjacent racks vibrate out-of-phase with respect to the rack being analyzed. Despite the above simplifying assumptions, targeted for accuracy and conservatism, a large menu of cases is run to foster confidence in the calculated safety margins. All safety analyses reported in the previous dockets (Table 6.1.1) over the past decade have relied on single rack 3-D model. From a conceptual standpoint, all aspects of the 3-D single rack model are satisfactory except for the fluid coupling effect. One intuitively expects the relative motion of the free standing racks in the pool to be poorly correlated, given the random harmonics in the impressed slab motion. Single rack analyses cannot model this interactive behavior between racks. However, as described later, analytical and experimental research in this field has permitted rack analyses to be extended to all racks in the pool simultaneously. Holtec International had successfully extended Fritz's classical two body fluid coupling model to multiple bodies and utilized it to perform the first two dimensional multi-rack analysis (Diablo Canyon, ca. 1987). Subsequently, laboratory experiments were conducted to validate the multi-rack fluid coupling theory. This technology was incorporated in the computer code DYNARACK which now could handle simultaneous simulation of all racks in the pool. This development marked a pivotal expansion in the rack structural modeling capability and was first utilized in Chin Shan, oyster Creek and - Shearon Barris plants (6.2.3). The 53 hole Pool Hulti-Rack (WPMR) 3-D analyses have corroborated thc uncanny accuracy of the single rack 3-D solutions in predicting the maximum structural stresses. The multi-rack analyses also serve to improve predictions of rack kinematics. In order to ensure utmost confidence in the results of structural safety analyses, we present results for both single rack 3-D and Whole Pool Multi-Rack (WPMR) 3-D analyses. The intent of this parallel approach is to foster added confidence and to uncover any peculiarities in the dynamic response which are germane to the structural safety of the storage system. 6-4 .l -l . .i In the following, we summarize _ the sequence of ' model' development i and analysis steps that are undertaken. Subsequent subsections model detail,-- limiting criteria for stress and provide displacement, and results of the analyses'. i
- a. Prepare three-dimensional dynamic models - of indivi' dual t fuel racks which embody all elastostatic characteristics and structural nonlinearities of the plant specific free standing rack modules.
i
- b. Perform 3-D dynamic. analyses on limiting module geometry types (from all those.- present in;the spent fuel pool) and include- various- physical conditions such as coefficient of friction, extent of cells contai(ning fuel !
-assemblios,-and proximity of_other racks).
- c. Perform -detailed ' stress . analysis for the limiting-case of all . the : dynamic analysis runs made in the foregoing steps. Demonstrate compliance :with ASME Code ' Section III, subsection NF (6.1.3) limits' on stress and displacement.
- d. Perform a degree-of-freedom (DOF)-reduction procedure on the single rack 3-D model such that k.inematic responses calculated by the Reduced DOF model- (RDOFM) are in agreement with responses obtained using the baseline single rack models of step-(b). The RDOFM is also truly three-dimensional,
- e. Prepare - a . whole pool multi-rack dynamic model whicI. -
includes the RDOFM's - of . all rack - modules in the' pool, and includes-all~ fluid coupling interactions among them, 4 as well as fluid coupling interactions between racks and pool walls.- This 3-D simulation is referred to as a- - -Whole Pool ~ Multi-Rack (WPMR)-model. l f. Perform 3-D Whole Pool Multi-Rack (WPMR) analyses to l demonstrate that . all kinematic criteria for the' spent fuel rack modules are satisfied, and that . resultant structure loads confirm the validity of the - structural qualification. The principal: kinematic- criteria are (i) _ i no rack-to-pool- wall impact, and : (ii) - no rack-to-rack ' impact in the cellular region of_the racks. c 6-5 _ _ . . _. . _.u . .. ._ -... _ ,.. __ _ ._,._.._. _ _.. ,_. _ ,.,._._._, 6.3- Artifleial Time-Histories. Section 3.7.1 of the SRP [6.1.2) provides guidelines for establishing seismic time-histories. Subsection 3.7.1.II.l.b gives applicable criteria for generation of time-histories from design response spectra. Response spectra for use in the structural qualifications are given-in [6.3.1]. A generated artificial time-history is acceptable if the response spectrum in the free field at the specified level of the site, obtained from the generated time-history, envelops the design response spectrum at the same location for all damping values used in the analysis. The acceptance criterion for spectrum enveloping is that no more than five points of the spectrum obtained from the time-history fall below, and no more than 10% below,- the design response spectrum. The SRP states that an acceptable method of comparison is to choose a set of frequencies such that each frequency is within 10% of the previous one. The nature of the spent fuel rack structure is such that primary response is to excitations above 5-8 HZ. Within the 5-33HZ range, discrete check points are established from the above 10% criterion. Generated artificial time-histories must also be statistically independent. Any two time-histories are considered to be statistically independent 12. their normalized correlation-coefficient is_.less than 0.15. The computer code GENEQ [6.3.2) is . used herein to generate-synthetic, statistically. independent time histories from the response spectra. For this plant, two different response spectra sets are considered i for the SSE event. These are designated as the " site specific" spectra for the SSE condition, and the " design basis" spectra for i the DBE condition. The " design basis" spectra can be divided by 6-6 .e- , , y-- ,# w - 7 e , - - . , . , _ , , - m - . , . , , --r.m.-..,,---,, -- 2.0 to obtain the OBE " design basis" spectra. There is no OBE for the site specific earthquake. The seismic design approach for the new spent fuel racks is to be based on the spectra giving higher stresses and displacements appiled to analyses. To establish the governing earthquake spectra set, three statistically independent time histories, which meet the enveloping requirements, are developed from each response spectra set provided. The governing spectra to be used in subsequent analyses is then established by applying each set of time histories to a typical rack' configuration. The evaluation leads to the followings
- a. SSE response spectra for " site specific" envelope the corresponding spectra for " design basis".
- b. Time histories developed from _ "alte specific" spectra give larger rack displacements and stresses than corresponding time histories developed from " design basis" spectra when both sets of time histories are - used as input loading functions for a typical spent fuel rack.
Based on the above, the SSE for the " site specific" spectra are set as governing for evaluation of structural behavior of the spent fuel racks. Two additional dets of three time histories have been developed from these governing spectra. All nine time histories (3 sets) so generated meet the test of statistical independence with each other and the test -that re-generated response spectrum from each time history meet the enveloping ! crf;.ria. Three independent sets of' time histories from the governing response spectra are developed to evaluate the sensitivity of rack response to different seismic events that all meet the seismic acceptance criteria. Since the numerical generation of time histories from a .given response spectra set does not lead to a unique set of time histories, the use of three different sets- of time histories in the fuel rack analyses provides increased assurance that conclusions reached concerning the safety of the structure are conservative. The results of the foregoing time history development are summarized in the following figures and tables. 6 4
- 4. . _ . ,- ,
Figures 6.3.1-6.3.3 and Figures 6.3.4-6.3.6 are reproduced from [6.3.1) and show the " site specific" and " design basis" response spectra, respectively. Figures 6.3.7-6.3.9 show the time histories generated from the " design basis
- spectra, and Figures 6.3.10-6.3.12 show the enveloping of the original response spectra by the regenerated response spectra. Table 6.3.1 gives the result of the correlation analysis of the set of time histories from the " design basis" spectra. The time histories for the " design basis" set of spectra meet the enveloping and the statistical independence criteria. Figures 6.3.13-6.3.15 show the initial " set 1" of time histories developed from the " site specific" spectra. Figures 6.3.16-6.3.18 illustrate the comparison of re-generated response spectra with the original spectra. Table 6.3.2 gives the results of the correlation analysis of the set of " site specific" time histories. The time histories for the " site specific" set of spectra also meet the enveloping and the statistical independence criteria.
To determine which set of time histories leads to a more severe set of stresses and displacements in spent fuel racks, two analyses have been carried out. The first is a comparison between the two sets of regenerated response spectra. Figures 6.3.19-6.3.21 show the two sets of spectra plotted on the same graph and demonstrate that the regenerated " site specific" set envelopes the regenerated " design basis" set at appropriate check frequencies, meaning that the " site specific" earthquake is more severe in all three orthogonal directions than the " design basis" earthquake. The second is a comparison single rack analysis. A spent fuel rack module, typical of the size and weight to be installed at sequoyah, has been subjected to both sets of time histories. For the comparison rack analysis, 24 structural damping is used for the " design basis" time history loading, and 41 structural damping is used for the " site specific" time history loading. These damping values are in accord with permissable damping levels for the events. The results of the analyses demonstrate that the " site 6-8 l specific" set of spectra leads to time histories which will govern ; the structural integrity- analysis of the -spent = fuel- racks. With l this conclusion available, then _ for purposes of evaluating - the effect of non-uniqueness of generated time hhtories, two additional sets of time- histories were generated from the " site specific" response spectra. Figures 6.3.22-6.3.24 show the time histories generated for " set 2", and Figures 6.3.25-6.3.27 compara re-generated spectra with original _ spectra for " set 2".-Similarly, Figures 6.3.28-6.3.30 and Figures 6.3.31-6.3.33 show the . time histories ~and the spectra comparison, respectively for " set 3". Table 6.3.3 shows the matrix of correlation coefficients-for the three sets of time histories generated from the " site specific" response spectra which is determined to govern _ rack structural analysis. It is clear that all three sots generated from the " site specific" spectra provided- in [6.3.1) meet the specified statistical independence test and the-enveloping criteria. These three sets were used in the rack dynamic analyses.. 6.4 Rack Modeling for Dynamic Simulations 6.4.1 General Remarks Spent fuel storage racks are Seismic Class I equipment. They are. .
- required to remain functional during and_after'an SSE ev<Mt . The l
racks are free standing; they are neither anchored - to the pool floor nor attached to the sidewalls. Individual rack modules.are not interconnected. Figure 6. 4.1 - shows a pictorial : view - of. -.a -
- typical module. The baseplate extends.beyond thatcellular-region L envelope ensuring that inter-rack impacts, if any, occur;first at the -baseplate elevation; - this area is structurally _ qual 3 fiable
- to =
withstand any large in-planeLimpact loads. A - rack may be completely -loaded--'with fuel assemblies (which corresponds to greatest- total _ mass), or it may be completely L empty. The coefficient of friction, p, between pedestal' supports ! and pool floor is- indeterminate . According to. Rabinowicz 6-9 -,en-r- -- .. - ,,---,n- , -, ,,-m,,-.,--- .n,:r-,,- ,e,,,,--,,,.,.,., ,..r,,,n,.n weewo m.m ,w ,n e. v I '6.4.1), results of 199 tests performed on austenitic stainless l steel plates submerged in water show a 'mean value of p to be 0.503 > vith standard deviation of 0.125. Upper and lower bounds (based on twice standard deviation) are 0.753 and 0.253, respectively. . Analyses are therefore performed for coefficient of friction values of 0.2 (1mr limit) and for 0.8 (upper limit), and for random friction values clustered about a mean of 0.5. The bounding values of p = 0.2 sno 0.8 have been found to bracket the upper limit of module response in previous rerack projects. Since racks are not anchored to the pool slab, not attached to the pool walls, and not interconnected, they can execute a wide variety of motions. Racks may slide on the pool floor, one or more rack support pedestals may momentarily tip and lose contact with l the floor slab liner, or - racks may exhibit a combination of l sliding and tipping. The structural models devoleped permit simulation of these kinematic events with inherent built-in conservatisms. The rack models also include components for simul atic.n of potential inter-rack and rack-to-wall impact phenomena. Lift-off of support pedestals and subsequunt liner i impacts are modeled using impact (gap) elements, and coulomb l friction between rack and pool liner is simulated by piecewise ! linear (friction) elements., Rack elasticity, relative to the rack base, is included in the model with linear. springs representing a beam like action. These special attributes of rack dynamics require strong emphasis on modeling of linear and nenlinear , springs, dampers, and compression only gap elements. The term " nonlinear spring" is a generic term to denote the mathematical element representing the esse where restoring force is not-linearly proportional to displacement. In the fuel rack , simulations, the coulomb fristion interface between rack support pedestal and liner _is typical of a nonlinear spring. i 6-10 ,- , _ - . . . . - . . -..,,.--,----..-w,_., - . _ . - , . - w.,-.. . . , - , - . . - - . ..- .-. ,,-,w,~ -r - - - ., 31 dynamic analyses of single rack modules require a key modeling assumption. This relates to location and relative motion of neighboring racks. The gap between a peripheral rack and adjacent pool wall is known, vith motion of the wall prescribed. However, another rack, adjacent to the rack being analyzed, !.s also free standing and subject to motion during a seismic event. To conduct the seismic analysis of a given rack, its physical interf ace with neighboring modules must be specified. The standard procedure in analys!.s e a single rack module is that neighboring racks move 180' out-of-phase in relation to the subject rack. Thus, the available gap before inter-rack impact occurs is 50% of the physical gap. This " opposed phase motion" assumption increases likelihood of intra-rack impacts and is thus conservative. However, it also increases the relative contribution of fluid coupling, which depends on fluid gaps and relative movements of bodies, making overall conservatism a less certain assertion. 3-D Whole Pool Hulti-Rack analyses carried out for Taiwan Power Company's Chin Shan Station, and for GPU 11uclogr's Oyster Creek Huclear Station demonstrate that single rack simulations predict , l smaller rack displacement during reismic responses. Nevertheless, l 3-0 analyses of single rack modules permit detalled evaluation of ( stress fields, and serve as a benchmark check for the much more involved, WPMR analysis. Particulars of modeling details and assumptions for 3-D Single Rack analysis and for Whole Pool Multi-Rack analysis are given in the following subsections. 6.4.2 The 3-D 22 D0F Model for Sincle Rack Module 6.4.2.1 Assumntions
- a. The - fuel rack structure is very rigid; motion is l captured by modeling the rack as a twelve degree-of-freedom structure. Movement of the rack cross-section at any height is described by six degrees-of-freedom of- the rack base and six degrees-of-f reedom at- the rack top. Rattling fuel assemblies within the rack are modeled by five 6-11 4
- - , , , - , - - - ,,, ,_ ., -, a _ _ _ . _ _ _ _ _ _ . __ _ _ _ _ _ _ ~ _ __. _.__ _ _ ._ ._ i lumped masses located at !!, .75I1, . 5 11 , . 2511, and at the rack base (!! is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement degrees <of-freedom. Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the basoplate level. The centroid of each fuel assembly . mass can be located off conter, relative to tho > rack ntructure centroid at that level, to simulato a partially loadod rack. b seismic motion of a fuel rack is characterized by random rattling of fuel assemblies in their individual storage locations. All fuel assemblies are assumed to move in-phase within a rack. Thic exaggerates computed dynamic loading on the rack structure and therefore yields conservative results.
- c. Fluid coupling betwoon rack and fuel assemblies, and betwoon rack and wall, is simulated by appropriato inertial coupling in the system kinetic energy. Inclusion of those effects uses the methods of [6.4.2) and [6.4.3) for rack / assembly coupling and for rack-to~ rack coupling, respectively. Fluid coupling terms for rack-to-rack coupling are based on opposed phase motion of adjacent modules,
- d. Fluid damping and form drag is conservatively noglected.
- o. Sloshing is negligible at the top of the rack and is neglected in the analysis of the rack.
- f. Potential impacts between rack and fuel assenblies are accounted for by appropriate " compression only" gap elements between masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at top and bottom of the rack in two horizontal directions. Bottom elemente are located at the basoplato elevation.
- g. Pedestals are modeled by gap elements in the vertical direction and as " rigid links" for transferring horizontal stress. Each pedestal support is linked to the pool liner by two friction springs. Local pedestal apring stif fnsas accounts for floor elasticity and for local rack elasticity- ,
juer above the peder:al. l 6-12
- h. Rattling of fuel assemblies inside the storage locations causes the gap between fuel assemblies and cell wall to change from a maximum of twice the nominal gap to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap.
6.4.2.2 Hodel Detallt > Figure 6.4.2 shows a schematic of the model. Si (i = 1,...,4) represent support locations, p1 represent absolute degrees-of-freedom, and qi represent degrees-of-freedom relative to the slab. I? is the height of the rack above the baseplate. Not shown in Fig. 6.4.2 are ges elements used to model pedestal / liner impact locations and impae locations with adjacent racks. Table 6.4.1 lists the degrees-of-freedom for the single rtuk m'odel. Translational and rotational degrees-of-freedom 1-6 and 17-22 describe the rack motion; rattling fuel masses (nodes 1*, 2*, 3*, 4*, 5* in Fig. 6.4.2) are described by translationa!. degrees-of-freedom 7-16. Ui(t) represents pool floor slab displacement seismic time-history. Figures 6.4.3 and 6.4.4, respectively, show inter-rack impact springs (to track potential for impact between racks or between rack and wall), and fuel assembly / storage cell impact springs at one location of rattling fuel assembly mass. Figures 6.4.5, 6.4.6, and 6.4.7 show the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending plane a shear and bending spring simulate elastic effects (6.4.4). Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model. Additional details concerning fluid coupling and determination of-stiffness elements are provided below. l 6-13 ( , .- . - -- - ~. . - - - . . - _ _ - ._ 6.4.2.3 Pluid Couclina Details The " fluid coupling effect* [6.4.2],(6.4.3) is described as follows: If one body (mass mi) vibrates adjacent to a second body (mass m2), and both bodies are submerged in frictionless fluid, then Newton's equations of motion for the '.wo bodies ares 2 (mi + M11) X1+M12 X2 " *PPlied forces on mass mi + 0 (X1) l H21 X1 + (m2 + H22) X2 = applied forces on mass n2 + 0 (X2 *) X,X2 1 denote absolute accelerations of masses mi and m2r respectively, and the notation O(X2 ) denotes nonlinear terms. ! l M11, M12, H21, and H22 are fluid coupling coefficients which l depend on body shape, relative disposition, etc. Fritz [6.4.3] gives data for Mij for various body shapes and arrangements. The fluid adds mass to the body (M11 to mass mi), and an external force proportional to c.cceleration of the adjacent body (maas m2). Thus, acceleration of one body affects the force field on another. This force field is a function of interbody gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a storage location cucuunters t.his effect. For example, fluid l coupling is between nodes 2 and 2* in Figure 6.4.2. The rack analysis also contains inertial fluid coupling terms which model the effect of fluid in the gaps between adjacent racks. Terms modeling effects of fluid flowing between adjacent racks are ! compeced assuming that all racks adjacent to the rack being analyzod are vibrating 1000 out of phase from the rack being l analyzed. Thus, the modeled rack is enclosed by a hydrodynamic mass computed as if there were a plane of symmetry located in the middle of the gap region. Rack-to-rack gap elements (Figure 6.4.3) l have initial gaps set to 50% of the physical gap to reflect this symnetry. l 6-14 n s - . _ _ . . - - ~ _ _ . _ . - - _ _ _ _ - _ ___ ___ _ _ _. _ _ . ___._ _- _ _ _ _ _ _ . . _ _ _ 6.4.2.4 Stif f nes t_ Element _E!cialla The cartesian coordinate system associated with the rack has the fallowing nomenclature: x = Borizontal coordinat. al-. " G..; short direction of rack rectangular pla < fem y = Horizontal coordinata t.a 7t om Acmq direction of the rack rectangular planfoim z = Vertical coordinato upward from the rock base Table 6.4.2 lists all spring elements used in the 3-D 22 Dor single rack model. If the simulation model is restricted to two dimensions (one horizontal motion plus vertical motion, for example), for the nurcoses of model clarification only, then a descriptive model of the simulated structure which includes gap and friction elements is shown in Figure 6.4.8. This simpler model is used to elaborate on the various stiffness modeling elements. Gap elements modeling impacts between fuel assemblics and rack have local etif,fneen KI in Figure 6.4.0. In Tabis 6.4.2, for example, gap elements 5 through 8 act on the rattling fuel mass at the rack top. Support pedestal spring rates Ks are modeled by elements 1 through 4 in Table 6.4.2. Local compliance of the concrete floor is included in Ks. Friction elements 2 plus 8 and 4 plus 6 in Table 6.4.2 are shown in Figure 6.4.8. Friction at support / liner interface is modeled- by the piecewise linear friction springs with suitably large stiffness Kg up to the limiting lateral load, pH, where N is the current compression load at the interface between support and liner. At every time step during transient analysis, the current value of N (either zero if the pedestal has lifted-off the liner, or a compressive finite value) is computed. Finally, support rotational friction springs KR reflect any rotational restraint that may be offered by the { 6-15 . - , - , . , , ~ . . , - - , - -- - ~, 'e foundation. The rotational friction spring rate is calculated using a modified Bousinesq equation [6.4.4) and is included to simulate resistive moment by the slab to counteract rotation of the rack pedestal in a vertical plane. The nonlinearity of these springs (friction elements 9, 11, 13, and 15 in Table 6.4.2) 1 reflects the edging limitation imposed on the base of the rack support pedestals and the shift in location of slab resistive load as the rack pedestal rotates. The gap element Ks, modeling the ef fective compression stiffness of the structure in the vicinity of the support, includes stiffness of the pedestal, local stiffness of the underlying pool slab, and local stiffness of the rack cellular structure above the pedestal. The previous discussion is li.mited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D motions and include all stiffness elements listed in Table 6.4.2. 6.4.3 Whel.t._ Pool Mult'l-Rack f WPMR) Model 6.4.3.1 poneral Ry m kg The single rack 3-0 (22 DOP) model outlined in the preceding subsection is used to evaluate structural integrity, physical stability, and to initially assess kinematic compliance (no rack-to-rack impact in the cellular region) of the rack modules. Prescribing the motion of the racks adjacent to the module being ~ analyzed is an assumption in the single rack simulations. For I closely spaced racks, demonstration of kinematic compliance is l further confirmed by modeling all modules in one comprehensive l simulation u,,ing a Whole Pool Halti-Rack (WPMR) - model.1 In WPHR i analysis, all racks are modeled, and_ their correct fluid interaction is included in the model. I 6-16 l 6.4.3.2 Whole Pool Pluid coupling I The presence of fluid moving in the narrow gaps between racks and between racks and pool walle causes both near and far field fluid coupling effects. A single rack simulation can effectively include only hydrodynamic effects due to contiguous racks when a certain set of assumptions is used for the motion of contiguous racks. In a Whole Pool Hulti-Rack analysis, far field fluid coupling effects ! of all racks are accounted for using the corroet model of pool fluid mechanics. The external hydrodynamic mass due to the presence of walls or adjacent racks is computed in a manner consistent with fundamental fluid mechanics principles (6.4.5) , using conservative nominal fluid gaps in the pool at the beginning of the seismic event. Verification of the computed hydrodynamic effect by comparison with experiments is also provided in (6.4.5). This formulation has been reviewed and approved by the Huclear Regulatory Commission during post-licensing multi-rack analyses for the Diablo Canyon Unit I and II raracking project. The fluid flow model used to obtain the whole pool hydrodynamic effect reflects actual gaps and rack locations. 6.4.3.3 Coefficients of Frletion To eliminate the last significant element of uncertainty in rack dynamic analyses, the friction coefficient is ascribed to the support pedestal /pcal bearing pad interface consistent with l Rabinowicz's data (6.4.1). Friction coef ficients, developed by a ! random number generator with Gaussian normal distribution l characteristics, are imposed on each pedestal of each rack in the pool. The assigned values are then held constant during the i l l i 6-17 entire simulation in order to obtain reproducible results.* Thus, the WPMR analysis can simulate the effect of different coefficients of friction at adjacent rack pedestals. 6.4.3.4 Mode 11qq_D.ttails r Figure 6.4.9 shows a planform view of the spent fuel pool which ' includes rack and pedestal numbering scheme and the global coordinate system used for the WPMR analysis. Table 6.4.3 gives details on number of cells per rack, and on rack and fuel weights.
