ML20212C150
| ML20212C150 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/31/1999 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20138D882 | List: |
| References | |
| NUDOCS 9909210155 | |
| Download: ML20212C150 (83) | |
Text
WESTINGHOUSE NON-PROPRIETARY CLASS 3 CRYSTAL RIVER UNIT 3 ENHANCED SPENT FUEL STORAGE ENGINEERING INPUT TO
{
LICENSE AMENDMENT REQUEST NO. 239 j
)
i l
hh O
,_. c) -
Ph2
%.'T '
August 1999 o:\\4878-non doc 1b-090199 Revision 0 J
TABLE OF CONTENTS 1.0 GENERAL INFORMATION................
..............................1-1 1.1 Purpose of Report........
.1-1 1.1.1 Current Status................
.................1-1 1.1.2 Criticality Analysis......
=1-1 1.1.3 Summary of Report..................................................... 1-2 1.2 Design Objectives.
......... 1 -2 1.3 Description of Replacement Spent Fuel Racks....
. 1-3 1.4 Fuel and Rack Handling =
....1-4 1.4.1 Fuel Handling Process.................
....... 1-4 1.4.2 Rack Handling.......
. 1-4 1.4.2.1 Installation Method......................
... 1-5 1.4.2.2 Rack Handling and Installation Sequence.
........ 1 -5 2.0 THERMAL-HYDRAULIC ANALYSIS..............
...................2-1 2.1 Decay Heat Calculations for Spent Fuel............
............... 2-1 2.2 Thermal Hydraulic Analysis for Spent Fuel Cooling.....
.......... 2-1 2.2.1 Local Fuel Bundle Thermal Hydraulics..........................
.2-1 2.2.1.1 Criteria..............
.2-2 2.2.1.2 Key Assumptions...
.2-2 2.2.1.3 Analytical Method and Calculations.
................ 2-2 2.2.1.4 Results.................
.......2-4 3.0 MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS........... 3-1 3.1 Description of Spent Fuel Pool.
.3-1 3.2 Structural Design of Racks..........................................
........3-1 3.3 Integrity of Fuel Racks Under Fuel Handling Accident Conditions.......
..3-1 1
3.3.1 Spent Fuel Handling Machine Load Scenario............................. 3-1 3.3.2 Fuel Assembly Drop Scenarios.................................
............. 3-1 3.3.2.1 Drop Orientations.....................
...................... 3-2 3.3.2.2 Acceptance Criteria......
3-2 3.3.2.3 Assumptions for Energy Dissipation...................... 3-2 3.3.2.4 Drop Analysis Results...................................
.3-3 3.3.2.4.1 Functional Capability of the Fuel Racks..... 3-3 3.3.2.4.2 Functional Integrity of the Spent Fuel Pool Structure and Liner................................... 3-4 August 1999 c:\\4878-non.tK1b-090199 il Revision 0
1 TABLE OF CONTENTS (Continued) i 3.4 Structural Analysis Procedures For Spent Fuel Storage Racks.........
... 3-4 3.4.1 Analysis Overview..
.......... 3-4 3.4.2 Synthetic Time Histories...
....... 3.5 3.4.3 Stress and Seismic Models...
...... 3-6 3A 3.1 Static Model of the Fuel Rack Structure..
..................3-7 3.4.3.2 Single Rad Dynamic Model............
.3-7 3.4.3.3 Whole Pool Dynamic Model...........
........3-9 3.4.4 Modeling Features......
,3-10 3.4.4.1 Damping........
... 3-10 3.4.4.2 Fluid Interactions...........................
. 3-10 3.4.4.3 Friction Coefficient..
........ 3 10 3.4.5 Seismic Evaluation.
.. 3-10 3.5 Structural Acceptance Criteria and Analysis Results for Spent Fuel Storage Racks......
............... 3-11 3.5.1 Structural Acceptance Criteria...........
..........3-11 3.5.1.1 Rigid Body Rack Sliding and Tilting Motions.....
. 3-11 3.5.1.2 Stress Limits.....
.3-11 3.5.2 Stress Analysis =
.3-11 3.5.3 Results for Single Rack Model Analysis.....
..... 3-11 3.5.4 Parametric Studies Results..
...... 3-12 3.5.4.1 Fuel Loading Configuration..
....... 3-12 3.5.4.2 Fuel Impact Stiffness.................................... 3-13 3.5.4.3 Friction Coefficients..
...................... 3-13 3.5.5 Whole Pool Multiple Rack Model Results........
......... 3-13 3.6 Materials, Quality Control and Special Construction Techniques...
.. 3-13 3.6.1 Constr uction Materials.....................
.. 3-13 3.6.2 Neutron Absorbing Material / Material Integrity...........
... 3-14 3.6.3 Quality Assurance..
.. 3-15 l
4.0 SAFETY EVALUATION.......................
.... 4-1
)
4.1 Degree of Subcriticality......
................4-1 4.2 Governing Codes for Design....
..... A-1 4.3 Ability to Withstand External Loads and Forces................................. 4-1 4.4 Ability to Ensure Continuous Cooling................................
........... 4-2 4.5 Provisions to Avoid Dropping of Heavy Loads...
..............................4-3 4.6 Material Compatibility.......
................................................................4-3 4.7 Radiological Considerations..
...................4-3 4.8 Ability of Racks to Withstand Accidental Lift Forces.............
......................4-3 4.9 Potential Fuel and Fuel Rack Handling Analysis...
.........................4-3 4.10 Concl u s ion..............................
.........................................................4-4 August 1999 o:\\4878-non. doc 1b-090199 iii Revision 0
a TABLE OF CONTENTS (Continued) i 5.0 COST / BENEFIT ASSESSMENT...................................................................................... 5-1 5.1 Need for Increased Storage Capacity................................................................. 5-1 5.2 Es tima t ed Cos ts................................................................................................. 5-3 5.3 Consideration of Alternatives...................................................................... 5-3 5.4 Resources Commi tted........................................................................................ 5-4 6.0 RADIOLOGICAL EVALUATION.................................................................
...6-1 6.1 Solid Radwaste......................................
..........................................6-1 6.2 Gaseous Radwaste.........................................
...6-1 6.3 Personnel Exposure....................................................
............................6-1 6.3.1 Personnel Exposure During Rack Replacement a nd Di sposal...................................................................................... 6-1 6.3.2 Personnel Exposure from Normal Operation with New Racks........................................................
=6-2 6.4 Rack Decontamina tion and Disposal............................................................. 6-3 7.0 CODES, STANDARDS, SPECIFICATIONS AND OTHER REFERENCES...
........ 7-1 i
t August 1999 o:\\4878.non doc:1b.090199 iv Revision 0 l
i
r LIST OF TABLES 3-1 Fuel Assembly Drop Scenarios Results.....................
.............................3-16 3-2 Load Cornbinations for Spent Fuel Racks...............
............ 3-17 3-3 Rack Minimum Margin to Allowable.......................................................................... 3-18 3-4 Rack Loading Configuration Factor of Safety Against Tilting.................................... 3-19 3 Comparison of Whole Pool Multiple Rack Model to Single Rack Model..................... 3-20 3-6 Impact of Fuel Assembly Stiffness on Fuel Assembly Loads...................................... 3-21 5-1 Spent Fuel Pool Capacity without Reracking 'B' Pool.............................................. 5-5 5-2 Components Stored in Spent Fuel Pool....................................
.................5-6 5-3 Spent Fuel Fool Capacity After Reracking 'B' Pool.......
...................................5-7 6-1 Gamma Isotopic Analysis of Spent Fuel Pool Water........................................................ 6-4 i
l August 1999 ones78-non.docntro90199 v
Revision 0
U LIST OF FIGURES 1-1 Crystal River 3 Spent Fuel Fool B Storage Rack Arrangement................
.............. 1 -7 1-2 Crystal River 3 Spent Fuel Pool B Storage Rack Module Cross Section.......................1-8 1-3 Crystal River 3 Spent Fuel Pool B Storage Rack Typical Cell Layout..............................1-9 2-1 Spent Fuel Pool Natural Circulation Model (Elevation View).......................................... 2-6 2-2 Spent Fuel Pool Natural Circulation Model (Plan View).............................................. 2-7 2-3 Spent Fuel Rack Inlet Flow Area (Plan View).......................................
...............2-8 3-1 Fuel Assembly Drop Orientation A......................................................................... 3-22 3-2 Fuel Assembly Drop Orientation B...........................................
................. 3-23 3-3 Fuel Assembly Drop Orientation C...............................................
..3-24 3-4 Overall Structural / Seismic Analysis Process.........................
=3...........
3-5 C-R-Unit 3 Synthetic Time History Acceleration, OBE Matching 2% and 4% Damped Floor Response Spectra................................
..................... 3-26 3-6 North-South Synthetic and Design Floor Response Spectra for OBE.......................... 3-27 3-7 North-South Synthetic and Design Floor Response Spectra for SSE.
............... 3-28 3-8 East-West Synthetic and Design Floor Response Spectra for OBE.................................. 3-29 3-9 East-West Synthetic and Design Floor Response Spectra for SSE............................ 3-30 3-10 Vertical Synthetic and Design Floor Response Spectra for OBE...............
...3-31 3-11 Vertical Synthetic and Design Floor Response Spectra for SSE...................................... 3-32 3-12 Power Spectra Density of North-South Synthetic Time History versus Crys tal Ri ver Ta rget PSD.......................................................................................... 3-33 3-13 Power Spectra Density of East-West Synthetic Time History versus Crystal River Ta rget PSD..............................................................................
- 3-34 3-14 Power Spectra Density of Vertical Synthetic Time History versus Crystal River Target PSD........................................................................................ 3-35 3-15 Single Rack Static Model of a 12x8 Rack......................................
..................... 3-36 3-16 ~ ANSYS Finite Element Model Representation of a 12x8 Single Rack............................. 3-37 3-17 ANSYS Finite Element Model Representation of a Fuel Assembly.............................. 3-38
' 3-18 ANSYS Finite Element Hydrodynamic Mass Model of a Fuel Assembly...................... 3-39 3-19 ANSYS Finite Element Model Representation of the Whole Spent Fuel Pool............... 3-40 1
4 i
August 1999 o:\\4s78-non. doc 1t>-090199 vi Revision 0
)
1.0 GENERhL INFORMATION
{
1.1 PURPOSE OF REPORT Florida Power Corporation (FPC) is pursuing the design and manufacture of new spent fuel storage racks to be placed into the Spent Fuel Pool B at Crystal River Unit 3. These new racks 1
willincrease the amount of spent fuel that can be stored in Pool B. The new racks are also needed to replace the existing racks containing the neutron absorbing material Boraflex, which is degrading, with new racks containing Boral as the neutron absorbing material. The racks are designed to store spent fuel assemblies in a denser array (9.11 inch centerline-to-centerline spacing) than the existing racks. Additionally, analyses are provided to demonstrate that i
Pool A can accommodate storage of new fuel in a checkerboard configuration with cells containing water only. Installation and use of new storage racks in Pool B, as well as revising the allowed fuel storage configurations in Pool A, will require an amendment to the Crystal River Urut 3 Improved Technical Specifications. The purpose of this report is to provide the technicalinformation required for NRC review of this amendment request.
1.1.1 Current Status There are two spent fuel pools at Crystal River Unit 3. Spent Fuel Pool A contains high-density storage racks capable of storing 542 fuel assemblics with a centerline-to-centerline spacing of 10.5 inches. Spent Fuel Pool B also contains high-density storage racks separated into two Regions. Region 1 is capable of storing 174 fuel assemblies, with a centerline-txenterline spacing of 10.60 inches. Region 2 is capable of storing 641 fuel assemblies with a centerline-to-centerline spacing of 9.17 inches. The total capacity of Pool B is 815 assemblies. The combined capacity of Pools A and B is 1357 assemblies. With this present spent fuel storage capa<t.y, FPC will lose full core reserve storage capacity after Refuel 18 in 2013. The Spent Fuel Fool A and B storage racks are designed to Seismic Class I criteria.
1.1.2 Criticality Analysis Holtec International performed criticality analyses of the Pool A storage racks (Reference 7.10.a). The analyses evaluated the storage of fresh fuel with an average enrichment of 5.0 weight percent. The Holtec report summarizes the results of the analyses that demonstrate the acceptability of a one-out-of-two checkerboard loading pattern of water cells and fresh 5.0 weight percent enriched fuel.
Criticality analyses of the replacement Pool B storage racks were also performed by Holtec i
International (Reference 7.10.b). The analyses evaluated storage of spent fuelin racks with Boral as the neutron absorbing material. The Holtec report summarizes the results of the analyses that demonstrate acceptable results for fuel with an initial enrichment of up to 5.0 weight percent and varying burn-up.