- In Whole Pool Hulti-Rack analysis, a reduced degree-of-freedom (RDOF) set is used to model each rack plus contained fuel. The rack structure is modeled by six degrees-of-freedom. A portion of contained fuel assemblies ~is assumed to rattle at the top of the rack, while the remainder of the- contained fuel 'is assumed as a distributed mass attached to the rack. The rattling portion of the !
contained fuel is modeled by two hor zontal degrees-of-freedom. Thus, the WPMR model involves all racks in the spent fuel pool with each individual rack modeled as an 8 degree of freedom structure. The rattling portion of fuel mass, within each rack, is chosen to insure reasonable agreement between displacement . l predictions from single rack analysis using a.22 DOF model and predictions from 8 DOF analysis under the same co ditions. Note that DYNARACK has the capability to change the coefficient of friction at any pedestal . at each instant of contact based on a random reading of the PC-clock cycle. However, exercising this option would yield results that could not be reproduced. Therefore, the random choice of coefficients is made only once per run.- 6-18 , .,.-.._m .,- -,- ,,. ,,....-..,.rwme,r-_, ,m.-. . . . , .~...._,-..i,.,__,. .,,.,....m.-..... c., -,.4 . g.. ,, ,,,,,,e.,-... i l 4 l 1 The Whole Pool Multi-Rack model includes gap elements representing ! ) compression only pedestals, representing impact potential at fuel , assembly-fuel rack interfaces, and at rack-to-rack or rack-to-wall ! locations at top and bottom corners of each rock module. Each , pedestal has two friction elements associated with force in the i vertical compression element. Values used for spring constants for l ! the various stiffness elements are equal to the values used in the ; 22 DOF model. 6.5 Accentance Criteria, stress L1=4tm, and Material , Proce rtish 6.5.1 heceptance criteria There are two sets of criteria to be satisfied by the rack modules:
- a. -Einematic Criteria The rack must be a physically stable structure and it i must be demonstrated that there are no inter-rack ;
impacts in -the cellular region. The criteria for , physical stability is that an isolated rack in water exhibit no overturning tendency when a seismic event of magnitude 1.1 x SSE'is applied [6.1.2].
- b. Stress Limit criteria
, Stress limits must not be exceeded under certain load combinations. The following loading combinations are 1 applicable [6.1.3). r Loadina Combination Service Level ! D+L Level A-D +'L + To D + L + To-+ E D + L + Ta + E Level 8 l D-+ L +-To + Pg. . D-t L + Ta + E' . . Level D D+L+Fd -The functional capability __of the fuel racks should. be demonstrated. 6-19 Abbreviations are those used in Section 3.8.4 of the Standard Review Plan and the " Review and Acceptance of Spe,nt Puel Storage and Handling Applications" sections D' = Dead weight.. induced internal moments (including fuel assembly weight) L = Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path). Fd = Force caused by the accidental drop of the heaviest load from the maximum possible height (see chapter 7 of this report). Pg = Upward force on t'he racks caused'by postulated stuck fuel assembly (see chapter 7). E = Operating Basis Earthquake (OBE) E' = Design Basis Earthquake (DBE or SSE) = Differential temperature induced loads (normal To operating or shutdown condition based on the most critical transient or steady state condition). a Differential temperature induced loads Ta (the highest temperature associated with the postulated abnormal design conditions). Ta and To cause local thermal stresses to be produced. For fuel rack analysis, only one scenario need be examined. The worst situation is obtained when an isolated storage location has a fuel I assembly generating heat at maximum postulated rate and i murrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing maximum possible temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary ! (thermal) stresses. For rack qualification, To, Ta are the same, l t 6-20 _ , _ _ , _ _ _ .,_m d 6.5.2 Stress Limits for various conditiona Stress lLnits are derived from the ASME Code , Section III, Subsection NF [6.1.3). Parameters and terminology are in accordance with the ASME Code. 6.5.2.1 Normal and Upset conditions (Level A or Level B1
- a. Allowable stress in tension on a not section is:
Ft = 0.6 Sy (Sy = yield stress at temperature) (Ft is equivalent to primary membrane stress)
- b. Allowable stress in shear on a net section ins i Fy = .4 S y
- c. Allowable stress in compression on a not section I
(k1)2 [1 - 2 /2Ce Sy F. = - 5 kl kl 3 3 (( } + [3 (- ;) /8Cc ] - [ (-- ) /8C e ]) 3 r r where: (2n2 g) 12 / Cc " [ ] Sy 1 = unsupported length of component k = length - roefficient which giver influence of boundary conditions; e.g. k' = 1 (simple support both ends) --2 (cantilever beam) = 1/2 (clamped at both ends) E = Young's Modulus r = radius of_ gyration of component 6-21 i l _ _ . _ _ _ . . . _ . __ __ _...__=_-_._ _ _ _ _ _._. _._ .__ ___. _ . _ _ _ k1/r for the main rack body is based on the full height and cross section of the honeycomb region.
- d. Maxiasum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry ist 0.60 S y Fb (equ=ivalent to primary bending)
- e. Combined flexure and compression on a not section satisfies fa Cmx fbx Cayf by
+ + <1 Fa DxFb x DFy yb where fa = Direct compressive. stress in the section fx b = Maximum flexural stress along x-axis fby = Maximum flexural stress along y-axis Cmx = Cay " 0.85 : fa Dx = 1 - F'ex l fa Dy=1-r'ay 12 n2 E F'ex,,y = kl 23 ( j r x,y 6-22 .-- m_ ,x-- , , -r.,..,..,-i,- ,r +,...r-.~.~ :--,,,. . ~ . . . - . . . . . . ~ - , - . . - . . . , , . . . . . . h t and subscripts x,y reflect the particular bending plane.
- f. Combined flexure and compression (or tension) on a net sections fa fx b iby
+ + < 1.0 0.6S y Fx b Pby The above requirements are to be met for both direct tension or compression. 6.5.2.2 L m .1 D Service Limita Section F-1370 (ASME Section III, Appendix F), states that limits for the Level D condition are the minimum of 1.2 (Sy /Ft) or (0.7Su /Ft) times the corresponding limits for the Level A condition. Su is ultimate tunsile stress at the specified rack
- design temperature. For example, if the material is such that 1.2S y is less than 0.7Su, then the multiplier on the Level A limits, to obtain Level D limits, is 2.0.
6.5.2.3 Dimensionless Stress Factors Stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value. Stress factors are only developed for the single rack analyses. The limiting value of each stress factor is 1.0 for OBE and 2.0 (or less) for the SSE condition. Stress factors reported are : R1 = Ratio of direct tensilt or compressive stress on a net section to its allowable.value (note pedestals only resist compression. R2 = Ratio of gross shear on a net section in the x-direction to its allowable value. R3 = Ratio cf maximum bending stress due to bending about the x-axis to its allowable value for the section. 6-23 l R4 = Ratio of maximum bending stress due to bending about the y-axis to its allowable value for the section. R5 = Combined flexure and compressive factor (as defined in 6.5.2.le above) R6 = Combined flexure and tension (or compression) factor (as defined in 6.5.2.lf) R7 = Ratio of gross shear on a net section in the y-direction to its allowable value. i 6.5.3 Material Properties Physical proporties of the rack and support materials, obtained from the ASME Boiler t. Pressure Vessel Code, Section III, appendices, are listed in Table 6.5.1. Maximum pool bulk temperature is less than 200*F; this is used as the reference design temperature f or evaluation of material properties. Stress limits for Level A,D, corresponding to conditions in section 6.5.2 above, are evaluated using given yield strength data. 6.6 Governina Equations of MotiQD Using the structural model for either 22 DOF single rack analysis, or the set of simplified 8 DOF models that comprise a Whole Pool Hulti-Rack model, equations of motion corresponding to each degree of freedom are obtained using Lagrange's Formulation (6.6.1). The system kinetic energy includes contributions from solid structures and from trapped and surrounding fluid. The final system of equations obtained have the matrix forms (M) (q") = (Q) + (G) where: l [M) - total mass matrix (including structural and fluid mass contributions); l 6-24 (q) - the nodal displacement vector relative to the pool slab displacement; (double prime stands for second derivatives with respect to time); (G) - a vector de on the given ground acceleration; pendent (Q) - a vector dependent on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom The equatiot.s can be rewritten ass (q') = [H]"1 (Q) + (H)~1 (G) This equation set is mass uncoupled, displacement coupled at each instant in time; numerical solution uses a central difference scheme built into the proprietary, computer program "DWIARACK" [6.6.2 - 6.6.5). As indicated earlier, this program has been used in the licensing effort for a considerable number of rcracking projects. DWIARACK has br en validated against exact solutions, experimental data, and solutions obtained using alternate numerical schemes (6.6.5). These nolutions are chosen to exercise all features of DDIARACK. It is demonstrated there that well known classical nonlinear phenomena (subharmonic resonance, bifurcation, stick-slip) can be reproduced using DniARACK. The application of DDIARACK to the spent fuel rack analysis requires the establishment of a time step to ensure convergence and stability of the results. DDIARACK utilizes the classical contral difference algorithm (6.4.4). Stability of the results is assured as long as the time step is significantly below the smallest period of the equivalent linear problem. Convergence is obtained by performing a series of rack analyses with different time steps to ascertain the upper limit on time step that will provide converged results. This is done by taking a typical rack module and subjecting it to the given time-histories using l 6-25 different integration time steps. Once an appropriate time step is determined, it is used in subsequent simulations. Results of the dynamic simulations are time-history response of all degrees-of-freedom of the particular model, and cf all forces . and moments at IJaportant sections of the structure. From these ; results, maximum movements and stresses can be ascertained for the event, and appropriate structural qualifications can be carried out. Where required, DYNARACK automatically tracks maximum values ! of dimensionless factors R1 to R7 defined above in Section 6.5, ! and reports results for the rack cross section just above the baseplate and for each pedestal cross section just below the baseplate. These are the critical sections which develop the ! highest stresses due to the geometry of a - fuel rack structure. From the archived results, time-histories of al1~ rack-to-rack fluid gaps, all rack-to-wall fluid gaps, and motion of any point ; on any rack can be generated. Sections 6.7 and 6.8 give the results obtaired from single and multi-rack analyses, respectively. The results demonstrate satisfaction of all requirements on structure and kinematic integrity. 6.7 Results of 3-D Nonlinear Analyses _of Sincle Rackg This section focuses on results from all 3-D single rack analyses. In the following section, we present results from the whole pool multi-rack analysis and discuss the sbnilarities and differences between single and multi-rack analysis. 6.7.1 Racks in the Fuel Pool-A summary of results of all analyses performed for racks in the l pool, using a single rack model, is presented in summary Tables -6.7.1 a 6.7.38. Table 6.7.1 lists all runs carried out. Table 6.7.2 presents the bounding results from~a n runs, and Tables 6.7.3 - 6.7.38 give details for each run. The tabular results for 6-26 - - - - - - - ._ . . , _ . . _ _ . _ , . . . . . . . . . . , . _ _ . _ , , , , . . . . . , . . _ , , , , _ . . - , , , . . .- .,.m , , . . . + , , _ . - . . - , . . . . , ,, l each run give maximax (maximum in time and in space) values of stress factors at important locations in the rack. Results are given for maximum rack displacements (see Section 6.4.2.2 for x,y ' orientation), maximum impact forces at pedestal-liner interface, and rack cell-to-fuel, rack-to-rack, and rack-to-wall impact forces. It is shown that no rack-to-rack or rack-to-wall impacts occur in the cellular region of the racks. In the single rack analysis, kinematic criteria are checked by confirming that no inter-rack gap elements at the top of the rack close (see Figure 6.4.3). By virtue of the symmetry assumption discussed in subeection 6.4.2.4, impact is assumed to occur if the i local horizontal displacement exceeds 50% of the actual rack-to-rack gap. Structural integrity at various rack sections is considered by computing the appropriate stress factors R1. Results corresponding to the SSE event yield the highest stress factors. Limiting stress factors for pedestals are at the upper section of the support and are to be compared with the bounding value of 1.0 (OBE) or 2.0 (SSE). Stress factors for the lower portion of the support are not limiting and are not reported. From Table 6.7.2, all stress factors are below the allowable limits. Analyses are performed using dimensions appropriate to racks with adjustable pedestals. Since rac'c heights are less and the support height is smaller for any rack with fixed pedestals, displacements and stress factors from racks with fixed pedestals are not controlling. overturning has also been considered. A multiplier of 1.5 on SSE horizontal earthquakes -(more conservative than required by the USNRC Standard Review Plan) is applied to an isolated rack-and the predicted displacementu examined. Horizontal displacements do not grow to such an extent as to imply ^'any possibility for overturning. l l 6-27 .- . - = _ - - - -- - _ . - - - _ - - _ . -- .. - Additional investigation of important structural items is carried out and results are summarized in Table 6.7.39. A discussion of these items follows: 6.7.1.1 Impact Analyses
- a. Imoact Leadina Detween Fuel,Ag.g.gmbiv and Cell Wall Local cell wall integrity is conservatively estimated from peak impact loads. Plastic analys/s is used to obtain the limiting impact load. Table 6.7.39 gives the limiting impact load and compares the limit with the highest value obtained from any of the single rack analyses. The limiting load is much greater than the load obtained from any of the simulations reported in Tables 6.7.1 - 6.7.38.
- b. Impacts Between Adjacent Racks No non-zero impact loads are found for the rack-to-rack gap elements (in the cellular r:cgion), or for the rack-to-wall elements; it is concluded that no impacts between racks or between racks and walls are likely to occur during a seismic event. This is confirmed by the Whole Pool Hulti-Rack results in Section 6.8.
6.7.1.2 Weld Stresses Wald locations subjected to significant seismic loading are at the bottom of the rack at the baseplate-to-cell connection, at the top of the pedestal support at the baseplate connection, and at. cell-to-cell connections. Results from dynamic analyses of single racks are surveyed and maximum loading used to qualify the welds.
- a. _Baseolate-to-Rack Cell Welds and Baseplate-to-Pedestal Welds Reference (6.1.3) (ASME Code Section III, Subsection NF) permits, for the SSE event, an allowable veld stress I =
.42 Su. A comparison of this allowable value . with the highest veld stress predicted is given -in Table 6.7.39. The highest predicted weld stress le less than the allowable veld st.ess value. 6-28 -- - -n..y , - . . - - - n ,,,,n.. - , . ,n -- -e r , r- r + - . - - . _ _ - - . _. - = _ _ _ - - - . , _ _ _ _ _ . . The weld between baseplate and support pedestal is checked using limit analysis techniques 6. 7.1) . . The structural weld at that location is consi(dered safe if the interaction curve between not force and moment is such that G = Function (F/Fy ,H/My) < 1.0 ' F, H are the limit load and moment under direct load obly [nd direct moment only. These values c' 'and on the configuration and on matertal yield strengths. F, H are absolute values of actual force and moments applied to the veld section. The calculated value of G for the pedestal / baseplate weld is presented in Table 6.7.39 and is less than the limit value of 1.0 This calculated value is conservatively based on instantaneous peak i loading. This value also conservatively neglects the l gussets that are in place to increase pedestal area and l inertia. J
- b. Cc11-to-cell Welds cell-to-cell connections are by a series of spot weldo along the cell height. Stresses in storage cell to storage cell welds develop along the length due to fuel assembly impact with the cell wall. This occurs if fuel as s en611es in adjacent cells are moving out of phase with one another so that impact loads in two adjacent cells are in opposite directione; this tends to separate the two cells from each other at the veld. Taole 6.7.39 gives results for the maximum allowable load that can be transferred by these welds based on the available weld area. An upper bound on the load required to be transferred is also given in Table 6.7.39 and is less than the allowable load. This upper bound value is obtained by using the highest rack-to-fuel impact load from Table 6.7.2 (for any simulation), and multiplying the result by 2 (assuming that two impact locations are supported by every weld connection).
6.7.2 Racks in the cask Pit Area The cask area of the fuel pool encompasses a pit area with a 12'x12' horizontal envelope. Analyses have been carried out for a 13x14 free-standing rack and for a 15x15 free-standing rack. As mentioned in Section 2, the 13x14 rack is intended to be placed in the cask pit temporarily during the re-racking stage to reduce the amount of fuel shuffling required. The 13x14 rack will be 6-29 relocated into the fuel pool at its appt spriate location at the completion of the re-rack. A 15x15 rack is also examined for , structural integrity. This rack la not part of the present rerack l campaign, but will be installed in the cask pit area at some : future date. To evaluate the rack in the cask pit, DYNARACK analysis is l performed using fluid gaps between rack and cask pit wall that reflects the actual dimensions of the cask pit area, and the rack envelope. The nominal rack-to-wall gap is 4.5* (for 15x15 spent fuel rack). Runs were carried out for coefficients of friction of .2 and .8 and for different rack fuel loading scenarios. From all analyses performed for a spent fuel rack in the cask pit area, the bounding structural and kinematic results are given in Table 6.7.40. While individual loads and stress factors increase somewhat from bounding values in the main pool region, detailed analyses shows that all structural and kinematic margins are acceptable and that either a 13x14 rack or a 15x15 spent fuel rack may be situated in the cask pit area without reducing the design margins reported herein. It is noted that the 15x15 rack, while heavier than the 13x14 rack, is not governing since the hydrodynamic coupling effect is increased due to the smaller gaps britween rack and cask pit walls. 6.8 Results from Whole Pool Multi-Rack Analyses Tables 6.8.1 - 6.8.2 show maximum corner absolute displacements at both the top and bottom of each rack in global x and y directions (refer to Fig. 6.4.9) from two multi-rack runs. The two runs have random friction coefficients in the range of 0.2 - 0.8 with mean value being 0.5 and governing earthquake time-histories for tho Site Specific SSE .and 50% of the site Specific SSE to conservatively represent OBE, respectively. In Table 6.8.3, the maximum displacement and pedestal vertical loads obtained from the 6-30 .- ,e. -,,,-r- *,, - SSE multi-rack simulation are compared with the limiting single rack analyses. The absolute displacement values are higher than those obtained from single rack analysis. Thus, it appears essential to perform Whole Pool Multi-Rack analyses to verify that racks do not impact or hit the wall. Figures 6.8.1 - 6.8.3 show the time-histories of rack-to-rack and rack-to-wall gaps at typical locations (see Figure 6.4.9 for locations). A survey of all of the rack-to-rack and rack-to-wall impact elements confirms that there are no rack-to-rack or rack-to-vall impacts in the cellular region of any rack in the spent fuel pool. The inter-rack gap elements in the whole pool analysis have an initia) gap equal to the actual gap. Tables 6.8.4 and 6.8.5 present peak pedestal compressive loads for all pedestals in the pool for each of the two WPMR analyses (see Figure 6.4.9 for pedestal locations). In addition to a report of maximum pedestal loads, the time-history of each pedestal load vector for each rack is archived for use in other analyses. l The Whole Pool Hulti-Rack analyses confirms that no new concerns l are identified; overall structural integrity conclusions are confirmed by both single and multi-rack analyses. 6.9 Bearino Pad Analysis To protect the slab from high localized dynamic loadings, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact on these bearing pads during a seismic event and pedesta) loading is transferred to the liner. Bearing pad I dimensions are set to ensure that the average pressure on the slab surf ace due to a static load plus a dynamic impact load does not exceed the American concrete Institute (6.9.1) limit on bearing pressures. Pedestal locations are set to avoid overloading of leak chase regions under the slab. Time-history results from dynamic 6-31 simulations for each pedestal are used to generate appropriate static and dynamic pedestal loads which are then used to develop the bearing pad size. Section 10 of [6.9.1) gives the design bearing otrength as fb " 4 (.85 fe') E where $ = .7 and fe' is the specified concrete strength for the spent fuel pool. E = 1 except when the supt eting surface is wider on all sides than the loaded area. In that case, E = (A2 /A1 ).5, but not more than 2. At is the actual loaded area, and A2 is an area greater than At and is defined in [6.9.1). Using a value of E > 1 inci.udes credit for the confining effect of the surrounding concrete. Bearing pads are sized so as to provide sufficient margin on average bearing pressure. Table 6.9.1 summarizes the limiting result. Where pads are placed over leak chases, the pads are sized to ensure that the average pressure is below the design bearing strength, and that the peak pressure is below 1.5 times the design bearing strength. 6.10 References for Section 6 [6.1.1) "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, and January 18, 1979 amendment thereto. [6.1.2) USNRC Standard Review Plan, NUREG-0800 (1981). [6.1.3] ASME Boiler f. Pressure Vessel Code, Section III, Subsection NF, appendices (1989). [6.2.1) USNRC Regulatory Guide 1.29, " Seismic Design Classification," Rev. 3, 1978. [6.2.2) Soler, A.I. and Singh, K.P., " Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Hotions", i Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984). l l 6-32 [6.2.3) Singh, K.P. and Soler, A.I., " Seismic Qualification of Free Standing Huclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (Harch 1991). [6.3.1) Sequoyah Civil Engineering Branch Reports, CEB-80-20 (Rev. 3), and SCG-CSG-87-206 (Rev. 1), Auxiliary / Control Building Seismic Response Spectra. [6.3.2) Holtec Proprietary Report - Verification and User's Manual for Computer Code GENEQ, Report HI-89364, January, 1990. [6.4.1) Rabinowicz, E., " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," HIT, a report for Boston Edison Company, 1976. [6.4.2) Singh, K.P. and Soler, A.I., " Dynamic Coupling in a closely Spaced Two-Hody System Vibrating in Liquid Hedium: The Case of Fuel Racks," 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982. [6.4.3) Fritz, R.J., "The Effects of Liquids on the Dynamic Hotions of Immersed Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172. [6.4.4) Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Veb M e Engineering," McGraw Hill, 1976. [6.4.5) Paul, B., " Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment", Holtec Proprietary Report HI-88243. , [6.6.1) " Dynamics of Structures," R.W. Clough and J. Penzien, i McGraw Hill (1975). [6.6.2) Soler, A.I., " User Guide for PREDniAl and DYNAHO", Holtec Proprietary Report HI-89343, Rev. 2, March, 1990. [6.6.3) Soler, A.I., " Theoretical Background for Single and Multiple Rack Analysis", Holtec Proprietary Report HI-90439, Rev. O, February, 1990. [6.6.4) Soler, A.I., DW1ARACK Theoretical Manual", Holtec Proprietary Report HI-87162, Rev. 1, January, 1988. l l 6-33 l (6.6.5] Soler, A.I., "DYHARACK Validation Manual, Holtee Proprietary Report HI-91700, Rev. O, October, 1991. [6.7.1) Singh, K.P., Soler, A.I., and Bhattacharya, S., ' Design Strength of Primary structural Welds in Free Standing Structu,es*, ASME, J_gurn. of Pressure Vessel Technoloay, August, 1991. [6.9.1} ACI 318-71, Building Code Requirements for Reinforced concrete, American concrete Institute, Detroit, Michigan, 1971. 6-34 ,-..-r.--.. . . , - - - . . ~ , , . .- , , , - - . . , . , , , . . . - . 0 Table 6.4.2 NUMBERING SYSTEN FOR GAP ELEMENTS AND FRICTION ELEMENTS 1 - I. Nonlinear Springs (can Elemental (64 Total) Number Hode Location Egpeription i 1 Support $1 3 compression only element ! 2 Support 52 E compression only element 3 Support 83 z compression only element 4 Support 54 Z comaression only element ' 5 2,2* X rack / fuel assed ly impact element 6 2,2* X rack / fuel assembly impact ' element 7 2,2* Y rack / fuel assembly impact element 8 2,2* Y rack / fuel assembly impact element 9-24 other rattling masses for nodes 1*, , , 4* and S* 25 Bottom cross- Inter-rack impact elements section of rack (around edge) Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements ' . Inter-rack impact elements . Inter-rack impact elements 44 Inter-rack impact elements 45 Top cross-section Inter-rack impact elements . of rack Inter-rack impact elements . (around edge) Inter-rack impact elements . Inter-rack-impact elements . Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements 64 Inter-rack impact elements l l 6-35 6 ..t' * " * - -
- y *'ve----v-wi'-- v 'r -ri-' 'sv- t --r e m V'te--t-e-mF"'-m*-+-* * - - 1 *
. . i Table 6.3.1 CROSS-CORRELATIO!1 COEFFICIE!1TS OF Tile SYllTilETIC TIME-!!ISTORIES BASED 011 T!!E TVA POOL "DESIGli BASIS" .9PECTRA P Epsults of Coefticient of Correlation DATA 1 TO DATA 2 = 3.640022E-02 DATA 1 TO DATA 3 = 7.739163E-02 DATA 2 TO DATA 3 = 1.046229E-01 !!ote: DATA 1: ACCELDBl.!!S DATA 2 ACCELDBl.EW DATA 3 ACCELDBl.VT 6-36 ,.,.w , ,. n- -- .-~ ,, r r , -n,... ., ,,,,-..,_,.n.- - - ., . w. , . - - , - - - , - - , , , . - - , - - b Table 6.3.2 CROSS-CORRELATION COEFFICIENTS OF THE SYNTHETIC TIME-HISTORIES (SET 1) BASED ON THE TVA POOL " SITE SPECIFIC
- SPECTRA Results of Coefficient of Correlation DATA 1 TO DATA 2 = 3.046205e-02 DATA 1 To DATA 3 = 7.548556e-02 DATA 2 TO DATA 3 = 1.121664e-01 Note: DATA 1: ACCELSSI.NS DATA 2: ACCELSSI.EW DATA 3: ACCELSS1.VT 6-37
O l l l l-Table 6.3.3 IGTRIX OF THE CROSS-CORRELATION COEFFICIENTS FOR THREE TIME-HISTORY SETS GENERATED FROM SEQUOYAH " SITE SPECIFIC" SPECTRA Time History Data Tile- ACCELS$1 ACCELSS1 ACCELSS1 ACCELSS2 Har.e NS EW VT NS ACCELSSI.NS ACCELSSI.EW 0.03046 ACCELSSI.VT 0.07549- 0.11217 ACCELSS2.NS 0.01219 0.03856 0.03679 ACCELSS2.EW -0.06769 0.01215 -0.08275 -0.08692 ACCELSS2.VT -0.0200*- -0.05163 0.01403 -0.03526 ACCELSS3.F4 0.03702 0.10223 0.09822 -0.07368 i ACCELSS3.EW + .03870 -0.009125 0.03869 -0.01383 l ACCELSS3.VT 0.006970 - '.07012 - 0.14385 0.00449 l i i 6-38 . . .-y ., . ~. - ..--c , ,- , , , . . , -- . - . . . . . . . . . . ... . . . . . . - . ~ . . .-. . . - - Table 6.3.3 (continued)- MATRIX OF THE CROSS-CORRELATION COEFFICIENTS FOR THREE TIME-HISTORY SETS GENERATED FROM SEQUOYAH " SITE SPECIFIC" SPECTRA Tine History Data File ACCELSS2 ACCELSS2 ACCELSS3 ACCELSS3 ACCELSS3 Nane EW VT NS EW VT ACCELSSI.NS ACCELSSI.EW ACCELSSI.VT ACCELSS2.NS ACCELSS2.EW ACCELSS2.VT 0.02227 ACCELSS3.NS 0.03775 -0.04327 ACCELSS3.EW -0.05712 0.11433 0.07693 ACCELSS3.vT 0.02334 0.07005 -0.01002 -0.06519 l I 1 l 6-39 . . - , . , -., - . . , , . .,e r. .- r ., . +,,-..--v , A s_ J A.s4+_ .-.-.JAa A 2_ L2.A..-A -. W-is.* 4 - A 4 J- aa._g. A. - s anJ z4, Table 6.4.1 DEGREES-OF-FREEDOM i I l Displacement Rotation Location Ux Uy U: Ox Oy Oz (Node) l 1 P1 P2 P3 94 95 96 2 p17 P18 P19 q20 921 922 Point 2 is assumed attached to rigid rack at the top most point. 2* P7 P8 3* pg pio 4 P11 P12 5* pl3 p14 1* P15 P16 where: Pi " 91(t) + Ul (t). i = 1,7,9,11,13,15,17 = qi(t) + U 2(t) i = 2,8,10,12,14,16,18 = l qi.t) + U 3(t) i = 3,19 Ui(t) are the 3 known carthquake displacements. 6-40 ._ar,.., . . . \ l l Table 6.4.2 NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS I. Upnlinear Sprinas (Gan Elementsi (64 Total) Number Rode Location Qgserietion 1 Support F1 Z compression only element 2 Support 82 Z compression only element 3 Support S3 Z compression only element 4 Support S4 Z compression only element 5 2,2* X rack / fuel assembly impact element 6 2,2* X rack / fuel assembly impact element 7 2,2* Y rack / fuel assembly impact element 8 2,2* Y rack / !uel assembly impact element 9-24 Other rattling masses for nodes 1*, 3*, 4* and 5* 25 Bottom cross- Inter-rack impact elements section of rack (around edge) Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements 44 Inter-rack impact elements 45 Top cross-section Inter-rack impact elements . of rack Inter-rack impact elements . (around edge) Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements . Inter-rack impact elements 64 Inter-rack impact elements l 6-41 O Table 6.4.2 (continued) NUMBERING SYSTEM FOR GAP ELEMEllTS MID FRICTIO!1 EL II. ELiction Elements (16 total) Number Node Location peserietion 1 Support S1 2 Support S1 X direction friction 3 Support S2 Y direction friction 4 Support S2 X direction friction 5 Support S3 Y direction friction 6 Support S3 X direction friction 7 Support S4 Y direction friction 8 Support S4 X direction friction 9 S1 Y direction friction 10 S1 X Slab moment 11 S2 Y Slab moment 12 S2 X Slab moment 13 S3 Y Slab moment 14 S3 X Slab moment 15 S4 Y Slab moment 16 S4 X Slab moment Y Slab moment = 6-42 _ _ - - - - - - - - - - - - - - ~ ~ ~ ^ ~ ' ~ ^ - .-. - . . . . = . . - . . . - . .. - - . . - . - - - . - . . - - . . . . - . . 1 l P Table 6.4.3 RACK NUMBERING AND WEIGHT INFORMATION. Rack No. of Weight of Weight'of h Cgilpi Rack. Ib. Fuel Assembly.-lb. 1 192 l26000 1550-2 168 24000 - 1550 3 182 - 26000 1550-4 182 26000 1550 5 182 26000 1550 6 168 24000 1550 7 182 26000- 1550 8 182 26000 1550 9 169 -24200 1550-10 156 22300 1550 11 169 24200 1550 12 169 24200 1550 6-42a , ,, , . , , , ,-[..w,- -., . . , , -,+y .-,m' . - , . , - -_.-y,-.- - , , , - - . - , ,, .,, , , - , _ .. _ ._... _ . _ _ _ __ _ ..- _ _ _ _ _ _ ..._. __..._ _. _ ._ _._ _. .. _ ._ .__ . _ .~_ _ ... _ _ _ _ . _. .._._ m Table 6.5.1 ; RACK MATERIAL DATA (200'F) Young's Yield Ultimate Modulus Strength Strength-Material E (psi) Sy (psi) Su (psi) 304 S.S.* 27.9 x 106 25,000' 71,000-Section III Table Table- Table Reference I-6.0 I-2.2 I-3.2 SUPPORT MATERIAL DATA-(200*F) Young's- Yield . . Ultimate Modulus Strength- Strength' Material E (psi) Sy_(psi) Su.(psi) L 1 SA-240, 27.9 x 106 '25,000 71,000 Type 304L* (upper part of support feet) 2 SA-564-630 27.9 x 106 106,300- 140,000 (lower part of - support feet; ! age hardened at l- 1100*F) Duallcertified to have chemical-composition of.304L-material L and' physical properties ~of'304 material. 6-43 4 .*z,---- ,--v. < , , , . ,.w- - y-.._.,-,,r --,,-m,, , - ~ . . , , , , - ..-.-,--,-,,..,,,..,,,,-.m,._,,,,,,,,,,,,...--,, 44-,,,,, .~ , .. ....,, Table 6.7.1 RESULTS OF SINGLE RACK ANALYSES List of All Runs l l Holtec Rack Fuel Fuel Loading Seismic Coefficient Motion I i -Run I.D. I.D. I.D. Condition Loading of Friction Mode do13x14c.rf8 A3 5-H17x17 Fully Loaded ACCELSS3 0.8 opposed 182 cells SSE phase do13x14c.rf5 A3 5-H17x17 Fully loaded ACCELSS3 0.5 opposed 182 cells SSE phase do13x14c.rf2 A3 5-H17x17 Fully loaded ACCELSS3 0.2 opposed 182 cells SSE phase , do13x14c.rh8 A3 5-Hi?x17 Half loaded - ACCELSS3 0.8 opposed 91 cells phase do13x14c.rh5 A3 5-H17x17 Half loaded ACCELSS3- 0.5 opposed 91 cells phase do13x14c.rh2 A3 5-H17x17 Half loaded ACCELSS3 0.2- -opposed-91 cells phase ! do13x14c.re8 A3 5-H17x17 13 cells ACCELSS3 0.8 opposed l phase do13x14c.re5 A3 5-H17x17 13 cells ACCELSS3 0.5 opposed phase do13x14c.re2 A3 5-H17x17 13 cells ACCELSS3 0.2 opposed I phase. di13x14c.rf8 A3 5-H17x17 Fully-Loaded ACCELSS3 - 0.8 in-phast 182 cells SSE di13x14c.rf5 A3 H17x17 Fully loaded ACCELSS3 0.5 - in-phase 182l cells SSE di13x14c.rf2 A3 5-H17x17 Fully-loaded- ' ACCELSS3' : 0.2- . in-phase
- 182 cells SSE di13x14c.rh8- A3 5-H17x17- Half loaded-- ACCELSS3 0.8- . in-phase 91 cells SSE-di13x14c.rh5 A3 5-H17x17 ' Half loaded ACCELSS3- 0.5 in-phase:
91-' cells - -SSE
- 6-44
-. -w w =e -.-r- ear e e v- w <-t,mr- 4 y -w ,.* m e m-,vo w s + o
- Sm--r t.-me wer -w yr w y v*w ee s t m vr y eswiwve s- Y y T w *' y- -n-'t +9 y -=-org "*
- e w**yr '-
= - _ - - . ._ . . - -. - _.. - - - -. Tabic 6.7.1 ( continued ) di13x14c.rh2 A3 5-H17x17 Half loaded ACCELSS3 0.2 in-phase 91 cells SSE di13x14c.re8 A3 5-H17x17 13 cells ACCELSS3 0.8 in-phase SSE di13x14c.rc5 A3 5-H17x17 13 cells ACCELSS3 0.5 in-phase SSE di13x14c.re2 A3 5-H17x17 13 cells ACCELSS3 0.2 in-phase SSE do12x14a.rf8 B1 5-H17x17 Fully Loaded ACCELSS3 0.8 opposed 168 cells SSE phase do12x14a.rf5 B1 5-H17x17 Fully loaded ACCELSS3 0.~ opposed 168 cells SSE phase do12x14a.rf2 B1 5-H17X17 Fully loaded ACCELSS3 0.2 opposed 168 cells SSE phase do12x14a.rh8 B1 5-H17x17 Half loaded ACCELSS3 0.8 opposed 84 cells phase do12x14a.rh5 B1 5-H17x17 Half loaded ACCELSS3 0.5 opposed 84 cells phase do12x14a.rh2 B1 5-H17x17 Half loaded ACCELSS3 0.2 opposed 84 cells phase do12x14a.re8 B1 5-H17x17 12 cells ACCELSS3 0.8 opposed phsse do12x14a.re5 B1 5-H17x17 12 cells ACCELSS3 0.5 opposed phase do12x14a.re2 B1 5-H17x17 12 cells ACCELSS3. 0.2 opposed phase dil2x14a.rf8 B1 5-H17x17 Fully Loaded ACCELSS3 0.8- in-phase 168 cells. SSE dil2x14a.rf5 B1 5-H17x17 Fully loaded ACCELSS3 0.5 in-phase 168 cells SSE - dil2x14a.rf2 B1 5-H17x17 Fully loaded ACCELSS3 0.2 in-phase 168 cells- SSE 6-45 -- , ,. - . . . - ~ e,..... - . - . - . . , - - - , . . - , . , - , - . . ~ _ . - - - -_ - . .. - = . . _ . - -. Table 6.7.1 ( continued ) dil2x14a.rh8 B1 5-H17x17 Half loaded ACCELSS3 0.8 in-phase 84 cells SSE dil2x14a.rh5 B1 5-H17x17 Half loaded ACCELSS3 0.5 in-phase 84 cells SSE dil2x14a.rh2 B1 5-H17x17 Half loaded ACCELSS3 0.2 in-phase 84 cells SSE dil2x14a.re8 B1 5-H17x17 12 cells ACCELSS3 0.8 in phase-SSE dil2x14a.ro5 B1 5-H17x17 12 cells ACCELSS3 0.5 in-phase SSE dil2x14a.re2 B1 5-H17x17 12 cells ACCELSS3 0.2 in-phase SSE l l I i 6-46 ..m- , ,- Table 6.7.2
SUMMARY
OF WORST RESULTS FROM ALL SINGLE RACK RUNS Item Run I.D.
- 1. Maximum total vertical pedestal load: 410,190 lbs. do13x14c.rf5
- 2. Maximum vertical load in any single pedestal: 197,715 lbs. dil2x14a.rf8
- 3. Maximum shear load in any_ single pedestal: 65,950 lbs. dil2x14a.rf8
- 4. Maximum fuel assembly-to-cell wall impact load: 299.5 lbs. di13x14c.rf8
- 5. Maximum rack-to-wall impact load at baseplat level: 0
- 6. Maximum rack-to-wall impact load at the top of rack: 0
, 7. Maximum rack-to-rack I impact load at baseplat level: 0
- 8. Maximum rack-to-rack impact load at the top of rack: 0
- 9. Maximum corner displacements Top corner in x direction:- 0.2473 in. dil2x14a.rf8 in y direction: 0.1421 in, dil2x14a.rh8 Baseplate corner in x direction: 0.0315 in._ dil2x14a.rf2 in y direction: -0.0353 in, dil2x14a.rh2
- 10. Maximum stress factors Above baseplate: 0.333 (R6) do12x14a.rf8 Support pedestals: 0.484 (R6)- dil2x14a.rf8 l
l 6-47
_ . _ - - _ _ ~ _ - . . - _ _ _ _ . _ _ _ _ _ _ _ _ . . - . . _ . . . _ _ ___e_ _ . - _ . _ . _ . _ _ .. Table 6.7.3
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: do13x14c.rf8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 182 cells loaded; Fuel.ce7troid X,Y: .0, .0 (in.) Coefficient of friction at the bottom of support pedestal: 0.8
$ Revision: 3.44 S
$Logfile: C:/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov b
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov $.
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical-pedestal load: 410189.9 (2) Maximum vertical load in any single pedestal: 174699.1 (3) Maximum shear load in any single pedestal: 49820.7 (4) Maximum fuel-to-cell impact per loaded cell: 263.1 (5) Maximum rack-to-wall impact at baseplate:- . 0 (6) Maximum. rack-to-wall impact _at. rack ~ top: . 0 (7) Maximum rack-to-rack impact at baseplate:- .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS.(in.) Location: X-direction - Y-direction Top corner: .1933 .1099 Baseplate corner: .0051 .0027 MAXIMUM STRESS FACTORS
- Stress factor: -R1 - R2 R3 - R4' R5 R6 R7-Above baseplate: .038 . 041 . 127 .19 2 -'- .217 .252 .039 Support pedestal: .254 . 106 . 136 .207 .348 ..378- -- .070-
- See Section 6.5.2.3 of-the Licensing Report.for definitions.
6-48
- , ,- a .vL. n,N - .- s ann , .,,.,,-,-,e n .,mm,ne..m,--+- .,,-,wwn-a n+,,,,,,,mame w,we
Table 6.7.4 l
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: do13x14c.rf5 Seismic Loading: ACCELSS3 Fuel Assembly I.D. c ni Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 182 cells loaded; Fuel centroid X,Y: .0, .0 (in.) Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.44 $
$Logfile: C:/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 410190.3 (2) Maximum vertical load in any single pedestal: 174699.5 l l (3) Maximum shear load in any single pedestal: 51349.9 (4) Maximum fuel-to-cell impact per loaded cell: 263.1 (5) Maximum rack-to-wall impact at baseplate: .0 f (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction l Top corner: .1933 .1099 l Baseplate corner: .0051- .0027 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .038 .041 .127 .192 .217 .252 .038 Support pedestal: .254 .110 .145 .215 .374 .401 .074
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-49
.~.- _ _ _
f Table 6.7.5
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE:'R13x14C Holtec Run I.D.: do13x14c.rf2 Seismic Loading ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) Fuel Loading: 182 cells loaded; Fuel centroid X,Y: .0, .0 (in.) Coefficient of friction at the bottom of support pedestal: 0.2 SRevision 3.44 $.
$Logfile: C:/ racks / dynamo / dynamo.fov $
$ Revision: 2.5 $ ,
$Logfile:- C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C./ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS-(lbs.) (1) Maximum total' vertical pedestal' load:= 404609.4 (2) Maximum vertical load in any single pedestal: 170804.6 (3) Maximum shear load in any single pedestal: 33208.0 (4) Maximum fuel-to-cell impact per ' loaded cell: 258.4 (5) Maximum rack-to-vall impact at baseplate: . 0 (6) Maximum rack-to-vall impact at rack. top: .0
-(7) Maximum rack-to-rack impact at baseplate:: .0 (8) Maximum rack-to-rack impact at. rack top: .0 L MAXIMUM CORNER DISPLACEMENTS-(in.)
Location: X-direction Y-direction P Top-corner: .1848= .1087 Baseplate corner: .0222 .0165 MAXIMUM STRESS FACTORS'* Stress: factor: .- R1 R2 - R3 -R4 R5 'R6 R7 Above baseplate: .037 .034 . 123 .165 .2021 .234- .030-Support pedestal: .248 .070 . 133 .137 .362 .383 .068-l
- See Section 6.5.2.3 of the Licensing Report for definitions._
6-50.
. - - - , . - - ,u+, ,. i. ; - ,:n- ,.,,,,.n--. ,-0..- - ,A -a -,--r-,gw- --r., s~,,,,, rw , .+ n wny-m,
Table 6.7.6
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: do13x14c.rh8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 91 cells loaded; Fuel centroid X,Y: .0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.8 SRevisioni"' 3.44 $
$Logfile: C:/ racks /dynam0/dynam0.fov S
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov S
$ Revision: 3.36 S
$Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (1bs.) (1) Maximum total vertical pedestal load: 211871.4 (2) Maximum vertical load in any single pedestal: 130853.4 (3) Maximum shear load in any single pedestal: 35804.1 (4) Maximum fuel-to-cell impact per loaded cell: 134.8 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MMIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top cornor: .1166 .1421 Baseplate corner: .0033 .0027 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .021 .020 .101 .096 .157 .183 .022 Support pedestal: .190 .058 .139 .113 .303 .324 .071'
- See Section 6.5.2.3 of the Licensing Report for definitions.
l l 6-51
. _ _ - - . - ~ - . - - . . . . - . - . - . - . - .. - ..- - . -. - . - . - -
+
.i Table'6.7.7
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holter Run I.D.: do13x14c.rh5 Seismic Loading: ACCELSS3 Puol Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) . Fuel Loading: -91 cells loaded; Fuel centroid X,Y: .0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.44 $-
$Logfile: C:/ racks /dynam0/ dynamo.fov $
SRevision: 2.5 $ SLogfile: C / racks /dynam0/dynasi.fov $ l SRevision: 3.36 -$ C:/ racks /dynam0/dynas2.fov SLogfile: $ DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 211871.9 (2) Maximum vertical load in any single pedestal: 130850.5 (3) Maximum shear load in any single pedestal: 33810.6 (4) Maximum fuel-to-cell impact per loaded-cell: 134.8 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.). Location: X-direction Y-direction
. Top corner: .1166 .1420 Baseplate corner: 0033 .0027 MAXIMUM STRESS FACTORS *
. Stress factor: . R1' R2. - R3- R4 .R5L R6 R7 Above baseplate: .021 - . 02 0, 101 .096 .157- 183
.022 Support pedestal: .190 .055 . 130 .108 .302- . 323 .066
- See Section-6.5.2.3 of the Licensing Report.for definitions.