I Calculations for both pools A and B were made with CASMO-3 and NITAWL-KENO-5 computer codes. Both normal and accident conditions were assessed. Credit for soluble boron 1
GeneralInformation August 1999 o:\\4878-non.dccit>.090199 11 Revision 0
normally present in the pool water was taken only for accident conditions. Results and additional details are provided in the attached reports (Reft.rences 7.10.a and.b).
1.1.3 Summary of Report This report follows the guidance of the NRC position paper entitled, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978, as revised by the NRC letter dated January 18,1979 (Reference 7.9.b) where appropriate. More recent guidance or revised consensus standards have superceded certain portions of that guidance. Section 1.0 provides generalinformation and a description of the new spent fuel racks. Two referenced Holtec International reports (References 7.10.a and b) provide the criticality analyses for rack replacement in Pool B and revised loading in Pool A of fresh 5.0 weight percent enriched fuel in a checkerboard configuration. They address both normal and abnormal fuel configurations. Section 2.0 addresses the thermal-hydraulic aspects of normal storage and handling of spent fuel as well as postulated failures with respect to the ability of the spent fuel pool cooling system to maintain sufficient cooling. Section 3.0 addresses mechanical, material and structural aspects and discusses the capability of the fuel assemblies, storage racks, and spent fuel pool system to withstand the effects of natural phenomena such as earthquakes and other service loading conditir as The design, procurement and fabrication of the storage racks comply with the v mliW Assurance Requirements of Appendix B to 10 CFR 50. Section 4.0 contains a bid c.mmary of the safety evaluations related to the design criteria, material compatibility, extent of sub-criticality, radiological considerations, accidental drop provisions and the ability to ensure continuous cooling. This section reflects that the spent fuel racks are safe and capable of performing their intended function. Section 5.0 addresses both the Cost / Benefit of the report as well as the environmental aspects of the effort. Section 6.0 addresses projected occupational radiation exposures, generation of radioactive waste, and rack decontamination and disposal. A list of applicable codes and standards as well as other references is provided in Section 7.0.
1.2 DESIGN OBJECTIVES The function of the spent fuel storage racks is to provide for storage of new and spent fuel assemblies while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from excess mechanical or thermal loading. A list of design criteria is given below:
1.
The racks are designed to meet the dimensional requirements of ANSI-N210-1976. The effective multiplication factor, K,, in the spent fuel pool is less than or equal to 0.95, in:luding all uncertainties and under all credible conditions.
2.
The racks are designed to allow coolant flow such that boiling in the water channels between the fuel assemblies in the rack does not occur. Maximum fuel cladding temperatures are calculated for various pool cooling conditions.
3.
The racks are designed to Seismic Class I requirements, and are classified as ANS Safety Class 3 and ASME Code Class 3 Subsection NF Component Support structures (Reference 7.3.b). The racks are designed to withstand loads that may result from fuel GeneralInformation August 1999 o:\\4878-non. doc 1b-090199 1-2 Revision 0
handling accidents and fmm the maximum uplift force of the fuel-handling crane without violating the structural acceptance criterion.
4.
Each storage position in the racks is designed to support and guide the fuel assembly in a manner that will minimize the possibility of application of excessive lateral, axial and bending loads to fuel assemblies during fuel assembly handling and storage.
5.
The materials used in construction of the racks are compatible with the storage pool environment.
6.
The new Spent Fuel Fool B storage racks are poison racks designed to store spent fuel based on variable criteria of initial enrichment vs. burn-up requirements. The Pool A racks will be qualified by this submittal to store fresh fuel in a checkerboard configuration with water holes in addition to spent fuel in compliance with existing Improved Technical Specification limits on initial enrichments and burn-up.
1.3 DESCRIPTION
OF REPLACEMENT SPENT FUEL RACKS The Pool B spent fuel storage rack arrangement is shown in Figure 1-1. This arrangement replaces the current two-region racks in Pool B, with a new set of racks all of the same design.
Therefore, specification of racks as " Region 1" or " Region 2" racks will no longer be a part of the Pool B design. The current capacity of Pool A is 542 assemblies. Full core reserve will be available throughout the rack replacement project. The Pool B racks are designed to accommodate irradiated fuel. The replacement racks will provide an additional 117 cell locations in Pool B, for a capacity of 932 assemblies. Thus, the total combined capacity of both pools will be increased from 1357 assemblies to 1474 assemblies. This will provide adequate storage capacity including full core reserve until the expiration of the CR-3 current operating license in the year 2016.
All of the Pool B racks are designed to accommodate irradiated fuel. The fuel to be placed in Pool B must comply with restrictions on enrichment and burn-up. This will be controlled administratively in accordance with the CR-3 Improved Technical Specifications in a manner similar to that currently used for Pool A.
The new Pool B storage rack modules are composed of individual storage cells made of austenitic stainless steel. The cells are welded to a base support assembly and to one another to form an integral structure. These modules utilize a neutron absorbing material, Boral, which is attached to each cell. The modules are neither anchored to the floor nor braced to the pool walls. The structure is shown in Figure 1-3.
Each module consists of three major sections. These are the leveling pad assembly, the base-plate assembly, and the cell assembly. Figure 1-2 illustrates these sections.
The major components of the leveling pad assembly are the leveling pad and the leveling pad screw. Each module is provided with leveling pads that contact the spent fuel pool floor. The leveling pad assemblies transmit the loads to the pool floor and provide a sliding contact. The General Information August 1999 o:\\4878-non. doc:1b-090199 1-3 Revision 0
r l
1
- leveling pad screw permits making the leveling adjustment of the rack remotely from above, through the cells during installation. The design allows this leveling adjustment to be made with the racks in place in the flooded spent fuel pool.
The base-plate assembly consists of the base plate, leveling pad support blocks and i
reinforcement plates, which am used for rack handling and installation.
- The major components of the cell assembly are the fuel assembly cell, the Boral (the neutron 1
absorbing) material and the austenitic stainless steel wrapper (see detail provided on -
Figure 1-3). The Boral is enclossi in the wrapper, which is attached to the outside of the cell by spot welding along the entire length of the wrapper. The wrapper comers are open to provide for venting of the cavity to the pool environment. On the periphery of each rack, some
{
msultant cells must be closed off on either one or two sides with a stainless steel panel. These panels contain Boral and a wrapper if they are adjacent to a fuel storage location that does not f
contain Boral.
j 1
1.4 FUEL AND RACK HANDLING I
~ Replacement of the existing racks with new racks will be performed only in Pool B. Load handling operations in the Spent Fuel Pool area will be conducted in accordance with FPC procedures and practices which implement the criteria of NUREG-0612, " Control of Heavy Loads at Nuclear Plants". Movements'of the high density racks within the Spent Fuel Pool area will be performed using a rack handling crane and special rigging.
1.4.1 Fuel Handling Process Fuel will be moved by FPC-qualified operators. The provisions of CR-3 FSAR Section 9.6.2.7, Fuel Handling System-Operational Requirements (Reference 7.11.0), will be complied with during all fuel moves. The permanent fuel handling equipment at CR-3 will P ved for making all fuel moves.
All fuel located in the four racks in the south end of Pool B will be moved to the racks in the north end of Pool B or Pool A. This will empty those four racks, allowing them to be removed from the pool, and replaced with new racks. This is described in section 1.4.2.2 below.
The plans for moving fuel within Pool B will take into account the spent fuel which will be discharged from the core during Refuel 11, to be conducted in Fall,1999, 1.4.2 Rack Handling All handling of the existing and replacement racks will be performed using the existing CR-3 cranes augmented with a rack handling crane with a minimum capacity of 20 tons, supported from the crane hook, and special rigging.
GeneralInformation August 1999 o \\4878-non. doc:1b-090199.
1-4 Revision 0
i
\\
1.4.2.1 Installation Method The installation of the new high-density racks will be accomplished in accordance with the following considerations and guidelines:
Rack installation will be performed in Spent Fuel Pool B only.
Rack handling operations in the Spent Fuel Pool area will be conducted in accordance with FPC procedures and practices which implement the criteria of NUREG-0612,
" Control of Heavy Loads at Nuclear Plants."
1.4.2.2 Rack Handling and Installation Sequence The rack installation will be performed in a flooded Pool B only. Safe load paths will be used throughout the rack installation effort with no movement of heavy loads over spent fuel without missile shields installed. The racks will be delivered to the Auxiliary Building loading bay, lifted using the existing Auxiliary building overhead crane, and moved adjacent to Spent Fuel Pool B. The installation sequence will alternate between removing existing racks and installing new racks.
A hoist with a minimum capacity of 20 tons will be attached to the Auxiliary Building overhead crane. A specially designed lift rig will be lowered into place over the racks to move and remotely install the racks. The current racks will be decontaminated to the extent possible prior to removal from the Spent Fuel Pool. No fuel rack will be moved over spent fuel without missile shields installed.
After removing the existing racks, the B Spent Fuel Pool floor will be inspected and debris removed prior to installing any mplacement racks.
The replacement racks will be transported from their on-site location to the loading bay under the floor hatch.
The racks will be lifted and moved to the spent fuel pool ama.
Prior to installation into the spent fuel pool, each new rack will be inspected for cleanliness and tolerances to confirm it is in an acceptable condition. The rac).s will be leveled prior to installation and adjusted as needed after final positioning.
The racks will then be lifted in the proper sequence and lowered into position in the spent fuel pool using a hoist with a minimum capacity of 20 tons.
Each rack will be positioned in the fuel pool within the rack to rack and rack to wall tolerances.
GeneralInformation August 1999 o:\\4878-non. doc:1b-090199 1-5 Revision 0
After the rack is lowered into the pool, a final verification of the tolerances is made and the position adjusted as needed to confirm that the required acceptance standards are met.
Following installation of the rack (s), a drag test is performed using a fuel assembly mockup to verify that a fuel assembly can be inserted into the fuel locations.
The following is the likely sequence of fuel and rack movements within Fuel Pool B for installation of the Irplacement high-density storage racks. The rack numbers referred to in the following section are identified on Figure 1-1.
1.
Move ~ he fuel from existing Racks 1,2,3 and 4 to existing Racks 7 and 8 and to the north t
most rows of existing Racks 5 and 6. This will remove all fuel from existing Racks 1,2, 3, and 4. Racks 1,2,3, and 4 will be prepared for removal.
2.
Remove existing Racks 1 and 2 from Pool B by lifting them straight up.
3.
Remove existing Racks 3 and 4 by moving them laterally to the South end of Pool B, one at a time and removing them.
4.
Remove debris, as necessary, to clean the bottom of the South end of Pool B. Repeat as necessary later in sequence.
5.
Install replacement Rack 3 by lowering the rack into the South end of Pool B and moving the rack laterally north into position.
6.
Install replacement Rack 4 by lowering the rack into the South end of Pool B and moving the rack laterally to the north into position.
7.
Install replacement Racks 1 and 2 into their positions in the South end of Pool B.
8.
Move all fuel assemblies in existing Racks 5,6,7, and 8, to replacement Racks 1,2,3, and 4, leaving the North most rows of Racks 3 and 4 free of fuel.
9.
Lift and remove existing Rack 8 by lifting the rack straight up out of the pool and moving it over the cask-loading pit.
- 10.
Remove the remaining existing racks by laterally moving the rack (s) clear of the surrounding racks and interfaces and lifting the rack (s) straight up out of the pool. The rack (s) is moved over the cask-loading pit.
11.
Install repla' cement Racks 5,6,7, and 8 by lowering the rack (s) into the East end of the Spent Fuel Pool, and moving the rack (s) laterally into its final location.
1 GeneralInformation August 1999 oA4878-non.docib-090199 1-6 Revision 0
l-1 390
'l 2.4 5. --
91.95 ' TYP
- 2.30 TYP 101.06 3.74
- 3.12 RACK 05 RACK 06 RACK 08 RACK 07 119.28 j
10 X 13 10 X 13 11 X 13 10 X 13 TYP 131 MOD 250-22 2 50
/N i
287 -
l l
- 8.60 RACK 04 RACK 03 2.44 36.00 TYP 82.84 12 X 9 12 X 9 2900 TYP-N l
CASK LDADING PIT 120.00 RACK 02 RACK 01 73.73 TYP 12 X B 12 X 8 3 00 I-11017 TYP
- - 2.50
- 8.71 234 36.00 120.00 j
i
\\
i I
l Figure 1-1 Crystal River 3 Spent Fuel Pool B Storage Rack Arrangement GeneralInformation August 1999 a\\4s7s-ron. doc 1b-090199 -
1-7 Revision 0
C I
~ 8 910 N084 f vP 8906 N0at TvP-l Nuh-CELL,1 D $2VAet CCLL ID 50uARC l
l-l l
e 9.18 f YP C[LL C1C $ PACING I
/
\\
l CCaNER CDVCA g l
5 I
L j
l
)
/
/
\\
/
e f
lI' CELL A!stMat v-f 020 WeA*PER t
- 43P MAI CLEARANCE l
-.073 BORAL 162 00 C73 CELL W4LL 163 73 I
e l
l I
l l
l l
l l en j
A,c g
.Au ei u,,,,
!q
!.. I J
=
,,7 3,3 l
1 N
h,N
(,,,,,,i t..,< An i
,,,,,,,,, /
J l-C. Ah, now.a c 4..
om-l..
C AN, now se t n.,,v.
irveu pan Assr /
\\ u-su,N n=cas S
4 LNJus SP'd WeddfP**MW i
l 1
Figure 1-2 Crystal River 3 Spent Fuel Pool B Storage Rack Module CrossSection I
GeneralIrdorrnation August 1999 a:\\4Bhondoc:1b-090199 1-8 Revision 0
{'_
911 SQUARE P]TCH
~
~
CELL CENTERLINE SPACING
~
~
VRAPPER PLATE 304L SST
.020 THK CELL VALL 304L SST g
j I-
.075 THK I
E \\
i
\\ -.is' g
g
(
_ BORAL SHEET g
.075 THK s >
~
~
~
~
TYP CELL INSID DIM.
107 THK I
BORAL VIDTH I
7.50
.1 I
l Figure 1-3 Crystal River 3 Spent Fuel Pool B Storage Rack Typical Cell Layout Generallnformation August 1999 o:\\4s7s-non. doc 1b090199 1-9 Revision 0
2.0 THERMAL-HYDRAULIC ANALYSIS 2.1 DECAY HEAT CALCULATIONS FOR SPENT FUEL The addition of high density spent fuel racks to Spent Fuel Pool B increases the total storage capability of Spent Fuel Pools A and B. They will accommodate f:tel discharges for the duration of the current operating license including reserved storage spaces for any full core discharge that may be required. The maximum heat load generated in the pools is based on off-loading the full reactor core at the end of a refueling cycle. The heat load calculations are based on a recent revision to the applicable FPC calculation (Reference 7.11.m) which is based on ANSI /ANS 5-1 (1994) (Reference 7.4.d) using our burn-up history and projected burn-ups. The maximum decay heat load is 29.6 x 106 Btu /hr. It is comprised of 26.6 x 106 Bru/hr. for the off-loaded full core plus 3.0 x 106 Btu /hr. for all the stored spent fuel assemblies in the spent fuel pools. For the off-loaded full core, a cooling period of 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> is assumed after reactor shutdown. This 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> will be the earliest time frame for isolation of the pool from the decay heat system. This is an assumption of the current bounding analyses. Shorter time frames could likely be supported for specific heat sink conditions, spcnt-fuel heat loads, etc.
Decay heat load calculations (Reference 7.11.1) have been generated to determine the maximum steady state temperature of the spent fuel pools for the established decay heat load of 29.6 x 106 Btu /hr. during a full-core off-load. Calculated pool temperatures are based on the design cooling water temperature of 95 F for the Nuclear Services Closed Cycle Cooling System. The calculated maximum pool temperature with both spent fuel cooling loops in operation is 157 F.
The spent fuel cooling system can maintain the spent fuel pool below 160 F for a full core off-Joad IS6 hours after shutdown with two complete trains of spent fuel cooling running in parallel. Should one of these trains fail the pool will not reach 190 F for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If a single pump can be configured to supply both heat exchangers, the temperature can be maintained below 190 F. If not, the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is sufficient time to perform the necessary repairs or provide an alternate source of water. Examples of these sources are utilization of the Fire Protection System (using pre-staged temporary connections and hoses) to maintain inventory or utilization of the Decay Heat Removal System that is (permanently) piped to the Spent Fuel Cooling System for cooling.
These temperatures are acceptable structurally as indicated in GAI Report 1949 (Reference 7.11.n). The structural capacity of the spent fuel pools is considered adequate for a steady state water temperature of up to 160*F or transient temperatures of up to 190 F in both pools.
2.2 THERMAL HYDRAULIC ANALYSIS FOR SPENT FUEL COOLING 2.2.1 Local Fuel Bundle Thermal-Hydraulics The purpose of this analysis is to verify that the proposed fuel rack design allows adequate coolant flow for the removal of decay heat generated by the spent fuel assemblies. The primary Thermal-Hydraulic Analysis August 1999 o:\\4878-non.docitr090199 2-1 Revision 0 m
f parameters of intemst are the maximum temperature of the pool water and fuel cladding at the outlet of the spent fuel assemblies.
l
'2.2.1.1 Criteria.
The criteria used to determine the acceptability of the design from a thermal-hydraulic viewpoint are summarized as follows:
1.
The fuel rack design and the spent fuel pool cooling system must allow adequate i
cooling of spent fuel assemblies by natural circulation. The coolant should remain subcooled at all locations within the pool when the cooling system is in operation.
j l
2.
If the cooling system is postulated not to be in operation, adequate cooling by natural circulation should limit fuel clad temperatures such that no structural failures would occur.
2.2.1.2 Key Assumptions 1.
. The nominal pool water level is approximately 24 feet above the top of the fuel storage racks.
2.
' A decay heat rate of 44.0 Btu /sec/ assembly is assumed for all available storage locations based on 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay time after reactor shutdown. Note: 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> is a current administrative limit for beginning fuel transfers to the pool.
3.
The maximum pool temperature is assumed to equal 160*F with one train of spent fuel pool cooling and no full core off-load or two trains and a concurrent off-load.
4.
. The maximum temperatum with only one single spent fuel train operating while a full core is off-loaded is assumed to be 190 F.
= 5. '
When the pool cooling is inadequate, the maximum temperature at the inlet to the cells is assumed equal to the saturation temperature of 212*F at the top of the pool.
2.2.1.3 Analytical Method and Calculations An analytical model is employed to determine the thermal-hydraulic conditions within the spent fuel storage cells being installed at CR-3. The model simulates the natural circulation
' flow within the spent fuel pool and determines the resulting fluid temperatures and flow velocities, which are then used to determine the cladding temperature for the stored fuel.
The model assumes that all down-flow occurs in the peripheral gap between the pool walls and the outermost storage cells, and that all lateral flow occurs in the space between the bottom of the racks and the pool floor. A multi-channel formulation is used to determine the variation in axial velocities through the various storage cells. The hydraulic resistance of the storage cells and the fuel assemblies is conservatively modeled by applying large uncertainty factors to loss Thermal Hydraulic Analysis.
August 1999 oms 78an. doc!b-090199 2-2 Revision 0
coefficients obtained from various sources. When necessary, the effect of the Reynolds Number on the hydraulic resistance is considered, and the variation in momentum and elevation head pressure drops with fluid density is also determined.
The solution is obtained by iterative solution of the conservation equations (mass, momentum and energy) for the natural circulation loops. The calculated flow velocities and fluid temperatures are used to determine the fuel cladding temperatures. An elevation view illustrating the modelis shown in Figure 2-1, where arrows indicate the flow paths. Note that each storage location corresponds to a row of storage locations that is located at the same distance from the pool walls.
The lateral flow area underneath the storage cells decreases as the distance from the wall increases as shown in Figure 2-2. This counteracts the decrease in total lateral flow due to flow branching upwards into the cells. This is significant because the lateral flow velocity affects both the lateral pressure drop underneath the cells and the turning losses that are experienced as the flow branches up into the cells. These effects are considered in the natural circulation analysis.
Fuel assemblies from the most recently discharged batch (" hottest" fuel assemblies) are assumed to be located in various rows in order to ensure that they may be placed anywhere within the pool, and still satisfy the two criteria discussed above. In order to simplify the calculations, each row of the model is assumed to be composed of storage cells having a uniform decay heat level. This decay heat level may or may not correspond to a specific batch of fuel. The most conservative analysis that can be performed is to assume that all assemblies in the pool have the same maximum decay heat rate. This maximizes the total natumi circulation flowrate, which leads to conservatively large pressure drops in the downcomer and lateral flow regions, which reduces the driving pressure drop across the limiting storage locations.
Since flow velocity strongly affects the temperature rise of the water and the heat transfer coefficient within a storage cell, hydraulic resistance is a significant parameter in the evaluation. In order to minimize hydraulic resistance, the design of the inlet region of the racks j
has been chosen to maximize this flow area. Each storage location has one large or several middle size flow openings as shown in Figure 2-3. The use of large flow holes or multiple holes significantly reduces the possibility that all flow into the storage racks can be blocked by debris or other foreign material that may get into the pool. In order to determine the impact of a partial blockage on the thermal-hydraulic conditions in the cells, an analysis is also performed for various assumed blockages.
Thermal-Hydraulic Analysis August 1999 c:)4878-non doc 1b- 090199 2-3 Revision 0
2.2.1.4 Results CASE A Inputs The rack inlet temperature is up to 190 F (allows for the failure of one train of SF cooling).
The decay heat output is equal to 44.0 Btu /sec/ assembly, corresponding to removal from the reactor at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown.
There is uniform decay heat loading in the poel - no credit for the lower actual decay heat input.
The peak rod produces 60 percent more heat output than the average rod.
All storage cells are filled.
Results The results of the analysis show that no boiling occurs within the storage racks when the conditions noted are present. The maximum clad surface temperature of the peak rod during pool operation under these conditions is equal to 246*F. Note, while no specific case was run, when conditions limit pool temperature to <160 F the pool will remain subcooled throughout all areas of the rack modules.
CASE B Inputs l
The rack inlet temperature is equal to 160 F.
Up to 80% of flow is Blocked The decay heat output is equal to 44.0 Btu /sec/ assembly, corresponding to removal
' from the reactor at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown.
There is uniform decay heat loading in the pool - no credit for the lower actual decay heat input.
The peak rod produces 60 percent more heat output than the average rod.
I 1
All storage cells are filled.
l I
I i
Thermal 4-lydraulie Analysis August 1999 l
o:\\4878an. doc:1t>-090199 2-4 Revision 0
Results The results of the analysis show that there would be no boiling in the water channels inside the cells. The' maximum clad surface temperature of the peak rod with 80% flow blockage is 236*F.
CASE C Inputs inadequate pool cooling implies that the temperature of the water at the inlet to the e
spent fuel racks is 212*F, which corresponds to the saturation temperature at the pool surface.
A nominal water level of 24 feet above the top of the racks is maintained.
There is uniform decay heat loading in pool - no credit for lower actual heat input The decay heat output is equal to 44 Btu /sec/ assembly, corresponding to removal from the reactor at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown.
The peak rod produces 60 percent more heat output than the average rod.
All storage cells are filled.
Results The results of this analysis show that due to the effects of natural circulation, the fuel cladding temperatures are sufficiently low to preclude structural failures and the maximum clad surface of the peak rod is equal to 267 E Thermal-Hydraulic Analysis August 1999 oM87B*on docib-090199 2-5 Revision 0
3 i
WATER SURFACE o
WATER HEAD ABOVE RACKS Pu n
n -
n n
a o
a a
n If U.
ROW j
X
/
1 2
3 4
5 6'
7 8
9
=
i 1
o n -
a n -
a n
- n n
a 5
-g-o 5
8 o
h
'h h
h h
h h
h h
n GJJJJJJJJ HEIGHT ABOVE FLOOR LATERAL FLOW REGION n
Figure 2-1 Spent Fuel Pool Natural Circulation Model (Elevation View)
Thermal-Hydraulic Analysis August 1999 OM84non. doc 1b-090199 2-6 Revision 0
IoE u
4 G
O Y
oc ROO LF EVO B
A T
H G
I E
H
/
o A
E R
m A
W O
y L
F L
e A
R E
R E
M T
w O
A C
L N
W a
O D
u M
IoE m
OoEzoOEnE i
moor E Rr 5
~h m1EmE m8 ZE,1n:nE25sfCD2 T."$.aE~
9a3ehEE
.ME=
>55 8
5a" Sl7 wN 5s.. g0
1 I
I I
I
(
i
- O
(,y I
I t
l l
- O oOn g-t o
L O
O O
1 j
l i
' Figure 2-3 Spent Fuel Rack Inlet Flow Area (Plan View)
- Thennal-Hydraulic Analysis August 1999 o:\\4878-non doc 1b-090199 2-8 Revision 0
r; 3.0 MECHANICAL, MATERIAL, AND STRUCTURAL 4
CONSIDERATIONS i
3.1 DESCRIPTION
OF SPENT FUEL POOL j
There are two individual spent fuel storage pools located within the fuel handling awa of the Auxiliary Building. Both pools are rectangular in plan (nominally): Pool A is 32' by 24' and Pool B is 33' by 24'. Both pools have a depth of 44'. A 10' by 10' cask storage pool is located in the southeastern corner of Pool B. The walls and bottom slab of the spent fuel pools have a nominal thickness of 5', except for the common wall between the two pools which is 4' thick.
i Walls forming the cask storage pool within Pool B are 3' thick. Wetted surfaces of the pools are lined with stainless steel to ensure watertight integrity.