52
, .e.-,.. --,_n., , - - - . , ,, _ . . , . . , ,...,,,.n,, ,,,,,,,.,,,.,..m.,,. ,,n --, . , , . ,.,y-m,vn.v.,,,r-
Table 6.7.8
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: do13x14c.rh2 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 91 cells loaded; Fuel centroid X,Y: .0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.2 SRevision: 3.44 S
$Logfile: C:/ racks /dynam0/dynam0.fov $
SRevision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $ DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 209401.5 (2) Maximum vertical load in any single pedestal: 124945.7 (3) Maximum shear load in any single pedestal:' 24980.2 (4) Maximum fuel-to-cell impact per loaded cell:. 134.8 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact et rack top: .0 (7) Maximum rack-to-rack impact at: baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0
~
MAXIMUM CORNE" DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: .1042 .1234 Baseplate corner: .0116 .0196 i i MAXIMUM STRESS FACTORS
- I Stress factor: R1 R2 R3 R4 R5 R6 R7 l Above baseplate: .020 .017 .101 .082 .148 .171 .018 i Support pedestal: .182 .046 .106 .091 .272 .288 .054
~
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-53
. . . ~ - -
l Table 6.7.9
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: do13x14c.re8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) Fuel Loading: 13 cells loaded; Fuel centroid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.8
~~
$ Revision: 3.44 S
$Logfile: C:/ racks /dynam0/dynam0.fov $
$ Revision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov S SRevision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 51847.9 (2) Maximum vertical load in any single pedestal: 31272.1 (3) Maximum shear load in any single pedestal: 10898.2 (4) Maximum fuel-to-cell impact per loaded cell: 200.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 l MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: .0318 .0382 Baseplate corner: .0008 .0007 l MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 l Above baseplate: .009 .005 .024 .022 .035 .040 .006 Support pedestal: .045 .021 .045 .040 .077 .083 .023
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-54
b l Table 6.7.10 [
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C l Holtec Run I.D.: do13x14c.rc5 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 13 cells loaded; Fuel centroid X,Y: .0, 58.5 (in.) l Coefficient of friction at the bottom of support pedestal: 0.5 SRevision: 3.44 $ SLogfile: C:/ racks /dynam0/ dynamo.fov S SRevision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $ $ Revision: 3.36 $ $Logfile: C:/ racks /dynam0/dynas2.fov S DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 51848.0 , (2) Maximum vertical load in any single pedestal: 31075.3 l (3) Maximum shear load in any single pedestal: 10816.9 l l (4) Maximum fuel-to-cell impact per loaded cell: 200.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximun rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) l l Location: X-direction Y-direction Top corner: .0318 .0382 Baseplate corner: .0027 .0022 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .009 .005 .024 .022 .035 .040 .006 Support pedestal: .045 .018 .044 .035 .078 .084 .022
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-55
] Table 6.7.11
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: do13x14c.re2 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 13 cells loaded; Fuel centroid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.2
$ Revision: 3.44 S
$Logfile: C / racks /dynam0/dynam0.fov $
$ Revision: 2.5 S SLogfile: C:/ racks /dynam0/dynasi.fov S
$ Revision: 3.36 S
$Logfile: C:/ racks /dynam0/dynas2.fov S DYNAMIC IMPACT LOADS (lbs.)
(1) Maximum total vertical pedestal load: 51848.0 (2) Maximum vertical load in any single pedestal: 29809.8 (3) Maximum shear load in any single pedestal: 5961.6 (4) Maximum fuel-to-cell impact per loaded cell: 199.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLAC1.%NTS (in.) Location: X-direction Y-direction Top corner: .0299 .0299 Baseplate corner: .0107 .0099 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6- R7 l Above baseplate: .009 .004 .018 .020 .032 .036 .004-l Support padestal: .043 .011 .025 .022 .065 .069 .013
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-56
^
1 Table 6.7.12
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: di13x14c rf8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 182 cells loaded; Fuel centroid X,Y: .0, .0 (in.) Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.44 $
$Logfile: C:/ racks /dynam0/dynam0.fov $
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynan2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 403715.3 (2) Maximum vertical load in any single pedestal: 153535.1 (3) Maximum shear load in any single pedestal: 26033.1 (4) Maximum fuel-to-cell impact per loaded cell: 299.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack-impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: .1054 .1278 Baseplate corner: _.0028 .0029 l MAXIMUM STRESS FACTORS
- c Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .037 .033 .150 .103 .182 .209 .033 Support pedestal: .223 .055 .111 .108 .287 .298 .057
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-57
Table 6.7.13
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: n13x14C Holtec Run I.D.: di13x14c.rf5 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 182 cells loaded; Fuel centroid X,Y: . 0, .0 (in.) Coefficient of friction at the bottom of support padestal: 0.5
~~~~ $ Revision: 3.44 S SLogfile: C:/ racks /dynam0/dynam0.fov $ $ Revision: 2.5 $ $Logfile: C:/ racks /dynam0/dynasi.fov $ $ Revision: 3.36 $ $Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (?.Ri I (1)' Maximum total vertical pedestal l'ad: 403715.4 (2) Maximum vertical load in any single pedestal: 153534.9 (3) Maximum shear load in any single pedestal: 26088.3 (4) Maximum fuel-to-cell impact per loaded cell: 299.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-tcalk"k impact at baseplate: .0 (8) Maximum rack-;~ rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: .1054 .1278 Baseplate corner: .0028 .0029 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .037 .034 .150 .103 .182 .209 .034 Support pedestal: .223 .057 .105 .111 .295 .309 .053
- See Section 6.5.2.3-of the Licensing Report for definitions.
, 6-58
i Table 6.7.14
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: di13x14c.rf2 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 1..., ; 1550.0 (lbs.) Fuel Loading: 182 cells loaded; Fuel centroid X,Y: .0, .0 (in.) Coefficient of friction at the bottom of support pedestal: 0.2
$ Revision: 3.44 S SLogfile: C:/ racks /dynam0/ dynamo.fov S
$ Revision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov $
SRevision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $ DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 403733.1 (2) Maximum vertical load in any single pedestal: 153289.6 (3) Maximum shear load in any single pedestal: 26645.1 (4) Meximum fuel-to-cell impact per loaded cell: 297.2 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: .1054 .1240 Baseplate corner: .0034. .0069 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 RS R6 R7 l Above baseplate: .037 .028 .149 .103 .178 .204 .031 l Support pedestal: .223 .054 .103 .106 .312 .a20 .053 l
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-59
4 Table 6.7.15
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtcc Run I.D. -di13x14c rh8 Seismic Loading: ACCELSS3 , Fuel Assembly I.D. and Weight: 5-M 17x17 ;- 1550.0 (lbs.). Fuel Loading: 91 cells loaded; hael centroid X,Y:- .0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.8
$ Revision: 3.44 $ I SLogfile: C:/ racks /dynam0/dynam0.fov $
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
.SRevision: 3.36 $
SLogfile: C:/ racks /dynam0/dynas2.fov $ DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load:: 210084.9 . (2) Maximum vertical load in any single pedestal: 115056.2-(3) Maximum shear load in any single pedestal: 32420.7 I4) Maximum fuel-tewaell impact per loaded-call: 128.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top:- .0 (7) Maximum rack-to-rack impact atib'aseplate: ' .0 (8) Maximum-rack-to-rack impact at rack top: .0 MAXIMUM CORNER. DISPLACEMENTS . (in. ) Location: X-direction. Y-direction-- Top corner: .0796 .1160_ Baseplate corner: .0021 .0020 ' MAXIMUM STRESS FACTORS *- Stress factor: R1 R2 R3 R4 RS 'R6 R7f Above baseplate: .021 .019 .077 -. 074 _ .112 .130 . 019 Support pedestal: .167 .036 .135- .071 .244' .260 .069
- See Section 6.5.2.3 of the Licensing-Report:for definitions.
6 ..- -- a . . _ . - . , _ . . . _ . - , , , . . ..-. - . , - - . . . , - ,
,8 i Table 6.7.16
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Ru.i I.D.: di13x14c.rh5 Seism Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 91 cells loaded; Fuel controid X,Y: .0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.5 l $ Revision: 3.44 $ 1 SLogfile: C:/ racks /dynam0/dynam0.fov $ SRevision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
SLogfile: C:/ racks /dynam0/dynas2.fov $ DYNAMIC IMPACT LOADS (lbs.) l (1) Maximum total vertical pedestal load: 210084.9 (2) Maximum vertical load in any single pedestal: 115049.3 (3) Maximum shear load in any single pedestal: 32429.6 (4) Maximum fuel-to-ce'll impact per loaded cell: 128.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction -Y-direction Top corner: .0796 .1160 Baseplate corner: .0021 .0020 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .021 .019 .077 .074 .112 .130 .019 Support pedestal: .167 .038 .134 .075 .247 .264 .069 l
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-61
Table 6.7.17
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run 1.D.: di13x14c rh2 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 91 cells loaded; Fuel centroid X,Y: . 0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.2
$ Revision: 3.44 $
$Logfile: C:/ racks /dynam0/dynam0.fov $
$ Revision: 2.5 $
$Logfile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 210084.3 (2) Maximum vertical load in any single pedestal: 113320.1 (3) Maximum shear load in any single pedoital: 20504.8 (4) Maximum fuel-to-cell impact per loadet cell: 128.4 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: .0817 .1078 Baseplate cor'.er: .0061 .0045 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .021 .016 .080 .069 .113_ .131 .016 Support pedestal: .165 .038 .088 .074 .223 .237 .045
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-62 _ ~ - - - - - -
Table 6.7.18
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C
'"~
Holtec Run I.D.: di13x14c.re8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 13 cells loaded; Fuel centroid X,Y: . 0, 58.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.44 $
$Logfile: C:/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 S
$Logfile: C:/ racks /dynam0/dynasi.fov S
$ Revision: 3.36 S SLogfile: C:/ racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 51684.3 (2) Maximum vertical load in any single pedestal: 33213.7 (3) Maximum shear load in any single pedestal: 12874.5 (4) Maximum fuel-to-cell impact per loaded coll:- 202.3 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 l l (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) i Location: X-direction Y-direction Top corner: .0346 .0414 Baseplate corner: .0009 .0008 , MAXIMUM STRESS FACTORS
- l Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .009 .005 . 025 .024 .040 .046 .005 Support pedestal: .048 .025 . 053 .050 .087 .095 .027
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-63 9 m -
- _ _~. .
Table 6.7.19
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: di13x14c.re5 Seismic Loading: ACCEL"SS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 13 cells loaded; Fuel centroid X,Y: . 0, 58.5 (in.) Coefficient of friction at the bottom of support pedrotal: 0.5-
$ Revision: 3.44 $
$Logfile: C:/ racks /dynam0/dynam0.fov $
SRev ion: 2.5 $ SLogrile: C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov 5 DYNAMIC IMPACT LOADS (lbs.)
(1)* Maximum total vertical pedestal load: 51684.3 (2) Maximum vertical load in any single pedestal: 33157.1 (3) Maximum shear load in any single pedestal:- 12700.2 (4) Maximum fuel-to-cell impact per loaded cell: 202.3 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at-rack top: .0 (7) Maximum rack-to-rack impact atl baseplate: -.0 (8)' Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.). Location: X-direction- Y-direction Top corner: -.0335 .0414 l Baseplate corner: .0012- .0011_- l MAXIMUM STRESS FACTORS-.* Stress factor: R1 - R2 R3' R41 ~ R5' R61 RR7 i Above baseplate: .009 -.005 ' 075
. . 023- =.040, .046' -.005 -:
L Support pedestal:- .048- .019 . 052 . 037 . 089 .096~ -.027 i
- =
- See Section 6.5.2.3 of the Licensing-Report for definitions. ,
l 6-64
,, - - - - - - r --- -im., .#v. ; -, . . , - , , . . . ~ . , %. .-xm, .~..'. ,#%,-. ,,-.m. c y, w . . n e r m y - ~~. w-w M,. .~ r --y-,,<,,.~ev
Table 6.7.20
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R13x14C Holtec Run I.D.: di13x14c.re2 Seismic Loading ACCELSS3 Fuel Assembly I.D. and Weight: S-H 17x17 ; 1550.0 (1bs.) Fuel Loading: 13 calls loaded; Fuel centroid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.2 Titevisions 3.44 S
$Logfilot C / racks /dynam0/dynam0.fov $ $ Revision: 2.5 $
SLogfiles c / racks /dynam0/dynasi. foe S
$ Revision: 3.36 $ $Logfile: c / racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (1bs.) c (1) Maximum to vertical pedestal load 51694.7 (2) Maximum vortical load in any single pedostal: 29338.2 (3) Maximum shear load in any singlo pedestal: 5392.8 (4) Maximum fuel-to-cell impact per loaded cell 205.1 (5) Maximum rack-to-wall impact at baseplates .0 (6) Maximum rack-to-well impact at rack top: .0 (7) Maxinum ract.-to-rack impac,t at baseplate: .0 (8) Maximum rack-to-rack impact at rack top .0
~
MAXIMUM CORNER DISPLACEMDITS (in.) Locations X-direction Y-direction Top corner .0252 .0283 Baseplate corner: .0058 .0066
~
' MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Abovo baseplato .009 .004 .018 .018 .031 .035 ,004 Support pedestals .043 .010 .022 .020 .059 .062 .011
- See section 6.5.2.3 of the Licensing Report for definitions.
6-65
Tablo 6.7.21
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS.FOR RACK MODULE: R12x14A lioltoc Run 1.D.* do12x14a.rf8 Soismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1350.0 (lbs.) ruel Loading: 168 cells loaded; ruel controid X,Y: . 0, .0 (in.) Coefficient of friction at the bottom of support podestal: 0.8
$ Revisions'~ 3.44 $
$Logfilot C / racks /dynam0/dynam0.fov S SRevision: 2.5 S
$Logfilot C / racks /dynam0/dynasi.fov S
$ Revision 3.36 $
$Logfile C / racks /dynam0/dynas2.foy $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical podestal load: 379261.0 l (2) Maximum vertical load in any single pAdostal: 193779.7 (3) Maximum shear load in any single pedestal: 53853.2 (4) Maxipum fuci-to-coll impact por loaded coll: 240.9 (5) Maximum rack-to-wall impact at basoplato .0 (6) Maximum rack-to-vall impact at rack top: .0 (7) Maximum rack-to-rack impact at basoplate .0 (8) Maximum rack-to-rack impact at rack top .0 l l i MAXIMUM CORNER DISPLACEMENTS (in.) Locations X-direction Y-direction . Top cornor: (
.2294 .1174 Basoplate corner .0064 .0030 MAXIMUMESTRESS FACTORS
- Stross factor: R1 R2 R2 R4 R5 R6 R7 l Above baseplates .039 .039 .136 .200 .288 .333
.036 Support pedestals .282 .110 .131 .215 .427 .453 .067
- Soo SEction 6.5.2.3 of the Licensini Report for definitions.
6-66
Table 6.7.22
SUMMARY
RESULTS OF 3-D SIllGLE RACK AllALYSIS FOR RACK MODULE: R12x14A Holtec Run I.D.: do12x14a.rf5 Seiamic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (ibs.) Fuel Loading: 168 calls loaded; Fuel controid X,Y: .0, .n (in.) Coefficient of friction at the bottom of support pedestal: 0.5
'~ $ Revision: 3.44 S SLogfilot C / racks /dynam0/dynam0.fov $ $ Revision: 2.5 $ $Logfile C:/radks/dynam0/dynasi.fov $ $ Revision 3.36 S $Logfile C:/ racks /dynam0/dynan2.fov $
DY11AMIC IMPACT LOADS (1bs.) (1) Maximum total vertical podestal load: 379260.7 (2) Maximun vertical load in any single pedestal '<3778.6 (3) Maximum shear load in any single pedestal: 53622.6 (4) Maximum fuel-to-cell impact per loaded coll: 240.9 (5) Maximum rack-to-wall impact at baseplates .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplates .0 (8) Maximum rack-to-rack impact at rack top .0
' MAX 2 MUM COlWER DISPLACEMElITS (in.)
Location: X-direction Y-direction Top cornor .2294 .1174 basoplate corner .0064 .0030 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Abovo basoplato: .039 .039 .136 .200 .288 .333 .036 Support pedestals .282 .106 .164 .208 .430 .464 .084
- Soo Section 6.5.2.3 of the Licensing Report for definitions.
6-67
Table 6.7.23
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODtJLE: R12x14A Holtec Run I.D.: do12x14a.rf2 Seismic Loading: ACCELSS3 Tuol Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (Ibs.) Fuel Loading: 168 colls loaded; "uol centroid X,Y: .0, .0 (in.) Coefficient of friction at the bottom of support podestal: 0.2
$Rovision: 3.44 S
$Logfile Ct/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile C / racks /dynam0/dynasi.fov $
$Rovision: 3.36 $
$Logfile C / racks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 369616.3 (2) Maximum vertical load in any single pedostal: 188559.2 (3) Maximum shear load in any single pedestal: 37706.9 (4) Maximum fuel-to-cell impact per loaded cell 226.9 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-vall impact at rack top: .0 (7) Maximum rack-to-rack-inpact at baseplato .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Locations X-direction Y-direction Top cornor .2260 .1098 Dacoplato cornor: .0112 .0125 MAXIMUM STRESS FACTORS
- Stross factor:- R1 R2 R3 R4 R5 R6 R7 Above basoplates .036 .032 .132 .195 .282 .326 .031 Support-podestalt .274 .071 .135 .139 .411 .435 .069
- Soo Section 6.5.2.3 of the Licensing Report for definitions.
6-68 _ _ _ _ _ _ _ - - _ _ - _ - i
. _ _ .. . - . - - _ _ . ..- . - . - _. --. .....-_=...- . ~... - _- . - - - _ _ . - . _ - _ _ .
1 Table 6.7.24
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R12x14A Holtec Run 1.D.: do12x14a.rh8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) ruel Loading: 84 cello loaded; Puni controid X,Y: .0, 31.5 (in.) l Coefficient of friction at the bottom of support pedestal: 0.8
$ Revision: 3.44 $
$Logfile C / racks /dynam0/dynam0.fov $
SRevision: 2.5 S j
$Logfile C / racks /dynam0/dynasi.fov $
SRevision: 3.36 $
$Logfile: C / racks /dynam0/dynas2.fov $
DYNAMIC TMPACT LOADS (1bs.) (1) Maximum total vertical pedestal load: 192005.1 (2) Maximum vertical load in any single pedestal - 120657.9 (3) Maximum shear load in any single pedestal: 39614.1 (4) Maximum fuel-to-cell impact por loaded cella 235.7 l (5) Maximum rack-to-wall impact at baseplater .0 ' i (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplates .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) I Location X-direction Y-direction l
- Top corner: .1467- .1701 Basoplate corner .0041 .0037 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplates . 020 .022 . 100 .105 .163 . 191 .024 Support pedestals . 107 .058 . 149 .114 .300 .321 .076
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-69 n- ,, -,- e , ---.--e --,e. - , . _ . - - - . - - - . - - - ~ , - , -- .,--.r- ,-. ,
Table 6.7.25
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R12x14A Holtec Run I.D.: do12x14a.rh5 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) Fuel Loadingt 84 cells loaded; Fuel centroid X,Y: .0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.5
$ Revision: 3.44 S $Logfile C / racks /dynam0/ dynamo.fov $
SAavision: 2.5 S
$Logfile: C / racks /dynam0/dynasi.fov $
$3evision 3.36 $
$Logfile: C / racks /dynam0/dynas2.fov $
~
DYNAMIC IMPACT LOADS (lbs.) (1)' Maximum total vertical pedestal load: 192048.6 (2) Maximum vertical load in any single pedestal: 128655.2 + (3) Maximum shear load in any single pedestal: 39130.3 (4) Maximum fuel-to-cell impact per loaded cell: 236.6 (5) Maximum rack-to-wall impact at baseplate .0 (6) Maximum rack-to-wall impact at rack top .0 (7) Maximum rack-to-rack impact at baseplates .0 (8) Maximum rack-to-rack impact at rack top .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top cornor .1467 .1702 Baseplato corner .0041 .0037 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplato: .020 .020 .100 .105 .163 .191 .023 Support pedestal:- .187 .069 .146 .134 .313 .338 .074
- See Section 6.5.2.3 of the Licenbing Report for definitions.
6-70
4 Table 5.7.26
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R12x14A Holtoc Run I.D'.: do12x14a.rh2 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 84 cells loaded; Fuel centroid X,Y: . 0, 31.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.2
$ Revision: 3.44 $ l SLogfile: C / racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile: C / racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
SLogfile C / racks /dynam0/dynas2.fov S DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load 195514.5 (2) Maximum vertical load in any single pedestal: 122639.1 (3) Maximum shear load in any single pedestal: 24170.3 (4) Maximum fuel-to-cell impact per loaded cell 188.9 l (5) Maximum rack-to-wall impact at baseplates .0 i (6) Maximum rack-to-wall impact at rack top .0 (7) Maximum rack-to-rack impact at baseplates .0 l (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location X-direction Y-direction Top corner .1451 .1398 Baseplate corner .0191 .0353 MAXIMUM STRESS FACTORS
- l l Stress factor: R1 R2 R3 R4 R5 R6 R7 l Above baseplater .021 .018 .092 .096 .155 .181 . 01(,
- Support pedestals .178- .050 .097 .098 .264 .279 .049
~
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-71
,-r - - - - , , , - . - . , , , , . , , , . - . . - - < ,, . . - - , . ,- - . - .
Table 6.7.27
SUMMARY
RESULTS OF 3-D SIl1GLE RACK AllALYSIS FOR RACK MODULE: R12x14A lloltec Run 1.D.: do12x14a.re8 Seismic Loading: ACCELSS3 ruel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 12 cells loaded; Fuel controid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.8 SRevision: 3.44 $
$Logfile: C / racks /dynam0/ dynamo.fov $
SRovision: 2.5 S SLogfile C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C / racks /dynam0/dynas2.fov $
DYllAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 47776.1 (2) Maximum vertical load in any single podontal: 35564.3 (3) Maximum shear load in any single podestal: 13984.2 (4) Maximum fuel-to-cell impact per loaded cell: 215.1 (5) Maximum rack-to-wall impact at baseplates .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate -.0 (8) Maximum rack-to-rack impact at rack top: .0
~
MAXIMUM COR11ER DISPLACEME!1TS (in.) Locations X-direction Y-direction Top corner: .0461 .0496 Bascplate corner: .0021 .0016 MAX 1 MUM STRESS FACTORS *
, Stress factor: P.1 R2 R3 R4 R5 R6 R7 Above baseplate: .009 .007 .027 .031 .048 .055 .007 Support podestalt .052 .022 .056 .043 .098 .107 .029
- See Sectic; .5.2.3 of~the Liconaing Report for definitions.
. 6-72 I
i
Table 6.7.28
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULEt R12x14A , Holtoc Run 1.D.: do12x14a.re5 Soismic Loadingt ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) Fuel Loading: 12 cells loaded; Fuel centroid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support podestalt 0.5
$ Revision: 3.44 $
SLogfilot Ct/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $ $Logfile C / racks /dynam0/dynasi.foy $ $Revisiont 3.36 $ $Logfile C / recks /dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 47776.1 (2) Maximum vertical load in any single pedestal: 35562.4 (3) Maximum shear load in any single pedestalt 12615.9 (4) Maximum fuel-to-cell impact por loaded cell?. 214.9 (5) Maximum rack-to-wall impact at baseplatet .0 (6) Maximum rack-to-wall impact at rack top .0 (7) Maximum rack-to-rack impact at baseplate .0 (8) Maximum rack-to-rack impact at rack topt .0 MAXIMUM CORNER DISPLACEME.NTS (in.) Locationt X-direction Y-direction Top corner: .0450 .0493 Basoplato corner .0025 .0023 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above basoplatet .009 .006 .028 .031 .048 .055 .007 Support pedestalt .052 .018 .049 .036 .093 .100 .025 See Section 6.5.2.3 of the Licensing Report for definitions.