I The spent fuel storage pools are supported on reinforced concrete walls extending downward l
to elevation 93', the top of the structural foundation mat. The bottom slab of the spent fuel storage pools is at elevation 118'. The operating floor at the spent fuel pools is at elevation 162'.
Above elevation 162', the spent fuel pool area is enclosed within the steel framed and metal sided Auxiliary Building structure. Missile shield structures are normally in place over the top of the spent fuel pools.
3.2 STRUCTURAL DESIGN OF RACKS l
The Pool B spent fuel replacement racks are described in Section 1.3.
3.3 INTEGRITY OF FUEL RACKS UNDER FUEL HANDLING ACCIDENT CONDITIONS I
The spent fuel handling machine load and fuel assembly drop scenarios were analyzed. Each of these scenarios is discus' sed below.
3.3.1 Spent Fuel Handling Machine Load Scenario An analysis was performed to demonstrate that a rack cell can withstand an axial upward or downward force of 500 pounds. This scenario could occur as a result of a jammed fuel assembly in the rack. The load is limited to 500 pounds because the Crystal River Unit 3 fuel handling equipment has a load limit switch that will stop fuel assembly movement in a storage j
position ifit exceeds 500 pounds. In the analysis, the load is applied to a cell assembly, the cell to cell weld, and the cell-to-base-plate welds. The highest stress,4.90 KSI, was found in the cell-to-cell welds. This compares to an allowable stress of 11.33 KSI (Service Limit Level B, Reference 7.3.b). The rack can maintain its integrity (with a safety factor of 2.3) in this scenario.
3.3.2 Fuel Assembly Drop Scenarios One storage situation was considered for the accidental drop of a fuel assembly onto or into the spent fuel racks. There are no plans for use of consolidated canisters in the pool utilizing this
- Mechanical, Material, and Structural Considerations August 1999 a.\\4sn-nondoc:1b-090199 3-1 Revision 0
spent fuel rack design. Therefore, no consolidated canister drop was analyzed. The fuel assembly assumed was a B&W 15x15 standard fuel assembly plus the handling tool, having a total maximum dry weight of 2,750 pounds. It was postulated to be dropped from a height of 24 inches above the top of the rack.
3.3.2.1 Drop Orientations Three orientations were considered. These were:
A.
Drop of a fuel assembly onto the top of a rack with the assembly in a vertical position, B.
Drop of a fuel assembly onto the top of a rack with the assembly in an inclined position, C.
Drop of a fuel assembly through an empty rack cell to the bottom of the rack.
These orientations are shown in Figures 3-1,3-2, and 3-3 respectively.
3.3.2.2 Acceptance Criteria Functional capability of the fuel racks is maintained.
e Functional integrity of the spent fuel pool structure, including liner is maintained.
3.3.2.3 Assumptions for Energy Dissipation Orientations A and B l
The kinetic energy of the dropped fuel assembly plus the handling tool is totally converted into the strain energy of the rack structure.
The fuel assembly falls freely in the spent fuel pool with hydrodynamic drag being considered.
The pool liner and floor flexibilities are not considered.
Orientation C The kinetic energy of the fuel assembly plus the handling tool is totally converted into e
the strain energy of the dropped fuel assembly.
The fuel assembly falls freely in the spent fuel pool with hydrodynamic drag being considered.
The pool liner and floor flexibilities are not considered.
Mechanical, Material, and Structural Considerations August 1999 0:kM78-non.d oc:1b-090199 3-2 Revision 0 L
3.3.2.4 Drop Analysis Results 3.3.2.4.1 Functional Capability of the Fuel Racks During fuel drop scenarios as defined by Orientations A and B, the fuel assembly impacts the fuel rack cells. These scenarios are shown in Figures 3-1 and 3-2. Orientation C is not considered here because the fuel assembly drops down within a cell and does not impact the f
fuel rack cells (refer to Subsection 3.3.2.4.3). It must be shown that the fuel racks will retain j
functional capability. To demonstrate that the fuel rack functional capability is not
{
compromised, elastic-plastic evaluations were performed. An energy relationship was written i
equating the kinetic energy and potential energy associated with the fuel assembly and the strain energy of the rack as it undergoes deformation. The fuel rack is assumed to behave in an I
elastic perfectly plastic manner. A finite element model (described in Section 3.4.3.1) was used to define the load at which the rack first reaches yieH. Eight cases were evaluated:
a, C i
The cell walls at the top of the rack are crushed in the area local to the impact defined by Orientation A and B in Figures 3-1 and 3-2. The plastic deformation is between 1.9 and 6.0 inches for the eight cases evaluated. The magnitude of the deformation depends on where the fuel assembly impacts the top of the fuel rack. The rack structure, however, remains capable of resisting loads from deadweight, thermal, and seismic events. In addition, the crushing does not impair the ability of the rack to maintain the fuel in a non-critical state. Therefore,it is concluded that the functional capability of the fuel racks is maintained during a fuel drop accident.
Mechanical, Material, and Structural Considerations August 1999 c:\\4878-non. doc 1b-090199 3-3 Revision 0
L 3.3.2.4.2 Functional Integrity of Spent Fuel Fool Structure and Liner Fuel drop Orientation C (Figure 3-3) was evaluated with respect to the integrity of the pool
- structure and liner. The impact velocity of the dropped fuel assembly was deternuned accounting for the effects of hydrodynamic drag. The drag coefficient was established based on a profile of a square rod dropping through water. The impact force on the pool liner and the supporting concrete floor was developed from the kinetic energy of the dropped fuel assembly.
The kinetic energy was converted to strain energy associated with the dropped fuel assembly lower end fitting. The impact load was then calculated from this strain energy.
The bearing load on the steel liner for the Orientation C drop was determined to be 506 kips, compared to an allowable bearing load of 688 kips (Reference 7.3.b). The punching load on the concrete was 506 kips, which is less than the allowable load of 2,322 kips for punching shear (Reference 7.4.a).
3.4 STRUCTURAL ANALYSIS PROCEDURES FOR SPENT FUEL STORAGE RACKS 4
The purpose of this section is to demonstrate the adequacy of the spent fuel rack design to store B&W 15x15 fuel assemblies under normal, accident and external event (seismic) loading conditions.
The stresses were checked against the design limits to ensure the structural adequacy of the design. The' dynamic models used considered potential lift-off, potential fuel impact with the rack cell, potential impact between individual racks, and the racks with the pool liner.
3.4.1 Analysis Overview The spent fuel storage racks are required to be Seismic Class I equipment. Therefore, they must remain functional during and after a Safe Shutdown Earthquake (SSE). The racks are free standing (neither anchored to the pool floor, to the pool wall, nor structurally interconnected)
- and the fuel is free to move inside the cell within the limits of the clearance between the fuel and cell. Therefore a nonlinear seismic analysis was performed. The modeling features and parametric considerations that were considered in the analysis are summarized below:
Modeling Features Parametric Considerations geometric non-linearities fuelloading configurations
- fuelimpact -
fuelimpact stiffness
- rackimpacts friction coefficients e
- racklift-off damping fluid interactions structural characteristics of the fuel and racks a
rack configurations (12x8,12x9,10x13,10x13 (modified) and 11x13)
Mechanical, Material, and Structural Considerations August 1999 i
o:M878-rmatdoc:1b-090199 3-4 Revision 0
1 The ANSYS code (Reference 7.11.a) was used for all the rack structural and seismic analyses.
l ANSYS is a general-purpose computer program for finite element analysis and design.
j Westinghouse has extensive experience in the use of ANSYS for nuclear component design and j
analysis. This experience includes the Wesflex dry cask storage and transportation system and major components of Nuclear Steam Supply Systems such as reactor vessels, reactor vessel internals, and steam generators.
The overall analysis process is presented in Figure 3-4. The steps in the analysis were:
Develop seismic time histories based on the Crystal River Unit 3 licensing basis.
I Develop dynamic finite element single rack and whole pool multiple rack models.
e Perform accident analyses for the fuel assembly and stuck fuel assembly scenarios.
l Perform seismic analyses using the ANSYS code.
e Develop fuel rack structural model for stress evaluation.
Perform stress evaluation of the racks using Reference 7.3.b.
3.4.2 Synthetic Time Histories One set of synthetic time histories (for the three orthogonal directions North-South, East-West and Vertical) was created for use in the non-linear seismic analysis of the racks. These seismic time histories are consistent with the Crystal River Unit 3 Auxiliary Building seismic design requirements at elevation 119', and were generated in accordance with NUREG-0800 Standard Review Plan, Section 3.7.1 (Reference 7.7.a).
Figure 3-5 presents the acceleration time histories for the Operating Basis Earthquake (OBE).
The time histories are 20 seconds long, and include a strong motion duration of 12 seconds. The Safe Shutdown Earthquake (SSE) time histories have been scaled-up from the OBE based on the Crystal River Unit 3 seismic design licensing basis (Reference 7.11.b). This design basis states that the SSE floor response spectra are two times the OBE, and the vertical floor response spectra are two-thirds of the horizontal spectra.
The time histories have been created using a computer code suite (References 7.11.c-f) developed, and verified in accordance with Westinghouse and Paul C. Rizzo Associates quality assurance procedures (Reference 7.11.g).
Figures 3-6 through 3-11 compare the computed and design floor response spectra in the North-South, East-West and vertical directions for both the OBE and SSE. The computed spectra meet the spectra enveloping criteria of Reference 7.7.
Mechanical, Material, and Structural Considerations August 1999 c:\\4878-non. doc 1b-090199 3-5 Revision 0 u
The power spectral densities (PSDs) associated with the ground motion used in the development of the synthetic time histories are given in Figures 3-12 through 3-14. These PSDs envelope the estimated target PSD associated with the licensing basis horizontal and vertical ground design response spectra in accordance with Section 3.7.1 of Reference 7.7.a.
For simultaneous application in the two orthogonal directions, the respective time histories must be statistically independent. This requirement is met if the correlation coefficient of the two orthogonal time histories is less than or equal to 0.16 as provided in Regulatory Guide 1.92 (Reference 7.11.k). The correlation coefficients for both the SSE and OBE for the three
{
combinations of directions are: 0.04 (North-South versus East-West),0.15 (East-West versus Vertical) and 0.16 (North-South versus Vertical). Based on these correlations,it is concluded that the time histories are statistically independent.
i 3.4.3 Stress and Seismic Models a, e I
l 1
Mechanical, Ma terial, and Structural Considerations August 1999 o:\\484non. doc 1b-090199 -
3-6 Revision 0
ac s
l i
5.4.3.1 Static Model of the Fuel Rack Structure
.a.g 1
J i
1 i
i 3.4.3.2 Single Rack Dynamic Model a, e I
Mechanical, Material, and Structural Considerations August 1999 a:\\4878-non.docit 090199 3-7 Revision 0 L_
7; EC i
)
Mechanical, Material, and Structural Considerations August 1999 c:\\4878-noniocibe90199 3-8 Revision 0
g.
l Ac l-3.4.3.3
_ Whole Pool Multiple Rack Dynamic Model LS l
l I
i l
l 1
i t
l i
i i
l Mechanical, Material, and Structural Considerations August 1999 0:\\#78-natdoc1b-090199 3-9 Revision 0
[7 l-3.4.4 Modeling Features 3.4.4.1-Damping 1.5 1'
l l
3.4.4.2 Fluid Interactions Sc
\\
i 3.4.4.3 Friction Coefficient a, C 3.4.5 Seismic Evaluation AS j
i Mechanical, Material, and Structural Considerations August 1999 oM878am.docit>090199 3-10 Revision 0
I a, c 3.S
- STRUCTURAL ACCEPTANCE CRITERIA AND ANALYSIS RESULTS FOR SPENT FUEL STORAGE RACKS 3.5.1 Structural Acceptance Cdteria The racks must satisfy the structural criteria outlined in NUREG-0800, Section 3.8.4, Appendix D, item (6) (Reference 7.7.b.) The structural acceptance criteria are comprised of
'(1) the limitation of rigid body rack sliding and tilting motions, and (2) maintaining stresses within acceptablelimits.
3.5.1.1 Rigid Body Rack Sliding and Tilting Motions The condition stated in part (b) of NUREG-0800, Section 3.8.4, Appendix D, item (6)
(Reference 7.7.b)is imposed. Any sliding and tilting motion will be contained within the geometric clearances between the racks.
3.5.1.2 Stress Limits The load combinations for determining the stresses in the racks are in accordance with Table 1 of NUREG-0800, Section 3.8.4, Appendix D, item (6) (Reference 7.7.b.) The stresses in the racks are evaluated to the limits of Reference 7.3.b for Class 3 supports. The load combinations and stress limits are listed in Table 3-2.
3.5.2 Stress Analysis The stress analysis of the racks is performed by linear elastic analysis in accordance with Reference 7.3.b. Material temperature, the effects of buckling, and the allowable increase of stress limits per service level listed in Table 3-2 are considered in the stress analysis. The results of the stress analysis are listed in Table 3-3.
4 3.5.3 Results for Single Rack Model Analysis The impact loads on the fuel assemblies due to the interaction with the rack during a seismic event have been determined. The maximum calculated seismic impact load for the SSE at fuel Mechanical, Material, and Structural Considerations August 1999 0:\\4878en. doc 1bD90199 3-11 Revision 0
spacer grid locations is 947 pounds. The impact loads have been considered in the stress analysis.
The results of the Single Rack Model stress analysis are presented in Tabb 3-3 for both the OBE and SSE. The results show that all induced stresses under the load con.' qations in Table 3-2 are less than the corresponding allowable stresses specified in the ASME Code,Section III, Division 1, Subsection NF (Reference 7.3.b), indicating that the rack design is adequate.
From the nonlinear time history analysis, the maximum rack displacement (absolute displacement) was determined to be 0.084 inches in the East-West direction and 0.080 inches in the North-South direction. From these results, it is concluded that the racks are spaced with ample clearance so that racks do not impact adjacent racks or the pool wall.
The seismic analyses performed employed nonlinear dynamic analyses. Sliding and uplifting (tilting) behavior was included in the model using gap / sliding elements, and nonlinear impact elements that have the capability to respond in compression or to represent uplift by not carrying any load. From the detailed nonlinear dynamic analyses performed,it was detennined that there is no sliding even using the conservatively low value of 0.2 for the coefficient of friction. Further,it was determined from the seismic time history analyses that l
there was no uplift of a fully loaded rack. The sliding and uplift (tilting) motion is less than the clearances, and there is no impact between adjacent rack modules or between a rack and the pool walls.
l 3.5.4 Parametric Studies Results Parametric studies were conducted to investigate fuel loading configurations, fuel impact stiffness, and friction coefficients on the seismic response of the fuel racks. The results of these pararnetric studies are discussed in the following sections.
3.5.4.1 Fuel Loading Configuration a, c The results are given in Table 3-4 for three different loading configurations; empty rack, quarter-l filled rack, and half-filled rack. No overtuming or sliding occurred in these loading j
configurations. The fully loaded rack configuration has been analyzed showing no sliding or 3
uplift. During rack loading, the fuel is placed such that overtuming is prevented on the rack side that is next to another rack in the pool. In summary, all of the rack configurations evaluated meet the factors of safety against tilting per NUREG-0800, Section 3.8.5.I1-5 (Reference 7.7.c).
I Mechanical, Material, and Structural Considerations August 1999 o:\\4878-nondoc1b-t90199 3-12 Revision 0 L
[
I
n 3.5.4.2 FuelImpact Stiffness a, c 3.5.4.3 Friction Coefficients As stated in Section 3.4.4.3, the upper and lower bounds of friction values are 0.8 and 0.2. The analyses performed used a coefficient of friction of 0.2, corresponding to the lower bound friction value. Since no sliding or uplift occurred with the minimum value, it was not necessary to perform analyses with the upper bound value, since the results would be the same.
3.5.5 Whole Pool Multiple Rack Model Results The WPMR model was created to evaluate the dynamic effects between racks in the spent fuel pool. The results of the WPMR model for both the SSE and OBE are presented in Table 3-5. The results are contrasted with the Single Rack Model analysis. As seen from the comparison, there is an amplification effect due to the interaction of adjacent racks and water that is not obtained from the single i ack model. Although the rack displacements are larger, no impact of the adjacent racks or pool wall occurred. The minimum rack-to-rack or rack-to-wall clearance is 2.3 inches. During the seismic event, this clearance between adjacent structures was reduced to 2.1 inches. In addition, no sliding or uplift (overturning) occurred. Further, the local impacts associated with fuel assemblies obtained from the WPMR analysis were similar to the single ratx analysis. The Single Rack Model stress analysis was performed with seismic loads that envelope the WPMR model results.
3.6 MATERIALS, QUALITY CONTROL, AND SPECIAL CONSTRUCTION TECHNIQUES 3.6.1 - Construction Materials The new fuel racks have been designed and will be constructed with consideration given to the mechanical and material qualifications, neutron poison, fuel handling qualifications, fuel climensional compatibility, and accident conditions. The principal construction material for the new racks will be annealed austenitic stainless steel. The rack designs, material selection and fabrication process will comply with the applicable ASTM Standards for service in the nuclear and the boric acid environments. The governing quality assurance requirements for fabrication of the racks are consistent with the quality assurance and quality control requirements of Mechanical, Material, and Structural Considerations August 1999 n:ws-non.docib-090199 3-13 Revision 0
10CFR50, Appendix B.' All materials used in the construction of the racks are compatible with the spent fuel pool environment. Surfaces that contact the fuel assemblies are made of annealed austenitic stainless steel. Other materials are corrosion resistant and will not adversely impact the stored fuel assemblies. Holes in the austenitic stainless steel wrapper are intended to vent off-gassing from the Boral exposed to the boric acid environment.
The racks will be made of 304 L Series stainless steel.
The materials'used in the construction of the spent fuel storage racks are not detrimental to the Spent Fuel Pool, Certified material test reports or material manufacturer's certification of compliance for e
materials used in construction including all weld material, confirm that chemical and mechanical properties meet applicable ASME material standards. In certain cases, ASTM material that meets ASME material standards is used.
Boral is used as a neutron absorber material and is discussed in Section 3.6.2. Boral is not used as a structural component.
_ Materials contacting the racks after final cleaning and installation will not consist of any unacceptable material containing halogens in amounts of 50 ppm or greater, including chlorinated cleaning compounds.
Any material that would form alloys or deposits on the fuel assembly will not be used in the fabrication of the racks.
Any carbon steel rack handling tool or rigging that has a protective coating and comes in contact with the pool water is compatible with the spent fuel pool environmental conditions.
3.6.2 Neutron Absorbing Material /MaterialIntegrity For primary nuclear criticality control in the new racks, a fixed neutron absorber integrated within the rack structure will be used. The absorber, trade name Boral, is a boron carbide and aluminum-composite sandwich. Boral is chemically stable and has a long history of application i
in spent fuel pool environments where it has maintained its neutron attenuation capability under thermal loads. Boralis manufactured under the control of a quality assurance program, which conforms to the requirements of 10CFR50, Appendix B. Boral applications do not generally mquire a surveillance program to periodically sample Boral coupons. Therefore, FPC does not consider such a program as required for the new racks. The initial examination will include visual inspection, as well as other tests determined necessary to verify that the performance of the Boralis consistent with the reported test results. Based on these results, FPC will take appropriate actions, if any, to assure material performance is acceptable
. throughout the life of the plant.
1
' Mechanical, Ma'erial, and Structural Considerations August 1999 c:\\4878-non. doc:1b-090199 3-14 Revision 0
The neutron absorbing materialis Boral manufactured by AAR Advanced Structures.
Boral is a metallic composite of a hot rolled (sintered) aluminum matrix containing boron carbide sandwiched between and bonded to type 1100 alloy aluminum. The boron carbide (B4C) is a stable chemical compound. The type 1100 alloy aluminum is a lightweight metal with high tensile strength that is protected 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 spent fuel pool environment.
The boron carbide contained in Boral conforms to ASTM C-750-89 nuclear grade Type 111 (Reference 7.A.b).
The Boral neutron absorbing material has a nominal B10 loading of 0.02 gm/cm2 (areal density), including a self-shielding penalty of 25%.
The B10 concentration is documented by the Boral material supplier certification data reports. Heat code traceability to the actual cell installed is maintained.
The material integrity of the spent fuel racks will be assured,in part, through monitoring of the spent fuel pool water quality. Currently, weekly chemical analyses are performed to monitor pH, boron, fluoride, chloride, silica, and turbidity. Other chemical analyses are currently completed on a quarterly basis including zeolites such as Aluminum, Calcium, and Magnesium. Currently, gamma-emitting radionuclide analysis is completed weekly, while tritium and alpha analysis is completed semi-annually for monitoring stored fuel cladding integrity.
3.6.3 Quality Assurance The overall quality assurance program conforms to the Westinghouse " Quality Management System (QMS)."
The spent fuel racks (including materials, welding and NDE) are fabricated in accordance with ASME Boiler and Pressure Vessel Code,Section III, Division 1, subsection NF for Class 3 Component supports (Reference 7.3.b).
NDE personnel qualification for inspections associated with rack fabrication is in accordance with SNT-TC-1A (1984) (Reference 7.4.c).
Certified material test reports in accordance with the requirements of NCA 3800 are required for all materials. The certified material test reports are in accordance with Article F-1000 of Section III of the ASME Code, and the rack drawings and specifications.
Mechanical, Material, and Structural Considerations August 1999 oM878-non. doc:ltr 090199 3-15 Revision 0
r-I Table 3-1 Fuel Assembly Drop Scenarios Results Weight of Fuel Assembly Drop With Tool Height Impact Velocity Kinetic Energy Drop Orientation (pounds)
(in.)
(in/sec.)
(Ib-ft)
A. Vertical drop onto rack 2,750 24 96 2,733 B. Inclined drop onto rack 2,750 24 96 2,733 C. Vertical drop through 2,750 186 268 21,300 cell onto pool floor Mechanical, Material, and Structural Considerations August 1999 o:\\4878 con.docib-090199 3-16 Revision 0
l Table 3-2 Load Combinations for Spent Fuel Racks Load Combination Basis D+L Level A service limits of NF D + L + To level A service limits of NF D + L+ To + E Level A service limits of NF D + L+ Ta + E Level B service limits of NF D + L + To + Pf Level B service limits of NF i
l D + L + Ta + E' Level D service limits of NF i
D + L + Fd The functional capability of the fuel raels should be demonstrated Notes:
Load combinations based on Reference 4 Dead Load D
=
Live Load L
=
E Loads generated by the Operating Basis Earthquake
=
Thermal effects and loads during normal operating and shutdown conditions based on To
=
the inost critical transient or steady state condition Highest temperature associated with the postulated abnormal design conditions J
Ta
=
Upward force on racks caused by a postulated stuck fuel rod (See Section 3.3.1 Spent Fuel Pf
=
Handling Machine Load Scenario)
Loads generated by the Safe Shutdown Earthquake E'
=
Force caused by the accidental drop of the heaviest load from the maximum possible Fd
=
height (See Section 3.3.2 Fuel Assembly Drop Scenarios) l l
i 1
I l
Mechanical, Material, and Structural Considerations August 1999 on4878-non. doc 1b-090199 3-17 Revision 0
o Table 3-3 Rack Minimum Margin to Allowable
- Level A Level B Level D ltem Actual Allowable Actual Allowable Actual Allowable Support Pads' O.04 1.00 0.239 1.00 0.345 1.00 Cells (Stress KSI) 0.05 15.70 17.99 20.88 22.08 25.56 Welds (Stress KSI)
- Cell to Cell 4.45 8.52 9.%
11.33 11.94 13.12
- Cell to Base Plate 1.71 8.52 8.33 11.33 11.41 13.12
- Cell to Wrapper 0.00 0.067 0.001 0.089 0.001 0.103 Notes:
Interaction ratio of axial and bending stress (no units) i i
j Mechanical, Material, and Structural Considerations August 1999 c:\\4878-non. doc 1b-090199 3-18 Revision 0
Table 34 Rack Loading Configuration Factor of Safety Against Tilting Rack Loading
Response
Seismic Factor of Limit of Factors Configuration Behavior Event Safety'"
of Safety Empty Tilting OBE 2.4 1.5 SSE 1.6 1.1 Quarter Filled *'
Tilting OBE 2.0 1.5 SSE 1.3 1.1 Half Filled *'
Tilting OBE 1.7 1.5 SSE 1.1 1.1 Factor of Safety is conservative since it is based on the initiation of uplif t not overturning.
Position of fuel during loading is such that overturning is prevented on the side that the fuelis placed.
Mechanical, Material, and Structural Considerations August 1999 c:\\4878+ondoc1b-090199 3-19 Revision 0
i Table 3 5 Comparison of Whole Pool Multiple Rack Model to Single Rack Model Rack Configurations Deflection (inches)*
Impact (Ibs.)
Single Rack Whole Pool Single Rack Whole Pool SSE 12x8 0.084 0.107 947 788 12x9 NA 0.101 NA 718 10x13 NA 0.183 NA 911 10x13 (modified)
NA 0.179 NA 744 11x13 NA 0.170 NA 674 OBE 12x8
.0.056 0.072 634 528 12x9 NA 0.068 NA 481 10x13 NA 0.123 NA 610 10x13 (modified)
NA 0.120 NA 498 11x13 NA 0.114 NA 452 Notes:
- Deflection at top of rack for the maximum horizontal acceleration.
NA -Not Analyzed (See Section 3.5.4).