6-73
,.________________.-___.__.___.--d
Table 6.7.29
SUMMARY
RESULTS OF 3-D SIliGLE RACK AllALYSIf: "7 RACK MODULE: R12x14A Holtec Run 1.D.: do12x14a.ro2 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (lbs.) Fuel Loading: 12 cells loaded; Fuel centroid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support pedestal: 0.2
$ Revision: 3.44 S
$Logfile: C / racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile C / racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/ racks /dynam0/dynas2.fov $
DYllAMIC IMPACT LOADS (lbs.) (1)' Maximum total vertical pedestal load: 47775.6 (2) Maximum vertical load in any single pedestal: 28775.2 (3) Maximum shear load in any single pedestal 5652.5 (4) Maximum fuel-to-cell impact per loaded cell: 204.2 (5) Maximum rack-to-wall impact at basoplates .0 (6) Maximum rack-to-vall impact at rack top: .0 4 (7) Maximum rack-to-rack impact at baseplate: .0 (8) Maximum rack-to-rack impact at rack top .0 MAX 1 MUM COR14ER DISPLACE.ME11T5 (in. ) Location X-direction Y-direction Top corner: .0436 .0305 Baseplate corner .0227 .0223 MAXIMUM STRESS FACTORS
- Stresa factor: R1 R2 R3- R4 R5 R6 R7 Above baseplate: .009 .004 .020 .022 .040 .046 .004 Support podental: .042 .010 .023 .020 .062 .065 .012 o
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-74 ________o_ _ _ _ _ _ _ _ _ _ _
Table 6.7.30
SUMMARY
RESULTS OF 3-D SIliGLE RACK ANALYSIS FOR RACK MODULE: R12x14A Holtec Run I.D.: dil2x14a.rf8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) Puol Loading: 168 cells loaded; Funi centroid X,Y: . 0, .0 (in.) Coefficient of friction at the boctom of support pedestal: 0.8
$ Revision: 3.44 S
$Logfile: C / racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$1ogfile: C:/ racks /dynam0/dynas2.fov $
DY11AMIC IMPACT LOADS (lbo.) (1) Maximum total vertical pedestal load: 387897.4 (2) Maximum vertical load in any single pedestal: 197715.3 (3) Maximum shear load in any single pedestal: 65950.4 (4) Maximum fuel-to-cell impact per loaded cell 221.9 (5) Maximum rack-to-Wall impact at baseplatet .0 (6) Maximum rack-to-Wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplates .0 (8) Maximum rack-to-rack impact at rack top .0 MAXIMUM CORNER DISPLACEMEtiTS (in.) Location: X-direction Y-direction Top corner: .2473 .1355 Baseplate corner:~ .0068 .0036 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplates .042 .038 .155 .201 .260 .301 .030 Support pedestals .287 .142 .131 .277 .453 .484 .067
- See Section 6.5.2.3 of the Licensing Report'Yar definitions.
6-75 ____.__j
Table 6.7.31 SU!OiARY RESULTS OF 3-D SIllGLE RACK ANALYSIS FOR RACK HODULE: R12x14A iloltoc Run I.D.: dil2x24a.rf5 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-11 17x17 ; 1550.0 (1bs.) Fuel Loading: 160 cells loaded; Fuel centroid X,Y: .0, .0 (in.) Coefficient of friction at the bottom of support pedestal: 0.5
$ Revision: 3.44 $
$Logfile: C:/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile C:/ racks /dynam0/dynasi.fov $
$ Revision: 3.36 $
$Logfile: C:/rackc/dynam0/dynas2.fov $
DYNAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 387894.7 (2) Maximum vertical load in any single pedestal: 197716.2 (3) Maximum shear load in any single pedestal: 62086.6 (4) Maximum fuel-to-cell impact por loaded cell 221.9 (5) Mr mum rack-to-vall impact at baseplates .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplates .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM COR!1ER DISPLACEME!1TS (in.) Location X-direction Y-direction Top corner: .2473 .1355 ' Baseplate corner: .0068 .0036 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate .042 .039 .155 .201 .260 .301 .030 Support pedestal: .287 .132 .122 .259 .446 .477 .062
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-76
-i
. _ _ . . I
- -_-.-. - _. . - _ -__~ _ - - . _ .- __.- _ - - .
Table G.7.32
SUMMARY
RESULTS OF 3-D SI!1GLE RACK ANALYSIS FOR RACK MODULE: R12x14A Eoltec Sun I.D.: dil2x14a.rf2 Seismic Loading: ACCELSS3 Fuel Ascombly I.D. and Weight: S-H 17x17 ; 1550.0 (1bs.) Fuel Loading: 168 cells loaded; Fuel centroid X,Y: . 0, .0 (in.) Coefficient of friction at the bottom of support pedestal: 0.2
$ Revision: 3.44 $
$Logfile C:/ racks /dynam0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile C:/ racks /dynam0/dynasi.foy $
$ Revision 3.36 $
$Logfile C:/ racks /dynam0/dynas2.fov $
DYllAMIC IMPACT LOADS (1bs.) (1) Maximum total vertical pedestal load 391835.0 (2) Maximum vertical load in any single pedestal: 187995.7 (3) Maximum shear load in any single pedestal: 36526.6 (4) Maximum fuel-to-cell impact per loaded cell 258.8 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplate .0-(8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: . 2461 .1146 Baseplate corner: .0315 .0122 l MAXIMUM STRESS FACTORS
- l Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplate: .043 . 033 .142 .188 .249 .288 .027 Support pedestal: .273 . 078 .126 .154 .398 .421 .064
~ ~ ~ ~
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-77
, - -n,-c,.- , , ,m- r - - - + . , . - - ,- .--,,4 + -
I Table 6.7.33
SUMMARY
RESUL', OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R12x14A iloltec Run I.D.: dil2x14a.rh8 Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-11 17x17 ; 1550.0 (1bs.) Fuel Loading: 84 cells loaded; Fuel centroid X,Y: .0, 31.5 (in.) Coefficient of friction at the sottom of uupport pedestal: 0.8 SRevision: 3.44 S SLogfile: C / racks /dynam0/ dynamo.foy $ SRevision: 2.5 $ SLogfile C:/ racks /dynam0/dynasi.fov $ SRevision: 3.36 S SLogfile C:/ racks /dynam0/dynas2.fov $ DYHAMIC IMPACT LOADS (lbs.) (1) Maximum total vertical pedestal load: 192434.5 (2) Maximum vertical load in any single pedestal: 130423.5 (3) Maximum shear load in any single pedestal: 38260.7 (4) Maximum fuel-to-cell impact per loaded cell 145.7 (5) Maximum rack-to-wall impact at baseplate .0 (6) Maximum rack-to-wall impact at rack top .0 (7) Maximum rack-to-rack impact at baseplates .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORNER DISPLACEMENTS (in.) Location: X-direction Y-direction Top corner: .1350 .1386 Baseplate corner: .0038 .0026 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplates .020 .022 .088 .106 .154 .179 .020 Support pedestal: .190 .069 .156 .135 .312 .337 .079
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-78
1 Table 647.34
SUMMARY
RESULTS OF 3-D SIl4GLE RAC); A14ALYSIS FOR RAC); MODULE: R12x14A
~
lloitec Run I.D.: dil2x14a.rh5 Seismic Loading: ACCELSS3 i Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) ; i Fuol Loading: 84 coils loaded; Fuel controid X,Y: . 0, 31.5 (in.) Coofficient of friction at the bottom of support pedestal: 0.5 IRovision: 3.44 $
$*,ogfile: C:/ racks /dynam0/ dynamo.fov $
SRovision: 2.5 S
$Logfile C:/ racks /dynam0/dynasi.fov $ i
$ Revision: 3.36 S l SLogfile: C / racks / dynamo,9ynas2.fov $ l l
DY14AMIC IMPACT LOADS (1bs.) l
)
(1) Maximum total vertical podestal load < 192434.5 l (2) Maximum vertical load in any single pedestal: 130433.9 (3) Maximum shear load in any single pedestal: 39131.6 (4) Maximum fuel-to-cell impact per loaded coll: 145.7 (5) Maximum rack-to-wall impact at baseplates .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack lepact at baseplate .0 (B) Maxir:.m rack-to-rack impact at rack top .0 MAXIMUM COR!JER DISPLACEME14TS (in.) Locations X-direction Y-direction i Top corner .1350 .1386 Basoplato corner: .0041 .0026
~~
l MAXIMUM STRESS FACTORS
- Stross factor: R1 R2 R3 R4 R5 R6 R7.
Above basoplate .020 .022 .088 .106 .154 .179 .019 Support pedestal .190 .065 .160 .127 .313 .338 .082
~
- See Section 6.5.2.3 of the Licensing Report for definitions.
6-79
-.- , . . . , - . . . . - . , ~ . - - . . . . - . --
I Table 6.7.35
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK MODULE: R12x14A Ubitec Run I.D.: dIT2x14a.rh2 Solimic Loading ACCELSS3 Tuol Accombly I.D. and Wolght: 5-H 17x17 ; 1550.0 (1bs.) i ruel Loading: 84 cells loaded; Fuel controid X,Y: .0, 31.5 (in.) l Coefficient of friction at the bottom of support pedantal: 0.2 TRevision: 3.44 $ SLogfilot C:/ racks /dynan,0/ dynamo.fov $
$ Revision: 2.5 $
$Logfile: C / racks /dynam0/dynasi.fov $
SRovisioni 3.36 $ SLogfile C / racks /dynam0/dynas2.fov $ DYHAMIC IMPACT LOADS (1bs.) (1)' Maximum total vertical pedestal load 192434.2 (2) Maximum vertical load in any single pedestal 120198.3 (3) Maxim'im shear load in any single pedestal: 23992.7 (4) Maximum fuel-to-cell impact per loaded cell: 152.2 (5) Maximum rack-to-wall impact at baseplates .0 (6) Maximun rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at baseplates .0 (B) Maximum rack-to-rack impact at rack top .0 MAXIMUM CORNER DISPLACEMENTS (in.) Locations X-direction V-direction Top corner: .1187 .1025 Baseplato corner .0250 .0293 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplates .020 .017 .074 .085 .137 .159 .017 Support pedestals .174 .047 .102 .092 .261 .277 .052
~
- Son Section 6.5.2.3 of the Licensing Report for definitions.
l 6-80 l
ll i l Table 6.7.36 l
SUMMARY
RESULTS OF 3-D SIllGLE RACK AllALYSIS FOR RACK MODULE: R12x14A l lloltoc Run 1.D.: dil2x14a.re8 Soismic Loading: ACCELSS3 Tuol Assembly 1.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) l Fuel Loading: 12 calls loaded; ruel controid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support podestal 0.8
~~
I GRovision: 3.44 $
$Logfile C /rscks/dynam0/ dynamo.fov $
$Rovision: 2.5 $
$Logfile C:/ racks / dynamo /dynasi.fov $
SRovision: 3.36 $
$Logfile: C / racks /dynam0/dynas2.fov $
DYl1AMIC IMPACT LOADS (lbs.) (1) Maximum total vertical podestal load: 47762.3 (2) Maximum vertical load in any singlo pedestal: 33715.5 ! (3) Maximum shear load in any single podestal: 9783.6 l l (4) Maximum fuol-to-cell impact por loaded coll: 206.5 (5) Maximum rack-to-wall impact at basoplato .0 (6) Maximum rack-to-wall impact at rack top: .0 (7) Maximum rack-to-rack impact at basoplates .0 (8) Maximum rack-to-rack impact at rack top: .0
- 1. MAXIMUM COR11ER DISPLACEME!iTS (in.)
Location: X-direction Y-direction l Top cornor: .0432 .0358 Baseplato cornor .0022 .0015 l KAXIMUM STRESS FACTORS
- i Stross factor: R1 R2 R3 R4 R5 R6 R7 Above basoplates .009 .006 .021 .029 .046 .052 .005 Support pedestal: .049 .021 .036 .041 .073 .079 .019
- Soo Section 6 5.2.3 of the Licensing Report for definitions.
l 6-81
--.-~,x. , . . . -
Table 6.7.37
SUMMARY
RESULTS OF 3-D SINGLE RACK ANALYSIS FOR RACK HODULE: R12x14A iloitec Run I.D.: dil2x14a.re$ Seismic Loading: ACCELSS3 Fuel Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) Fuel Loading: 12 cells loaded; Fuel centroid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support pedestal 0.5
$ Revision: 3.44 $
$Logfile C / racks /dynam0/ dynamo.foy $
$ Revision: 2.5 $
$Logfile: C / racks /dynam0/dynasi.fov $
SPevision: 3.36 $
$Logfile C / racks /dynam0/dynas2.fov $
DYHAMIC IMPACT LOADS '1bs.) (1) Maximum total vertical pedestal load: 47762.4 (2) Maximum vertical load in any single pudostal: 33715.1 (3) Maximum shear load in any single pedestal: 10270.1 (4) Maximum fuel-to-cell impact per loaded cell 206.5 (5) Maximum rack-to-wall impact at baseplate: .0 (6) Maximum rack-to-wall impact at rack topt .0 (7) Maximum rack-to-rack impact at baseplate .0 (8) Maximum rack-to-rack impact at rack top: .0 MAXIMUM CORHER DISFLACEMENTS (in.) Locations X-direction Y-direction Top corner: .0411 .0357 Basoplate corner: .0050 .0039 MAXIMUM STRESS FACTORS
- Stress factor: R1 :R2 R3 R4 R5 R6 R7 Above basoplates .009 .006 .021 .029 .046 .052 .005 Support pedestals .049 .019 .035 .038 .077 .083 .018
- See Secticn 6.5.2.3 of the Licensing Report for definitions.
i
- 6-82 1
Table 6.7.38
SUMMARY
RESULTS OF 3-D SIliGLE RACK ANALYSIS roR RACK MODULE: R12x14A Holtoc Run I.D.: dil2x14a.ro2 Seismic Loadingt ACCELSS3 ruol Assembly I.D. and Weight: 5-H 17x17 ; 1550.0 (1bs.) Puol Loading: 12 cells loaded; Puel controid X,Y: .0, 58.5 (in.) Coefficient of friction at the bottom of support podestalt 0.2 TRevisient 3.44 S ' SLogfilet Ct/ racks /dynan0/dynam0.foy $
$ Revision: 2.5 $
SLogfilet C / racks /dynam0/dynani.foy $
$ Revision 3.36 $
SLogfilot Ct/ racks /dynam0/dynas2.fov $ DYNAMIC IMPACT LOADS (1bs.) (1) Maximum total vertical podestal load: 47765.1 (2) Maximum vertical load in any single pedestalt 29206.2 (3) Maximum shonr load in any single podestal 5833.5 (4) Maximum fuel-to-coli impact por loaded coll: 201.6 (5) Maximum rach-to-wall impact at baseplates .0 (6) Maximum rack to-wall impact at rack topt .0 (7) Maximum rack-to-rack impact at baseplates .0 (8) Maximum rack-to-rack impact at rock topt .0 MAXIMUM CORllER DISPLACEMENTS (in.) Locations X-direction Y-direction Top cornert .0421 .0317 Basoplato cornert .0186 .0208 MAXIMUM STRESS FACTORS
- Stress factor: R1 R2 R3 R4 R5 R6 R7 Above baseplatet .009 .004 .018 .024 .040 .046 .004 Support pedestalt .042 .012 .024 .023 .064 .067 .013 ,
- Sco Section 6.5.2.3 of the Licensing Report for definitions, 6-83
Table 6.7.39 COMPARISOli OF CALCULATED AND ALLOWABLE LOADS / STRESSES AT IMPACT LOCATIOliS AND AT WELDb Value Item / Location Calculated Allowable Fuel assembly / 299.5 2778 cell wall 12npact, lbs. Rack / Baseplate weld 0830 29820 psi Pedertal/ Baseplate .757 1.0 weld (dimensionless limit load ratio) l Cell / Cell spot wolds 604. 11710 lbs. (2 welds per side of box) lbs. l L 1 6-84
l t i Table 6.7.40 . STRUCTURAL AND KINEMATIC RESULTS FOR SPEllT FUEL RACKS LOCATED III T1!E SEQUOYAll CASK PIT AREA 1113 13x14 Rack 15x15 Rack Total vertical load 428000 lb. 299200 lb. on slab in cask pit Max. Load from ono 281700 lb. 157400 lb. podestal Fuel-to-cell wall 274 lb. 167 lb. peak impact load
- Maximum rack displacement .3541" .2034" in either x or y direction at top corner of rack
- Podestal stress factor .700 .542 Stress factor above baseplate .416 .198 6-85
, yw e -- ap , -9 *-y*-'- '
+i-w -. - - - -- .a q y ., m
Table 6.8.1 t MAXIMUM CORNERABSOLUTE DISPLACEMENTS AT DOTl! Ti!E TOP AND BOTTOM OF EACH RACK IN GLOBAL X AND Y DIRECTIONS FROM Wi! OLE POOL MULTI-RACK RUH: WPMRSEQS.RTR (12 Racks in the pool; cof.= random with mean=0.5; fully loaded with reg. fuel; site specific SSE.) rack uxt uyt uxb uyb 1 .1585E+00 .1422E+00 .4760E-01 . 5739E-01 2 .1843E+00 .1120E+00 .2042E-01 . 1175E-01 3 .1370E+00 .1289E+00 .2861E-01 . 3000E-01 4 .1481E+00 .1326E+00 .2568E-01 . 2446E-01 5 .3752E+00 .4058E+00 .3176E+00 . 3303E+00 6 .2301E+00 .1268E+00 .7972E-01 . 4803E-01 7 .2480E+00 .1905E+00 .1098E+00 . 8650E-01 8 .1784E+00 .2294E+00 .1213E+00 . 1064E+00 9 .1401E+00 .1289E+00 .2429E-01 . 3086E-01 10 .1802E+00 .1293E+00 .5306E-01 . 2179E-01 11 .3444E+00 .3075E+00 .2724E+00 . 2051E+00 12 .1206E+00 .2198E+00 .3913E-01 . 7292E-01
$ Revision: 1.7 $
SLogfilot C / racks /multirac/maxdisp.fov S I l 6-86
,nn- , - . - , . - -.- , - . . . . , , . . . , , , , , - . , . . , , ,. . . , , . . , -.,,.,,.._.e.. ,, . . , . , , . . - , , .
Table 6.8.2 MAXIMUM CORNERABSOLUTE DISPLACEMENTS AT BOTH Ti!E TOP AND BOTTOM OF EACH RACK IN GLOBAL X AND Y DIRECTIONS FROM Wi! OLE POOL MULTI-RACK RUNI WPMRSEQO.RFR (12 Racks in the pool; cof.= random with mean=0.5; fully loaded with reg. fuel; sito specific OBE.) rack uxt tyt uxb uyb 1 .8605E-01 .7823E-01 . 3488E-01 .3092E-01 l 2 .9033E-01 .6231E-01 . 5232E-01 .7829E-02 l 3 .5574E-01 .7389E-01 . 3237E-01 .8425E-02 ' 4 . 6933E-01 .7414E-01 - . 2788E-01 .1844E-01 ! 5 .7985E-01 .9177E-01 . 3973E-01 .4429E-01 l 6 . 1247E+00 .6698E-01 . 1118E+00 .2000E 7 . 8175E-01 .8923E-01 . 7834E-01 .4651E-01 8 . 8614E-01 .8378E-01 . 2441E-01 .1741E-01 i 9 . 1020E+00 .7105E-01 . 3627E-01 .3118E-01 10 . 5826E-01 .7308E-01 . 2958E-01 .1379E-01 1 11 . 6673E .1107E+00 . 5031E-01 .3889E-01 12 . 7988E-01 .9590E-01 . 4120E-01 .3240E-01
$ Revisions 1.7 $
SLogfilan t -Ct/ racks /multirac/maxdisp.foy $ i 6 4
.n,- ,-
, , - , , - - -.- ...,._....--,---.,,-.,...e. .n, ,n., n . - . , , . . , . ~ , , . ~ . , . .-,,r .- . , , ..n,,,
.1 Table 6.8.3 MAXIMUM RACK DISPLACEMENT 3 AND PEDESTAL LOADS IN SINGLE RACK ANALYSES AND IN WPMR ANALYSES ( Site Specific SSE-SS3; Reg. fuel, fully loaded ) Maximum RacTc Maximum Pedestal
.Run I.D. Remarks Corner Displacement Vertical Load (in.) (lbs.) .
dil2x14a.rf8 Single rack 0.2473 in X dir. 197715 (Foot-2) analysis; Rack-2, (B1) ; cof.= 0.8 dil2x14a.rf5 Single rack 0.2473 in X dir. 197716 (Foot-2) analysis; Rack-2, (B1) ; cof.= 0.5 dil2x14a.rf2 Single rack 0.2473 in X dir. 187996 (Foot-2) analysis; Rack-2, (B1) ; cof.= 0.2 wpmrsoqs.rfr WPMR analysis; 0.4028 in Y dir. 193000 cof.= random (Rack-7, Foot-4) (nean cof.=.5) cof.= coefficient of friction between pedestal and pool liner. 6-88
._, ,,_, . . , , -, - - - . , . ,.~-- - - ,- ---- - - --
a i i ! Table 6.8.4 ; MAXIMUM RACK PEDESTAL VERTICAL LOADS FROM WHOLE POOL MULTI-RACK ANALYSIS ( cof.= random; regular fuel, fully loaded; ! coismics sito specific SSE-ACCELSS3.) l
. [
RACK AND MAX. FORCE TIME PEDESTAL No. Ibs. f i RACK-it i 1.265D+05 1.326D+01 j 2 1.488D+05 1.114D+01 ' 3 1.706D+05 1. 358D+01 4 1.614D+05 1.282D+01 RACK-2: i 1 1.304D+05 1.583D+01 2 1.518D+05 1.801D+01 . 3 1.551D+05 1.424D+01 l 4 1. 4 00D+ 05 1.646D+01 RACK-3 ; 1 1.544D+05 1.582D+01 2 1.608D+05 1.1070+011 t 3 1.638D+05 1.608D+01 4 1.579D+05 1.281D+01 RACK-4: 1 1.502D+05- 4.422D+00 2 1.588D+05 4.656D+00 3 1.508D+05 1.546D+01 1 4 1.598D+05 .1.281D+01 RACK-St i 1.554D+05 9.569D+00 2 1.359D+05 1.753D+01 1.566D+05. 1 3 1.154D+01 4 '1.730D+05 9.126D+00 RACK-6: 1 1.482D+05 1.686D+01 2 1.604D+05 1.475D+01 3- 1.432D+05 1.203D+01-4- -1.403D+05- 1.625D+01 i RACK-7 1 1.513D+05 1.116D+01 "
.2 1.637D+05 1.246D+01 >
3 .1.634D+05 1.415D+01 l 4 1.930D+05 1.282D+01 i RACK-8 1 1 1.677D+05 -1.185D+01 ! 2 1.549D+05 6.286D+00 3 1.553D+05 1.203D+01 4 1.833D+05- '1.282D+01 i 6 ' l , - u-. - _ _ _ . . . . _ .. .. _ __ _ .., _ _.. _ _ _ ._,.,,_-_..u.__.__,__,.__ ,__,...,...._.m..... . . _ . . . . , ,
l i [ 3 Table 6.8.4 ( continued ) 4
~
RACK-9: 1 1.390D+05 1.185D+01 2 1.262D+05 1.114D+01 3 1.423D+05 4.125D+00 4 1.4270+05 6.497D+00 RACK-103 1 1.361D+05 1.798D+01 2 1.249D+05 1.810D+01 3 1.253D+05 1.757D+01 4 1.383D+05 9.149D+00 RACK-11: 1 1.501D+05 1.584D+01 2 1.323D+05 1.37ts+01 3 1.587D+05 1.622D+01 4 1.386D+05 1.647D+01 RACK-122 1 1.424D+05 1.789D+01 2 1.534D+05 6.150D+00 3 1.495D+05 1.760D+01 4 1.518D+05 5.743D+00 l I 6-90 l
- - . - - _ . - . . - . - - _ , - - . - .-.--.,a- _- . . . - . . - . - . - .
- - - - . _ - _ - - - . - . = - .. _ - - - -..-.- - - ._- ._-
Table 6.8.5 1 MAXIMUM RACK PEDESTAL VERTICAL LOADS ! TROM WHOLE POOL MULTI-RACK ANALYSIS l ( cof.= random; regular fuel, fully loaded; i seismict site specific ODE-1/2 ACCELSS3.) PACK AND MAX. TOHCE TIME PEDESTAL HO. Ibs. RACK-It 1 1.016D+05 1.516D+01 2 1.050D+05 4.661D+00 3 1.260D+05 1.546D+01 4 1.076D+05 1.570D+01 RACK-22 1 8.785D+04 1.577D+01 2 9.927D+04 4.660D+00 ; 3 1.002D+05 1.607D+01 l 4 9.712D+04 1.281D+01 i RACK-3: 1 8.837D+04 4.415D+00 2 1.045D+05 1.245D+01 3 1.183D+05 1.607D+01 ' 1.008D+05 1.327D+01 RACK-4: 1 1.031D+05 1.583D+01 2 1.179D+05 4.649D+00 3 1.102D+05 1.554D+01 4 9.850D+04 1.280D+01 RACK-5 1 1.189D+05 1.116D+01 2 1.074D+05 4.557D+00 3 1.206D+05 1.088D+01 4 1.235D+05 1.013D+01 RACK-63 1 1.005D+05 1.118D+01 2 1.035D+05 9.852D+00 3 9.676D+04 1.202D+01 4 1.032D+05 1.279D+01 RACK-73 1 1.001D+05 4.304D+00 2 1.034D+05 1.246D+01 3 1.154D+05 1.154D+01 4 .1.180D+05 1.279D+01 RACK-8 1 1.274D+05 1.121D+01 2 1.227D+05 4.562D+00 3 1.100D+05 1.089D+01 ' 4 1.090D+05 1.013D+01
, 6-91 1
. - . . . . _ . . - - -- - - .- . -. . - - - - . =
Table 6.8.5 ( cohcinued ) RACK-9 1 9.858D+04 1.185D+01 2 8.926D+04 1.114D+01 3 9.899Dv04 1.620D+01 4 1.121D+05 6.490D+00 RACK-10 1 9.021D+04 1.128D+01
; 2 8.087D+04 6.214D+00 3 8.161D+04 1.546D+01 4 9.675D+04 6.491D+00 RACK-11 1 1.017D+05 1.784D+01 2 7.854D+04 1.607D+01 3 9.579D+04 1.754D+01 4 1.068D+05 6.455D+00 RACK-121 1 1.100D+05 1.115D+01
. 2 1.034D+05 1.684D+01 3 9.038D+04 1.154D+01 4 1.095D+05 1.721D+01 l
l l l 6-92
,. - , . . , , . . - - . - , , . - ~ ,.
} Table 6.9.1 AVERACE BEARING PAD PRESSURE - COMPARISON OF CALCULATSD AND ALLOWABLE STRESSES f
-t STRESS fosil Max. Load Ead Sig_q fib.1 Calculated Allevable 17.5 X 17.5 193,000 630 2975 (no leak (average) ebase)
(based on concrete strength fe'
- 5000 psi)
'7.5 x 17.5
. 193,000 1847.5 2975 x 1.5 (peak) 4 factor E = 1 1
6-93
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es - - - - TVA, SEQUOYAH NUCLEAR PLANT, '
- Synthetic NS-SSE Acceleration Time History ( Set-3 ),
ci: Developed based on Site Specific Seismic Spectrum, 0.02 Damping
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~
the Original VT-SSE Site Specific Spectrum. 02 Damping. j -
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' FIGURE 6.3.33
i TYPICAL CELL WALLS
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N BASE? LATE FIGURE 6.4.1- PICTORIAL VIEW OF RACK STRUCTURE
. 6 -127
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-FIGURE 6.4.2 SCilEMATIC MODEL FOR DYNARACK 6-128
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f FIGURE 6.4.4 FUEL-TO-RACK IMPACT SPRIT.T.S
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- SSE, FrictionFully loaded with coefficient regular fuel
= random ( rneon assemblies-).
= 0.5 O~
o ...,.....,.........,......... .........i 6.00- 5.00 10.00 15.00 20.00 Iime, Sec. FIGURE 6.8.1 o
4 I i i i ! j O GAP _ TIME HISTORY. SEQUOYAH Unit 1 and 2. TVA. W-
. Gap between Rock-1 and Rock-5. Position-2. South Side, Top. !
"- SSE, Fully loaded with regular fuel assemblies; !
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w - - 4 f 00 W oo: O o.: O o: R2 GAP TIME HISTORY, SEQUOYAH Unit ' 1 and 2, TVA. ci: Gop between Rock-5 and South Wall, Position-1, East Corner, Top,
- SSE, Fully loaded with regular fuel assemblies-
- Friction coefficient = random ( mean = 0.5 ).
O-O, ...e a . . i .;i 4 . i .g ......... .. ..... 20.00 0.00 5.00 10.00 15.00 Ime, sec. FIGURE 6.8.3 D
7.0 ACCIDERT_MALYSILAND_HIEELLAllEQ111_SIEUCTUPAL EVALVAUDER 7.1 Introduction This section provides results of accident analyses and miscellas.*ous evaluations perform 2d to demonstrate regulatory complianes of the now fuel racks. Accident events considered are taken from Ref. (7.1.1). The following limiting accident and miscellaneous structural evaluations are considered: Refueling accident - dropped fuel assembly, or fuel handling tool plus fuel assembly, or drop of gate
- Local cell well buckling
- Structural adequacy of the impact shie H for cask pit Analysis of welded joints due to isolatud hot cell 7.2 Refuelina Accidents The Updated Final Safety Analysis Report (UFSAR) (7.2.1) states that the fuel hoist system is supplied with redundant load support paths such that the failure of any one component will not result in freo f all of the fuel assembly. The combined weight of a fuel assembly plus gripper is approximately 2100 lb. Controls on crane movement are such that the top of an active fuel assembly can only be raised to wit.hin 10 feet of the top of normal water level.
Despite the controls imposed on the crane, a conservative accident evaluation of the fuel racks should include the effect of a fuel ( assembly separating from the gripper. Drop accidents focusing on the integrity of the rack structure due to such drops are considered below. 7-1
7.2.1 DIsppad_EncLAutFhlY The consequences of dropping a fuel assembly as it is being moved over stored fuel is discussed below. Based on the highest lift of a fuel assembly (7.2.1), the maximum distance from the bottom of a fuel assembly, travelling over fuel racks, to the top of the rack is 36 inches.
- a. DImpped Puel Assembly. Accident 1 A 2100 lb. fuel assembly plus gripper in dropped from 36" above the top of a storage location and 12npacts the base of the module. Local failure of the baseplate is acceptable; however, the rack design should ensure that gross structural failure does not occur and the suberiticality of the adjacent fuel assemblies is not violated. Calculated results show that there will be no change in the spacing between cells. Local deformation of the baseplate in the neighborhood of the impact will occur, but the dropped assembly will be contained and not impact the liner. We show that the maxistun movement of the baseplate toward the liner af ter the 12npact is less tha 1.98". The load transmitted to the liner through the support by such an accident is well below the loads caused by selsmic events (given in Section 6).
- b. QImpped Puel Assembly Accident II One fuel assembly plus the gripper (2100 lb.) is dropped from 36" above the top of-the rack and impacts the top of the rack. Permanent deformation of the rack is acceptable, but is required to be limited to the top region such that the rack cre is-sectional geometry at the level of the top of the active fuel (and below) is not altered. It is shown that damage, if it occurs, will-be restricted to a depth of 6.1" below the top of the rack. This is above the active fuel region.
- c. Drop of 3820 lb. Cate We consider the drop of a 3820 lb. mass from 24" above the top of the rack. Anklysis shows that this drop causes less damage to the rack than accident case b.
a 7-2
l l 7.3 Local Buckling _.of Puel cell Walla This subsection and subsection 7.5 present details on the secondary stresses produced by buckling and by temperature effects. The allowable local buckling stressna in the fuel cell walls are obtained by using classical plate buckling analysis. The following formula for the critical stress has been used based on a width of cell "b" (7.3.1): n2 Et2 Ocr " 12 b2 (1.p) 2 acr is the limiting vertical compressive stress in the tube, E= 27.9 x 106 psi, p = 0.3, (Poison's ratio), t= .060" (away from a pedestal), b = 8.75". The factor p is suggested in (Ref. 7.3.1) to be 4.0 for a long panel. Near a pedestal, additional cell wall strength is provided by added strip material which increases the ef fective thickness of the region prone to huckling to .1045" in the highly loaded region. For the given data, acr = 14387 poi It should be noted that this stability calculation is based on the applied stress being uniform along the entire length of the cell wall. In the actual fuel rack, the compressive stress comes from consideration of overall bending of the rack structures during a seismic event and as such is negligible at the rack top and maximum at the rack bottom. It is conservative to apply the above l equation to the rack cell wall if we compare a tc with the maximum l compressive stress anywhere in the cell wall. As shown in Section 6, the local buckling stress limit is not violated anywhere in the body of the _ rack modules. The maximum compressive stress in the 7-3 _ _ _ , , .._s , , , , - . . - - . -- -
t outermost cell is obtained by multiplying the limiting value of the stress factor R6 (for the cell cross-section just above the baseplate) by the allowable stress. Thus, from Table 6.7.2, o = R6 x allowable stress = .333 x 25000 = 8325 psi under faulted conditions. i 7.4 Analysis of the Impact Shield for cask Pit To maximize the storage capacity of the spent fuel pool, a spent fuel storage rack containing 225 cells (15x15 cells) is proposed to be installed in the 12'x12' cask loading area of the cask pit of the Sequoyah spent fuel storage pool. After installation of the rack in the cask pit, the pit will be aquipped with a removable impact shield to prevent accidental dropping of any object on the fuel rack. The proposed impact shield in shown in Figure 2.4.16. It consists of panel coverplates attached to a frame made of wide flange beams. This shield is deoigned to withstand a total load of 288,000 lbs. uniformly applied on the whole shield, or a total load of 70,000 lbs. uniformly applied on one of the panel plates. The panel plate thlekness is determined by a limit load analysis, and the dimensions of the wide flange beams are chosen so that the maximum stresses in the frame for the pontulated load cases are within the corresponding allowables. The ANSYS finite element program is used to perform the frame stress analysis. The results are summarized below: l (1) Panel plate can resist a uniform load of 70,000 lbs. on one panel or a concentrated load of 7952 lbs. applied at l any point without sustaining a plastic collapse. (2) Haximum direct plus bending stress in the frame beams is 51961 psi, which is below 90% of the ultimate material l strength. Maximum average shear stress is 2850 psi, which is less than the postulated allowable (36,000 psi). (3) Maximum average compression stress on concrete wall at the bearing locations is 329 pai, which is considerably lower than the allowable (2975 psi). 7-4
,, - ~ . - . , , ,
_ _ _ _ - -.__ _ _ _ _ _ ~- __ _ ___ _ _ _. _ . _ _ _ _ - _ - - - _ . . Figure 7.4.1 shows the allowable drop height as a function of the i heavy load weight with the cross-sectional area of the load as the parameter. From this figure, the allowable drop height for any heavy load can be ascertained by interpolation. 1 7.5 Analysis of Welded Joints in Rack due to Isolated Hot J Cell 1-In this subsection, in-rack welded joints are examined under the loading conditions arising from thermal effects due to an isolated hot cell. A maximum thermal gradient between cells will develop when an isolated storage location contains a fuel assembly emittine maximum postulated heat, while the . surrounding locations are empty. We can obtain a conservative - estimate of weld stresses along the length of an isolated hot cell by considering a bem l strip (a cell wall) uniformly heated .ind austrained_ from growth l along one long edoa. The strip is subject to a uniform temperature rise AT = 55.09'F. The temperature rise has been calculated from i the difference of the maximum local water temperature and bulk water temperature in the spent fuel pool. (see Tables 5.5.1 and 5.7.1). Then, using a shear beam theory, we can calculate an estlanate of- the maximum value of. the average shear stress in the strip (see Figure 7.5.1)'. The final result for wall-maximum shear stress, under conservative restraint assumptions is given as [7.5.1): E a AT tmax " - where a = 9.5 x 10-6 in/in 'F. 7-5 ;
,--vs,--- ...E.,..Mmm.,-y,---, ,-I,',-,--n...-..-,7,-,,-,,r, ,w,,,, - , - , .,,-.,qe. e.,-,,. cry.w,,,--,.- .y~,,..--
k-Therefore, we obtain an estimate of maximum weld shear stress in an isolated hot cell as raax = 15684 poi Since this is a secondary thermal stress, it is appropriate to cornpare this to the allowable weld shear stress for a faulted event r < .42Su = 29820 psi. In the fuel rack, this maximum stress occurs near the top of the rack and does not interact with any other critical stress. 7.6 References for Section 7 (7.1.1) TVA Specification 3954-3QNP-90, Revision 1, p. 42. [7.2.1) TVA Sequoyah Nuclear Plant Updated Final Safety Analysis Report", April, 1991, Section 9.1. [7.3.1) " Strength of Haterials", S.P. Timoshenko, 3rd Edition, Part II, pp 194-197 (1956). [7.5.1) " Seismic Analysis of High Density Fuel Racks, Part III-Structural Design Calculations - Theory", HI-89330, Revision 1, 1989.. k i 7-6 I a
--: a :
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PIGURE 7.3.1 LOADING ON RACK WALL 7-7
Note: These curves are not applicable if there are static loads resting on the impact saileid. 10 's- - 10 's 3 -
- -Impact Over An Area Ev 10 *=_ of' 144 sq.ft.
E _ en y- u e _ o. 10 's _ 8 Impact Over An Area g - of 8 sq.ft. o - I
~o 2
.g 10 i impact Over ' An Arece of .1 sq.ft.
l _ 10 ' ' ' ' * . 'n * * ' ' ' ' s. ' ' '
s
! 1 10 10 .' 10 ' Allowable,. Height Above Shield Surface, in. FIGURE.7.4.1 , RELATIONSHIP BEDfEEN LOAD, ALLOWABLE HEIGHT AND INPACT AREA FOR DROPPED OBJECT m
.- . m -.
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i l
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l 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDIRATIONS l 8.1 Introduction Sequoyah spent fuel pool is a sa.toty related, seismic category i;einforced concrete structure. In this section, the analysis to j demonstrate the structural adequacy of the pool structure, as j required by Section IV of the USNRC OT Position Paper (8.1.1) is abstracted. It is racognized that, from the pool structure '
- integrity standpoint, V proposed raracking of the sequoyah spent l fuel pool with free-standing racks will result in reduced loading to the pool walls, which are the only regions of the pool subject to loadings comparable to their structural strength. l The Sequoyah pool slab is over 20' thick reinforced concrete supported on rock, and is therefore not a candidate region of ;
potential deficiency in structural strength. The pool walls in the Sequoyah pool presently serve as the ultimate anchors against* 1ateral motion of the racks. A gridwork of beams spanning the entire pool planform undergirds ths existing racks and serves to
- transfer shear loads to the pool walls. This inertial loading will be eliminated in the raracking operation unburdening the walls of i an important loading source. In vi w of the reduction in loading, it would be adequate to refer to the prior pool' structural i qualification performed by TVA= and to dispense with a new analysis.
However, in the interest of revalidation of prior work, a , reanalysis of the pool structure for the new loadings is performed. i The original design bases are- retained, namely the Design Basis Earthquake evaluated for the N-S and vertical and for the E-W , and vertical seismic loadings - separately. The original design ., basis ACI Code (8.1.2) and the. Working Stress Design (WSD) Method are also invoked to- maintain consistency- with the previous.
~
qualification effort. . However, the method of analysis .is . upgraded from the-classical static analyses to the Response Spectrum Method.
'the latter methodology enables a realistic characterization.of l
l 8-1
nydrodynamic sloshing loads during seismic event s which could only be incorporated with a large margin of uncertainty in the static analysis procedures. Pool structural loading involves the following discrete conponents: a) Static Loadina
- 1) Dead weight of pool structure plus pool water (ine.luding hydraulic pressure on the pool walls).
Com aining the hydrostatic and structure dead weight is in conformance with (8.1.2).
- 2) Dead weight of rack modules and fuel assemblies stored in the modules.
b) Dynamic Loadina
- 1) Vertical loads transmitted by the rack support pedestals to the slab during a DBE or OBE seismic event.
- 2) Inertia loads due to the slab, pool walls and' contained wate. mass and sloshing loads which arise during a seistic event.
- 3) Hydrodynamic loads caused by rack notion in the pool during a seismic event.
c) Thermal Loadina
- 1) Mean temperature rise and temperature gradient across the pool slab and - the pool walls due to temperature differential between the pool water and the atmosphere external to the slab and walls. Both normal and pool boiling (loss of cooling accident) conditions are consideref.
The Sequoyah spent fuel pool region is pictorially illustrated in Figure 8.1.1. The pool structure is analyzed using the fini'.e element method. The results for the above load components are combined using load combinations mandated by ACI [8.1.2] using the
" working stress" design method. It is demonstrated that for the critical load combinations, design criteria requirements are maintained when the fuel pool is assumed to be fully loaded with high density fuel racks with all storage locations occupied by fuel 8-2
s - ar,semblies. The general purpose finite element code ANSYS (8.1.3] is utilized to perform the analysis. The critical regions examined for the fu6 pt-1 are the slab and the critical wall sections adjoining th. p. . slab. Both moment and shear capacities of the cru U 4 wall regions are checked for structural integrity. Local punchi.ng and bearing integrity of the slab in the vicinity of a rex inodule support peAstal pad is evaluated. Structural capacity evaluations are carried out in accordance with the requirements of the American concrete Institute (ACI) (8.1.2]. 8.2 General Features of the Model The fuel pool model is constructed using information from plant auxiliary building structural drawings. A description of the pool structure modeled for analysis is given in the following: ' The fuel pool slab is a 20.71' thick reinforced concrete clab supported on rock with irside dimensions 53'-1/4" long and 31' 1/2" wide. The top of the slab is located at elevation 685.71' and its long direction is considered aligned along the plant North-South direction. The East edge of the slab has a 7' thick vertical reinforced wall which rests on the slab and which extends above the slab to level 734'. The West edge of the slab has a 6' thick wall from the slab to level 734' The West wall separates the fuel pool from the fuel transfer canal. The canal is not modelled; however, the discontinuity in the wall structure in the center of the West wall is included. The North wall is a 6'-11-1/2" thick wall extending from the slab to level 734'. The South edge of the slab has a 7' thick wall extending up to level 734'. The cask area is located in a 12' x 31'-8-1/2" space at the-north end of the pool. The cask area is separated from the spent fuel storage area by a l'-6" thick wall aligned along the East-West direction. This dividing wall estends to elevation 734', and has an opening to 8 . _ _ _ _ _ _ _ - - - _ _ _ _ .