Mechanical, Material, and Structural Considerations August 1999 o;\\4878-non. doc 1b-090199 3-20 Revision 0 I
o
i Table 3-6 Impact of Fuel Assembly Stiffness on Fuel Assembly Loads
.!kS l
I i
Mechanical, Material, and Structural Considerations August 1999 o:\\484non.docib-090199 g3 Revision 0
I t
/G?=
TOOL
/
IMPACT AREA 12x8 STORAGE RACK Rz
,4'sk 4
/4 hf Figure 3-1 Fuel Assembly Drop Orientation A Mechanical, Material, and Structural Considerations August 1999 o,\\4878-non. doc 1b-090199 3-22 Revision 0
15x15 FUEL ASSEMBLY W TH HANDLING 0
s s
IMPACT AREA #
12kB STORAGE RACK sd g
NQ W, W.
R Figure 3-2 Fuel Assembly Drop Orientation B l
Mechanical, Material, and Structural Considerations August 1999 o:\\4878 non. doc 1b-090199 3-23 Revision 0
1 i
15x15 FUEL ASSEMBLY WITH HANDUNG TOOL I
l 12x8 STORACE RACK s
g/
q
/,4 IMPACT AREA I
l Figure 3-3 Fuel Assembly Drop Orientation C l
l Mechanical, Material, and Structural Considerations August 1999 a\\ 4878-non. doc 1b-090199 3-24 Revision 0 j
Seismic Analyses wheie r i M.iupie n=k M. dei a,
seisade Tlaw Himertes sinsi. Rock Dynande Medes Accidents Fuel Asseusly Drop Stress Evaluation per ASME Code Criteria Fuel mock Structural Medel racks Stuck Feel Annenddy
- Overturning /stkling of racks Figure 3-4 Overall Structural / Seismic Analysis Process Mechanical, Material, and Structural Considerations August 1999 o:\\4878-non. doc:1b-090199 3-25 Revision 0
[
NORTH SOUTH DIRECTION
- o j -30 0
2 4
6 8
10 12 14 16 18 20 Time (seconds)
Ref F-PCR24 EAST-WEST DIRECTION i
l l
Time (seconds)
Ref F-PCR28 VERTICAL DIRECTION 30 g 20 -
I f
~
g
- 3o 0
2 4
6 8
10 12 14 16 18 20 Time (seconds)
Figure 3-5 Crystal River Unit 3 Synthetic Time History Acceleration, OBE Matching 2%
and 4% Darnped Floor Response Spectra *
' Spectra at Auxiliary Building Spect Fuel Pool Elevation 119"-0' Mechanical, Material,and StructuralCc +iderations August 1999 a:\\48&non.dac:1b-090199 3-26 Revision 0
F
]
0.4 0.35 Danphga%
r ZPA=0.05G J
0.3 1 1180'
{ 0.25 q
]
1 I
3 0.2 M
I lii 0.15 0.1 0.05 0
O.1 1
10 100 Frequency (Hz)
Ref F-PCR25 Figure 3-6 North-South Synthetic and Design Floor Response Spectra for OBE Mechanical, Material, and Structural Considerations August 1999 c:\\4 878-non. doc'1b-090199 3-27 Revision 0 L
l t
l 0.6 i
l 0.5 1
DAWNG=4%
ZPA4.1G EL.119' 0*
g 0.4 8li 5
3 0.3 y
N 0.2 0.1 -
0 O.1 1
10 100 Frequency (Hz)
Ref F-PCR26
/
l Figure 3-7 North-South Synthetic and Design Floor Response Spectra for SSE i
I l
Mechanical, Material, and Structural Considerations August 1999 0:\\4878-non. doc 1b-090199 3-28 Revision 0
I' l
I l
l l
l l
0.5 i
PGA 0.oso l
DAhPNG-2%
l g,4 FLOOR EL 119'-0*
5 0.3-
}
5 a 0.2 -
I u
0.1 -
(
0 0.1 1
10 100 Frequency (Hz)
Ref F-PCR29 Figure 3-8 East-West Synthetic and Design Floor Response Spectra for OBE l
\\
\\
l Mechanical, Material, and Structural Considerations August 1999 l-0:\\ 4878-nondoc1b-090199 3-29 Revision 0 i
~
p I
i-L 1
i l
l
- 0.6
O 0.5-FLOOR EL 119'-0" g 0.4 -
.1li e
~MA' 0.3 -
L 0.2-0.1 -
)
0 0.1 1
10 100 Frequency (Hz)
Ref F-PCR30 i
Figure 3-9 East-West Synthetic and Design Floor Response Spectra for SSE Mechanical, Material, and Structural Considerations August 1999 on4878-non. doc 1b-090199 3-30 Revision 0 I
L.
0.3 f
PGA=0.033G 0.25 -
meg FLOOR EL 119'-0" 0.2 -
8
' ii k
{ 0.15 -
1
)
1 0.1 -
0.05 -
1 0
0.1 1
10 100 Frequency (Hz)
Ref F-FCR33 i
i Figure 3-10 Vertical Synthetic and Design Floor Response Spectra for OBE Mechanical, Material, and Structural Considerations August 1999 c:\\4878-non. doc:1b-090199 3-31 Revision 0 I.
r 0.4 RIA=0.067G DAMPNG=4%
FLOOR EL 119'-0" 0.3 -
5 11 0 I
.2 -
y 8*
0.1 l
0 0.1 1
10 100 Frequency (Hz)
Ref F-PCR34 Figure 3-11 Vertical Synthetic and Design Floor Response Spectra for SSE Mechanical, Ma terial, and Structural Considerations August 1999 oM8Enondoc1b-090199 3-32 Revision 0 m.
l i
l l
l l
100.0000 PGA = 0.05g 10.0000 PSD of Ground Trne Hstory Used to Develop Synthetic Floor Trne Hstory
^^
1.0000 0.1000 s_
Estimated Site-specNic f
0.0100 Target
- PSD 0.0010 I
0.0001 1
0.0000 O.1 1
10 100 Frequency (Hz)
Ref F-PCRll 1
i Figure 3-12 Power Spectra Density of North-South Synthetic Time History Versus Crystal River Target PSD Mechanical, Material, and Structural Considerations August 1999 0:\\4878-non. doc 1b-090199 3-33 Revision 0
100.0000 RSA = 0.05g MD of Gud Tme Hstory used to Develop 10.0000 Synthetic Floor Trne Hstory 1.0000 0.1000 a
m Estmated Site-spectc E
" Target
- PSD 6
2 0.0100 0.0010 0 0001 0.0000 O1 1
10 100 Frequency (Hz)
Ref F-PCR14 Figure 3-13 Power Spectra Density of East-West Synthetic Time History Versus Crystal RiverTarget PSD Mechanical, Material, and Structural Considerations August 1999 o:\\4878-non. doc:1bo90199 3-34 Revision 0
)
100.0000 PGA = 0.0Sg
~
PSD of Ground Time Hstory Used to Develop Synthetic Floor Trre Hstory 1.0000 C
0.1000 Y
i; Fr
[
0.0100 Estmated site-specif
" Target' PSD 0.0010 0.0001 O.0000 0.1 1
10 100 Frequency (Hz)
Ref F-PCR17
)
i Figure 3-14 Power Spectra Density of Vertical Synthetic Time History Versus Crystal River Target PSD Nechanical, Material, and Structural Considerations August 1999 cx\\4r78 eon. doc 1b-090199 3-35 Revision 0 L.
r-a,C 1
1 Figure 3-15 Single Rack Static Model of a 12x8 Rack
. Mechanical, Material, and Structural Considerations August 1999 c\\48%cndocib-090199 3-36 Revision 0 6.
a,c 1
1 1
l-l l
i i
Figure 3-16 ANSYS Finite Element Model Representation of a 12 x 8 Single Rack 1
Mechanical, Material, and Structural Considerations August 1999 3
a.\\48&non.doctb-090199 3-37 Revision 0
g l
t.
1C l
l-t i
i FigureM7 ANSYS Finite Element Model Representation of a Fuel Assembly l'
Mechanical, Materi al, and Structural Considerations August 1999 a:4878*>ndoc1bo90199 3-38 Revision 0 t
7.-.
a,c i
I i
l I
i l
t i
i l
Figure 3-18 ANSYS Finite Element Hydrodynamic Mass Model I
1
- Mechanical, Material and SWctural Considerations August 1999 a\\ manondoe:1M90199.
3-39 Revision 0
a,C l
l
)
i i
1 l
i Figure 3-19 ANSYS Finite Element Model Representation of the Whole Spent Fuel Pool Mechanical, Material, and Structural Considerations August 1999
&\\4sanon.docib 090199 -
3-40 Revision 0 L1
T-~
4.0 SAFETY EVALUATION 4.1 DEGREE OF SUBCRITICALITY j
i l
As demonstrated in References 7.10.a and.b the design of the racks is such that K,, remains less j
than or equal to 0.95 under all conditions, including fuel-handling accidents. The close spacing i
of the racks precludes insertion of fuel assemblies in other than design storage locations, except in an area south of the racks whem a fuel assembly may be inserted between the pool wall and the racks. Such inadvertent insertion of a fuel assembly into this location, or the placement of a fuel assembly across the top of a fuel rack, is considered a postulated accident, and as such, realistic initial conditions such as boron in the water can be taken into account. This condition has an acceptable K,, of less than or equal to 0.95.
4.2 GOVERNING CODES FOR DESIGN Design of the racks in accordance with applicable USNRC Regulatory Guides, Standard Review Plans and the OT Position for Review and Acceptance of Spent Fuel Handling Applications, ensures adequate safety under normal and postulated accident conditions. These are listed in Section 7.0 and referenced in the text.
4.3 ABILITY TO WITHSTAND EXTERNAL LOADS AND FORCES The racks are Nuclear Safety Class 3 and Seismic Class I structures and are designed to withstand all design basis loads. The racks have been evaluated for loads and load combinations specified in USNRC SRP 3.8.4 (Reference 7.7.b) and are adequate for all postulated loading. Additianally, the racks are designed with adequate energy absorption capability to withstand the impact of a dropped fuel assembly from the maximum lift height of the fuel-handling machine. Loads considered in the evaluation of the racks included dead
(
weight, seismic loads, and thermal loads. Seismic loading includes both the Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE). Development of the above loads considered the submerged weight of the storage racks. The heat generated by a fR core off load governed thermalload development. Accident thermalload conditions were defined as a full core off load combined with inadequate spent fuel pool cooling.
The pool structure (i.e., Auxiliary Building) also has been evaluated for the loads and load combinations to verify the ability of the fuel pool to withstand the loading of the new fuel storage racks. This analysis also verified that the existing seismic analysis of the Auxiliary Building would not be affected by the addition of the new storage racks.
Interaction of the free standing fuel storage rack structure and the spent fuel pool structure were accounted for by the application of a dynamic load factor to the submerged weight of the racks and fuel assemblies.
Safety Evaluation August 1999 o:\\4578-non.docitr090199 4-1 Revision 0
-l Detailed information on the evaluation of the new racks is contained in this report. Detailed information on the pool structure's capacity to resist the imposed loads is documented in GAI Report 1949 (Reference 7.11.n) and FPC design calculation S99-0166 (Reference 7.11.p).
4.4.
ABILITY TO ENSURE CONTINUOUS COOLING One of the major features for ensuring continuous cooling to the spent fuel pools is the provision of two independent and redundant mechanical trains of spent fuel cooling which can be supplied by on-site emergency power. The partial or complete loss of spent fuel cooling is highly unlikely for the following reasons:
1.
Primary flow-path piping and components in the spent fuel cooling loops are Seismic Class I.
2.
The spent fuel cooling system contains redundant pumps and heat exchangers.
3.
Electrical components are supplied from independent power sources, which have the capability of being powered by the associated emergency diesel generator.
4.
The systems that provide the ultimate heat sink for the spent fuel cooling heat exchangers are also Seismic Class I systems and are provided with redundant pumps and heat exchangers.
In addition to the Spent Fuel Cooling System, there are four supplemental means of providing inventory and cooling to the spent fuel pools:
1.
The Decay Heat Removal System is designed so that it can be aligned to cool the spent fuel pools.
2.
The forced ventilation system above the pools will enhance the cooling effects of pool surface evaporation.
3.
The Borated Water Storage Tank (BWST) water can be used for pool water makeup as well as for its cooling effect.
4.
The fire service system can be used to provide make-up water. Necessary equipment is pre-staged to support this capability.
Delay in the transfer of spent fuel assemblies into the fuel pools significantly impacts heat loads. The decay heat analysis assumes that the pools are not separated from the decay heat system until 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> after reactor shutdown when a full core is discharged. No credit is taken for additional decay or other conservatisms in this bounding analysis. Specific analysis based on a specific set of conditions may be able to support a shorter interval.