transfer fuel assemblies extending down to level 702.12' . All wall modeling is done to level 734'; free edges are assumed at this level. Figure 8.1.1 shows a schematic of the above geometry which is modelled and analyzed using the finite element mithod. The pool is assumed to be loaded with 12 high density fuel racks having a total of 2091 cells. For analysis purposes, each cell is assumed to contain a 1550 lb. Weight fuel assembly. Figure 8.2.1 shows a layout of the entire 3-D finite element model. The hydrodynamic loads arising from the rack in the cask pic are not modelled in this analysis. Neglect of any horizontal loads from this rack in the cask pit yields a more conservative analysis of the intermediate wall between the main pool and the cask pit. Hydrodynamic loads on the other bounding walls of the cask pit due to ti..s rack are less than corresponding loads on walls in the main pool area. Therefore, results obtained for the walls in the main pool area control the integrity analysis. The gridwork in different . regions shows the totality of elements used. The finite element model is only of the five vertical walls; the slab is not modeled but its effect on the vertical walls is considered by assuming complete mechanical fixity at all wall-slab interfaces. In this plant, the depth of the slab (20.71') is- such that it may be considered as a rigid body for the purposes of structural re-qualification of the vertical walls; the slab is only structurally examined to demonstrate satisfaction of local punching and bearing requirements. Growth of the slab is considered in the analysis of the fuel pool walls under thermal loading. Fluid sloshing effects are included by using a fluid model based on masses and springs in accordance with (8.2.1). The sloshing fluid mass -is connected to the slab valls by weak springs tuned to reproduce the sloshing. frequency. The remainder of the fluid mass is coupled to the structure using stiff springs. 8-4
The finite element model is constructed using the ANSYS classical shell element STIF63 in the ANSYS finite element code. The shell element thickness in the various regions of the structure is the actual thickness of the structure at the location. The finite element model is prepared for the analysis of both meenanical load and thermal load. Tho effects of structural reinforcement and the properties of the concrete (cracked or uncracked) are accounted for in the finite element model by establishing an appropriate effective modulus for each shell element. Effective moduli are defined for each local in-plane axis for the shell elements. The different moduli reflect the fact that different reinforcement geometries may be used in perpendicular directions of the plate-like sections when the different concrete section assumptions (cracked or uncracked) are applied to the structure. Only major reinforcement which affects the plate and shell behavior of th'e structure is incorporated into the definition of the effective. moduli; additional local reinforcement in various areas of the pool structure are neglected in the defining of the effective moduli. However, such local reinforcement may be accounted for in the stress evaluation af ter results are obtained for local bending moments. The variation of the extent of reinforcement is taken into account in the finite element model by defining different material types as necessary to reflect the varying values of effective moduli in different regions. Uncracked section properties are assumed for the initial mechanical load analyses. For the thermal analyses, it is shown that the therral gradients will always yield a cracked section if the uncracked stiffness is used; that is, an iterative solution is used to show that cracked section properties should be used for all finite element analyses for thermal loading. The effective properties for the elements used in the finite element model are calculated using standard procedures for reinforced concrete sections to define equivalent effective homogeneous materials having the appropriate stiffness and strength, 8-5
, , _ _ _ _ , , _ , . _ , _ _ _ _ , _ _ _ _ _ - - - - - - ~'
8.3 Loadino Conditions i To evaluate response due to the dif ferent load mechanisms outlined in Section 8.1, the following finite element analyses are carried out. Loading cases are defined below which enable us to obtain the moments and shears for required combined loadings (8.1.1) by linear combination. Case 1: Dead loading from concrete, reinforcement and 41.29 feet of hydrostatic head. The loading is applied as a 1.0g vertical gravitational load for the structure and a surface pressure on the walls to simulate the hydrostatic head. The cask pit is assumed to be full of water. The fuel transfer canal is assumed empty for this analysis to maximize west wall lateral loading. Cace 2: Saismic horizontal loading due to pool structure mass and contained water mass. A response spectrum analysis is performed with the contained fluid modelled as impulsive mass and rigid mass (8.2.1). The input response spectra loading are the two - horizontal-design basis seismic response spectra at 5% damping. Subsequent load combinations reflect the postulated 2-D analysis carried out for each horizontal direction separately. Case 3: Seismic horizontal load due to hydrodynamic effects from fluid coupling caused by rack motion relative to the walls. The level of horizontal pressure loading is obtained from the results of spent fuel rack analyses outlined in Section 6. This analysis is carried out for the SSE pressure distribution. Due to the small contribution from this effect, the i loading is also used for the OBE condition in the load combinations. For conservative analysis of the intermediata wall, as noted previously, no rack is assumed in the cask pit. Case 4: A mean temperature rise plus a thermal gradient applied across the walls to sin. late the heating effect of the water in the pool. This gradient is calculated based on appropriate- surface heat transfer coefficients -and on maximum wall temperature deduced from the pool bulk temperature calculations for the licensing basis as discussed in Section 5. The cask area is assumed to be full and the fuel transfer canal is assumed to be empty for thermal stress analysis. 8-6 A - - _ - _ - _ - _ _ _ _ _ _ - - _ -
Case 5: A mean temperature rise plus a thermel gradient applied across the walls to simulate the case of pool boiling. For this abnormal case thermal analysis, the cask area is assumed to-be full and the fuel transfer canal is assumed to be empty. For subsequent discussion of structural integrity checks using various mandated load combinations, the above individual finite element load cases are referred to as load cases 1-5, respectively. For the pool structural analysis, the following load combinations are evaluated: Load combinations Allowable WSD Stresses ( Case I = D (concrete) Case I = D + T, f, = .. 44f5fy ',(steel) f, = Case II = D + E f, = 0. 4 5 f ', f, = 0. 50 f y Case II = D + E + T, f, = 0. 4 5 f ', f, = 0. 50 f y - Case III = D +-E' f, = 0.75-f', f = 0.90 f Case III = D + E' +T u f = 0.75 f l, _ f, = 0.90 f y Case IV = D-+ Ta f , = 0 . 7 5 f ' ,. f, = 0.90 f y In the above loading combinations, the load conditions are defined as follows: D = Dead weight and: hydrostatic load E' =-Design Basis Earthquake (DBE)- E = Operating Basis Earthquake (OBE) T, = Normal Thermal Load T, = Accident Thermal Load -(pool boiling) , WSD'is the abbreviation for-Working Stress Design Method. 8-7
The appropriate load conditions are formed from the results of finite element cases defined at the beginning of this subsection as follows: D = case 1 E' = 1.0 x case 2 + 1.0 x case 3 E = OBE amplifier (1/2) x case 2 + 1.0 x Case 3 T, = case 4 Tg = case 5 Load combinations are formed using absolute values where necessary , o as to maximize critical combined stress resultants. B.4 Results of Analyses The ANSYS postprocessing capability is used to form the appropriate , load combinations identified above and to establish the critical bending moments in various sections of the pool structure. For this plant, the following limit strengths for concrete and for reinforcement are used in the computation of allowable stresses (8.4.1]. concrete f,au 5000 psi (compression) (8.4.2] reinforcement = f y = 60000 psi (tension / compression) In each section, we define the safety margin for bending as the minimum allowable bending moment divided by the calculated bending moment (from the ANSYS-postprocessing of the required load-cases). Table 8.4.1 summarizes the results obtained from the finite element analyses and shows minimum safety margins on each section of the structure. The floor slab is also checked against local bearing strength in the vicinity of highly loaded rack pedestals. It is shown that the average bearing loads under a pedestal subjected to an instantaneous peak load meets the requirements of [8.1.2]. 8-8 I l
8.5 Pool Liner The pool liner is subject to in-plane strains due to movement of the rack support feet during the scismic event. Calculations are made to establish that the liner will not tear or-rupture under all loading conditions in the pool and that the liner can withstand DBE and OBE seismic events without fatigue failure. An ANSYS enalysis of a linet plate section subjected to vertical and horizontal static pedestal loading is carried out. The maximum stress in the liner and liner weld during a DBE seismic event is 22994 psi which is less than the ultimate strength of .the liner material, i.e., tear or rupture failure will not occur. The time history result for the pedestal loading, obtained from the WPMR analysis, is then used to evaluate the number of stress cycles to be expected in the liner for each seismic event. The cumulative damage factor (CDF) (limiting code value 1.0) is computed and shown to be less-than.
.0105 in critical regions of the liner and attachment locations.
The number of stress cycles used in the CDF evaluation is based on 1 DBE and 20 OBE events. Due to the presence of a rack in the cask pit, the cask pit liner is also subjected to in-plane strains. The cumulative damage factor is less-than 1.0 in the cask pit. 8.6 Conclusions Critical regions affected by loading the fuel pooi completely with high density racks are examined for structural integrity under _ bending and shearing action. It in determined that adequate safety _ factors exist assuming that all racks are fully loaded with fuel and that the factored load combinations are checked against the appropriate structural design strengths. The analysis is carried - out assuming that no rack is present in-the cask pit area. . Results are conservative in that ; the-- neglect of the-'sme11 hydrodynamic pressure load from a rack in-the cask pit will:cause'a higher net loading on the 18" intermediate wall than will actually be present. It is also shown that local frictional loading on the liner-in both 8-9
--_--_---__a l
the main pool and in the cask pit results in stresses that are low ecough so that liner fatigue is not a concern. 8.7 References for Section 8 (8.1.1) OT Position fot' Review and Acceptance of Spent Fuel Handling Applications, by B.K. Grimes, USNRC, Washington, D.C., April 14, 1978. (8.1.2) ACI 318-63, 318-71, Building Code Requirements for Reinforced Concrete, American Concrete Institute, Detroit, Michigan. (8.1.3] ANSYS User's Manual, Swanson Analysis Rev. 4.4A, 1990. (8.2.1] " Nuclear Reactors and Earthquakes, U.S. Department of Commerce, National Bureau of Standards, National Technical Information Service, Springfield, Virginia (TID 7024). (8.4.1) TVA Sequoyah Nuclear Plant Updated Final Safety Analysis Report, April, 1991, Table 3.8.4-1 (sheet 3). . (8.4.2) SQN-DC-V-1,1, " Design of Reinforced Concrete Structures", K12, 1989. y 8-10 i
Table 8.4.1 SAFETY FACTORS FOR BENDING OF POOL STRUCTURE REGIONS REGION SAFETY MARGIN Intermediar.e Wall 1.11 East Wall 1.33 South Wall 1.18 West Wall 1.18 Above the limits prescribed by the . working stress design method. -g 9 V 8-11 < i
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T 9.0 RADIOLOGICAL EVALUATION 9.1 Fuel Handlina Accident The potential radiological consequences of a fuel handling accident in the Sequoyah Auxiliary Building have been dete" mined. 9.1.1 Assumptions and Source Term Calculations Evaluations of the accidents were based on fuel of 5.0 vt% initial enrichment burned to 60,000 Mad /mtU. The reactor was assumed to have been operating at 3565 Mw thermal power prior to shutdown; this yields a (conservative) specific power of 40.00 kw/kgU. Except for fuel enrichment and discharge burnup, the assumptions used in the evaluations are the same as those previously reviewed and accepted by the U.S. Nuclear Regulatory commission. As in the UFSAR evaluation, the fuel handling accident was conservatively ac med to result in the selease of the gaseous fission products ccntained in the fuel / cladding gaps of all the rods in the peak-power fuel assembly at the time of the accident. Gap inventories of fission products available for release were estimated using the release fractione identified in (1) NUREG/CR-5009' and (2) Regulatory Guide 1.25 .2 NUREG/CR-5009 has confirmed that the Reg Guide 1.25 assumptions remain conservative for extended burnup except for I-131, for which the release fraction is reported to be 20% higher. Dose calculations were D.A. Baker, et al., " Assessment of the Use of Extended Burnup Fuel in Light Water Power Reer tors", NUREG/CR-5009 (PNL-6258) , Pacific Northwest Laboratory, February 1988. 2 Regulatory Guide 1.25 (AEC Safety Guide 25), " Assumptions Used For Evaluating The Potential Radiological Consequences Of A Fuel Handling Accident In The Fuel Handling And Storage Facility For Boiling And Pressurized Water Reactors." 9-1
0 ! performed for both sets of release fractions, and for a fuel
. - cooling time of 100 hours.
The gaseous fission products that have significant impacts on the off-site doses following short fuel cooling periods are the short-lived nuclides of iodine and xenon, which reach saturation inventories <aring in-core operation. These inventories depend primarily on the fuel specific power over the few months immediately preceding reactor shutdown. In the highest power assembly, the specific power and hence the inventory oc iodine and xenon will be proportional to the peaking factor (assumed tn be 1.65 per Reg Guide 1.25). At the conservative (short) cooling time of 100 hours used in the sequoyah calculations, most of the thyroid dose comes from Iodine-131, while most of the whole -body dose comes from Xenon-133. At longer cooling times, Iodine-131 remains the dominant d isotope for thyroid dose, while the major contributor to whole-body dose becomes Krypton-85 (the sht er-lived Xenon-133 having decayed to very low levels) . The '.,ses after long cooling periods are so low as to be insignificant compared to the doses calculated here for the very short cooling time of 100 hours. ~ Though the single iodine and xenon isotopes are the major contrib". ors to off-site doses, the contributions from other isotopes are calculated and included in the overall dose values. The present evaluation uses values for atmospheric diffusion factor (y/Q) and for filter efficiencies that have previously been reviewed and necepted. Cere specific inventories (Ctries per metric ton-of uranium) of fission products were estimated with 3 the ORIGEN-2 code , based upon parameters stated earlier (specific power of 40.00 Mw/mtU, initial enrichment of 5. 0 wt% U, 3 QENL Isotope GMeration and Depletion, ORNL/TM-7175, Oak Ridge National Laboratory, July, 1980. 9-2
burnup of 60,000 Mwd /mtu, and a cooling time of 100 hours). The results of the ORIGEN calculations for isotopes of interest are given in Table 9.1, while the percentages of the core inventories ' released from the fuel to the fuel rod gapc under the assumptions of NUREG/CR-5009 and Reg Guide 1.25 are listad in Table 9.2. The following equation, from Reg Guide 1.25, was used to calculate the thyroid dose (D, in rads) from the inhalation of radiolodine. Values for many of the terms in the equation are given in Table 9.2 and Table 9.3. Dose = E - , where DF, DFg F, = fraction of fuel rod iodine Rg = dose conversion-inventory in gap space factor (rads per curie). I n = core iodine radionuclide inventory at time of the y/Q= atmospheric diffusion accident (curies) factor (see per cubic F= fraction of core damaged ao as to release iodine in'the DF # =-effective iodine rod gap decon. factor for pool water P= core peaking factor
- DFg= effective-iodine l B= breathing rate (cubic meters -decon.: factor for j per second) i
-filters The equations given on the following page were used to calculate the external whole-body dose from beta and gamma radiation in the cloud of radionuclides released in the -fuel-handling ' accident.
The equations contain several of the terms defined above, i l Dose, = I O.23 (y/Q) FPG g E p.g. I 1 9-3 __ m - ..-e. * + m--r- "
l Dose y = II 0.25 Q/Q) FPG i E y,i. In these expressions, G i is the gap inventory of the gaseous radionuclides of xenon and krypton, while the E sab eeipt term is the average energy per disintegration of each radionuclide (in Mev per disintegration, as given in Table 9.3). These functions assume the noble gas decontamination factors in water and the charcoal filters are 1.0. The gap inventories of radioiodine make negligible contributions to the whole body doses, D, and D, , because cf the large decontamination factors appropriate to the iodines. 9.1.2 Results The sequoyah site boundary doses from tha specified fuel handling accident ars tabulated below. The doses, which are based on the release of all gaseous fission product activity in the gaps of all the fuel rods in the highest power assembly, are given for NUREG/CR-5009 and for Reg Guide 2.25 source-term acumptions. NUREG REG G Thyroid dose, rad = 16.17 13.60 Whole-body dose, rem Beta dose, D. = 1,74 3.59 Gamma dose, D, = 0.40 0.76 Whole body total = 2.14 4.15 These potential doses are well within the exposure guideline values of 10 CFR Part 100, paragraph 11. As defined in Standard Review Plan 15.7.4, Radiolocical Consecuences of Fuel Handline Accidents, "well within" means 25 percent or less of the 3 0CFR100 guidelines, or values of 75 rad for thyroid doses and 6.25 rem 9-4
for whole-body doses. The potential doses at Sequoyah from the conservative scenarios presented here easily meet the criteria for "well within." 9.2 Solid Radvaste The necessity for resin replacement is determined primarily by the requirement for water clarity, and the resin is normally changed about once a year. No significant increase in the volume of solid radioactive wastes is expected with the expanded storage capacity. During reracking operations, a small amount of additional resins may be generated by the pool cleanup system on a one-time basis. 9.3 Gasenus Releases Gaseous releases from the fuel storage area are combined with other plant exhausts. Normally, the contribution from the fuel storage area is negligible compared to the other releases and no significant increases are expected as a result of the expanded storage capacity. l 9.4 Personnel Exposures During normal operations, personnel working in the fuel storage arca are exposed to radiation from the spent fuel pool, operating experience has shown that the area radiation dose rates, which originate primarily from radionuclides in the pool water, are generally 1 to 3 nrem/hr, with an occasional reading of 5 mrem /hr. Dose rates on the pool bridge crane platform are 4 to 5 mrem /hr. These doses may temporarily increase slightly during refueling operations. 9-5 6 r
t Radiation levels in zones surrounding the pool are not expected to be significantly af fected. Existing shielding around the pool { twater depth and concrete walls) provide more than adequate @ pt "+ 9 tion, despite the slightly closer approach of the new racks f r lo ae walls of the pool. Representative concentrations of radionuclides in the pool water are shown in Table 9.4. During reracking operations, the con-centrations might be expected to increase due to crud deposits spalling from spent fuel assemblies. However, experience to date has not indicated a major increase as a consequence of reracking, operating experience has also shown that thcre have been negligible concentrations of airborne radioactivity and no increases are expected as a result of the expanded storage capacity. Area monitors for airborne activities are available in the immediate vicinity of the spent fuel pool. No increase in radiation exposure to operating personnel is expected; therefore, neither the current health physics program nor the area monitoring system needs to be modified. 9.5 Anticipated ErnogltLe Durina Rerackina All of the operations involved in reracking will utilize detailed procedures prepared with full consideration of ALARA principles. Similar operations have been performed in a number of facilities in the past, and there is every reason to believe that reracking . can be safely and efficiently accomplished at the secuoyah plant, with rinimum radiation exposure to personnel. 9-6 I M
Total- occupational exposure for the raracking operation is estimated to be between 6 and 12 person-rem, as indicated in Table 9.5. While individual task efforts and exposures may differ from those in Table 9.5, the total is believed to be a reasonable estimate for planning purposes. It will be necessary to use divers to remove certain underwater appurtenances. Careful monitoring and adherence to procedures vill assure that the radiation dose to the divers will be maintained ALARA. The existing radiation protection program at Sequoyah is adequate for the reracking operations. Where there is a potential for significant airborne activity, continuous air samplers vill 'be in operation. Personnel vill wear protective clothing and, if necessary, respiratory protective equipment. Activities will be governed by a Radiation Work Permit, and personnel monitoring equipment will be assigned to each individual. As a minimum, this will include thermoluminescent dosimeters and pocket dosimeters. Additional personnel monitoring equipment (i.e. , extremity badges or alarming dosimeters) may be utilized as required. Work, personnel traffic, and the movement of equipment vill be monitored and controlled to minimize contamination and to assure that exposures are maintained ALAPA. In reracking, the existing storage racks will be removed, decon-taminated as much as possible by washing and wipe-downs, and packaged for shipment. Shipping containers and procedures will conform to Federal DOT regulations and to the requirements of any state through which the shipment may pass, as set forth by the State DOT office. 9-7
z Table 9.1 RESULTS OF ORIGEN-2 CALCULATIONS FOR RADIONUCLIDES OF IODINE, KRYPTON, AND XENON AT 100-HotIRS' COOLING TIME Radionuclidg curies ner atU + I-131 5-7.564 x 10 3 I-132 6.280 x 10 I-133 7.573 x 10' I-135 5.460 x 10' Kr-85 1.606 x 10' Xe-131m 1.140 x 10' Xe-133 1.439 x 10' Xe-133m 2.691 x 10' Xe-135 2.627 x'10 3 l l l l Because of -their small inventories,_ and because the small inventories are-not offset by large adose: conversions", these isotopes-were'not included in the dose alculations. The-same is true for nine additional ~ isotor us of - iodine, . krypton, and xenon.- 9-8
Table 9.2 DATA AND ASSUMPTIONS FOR THE EVALUATION OF TIfE FUEL HANDLING ACCIDENT Core power la'el, Mw(t) 3565 Puel enrichment, vtt U 5.0 Fuel burnup, Mwd /mtU 60,000 Specific power, kw/kgU 40.00 Power peaking factor 1.65 Number of failed fuel All rods in 1 of rods 193 assemblies Core inventory released NUREG/CR- _ Reg Guide to gap, % 5009 1.25 Iodine-131 12, 10 Other iodines 10 10 Krypton-85 14 30 Xenon-133 5, 10 Other xenons 10 10 Iodine composition, % Elemental 99.75 Organic 0.25 Pool decontamination factors Elemental iodine 133 organic iodine 1 Noble gases 1 Filter decontamination factors t Elemental iodine 20 l Organic iodine 20 Noble gases 1 Atmospheric diffusign 4 factor (y/Q), sec/m 8.59 x 10 Breathing rate, m 3 /sec 3.47 x 10 l l
- No release .'raction given: assumed same as Reg Guide 1.25, 9-9
.~. -e
Table 9.3 RADIONUCLIDE PROPERTIES USED IN THE FUEL HANDLING ACCIDENT ANALYSIS Dose-Conversion, Nuclide Rads / Curie Ep (Mev) E, (Mev) , Iodine-131 1.48 x 10' ----- ----- Iodine-132 5.35 x 10' ----- ----- Iodine-133 4.0 x 10 3 Krypton-85 ----- 0.223 0.002-Xenon-131m ----- 0.140 0.164 Xenon-133 ----- 0.146 0.030 Xenon-133m ----- 0.155 0.033 9 - 10 __Z______ ___.....__m__ . _
s Table 9.4 REPRESENTATIVE CONCENTRATIONS OF RADIONUCLIDES IN THE SPENT FUEL POOL WATER I Concentration, Nuclide uCi /ml-5 Co 3.x 10 Co-60 4 x 10Y 5 Cs-134 1 x 10 Cs-137 3_-xl.10 9 4
- _ _ _ _ _ _ _ _ _ ___________i.1__.. ____._m.___ _ _ _ _ _ _ _ _ _ _ _
Table 9.5 PRELIMINARY ESTIMATE OF PERSON-REM EXPOSURES DURING RERACKING Estimated Number of Person-Rg Sigg Personnel lipm Expogure.__ Remove empty racks 5 40 0.5 to 1.0 Wash and Decon racks 3 10 0.08 to 0.2 Clean and Vacuum Pool 3 25 0.3 to- 0.6 Remove underwater 4 5 0.4 to 0.8 appurtenances Partial installation 5 20 'O.25 to 0.5 of new rack modules Move fuel to new racks 2 150 . 0.8 to 1.5 Remove remaining racks. 5 120 1.5 to 3.0 Wash and Decon racks 3 30 0.2 to 0.4 Install remaining new 5 35- 0.4 -to 0.8 rack modules Prepare old racks for 4 80 1.0 to 2.0 shipment J Total Exposure, person-rem 6 to 12 4
- Assumes minimum dose rate of 2-1/2 mrem /hr (expected) to a maximum of 5 mrem /hr except for-pool vacuuming operations, uhich assume 4 to 8 mrem /hr, and diving operations, which assums-20 to 40 mrem /hr.
9 - 12
0 10.0 BORAL SURVEILLANCE 10.1 Purpose Boral'", the neutron absorbing material incorporated in the Sequoyah spent fuel storage racks to assist in controlling reactivity, consists of finely divided particles of boron carbido (B5 c) uniformly distributed in type 1100 aluminum powder, clad in type 1100 aluminum and pressed and sintered in a hot-rolling process. Tests simulating the radiation, thermal and chemical envhonment of the spent fuel pool have demonstrated the stability aM chemical inertness of Boral (References 10.1.1-10.1.3). The accumulated dose to the Boral over the expected rack lifetime is estimated to be about 3 x 10" to 1 x 10" rads depending upon how the racks are used and the number of full-core off-loads that may be necessary. As , indicated in the aforementioned raferences, the laboratory and test reactor data have confirmed the ability of this material to withstand equivalent gamma dosages which are an order of magnitude higher than those expected in thct spent fuel pools. Based upon the accelerated test programs, Boral is considered a most satisf actory material for reactivity control in spent fuel storaga racks and is fully expected to fulfil its design function over the lifetime of the racks. Tabic 2.3.1 lists U.S. and overseas applications of Boral in light water reactor spent fuel pools. As can be inferred from that table, Boral's in-pool experience is over twenty years, and the cumulative underwater experience is in excess of 150 pool ys re. Selected properties of the Boral will be pre-characterized with the objective of providing data if necessary to assess the continued integrity of the Boral panels using certain basic tests and acceptance criteria. Representative coupon samples will be available to monitor performance of the absorber material without disrupting the integrity of the storage system. The principal D 10 - 1 1 s
.- . - -. . .. -- . - - - - .. .-. _~
parameters which would be measured and monitored are the thickness (to monitor for swelling) and neutron attenuation (to monitor for the continued presence of boron in the Boral) . 10.2 COUPON SURVEIT.TANCE EQUIPMENT 10.2.1 Couoon Descriotion The coupon surveillance equipment consists of coupons suspended on a mounting (called a " tree"), placed in a designated cell, and surrounded by spent fuel. Coupons may be removed from the array and certain physical and chamical prope.rties measured from which the stability and integrity of the Boral in the storage cells may be inferred. Each surveillance coupon will be approximately 6 inches wide and 12 inches long, weighing approximately 325 grams. A total of-12 test coupons will be provided with each coupon mounted in a stainless steel jacket, simulating as nearly as possible, the actual-in-service geometry, physical mounting, materials, and flow conditions of the Boral in the storage racks. The jacket (of the same alloy used in manufacture of the racks) will be closed by screws- to allow j easy opening with minimum possibility of mechanical damage-to the Boral specimen inside. In mounting the coupons on the-tree, the coupons will be positioned axially within the central 8 feet of the fuel zone where the gamma flux is expected to be . reasonably uniform. Each coupon will be carefully pre-characterized prior to insertion in the pool-to provide reference initial values for comparison with
-measurements made after irradiation. As -a minimum, the surveillance coupons will' be pre-characterized for weight, dimensions (especially- thickness) and neutron attenuation. In addk. ion, two coupons (which need not be jacketed) will be 10 - 2
+
- - r i- w --,ry- r-%,.. vi-- , , . ~ , ,-~,- err - --
E preserved as archive samples for cosparison with subsequent test coupon measurements. Wet chemical annlyses of samples from the same lot of Boral will be saved by TVA for comparison. 10.2.2 Surveillance C2gpon Location To assure that the coupons will have experienced a slightly higher radiation dose than the Boral in the racks, the coupon tree will be surrounded by freshly discharged fuel assemblies at each of the first five refuelings following insts11ation . of the~ racks. . Beginning with the fifth load of spent fuel, the fuel assemblies will remain in place for the remaining lifetime of the racks. Evaluation of the coupons can provide information of the effects of the' radiation, thermal and chemical environment of the pool and by inference, comparable information on the Boral panels in the racks. The coupons will accumulate more radiation dose than the expected lifetime dose for normal storaa cells. Coupons which have not been destructively analyzed by wet-chemical processes, may optionally be returned to the storage pool and re-mounted on the tree. They will-then be available for subsequent-investigation of defects, should any be found. l 10.2.3 Measurements l Coupon measurement may ' be used to monitor changes -in physical properties of the Boral absorber material by performing the j following measurements: l
- i. o Visual Observation and Photography, o Neutron Attenuation, l
10 - 3 T = . - - - - - ww +-we w m ew- ~w n 14- -r y v- ---e,.w mm =% ew - , y
s o Dimensional Hemsurements (length, width and thicknoss), o Weight and Specific Gravity, and o Wet-chemical analysis (optional). The most significant measurements are thicknesq (to monitor for swelling) and neutron attenuation * (to confirk the concentration of Boron-10 in the absorber material) . In the event loss of boron is observed or suspected, the data may be augmented by wet-chemical analysis (a destructive gravimetric technique for total boron only). 10.2.4 Measureagnt Accentance Criteria Acceptance criteria for neutron attenuation and thickness measurements are as followst o A decrease of no more than 5% in Boron-10 con-tent, as determined by neutron attrmuation, is acceptable. (This is tantancant to a requirement for no loss in bor n within the accuracy of the measurement.) o An increase in thickness at any point should ' not excted 10% cf the initial thickness at th t point. Neutron attenuation measurements are a precise Instrumental method of chemical analysis for Boron-10 content using a non-destructive technique in which the percentage of thermal neutrons transmitted through the panel is measured and compared with pre-determined calibration data. Boron-10 La the nuclide of principal interest since it is the isotope responsible for neutron absorption in the Boral panel. 10 - 4
Changes in excess of either of these two criteria requires engineering evaluation which may prescribe early retrieval and measurement of one or more of the remaining coupons to provide cor-roborative evidence tnat the indicated change (s) is re ?.. If th6 deviation is determir.sd to be renl, an engineering evaluation shall be performed to identify further testir.g or any corrective action that may be necessary. The remaining meas"msar . parameters serve a supporting role and provide early indications of the potential onset of Boral degradation that would suggest a need for further attention. These include (1) visual or photographic evidence of unusual surface pitting, corrosion or edge dotarloration, or (2) unaccountable weight loss in excess of the measurement accuracy).
10.3 REFERENCES
10.1.1 " Spent Puel Storage Module Corrosion Report", Brooks & Perkins Report 554, June 1, 1977 10.1.2 " Suitability of Brooks & Perkins Spent Puel Storage Module for Use in PWR Storage Pools", Brooks & Perkins Report 570, Ju]y 7, 1978 10.1.3 "Boral Neutron Absorbing /Shiciding Material , Product Performance Report", Brooks & Perkin.3 Report 624, July 20, 1942 10 - 5 f _ _ - - _ _ _ _ _ - _ _ _ _ _ _ . i
11.O DNIRONMENTAL COST /BIl{IPIT ASSISS}iI2fI 11.1 Introduction Article V of the USNRC OT position paper (11.1) requires the submittal of a cost / benefit analysis for the chotan fuel storage capacity enhancement method. This section discusses factors considered by TVA before unlocting raracking as the mout viable alternative. 11.2 Iracerative for Increasinn Socnt F)1el Storace The specific need to increase the limited existing spent fuel storage capacity at Sequoyah is based on the cont 4nually increasing inventory in the spent fuel pool and the advisability of maintaining full-core off-load capability. Reference is made to Tables 1.2 1 and 1.1.2 of Section 1 wherein the current and projected fuel discharges in tla Sequoyah spent fuel pool are tabulated. It is seen that the Sequoyah fuel pocl will lose the capacity to accept a discharge of one full core (193 fuel assemblies) in 1996. The capacity to accept a normal discharge batch could be lost soon thereafter (ca. 1998). The projected loss of storage capacity in the Sequoyah pool would affect TVA's ability to operate both sequoych reactors. There are no commercial independent spent fuel storage facilities operating in the U.S. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocssing represents an added cost for the nuclear fuel cycle which already includes the NWPA Nuclear Waste Fund fees. In any event, there are no domestic reprocessing facilities. TVA does not hsve an cxisting or planned contractual arrangement for third-party funi storage or fuel reprocessing. There are no acceptable alternatives to 11-1 I
developing additional onsite spent fuel storage capacity for the Sequoyah Nuclear Plant. Replocement power costs average $320,000 per day per unit; shutting down the Sequoyah Nuclear Plant is many tions more expensive than increasing onsite spent fuel storage capacity. 11.3 Aparaisal of Alternatiye Ontions TVA has determined that reracking is the most viable option for sequoyah in comparison to other spent fuel storage alternat.ives, s The key considerations in evaluating the alternative options were
- Minimize-the effects on plant systems and operations by reducing the amoun'; of fuel handling as well as the attendan*. potential impacts o7 safety and ALARA.
Maturity of the technology and the extent of industry experience.
- Maximize flexibility tot
- 1. Implement subsequent actions for further increasing onsite spent fuel storage capacity.
- 2. Interface with Department of Energy technology choices for shipment, storage, and ultimate disposal of the spent fuel.
- Minimize overall capital and o&M costs.
Roracking was found by TVA to be the most attractive option with respect to each of the foregoing criteria when compared to the following alternative technologies. Wet Storace
- 1. Roracking.
- 2. Double-tiered racks.
- 3. Pod consolidation.
11-2
- 4. Transshipment (pool-to-pool).
- Dry Storagg
- 1. Metal casks.
- 2. Concrete casks.
- 3. Concrete vaults.
- 4. Dual-purpose casks.
11.4 Proiect cost Estimate The total cost for the Sequoyah rarack project is estimated to be approximately $10 million and includes engineering design, unterial procurement, fabrication, installation, disposal of the existing spent fuel racks, and an allowance for contingencies. Comparative estinates of the costs par incremental fuel assembly storage space for the alternative technologies in 1990 dollars are Roracking $12,000 Rod Consolidation $14,000-27,000 Transshipment $14,000-18,000 Metal Casks $36,000-52,000 Concrete Casks $18,000-38,000 Concrete Vaults $18,000-30,000 11.5 Resource ,Cpmmitment The expansion of the spent fuel pool capacity is expected to require the following primary resources: Stainless steel 300 tons Boral neutron absorber 25 tons,.of which 13 tons is boron carbide powder and 12 tona are-aluminum. 11-3
The requirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than .001%) percent). Although the fraction of World production of boron carbide required for the fabrication is somewhat higher than that of stainless steel or aluminum, it is unlikely that the commitment of boron carbide to this project vill affect other alternatives. Experience has shown that the production of boron carbide is highly variable god depends upon need, and can easily be expanio4 to accommodate worldwide needs. 11.6 Environmental congiderati.2at Due to the additional heat-load arising from increased spent fuel pool inventory, the anticipated maxinum bulk pool temperature vill rise by less than 10*F due to the proposed increase in the spent fue'l inventory in the spent fuel pool. The total heat-load for the emergency core off-load (worst case) is less than 50 million BTU /HR, which is less than one percent of the total plant heat loss to the environment. The increased bulk pool temperature vill result in an increase in the pool water evaporation rate. This increase is within the capacity of both existing Sequoyah HVAC systems. The not result of the increased heat loss and water vapor emission to the environment is negligible. In Section 9 of this report, an assessment of the impact of reracking on pool radwaste volume is considered. During installation of the new racks a relatively small amount of additional resins may be generated by the pool cleanup system on a one-time basis. It is concluded that the effect of the propound capacity increase is insignificant once the reracking operations are completed. 11e4
volume reduction and disposal of the existing Sequoyab racks, together with the Floor Support Grid Structure and miscellaneous hardware, will add 1100 cubic feet to the burial volume, which is less than 15% of the plant annual LSA waste output. 11.7 RErlEDiCES 11.1 "OT Position Paper for Review and Acce Puol Storage and 11andling Applications",ptance of Spent dated April 14, 1973, and January 18, 1979 amendment thereto. 1 11-5 t 4
...____.._)
ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN T*-92-0 0 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS
- - _ - ~ . - .
- _ - - - - - . - . .- -.- . . ~ - - - - -.- ~_- - .. .
Enclosure 3 SIGNIFICANT llAZARDS EVA13JATION 4 TVA has evaluated the proposed technical specification (TS) changes and has determined that they d: not represent a significant hasards consideration based on criteria established in 10 CFR 50.92(c). j Operation of Sequoyah in accordance with the proposed amendment will nott , (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The following potential reenarios were considered '
- 1. A spent-fuel assembly drop.
- 2. Drop of the transfer canal gate or the divider gate in the spent-fuel pool.
- 3. A seismic event.
- 4. Loss of cooling flow in the spent-fuel pool.
- 5. Installation activities.
The probability of on.currente of the first four is not affected by the racks themselves; therefore, the modification cannot increase the associated probabilities. With regard to installation activities, the existing Sequoyah TSs , prohibit loads in excess of 2100 pounds from travel over fuel assemblies in the storage pool and require the associated crane interlocks and physical stops bu periodically demonstrated oporrble. During , installation, racks and associated handling tools will be moved over the ! spent-fuel pool but movement over fuel will be prohibited. All installation work in the spent-fuel-pit area will be controlled and performed in strict accordance with specific written proceduren. NBC regulations provide that, in lieu of providing a single failuro-proof ' crane system, the control of heavy loads guidelines can be satisfied by establishing that the potential for a heavy load drop is extremely small.. Storage rack movements to be' accomplished with the Sequoyah auxiliary building crane will conform with NUREG 0612 guidelines, in that the probability of a drop'of s. storage rack is extremely small. The crane _has a tested'espacity of 80 tor,J. The maximum weight;of any. existin6 or replacement _ storage rack and its associated handling tool is less than 15 tons. Therefore, there is ample safety factor margin for movements of the storage racks by the auxiliary building crane. Special lifting devices, which have redundancy or a rated capacity sufficient to maintain adequate safety factors,_will also be utilized in the movements of the otorage racks. In accordance with NUREG-0612, Appendix B, the safety margin ensures that the probability of a load drop la extremely-low.
- , , ~ - - . . - , -n -- -
-2 Load travel over fuel stored in the cask loading area of the cask pit will be minimited and, in any case, will be prohibited unleso an impact shield, which has been specifically designed for this purpose, is covering the area. Loads that are permitted when the shield is in place must mest analytically determined weight, travel height, and cross-sectional area criteria that preclude penetration of the shield. A TS has been proposed that incorporates the previously nentioned load criteria.
A fuel movement and rack changoout sequence has been developed that illustrates that it will not be necessary to carry existing or new racks over fuel in the cask loading area or any region of the pool containing fuel. A latorst-free zone clearance f rom stored f uel shall be maintained. Accordingly, it is concluded that the proposed installation activities citi not significantly increase the probability of a load-handling accident. The consequences of a load-handling accident are unaffected by the proposed installation activities. The consequences of a spent-fuel assembly drop were evaluated, and it was determined that the racks will not be distorted such that they would not perfonn their safety function. The criticaalty acceptance criterion, Ke gt 1 0.95, is not violated, and the calculated doses are.well within 10 CFR Part 100 guidelines. Thus, the consequences of this type of accident are not changed from previously evaluated spent-fuel assembly drops that have been found acceptable by NRC. The existing TSs permit the transfer-canal bate and the divider gato in the spent-fuel pool to travel over fuel assemblies in tha spent-fuel pool. Analysis showed that this drop caused less damage to the new racks than the fuel assembly drop when it impacts the top of the rack. Rack damage is restricted to an area above the active fuel region. The consequences of a seismic event have been evaluated. -The new racks are designed and will be fabricated to meet the requirements of applicable portions of the NRC regulatory guides and published _tandards. The new free-standing racks are designed to maintain their integrity during and after a seismic event. Thus, the consequences of a seismic event are not significantly changed from previously evaluated events. The spent-fuel pool system is a passive system with the exception of the fuel pool cooling train and heating, ventilating, and air-conditioning (HVAC) equipment. Redundancies in the cooling train and.HVAC hardware are not reduced by the planned fuel storage densification. The potential increased heat load resulting from any additional storage of spent fuel is well within the existing system cooling capacity. Therefore, the probability ol' occurrence or malfunction of safety equipment leading to the loss of cooling flow =in the spent-fuel pool is not significantly affected. Furthermore, the consequences of a loss of cooling system flow in'the spent-fuel pool have been evaluated, and it was found that SuffiClent time is still available to-provide an alternate means for cooling in the event of a failure in the cooling system and.that the-effect on the pool structure because of thermal loadings is acceptable. Thus, the consequences of this typs of incident are not significantly increased from previously-evaluated loss of cooling system flow malfunctions.
- - _- - . - - -- __- - -.- - .---- .. .. _ - ~ _ _ _ .
3 (2) Create the possibility of a new or different kind of accident from any accident previously analyzed. The proposed modification has been evaluated in accordance with the guidance of the NRC position paper entitled, "0T Position for Review and Acceptance of Spent-Fuel Storage and Handling Applications"t appropriate NRC regulatory guideal appropriate NRC standard review planst and appropriate industry codes and standards. Proven analytical technology was used in designing the planned fuel storage expansion and will be utilized in the installation process. Basic reracking technology has 4 been developed and demonstrated in over 80 applications for fuel pool capacity increasen that hve already received NRC staf f approval. The TSs for the existing spent-fuel storage racks use burnup credit and fuel assembly administrative placcment restrictions for criticality control. The change to three-zone storage in the spent-fuel pool is described in the proposed change to the denign features section of the TSs. Additional evaluations were required to ensure that the criticality criterion is maintained. These include the. evaluation for the limiting criticality condition, i.e., the abnormal-placement of-an unirradiated (fresh) fuel assembly of 4.95 weight percent enrichment into a storage cell 1ccation for irradiated fuel meeting the highest rack design burnup criterion. The evaluation for this case shows that the reactivity would exceed the limit in the abr uce of soluble boron. Soluble boron, for which credit is permitted under these abnormM. conditions, ensures that reactivity is maintained substantially less than the design requirement. Calculations indicate that a soluble poison concentration of 685 parts per million (ppm) boron would be required to limit the maximum reactivity to a ke gg of 0.95, including uncertainties. This is less than the existing and proposed TS requirements of 2000 ppm. ; It is not physically possible to install a fuel assembly outside and adjacent to a storage module in the-spent-fuel storage pool. However,- for a storage module installed in the cask loading area of the cask pit, there would be sufficient room for such an extraneous assembly. . The module in this-area is administrative 1y limited by the proposed TS change to spent fuel only, and calculations show that the maximum k erg remains well below the 0.95 limit under this postulated accident condition, even in the absence of soluble boron. To provide reactivity con sol assurance for the abnormal placement of a fresh assembly in the cask loading area module, a modificasion to the existing TS has been proposed that requires boron concentration measurements while handling fuel in that area. Although these changes required addressing additional aspects of a
-previously analyzed accident, the possibility of a previously unanalyzed accident is not created. It is therefore concluded that the proposed-reracking does not create the possibility of a riew or different kind of accident from any previously analyzed.
(3) Involve a significant reduction in a nergin of safety. The design and technical review process applied'to the reracking modification included addressing the following areas:
- 1. Nuclear criticality considerations.
- 2. Thermal-hydraulic considorations.
- 3. Mechanical, material, and structural considerations.
The established acceptance criterion for criticality is that the neutron multiplication factor shall be less than or equal to 0.95, including all uncertainties. The results of the criticality analysis for the new rack design demonstrato that this criterion is satisfied. The methods used in the criticality analysis conform to the appilcable portions of NRC guidance and industry codes, standards, and specifications. In meeting the acceptance criteria for criticality in the spent-fuel pool and the cask loading area, such that kett is always lors than 0.95 at a 95/95 percent probability tolerance level, the proposed amendment does not involve a significnr.c reduction in the margin of safety for nuclear criticality. Conservative methods and assumptions were used to calculate the maximum fuel temperature and the increase in temperature of the water in the spent-fuel-pit area. The therrral-hydraulic evaluation used methods previously employed. The proposed storage modification will increase the heat load in the spent-fuel pwl, but the evaluation shows that the existing spent-fuel cooling system will maintain the bulk pool water temperature at or below 180 degreer Fahrenheit. Thus it is demonstrated that the worst-case peak value of the pool bulk temperature is considerably lower that. the bulk boiling temperature. Evaluation also shows that maximum local water temperatures along the hottest fuel assembly are below the nucleate boiling condition value. Thus there is no significant reduction in the margin of safety for thermal hydraulic or spent-fuel cooling considerations. The mechanical,rnaterial, and structural design-of the new spent-fuel racks is in accordance with applicable portions of "NRC OT Position for-Review and Acceptance of Spent-Fuel Storage and Handling Applications," dated April 14, 1978 (as modified January 18, 1979), as well as other applicable NRC guidance and industry codes. The primary safety function of the spent-fuel racks is_to maintain the fuel assemblica in a safe configuration through all normal and abnormal loading conditions. Abnormal loadings that have been evaluated with acceptable res'its and discussed previously include the effect of an earthquake and che impact because of the drop of a fuel assembly. The rack materials used are compatible with the fuel astemblies and the environment in the spent-fuel , pool. The structural desiga for the new racks provides tilting, deflection, and movement margans-such that the racks do not impact each otM r or the spent-fuel-pit walls in the active fuel region during the pe culated seismic events. Also the spent-fuel assemblies themselves remain intact and no criticality concerns exist. In addition, finite element analysis methods were used to ev61uate the continued structural acceptability of the' spent-fuel pit. The analysis was performed in accordance with " Building Code Requirements for Reinforcad Concrete" (API 318-63. 77).- Therefore, with respect to-mechanical, material, and structural considerations, there is no significant reduction in a margin of safety.
5-In summary, the proposed spent-fuel storage modifications do not
- 1. Involve a significant increase in the probability of consequences of an accident previourly evaluatedt or
- 2. Create the possibility of a new or dif ferent kind of accident f rom any accident previously evaluatedt or
- 3. Involve a significant reduction in a margin of safety.
Therefore. TVA has determined that the proposed amendments as described do not involve significant hazard considerations and that the criteria of 10 CFR 50.91 have accordingly been met. TVA has also reviewed the NRC examples of licensing amendmente cons.idered not likely to involve significant tiv ds considerations as provided in the final adoption of 10 CFR 50.92 ps.iished on page 7751 of the Federal Register, Volume 51, No. 44, March 6, 1986. Example (X) provides four criteria that, if satisfied by a reracking request, indicate that it is likely no significant hazards considerations are involved. The criteric and how TVA's amendment request for Sequoyah complies are indicated below. Criterion (1): The starage expansion method consists of either replacing existing racks with a design that allows closer spacing between stored spent-fuel assemblies or placing additional racks of the original design on the pool floor if space permits. Proposed Amendment: The Sequoyah Nuclear Plant reracking involves-replacing the existing racks with a design that allows closer spacing between stored fuel assemblies and also provides additional rack storage on the pool floor where space permits. C-iterion (2): The storage expansion method does not involve rod concolidation or double tiering. Proposed Amendment
-The Seguoyah racks are not double tiered, and all racks will sit on the floor of the spent-fuel pool. Additionally, the amendment application does not fnvolve consolidatien of spent fuel. 's o
Criterion (5): ' The ke gg of the pool is maintained less than or equal to 0.95. Proposed Amendment The ucsign of the new spent-fuel racks contains a neutron absorber, Botal, to allow cit'se storage of spent-fuel assemblics while ensuring that the kegg remains less than 0.95 under all normal operating . conditions with unborated water in the pool and less than 0.95 under ! abnormal conditions with soluble boron in the pool. Ortcerion (4): No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary to justify the expansion. Proposed Amendment The construction processes and analytical techniques used in the f abrication and design are substantially the same as those of numerous other rack installations. Thus, no new or unproven technology is utilized in the construction or analysis of the high-density, spent-fuel racks at Sequeyah. TVA's Contractor, !!altec Internacional, has previously supplied licensable racks of very similar design for about 10 other reracking projects.
..- .. . . . . - , . . . - , _ - . - . ~ , . . ., - - . . . - - .,
Enclosure 4 PROPOSED ENVIRONMENTAL IMPACT EVALUATION SEQUOYAH NUCLF.AR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 The proposed change request does t4ot involve an unreviewed environnental question because operation of Sequoyah Units 1 and 2 in accordance with this change would not:
- 1. Result in a significant increase in any adverse environmental tmpact previously evaluated in the Final Environmtntal Statement (FESI as modified by the staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals., **
decisions of the Atomic Safety and Licensing Board.
- 2. Result in a significant chrnge in effluents or power levels.
- 3. Result in mattera not previously reviewed in the licensing basis for Sequoyah that may have a significant-environmental impact.
Note: The bases for the above statements are discussed in Section 9 (Radiological Evaluation) and Section '11 (Environmental Cost./ Benefit Assessment) of Enclosure 2 - Spent-fuel Pool Modification for Increased Storage Capacity. 9
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