Safety Evaluation August 1999 OA4878-non. doc:lb490199 4-2 Revision 0 L
4.5 PROVISIONS TO AVOID DROPPING OF HEAVY LOADS t
The NRC approved Phase I of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", for Florida Power Corporation by letter dated July 13,1984. The objectiveidentified in
{
Section 5.1 of NUREG-0612 for providing " maximum practical defense in depth" is satisfied by PhaseI compliance.
FPC is committed to prohibit loads greater than 2750 pounds (the nominal weight of a fuel assembly and handling tool) from being transported over spent fuel in the Spent Fuel Pool unless the missile shields are in place (Reference 7.11.o). During the removal of the old racks and placement of the new racks in Pool B administrative controls will be implemented to pmvent the racks from being moved over fuel assemblies unless the missile shields are in place.
The Auxiliary Building overhead crane was reviewed and a determination made that sufficient capacity exists to reduce the likelihood of a load handling accident. No additional restrictions on load handling operations in the Spent Fuel Pool area beyond those already in-place through existing plant procedures are necessary.
4.6 MATERIAL COMPATIBILITY As discusred in Section 3.6.2 all materials used in construction of the racks are compatible with the storage pool environment. All rack surfaces that come into contact with the fuel assemblies are made of annealed austenitic stainless steel. The materials are corrosion resistant and will not contaminate the fuel assemblies or the pool environment.
4.7 RADIOLOGICAL CONSIDERATIONS Tim inventory of spent fuel in the new high-density racks will be increased. FPC has determined, as discussed in Section 6.3.2, that the installation and use of the racks will not change the radiological consequences of normal operations in the Spent Fuel Pool area.
l
.~
. ABILITY OF RACKS TO WITHSTAND ACCIDENTAL LIFT FORCES 4.8 Analysis of the fuel racks demonstrates that the racks can withstand a maximum uplift load frorn the fuel-handling machine without violating the criticality acceptance criterion. The
. resulting stresses from this load are within acceptable limits.
4.9 POTENTIAI FUEL AND FUEL RACK HANDLING ACCIDENT The Fuel Handling Accident outside the Reactor Building is postulated as the dropping of a fuel
. assembly into the spent fuel storage pool that results in damage to a fuel assembly and the release of the volatile gaseous fission products. The Fuel Handling Accident analysis was updated in 1978 to reflect damage to 208 fuel pins and the results incorporated into the FSAR.
As documented therein, the results of the analysis demonstrate that the applicable 10CFR100.11 dose acceptance criteria are satisfied. The radiological consequences of this accident are based Salety Evaluations.
August 1999 oms 7 sam.docibOM99 4-3 Revision 0
solely on the failure of the dropped assembly. The current analysis assumes that all 208 fuel pins in the dropped assembly are damaged. Thus, the installation of the high-density racks in the spent fuel storage pool has no effect on the dose analysis or related conclusions.
4.10 CONCLUSION On the basis of the design requirements presented in this report, CR-3 and industry operating experience with high density fuel storage, and other material provided or referenced in this i
report,it is concluded that the proposed Pool B rack replacement and alternate new fuel loading configuration in Pool A will continue to provide safe spent fuel storage. Further, it is consistent with the facility design and operating criteria as provided in the Crystal River Updated FSAR and Operating License as well as generic regulatory guidance.
1 i
1 l
i Safety Evaluation August 1999 o:\\4878-non.docib-090199 4-4 Revision 0 m
y 5.0 COST / BENEFIT ASSESSMENT The cost / benefit of the proposed rack replacement is presented in the following sections.
5.1 NEED FOR INCREASED STORAGE CAPACITY Increased spent fuel storage capacity at Crystal River Unit 3 is needed to ensure adequate storage through the expiration of the current operating license in the year 2016. Based on the present storage configuration, a full core reserve will not be available after 2013 (end of cycle 18). FPC has completed a detailed evaluation of alternative spent fuel storage technologies and techniques currently available, or under development. Considerations of design, implementation, licensing, plant modifications and life cycle cost were identified. This evaluation resulted in the development of a Long Term Strategic Plan for on-site Spent Fuel Storage.
The FPC strategic plan includes three major phases, which are proposed to be implemented over the next ten years. This plan maintains full core reserve, and provides a workable solution to be prepared to ship fuel to the Departmcnt of Energy, when our allocation permits shipment.
The current proposal is summarized below.
Phase 1 (1998-2001) - This phase is the replacement of the racks in Pool B with higher-capacity racks using Boral as the neutron poison. This resolves the existing issues of Boraflex degradation and consequential silica management, as well as providing the best possible long term economic position for FPC in a competitive environment. This will increase the available cell locations, enabling FPC to maintain full core reserve through the end of +he current operating license. New fuel assemblies will be stored in Pool A using a checkerboard loading pattern.
Phase 2 (estimated 2001-2003) - This includes the implementation of plant modifications t
which will allow the handling of Multi-Purpose Canisters (MPCs) and upgrading the fuel building crane to single failure proof with a 125-ton capacity.
Phase 3 (estimated 2003-2009) - This includes design, construction and licensing of an on-site dry storage and shipping facility, which incorporates the use of MPCs. This facility will provide an attemative plan to accommodate on-site storage to maintain full j
core off load capability through the current operating license, while providing a method to ship fuel off-site to DOE or another facility. This facility has the design flexibility for j
future expansion, should DOE not accept fuel prior to the plant shutdown.
This phased approach optimizes FPC ability to meet spent fuel storage and shipping needs through the existing plant life cycle. Additionally,it provides flexibility to implement supplemental spent fuel storage,if plant life extension is pursued. The schedule achieves the ultimate objective, of being able to ship fuel off-site to the Department of Energy (DOE), or an interin facility (estimated in the 2007/2008 time frame), while minimizing the total spent fuel l
life cycle cost.
Cost Benefit / Assessment August 1999 oA4878 con.docitro90199 5-1 Revision 0
The following factors were considered in the development of this phased approach:
Florida Power Corporation (FFC) currently has no contractual arrangements with a.
any fuel reprocessing facility or interim off-site, away from reactor dry storage facility.
b.
Crystal River 3 has two adjacent spent fuel pools on site, Pool A and Pool B. The racks in both pools have previously been replaced. The total number of cell locations currently available in both pools is 1357. Of this number,542 cell locations are located in the Pool A, and 815 cell locations are located in Pool B.
Because of physical interferences in some areas some cell locations are not accessible with the fuel handling bridge. These other locations are accessible by utilizing a manual fuel-handling tool. Additionally, there are six cell locations derignated for failed fuel assemblies.
The projected refueling schedule for future fuel cycles at CR-3 is shown in c.
Table 5-1. The table shows the projected approximate dates of refueling as well as projec+ed number of fuel assemblies that will be transferred into the spent fuel pools, until the total existing capacity is reached. All calculations in the table for loss of full core reserve are based on the number of usable cells, not the total number of cells in the pools.
d.
The Crystal River 3 spent fuel pools currently contain a total of 680 spent fuel assemblies.
Currently, the storage of components other than fuel has not significantly affected e.
the total number of available storage locations in each pool. These components are often inserted within the fuel assemblies. An itemized list of components stored in each pool is given in Table 5-2. Fifteen devices are stored in existing rack locations. Of these, fourteen items could be removed, compacted and disposed of.
The 58 (fifty-eight)-rod container contains a few damaged pins, and therefore must remain in the pool.
f.
Adoption of this proposed spent fuel storage expansion will extend the time period that spent fuel assemblies could be stored on site. Until interim off-site storage facilities are available, spent fuel assemblies at CR-3 must remain on site.
g.
Table 5-3 presents the spent fuel storage capacity after each cycle following this rack replacement project. Based on the current FPC fuel management reload policy, full core reserve will be maintained through the end of the operating license in 2016 (after Operating Cycle 19).
Cost Benefit / Assessment August 1999 oM878-non.docit*090199 5-2 Revision 0
5.2 ESTIMATED COSTS The costs associated with the proposed Spent Fuel Pool B rack replacement are estimated to be approximately $5.5 million dollars. This figure includes the cost of:
1.
engineering studies of spent fuel disposal alternatives, 2.
design, fabrication, and installation of new spent fuel storage racks, and
' 3.
removal and offsite disposal of the existing spent fuel storage racks as low level radioactive waste.
Note: The estimated values for uncertainty factors, cost escalation, and allowance for funds used during construction (AFUDC) are not included in this estimate.
5.3 CONSIDERATION OF ALTERNATIVES 1.
Reprocessing. No commercial reprocessing facilities able to meet FPC's needs are currently operational. No such facilities are expected to be operational in the foreseeable future.
i
' 2.-
Storage (in an Independent Spent Fuel Storage Installation (ISFSI)). As of March 1999, the two proposed away from-the-reactor-site private fuel storage facilities have not been
- licensed by the NRC, nor has construction started. There are no firm commitments by either commercial firms or government agencies to construct or operate an interim,-
IS3I facility. Furthermore, cost and schedule considerations make an on-site ISFSI ixility a long-term solution that will not meet the current needs of Crystal River Unit 3 for near term spent fuel storage needs.
3.
Transshipment (shipping spent fuel to other facilities within the FPC system). Crystal River Unit 3 is the only nuclear unit of Florida Power Corporaticn. Therefore, transshipping spent fuel to other facilities within the FPC system is not an available option.
4.
Decommissioning. The current annual growth rate for energy demand in the state of Florida is estimated at 1.7%. Generation from Crystal River 3 is essential to meet curmnt demands. Therefore, additional capacity would be required to replace Crystal River 3 if it were retired early. The replacement energy costs for the premature shutdown of Crystal River Unit 3 for a two-year period would have a net present value in excess of
. $250,000,000 (1999 dollars). Additionally, a substantial capital investment would be requimd for new capacity (roughly $400,000,000 or $500/kW based on FPC's most curmnt experience). Themfom, permanent shutdown of CR-3 would place a heavy financial burden on Florida residents within FPC's service area.
Cost Benefit / Assessment August 1999 o:\\4878-nortdocit>090199 -
5-3 Revision 0
5.4 RESOURCES COMMITTED Replacement of the racks in the spent fuel pool will not result in any irreversible and irretrievable commitments of water, land or air resources. Increasing the density of on-site fuel storage capacity results in more efficient use of the land area now used for spent fuel storage.
The materials used for new rack fabrication are discussed in Section 3.6.1. These materials are not expected to foreclose alternatives available with respect to any other licensing actions intended to relieve the possible shortage of spent fuel storage capacity.
1 Cost Beneta/Aawar.:.;
August 1999 i
c:\\4878-non doc 1b-090199 5-4 Revision 0 1
Table 5-1 Spent Fuel Pool Capacity Without Reracking B Pool Refueling No. Assyin Pool Capacity Date After Refuel Cycle No.
Reload Size After Refuel October 1999 752 11 72 605 l
October 2001 820 12 68 537 October 2003 888 13 68 469 October 2005 956 14 68 401 October 2007 1024 15 68 333 October 2009 1092 16 68 265 October 2011 1160 17 68 197 October 2013 1228 18 68 129' October 2015 12 %
19 68 January 2017 1473 20 l
- Full Core Reserve Lost Cost Benefit / Assessment August 1999 c:\\4878-non. doc 1b-090199 5-5 Revision 0
Table 5-2 Components Stored In Spent Fuel Pool Component Number Orifice Assemblies 2
Burnable Poison Rod Assemblies 163 Gamma Sample Holder 1
ControlRod Assembly 61
" Dummy" Test Fuel Assembly 1
Control Component Storage Assemblies 3
Primary Sources 2
Axial Power Shaping Rods 8
)
58-rod Container 1
Old cages from the re-caged fuel assemblies 5
Re-cage trash container 1
In-core trash container 2
1 l
Cost Benefit / Assessment August 1999 o:\\4878-non.dx.lb-090199 5-6 Revision 0 i
j
i 1
Table 5-3 Spent Fuel Pool Capacity After Reracking B Pool Refueling No. Assy in Pool Capacity Date After Refuel Cycle No.
Reload Size After Refuel October 1999 752 11 72 722 October 2001 820 12 68 654 October 2003 888 13 68 586 October 2005 956 14 68 518 October 2007 1024 15 68 450 October 2009 1092 16 68 382 October 2011 1100 17 68 314 October 2013 1228 18 68 246 October 2015 12 %
19 68 178 January 2017 1473 20 177 1
J Cost Benefit / Assessment August 1999 a:\\4s78-non.docibo901w 5-7 Revision 0
6.0 RADIOLOGICAL EVALUATION i
6.1 SOLID RADWASTE No significant increase in volume of solid radioactive wastes generated annually is expected as a result of the increased spent fuel pool capacity. It is estimated that an additional 30 cubic feet of solid materials (resins, filters, debris) will be generated by the spent fuel pool cleanup system, pool vacuuming, and equipment removal during the actual rack replacement process.
- The existing racks will be further processed for disposal as described under section 6.4.
6.2 GASEOUS RADWASTE 1
There has been no measured Krypton-85 release from the fuel building ventilation system for the last two years. The Semi-Annual Radioactive Effluent Release Report includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from Crystal River Unit 3 as outlined in Regulatory Guide 1.21 (Revision 1,1974). No significant increase in gaseous effluents is expected as a result of increased spent fuel storage.
6.3 PERSONNEL EXPOSURE 6.3.1 Personnel Exposure During Rack Replacement And Disposal This section discusses personal exposure control measums and the exposure to personnel expected from the emptying, removal, decontamination, and disposal of the existing racks, and the installation of the new racks. Personnel exposure from the rack replacement activity is expected to be no greater than 15 man-rem.
The radiation protection aspects of the spent fuel pool modification are the responsibility of the Plant Radiation Protection Manager. Gamma radiation levels in the pool area are constantly j
monitored by the CR-3 Area Radiation Monitoring System, with high level alarm feature.
Personnel working in radiological controlled areas wear protective clothing and respiratory j
protective equipment, depending on work conditions, as required by the applicable Radiation i
Work Permit. Personal dosimetry is assigned to and worn by all personnel in any j
radiologically controlled work area. At a minimum, this dosimetry consists of a thermoluminescent dosimeter (TLD) and self-reading, dose-rate alarming dosimeter. Additional personal monitoring equipment, such as extremity badges, is utilized as required.
j 1
Contamination control measures are used to protect persons from internal exposures to radioactive material, and to prevent the spread of contamination. Work, personnel traffic, and the movement of material and equipment in and out of the radiologically controlled area,is controlled to minimize contamination problems. The station radiation protection staff closely monitors and controls all aspects of the work, so that personnel exposures, both internal and external, will be maintained as low as reasenably achievable (ALARA).
Radiological Evaluation
- August 1999 c:\\es7s-non. doc 2-090199 61 Revision 0
6.3.2 Personnel Exposure From Normal Operation With New Racks The following discussion addmsses expected increases in the doses to personnel from radionuclide concentrations in the spent fuel pool due to the proposed expansion of the spent fuel storage capacity.
The values of radionuclide concentrations from recent gamma isotopic analyses of a.
spent fuel pool water are shown in Table 6-1.
- b.
.CR-3 operating experience shows dose rates of 0.5 to 2.0 mrem / hour either at the edge or above the center of the spent fuel pools regardless of the quantity of fuel stomd. This is not expected to change with the proposed rack mplacement because radiation levels above the pool am due primarily to radioactivity in the water, which experience shows to retum to a level of equilibrium. Stored spent fuel is so well shielded by the water above the fuel, that dose rates at the top of the pool from this source are minimal.
There are no routine concentrations of airbome radioactivity from the spent fuel c.
pools. The spent fuel pool ventilation system provides a continuous purge of air across the top of the spent fuel pools and cask-loading pit. The system includes a continuous row of supply diffusers along the southside of the pools, and a continuous now of exhaust outlets along the northside of the pools. Additionally, a continuous exhaust flow is maintained from the enclosed top portion of the pools when the missile shields are in place. All exhaust flow is directed to the main Auxiliary Building filter system, where it passes through roughing, HEPA, and charcoal filters before being discharged through the plant vent. The proposed rack replacement is not expected to incmase this activity.
d.
CR-3 operating plant experience with high-density fuel storage has shown no noticeable increases in airbome radioactivity above the spent fuel pool, or at the site boundary.
As stated in Sections 6.1 and 6.2, and based on CR-3 operating experience with e.
high-density fuel storage racks, no significant increases in the radwaste generated by the spent fuel pool cleanup system are expected. Further, no significant i
increase in the radioactivity levels in the spent fuel pool water, nor any significant increase in the annual man-rem exposure due to the increased fuel storage, including the changing of spent fuel pool cooling system resins and filters are expected. The spent fuel pool filters and demineralizer are located in a shielded I
cubicle in an area that is not madily accessible, and thus, will not present any radiation hazard if they become contaminated.
f.
There has been no buildup of radioactive material, e.g., Cobalt-58 or -60, along the sides of the pool. The highest possible water level is maintained in the spent fuel pool to keep exposure as low as reasonably achievable. Such buildup could be Radiological Evaluation August 1999 a\\ 487Beon. doc 2-OlOl99 '
6-2 Revision 0
easily washed down or hydrolyzed from the spent fuel pool walls around the pool edge if present.
g.
During normal operation, the radiation zone designation of areas around the sides of the pools will not change due to rack replacement. Two considerations to reduce radiation exposure through adequate shielding were included in the analysis of the spent fuel pool. These were the concrete shielding surrounding the pool and the minimum water depth above the fuel assemblies. The shielding is adequate to maintain the radiation level below tolerance dose levels for normal contamination of the pool water by particulates. Similarly, gaseous activity coming out of solution from the pool water is picked up by the " push-pull" type ventilation system over the pool.
CR-3 operating experience has shown no increase in normal personnel exposure due to the increased fuel storage with high-density racks. Therefore, aside from the one-time exposure of less then 15 man-rems expected from the rack replacement, no increase in the annual man-rem from normal operation is expected at Crystal River Unit 3 as a result of the increased storage capacity of the spent fuel pools with the higher density racks.
The existing Crystal River health physics program did not have to be modified as a result of the previous increase in storage of spent fuel. It is not anticipated that the health physics program will require any modification for the proposed increase in fuel storage capability.
6.4 RACK DECONTAMINATION AND DISPOSAL There are eight spent fuel storage rack modules that will be removed from Spent Fuel Pool B.
The total weight of these eight racks is approximately 94,600 pounds.
. The racks will be decontaminated at Crystal River 3 to the extent necessary to permit transport of the racks to a waste-processing vendor. Following decontamination, the racks will be placed into approved licensed containers and transported to a disposal facility for volume reduction and disposal by an independent waste processing vendor experienced in this activity. The racks will be entombed at a NRC licensed facility. The applicable NRC requirements will be adhered to for transport, receipt, storage, and processing of radioactive contaminated equipment and material.
1 Radiokgical Evaluation August 1999 o:\\4878 mon.d oc1b-090199 6-3 Revision 0
l
{
Table 6-1 Gamma Isotopic Analysis of Spent Fuel Pool Water ISOTOPES CONCENTRATION uci/cc Cobalt - 58 1.30 x 10" Cobalt - 60 3.00 x 10' Cesium - 134 5.30 x 10' 4
Cesium-137 5.30 x 10 Antimony - 125 1.20 x 10*
TotalGamma Activity 7.34 x 10*
l l'
)
Radiological Evaluation August 1999 oms 78-non.doctb-090199 6-4 Revision 0
m 7.0 CODES, STANDARDS, SPECIFICATIONS AND OTHER REFERENCES 7.1 AMERICAN NATIONAL STANDARDS INSTITUTE (ANSI /ASME NQA-2-1983):
N45.2.1-1980," Cleaning of Fluid Systems and Associated Components for Nuclear a.
Power Plants."
b.
N45.2.2-1978, " Packaging, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power Plants."
i 7.2 American National Standards Institute /American Society of Mechanical Engineers (ANSI /ASME):N45.2-1971 (ASME NQ-1-1994), " Quality Assurance Program Requirements for Nuclear Facilities."
l 7.3 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, 1989 Edition (No Addenda):
a.
Section II, " Material, Specifications."
b.
Section III, Division 1, " Nuclear Power Plant Components," Subsection NF,
" Component Supports," Class 3,1989, (No Addenda).
c.
Section V, " Nondestructive Exarnination."
d.
Section IX, " Welding Qualifications."
7.4 OTHER CODES AND STANDARDS American Concrete Institute Code Requirements for Nuclear Safety Related a.
Concrete Structures ACI 349-85,1990 Supplement.
b.
American Society of Testing Materials (ASTM) C-750-89, Specification for Nuclear-Grade Boron Carbide Powder Type III.
c.
American Society of Nondestructive Testing (ASNT) SNT-TC-1 A (1984),
Recommended Practices.
d.
ANSI /ANS 5.1-1994, " Decay Heat Power in Light Water Reactors."
7.5 CODE OF FEDERAL REGULATIONS a.
10CFR50 Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."
Codes, Standants, Specifications and Other References August 1999 0:\\4878-non. doc 1b- 090199 7-1 Revision 0 L-
n
)
b.
10CFR21," Reporting of Defects and Noncompliance."
7.6 NRC REGULATORY GUIDES 1.29, " Seismic Design Classification," Rev. 3, September 1978.
a.
b.
1.31, " Control of Ferrite Content in Stainless Steel Welding Metal," Rev. 3, April 1978.
1 1.38, " Quality Assurance Requirements for Packaging, Shipping, Receiving, c.
Storage, and Handling of Items for Water-Cooled Nuclear Power Plants," Rev. 2, May 1977.
d.
1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Rev. 3, November 1978, Section 9.1.2, " Spent, Fuel Storage."
1.92, " Combining Modal Response and Spatial Components in Seismic Response
{
e.
Analysis," Rev.1, February 1976.
]
7.7 NRC STANDARD REVIEW PLAN (NUREG-0800) 1 Section 3.7.1 " Seismic Design Parameters," Rev. 2, August 1989.
a.
b.
Section 3.8.4, "Other Seismic Category I Structures," Rev.1, July 1981.
c.
Section 3.8.5 II-5, " Foundations," Rev. 2, August 1989.
d.
ection 9.1.2, " Spent Fuel Storage," Rev. 3, July 1981.
j l
7.8 NRC BRANCH TECHNICAL POSITIONS APCSB 9.2, " Residual Decay Energy for Light Water Reactors for Long Term a.
Cooling."
i b.
RDT Standard F6-6T," Welding of Structural Components," latest revision.
7.9 OTHER NRC GUIDANCE DOCUMENTS US. Nuclear Regulatory Commission (USNRC), NUREG-0612, " Control of Heavy a.
Loads at Nuclear Power Plants," July 1980.
b.
US. Nudear Regulatory Commission (USNRC) Letter, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978, with revisions dated January 18,1979.
Codes, Standards, Specifications and Other References August 1999 c:\\ 487&non. doe:n>co199 7-2 Revision 0 t
7.10 CRITICALITY ANALYSES Holtec Report HI-992285, " Criticality Safety Evaluation of the Crystal River Unit 3 a.
Pool A for Storage of 5% Enriched Mark B11 Fuelin a Checkerboard Arrangement with Water Holes," August,1999.
b.
Holtec Report HI-992128, " Criticality Safety Analysis of the Westinghouse Spent Fuel Storage Racks in Pool B of Crystal River Unit 3," April,1999 7.11 REFERENCES ANSYS Version 5.5, Swanson Analysis Systems IP Inc., August 1998.
a.
b.
Crystal River Unit 3 Environmental and Seismic Qualification Program Manual, Florida Power Corporation, Rev.10, October 1998.
" HIST 1-A Program for Development of Artificial Time History Compatible with c.
Smoothed Design Ground Response Spectra," File No.91-906, Paul C. Rizzo Associates,1991.
d.
"INTEG-A Program to Integrate an Acceleration Time History with the Newmark-Beta Method," File No.91-906, Paul C. Rizzo Associates,1991.
i
" QUAKE-A Program to Compute HUSID Times, Power Spectral Density and e.
Ferform Interpolation of a Given Time History," File No. 96-9006, Paul C. Rizzo Associates,1996.
f.
"3-Mass-A Program that Calculates Dynamic Response for a Three-Degree of Freedom System by Modal Superposition," File No. 98-9006, Paul C. Rizzo Associates,1998.
g.
Quality Assurance Manual, Paul C. Rizzo Associates,1996.
h.
Letter FPM-98-548, Fuel Assembly Data for Spent Fuel Fool Rack Replacement, Rev. 2, Framatome Cogema Fuels, November 12,1998.
i.
"ANSYS Acoustics and Fluid-Structure Interaction, ANSYS Version 5.0 Tutorial,"
Swanson Analysis Systems IP Inc., June 1992.
j.
Fritz, R, "The Effect of ' Liquids on the Dynamic Motions of Immersed Solids,"
ASME Journal of Engineering For Industry, Volume 94, pp 167-173,1972.
k.
" Friction Coefficients of Water-Lubricated Stainless Steels for a Spent Fuel Rack l
Facility," E. Rabinowicz, Massachusetts Institute of Technology, a report for Boston Edison Company,1976.
i l
Codes, Standards, Specifications and Other References August 1999 c:T4873-non.d ocbO90199 7-3 Revision 0 L
1.
FPC Calculation M97-0014, R1 " Spent Fuel Pool Temperature Rise From Fuel in the Pool" FPC Calculation F97-0014, R1 " Spent Fuel Pool Heat Load and Heat Exchanger m.
Performance at Reduced SW Flow" GAI 1949 " Structural Evaluation of High Capacity Racks", November 1994.
n.
FSAR Section 9.6.2.7, Fuel Handling System - Operational Requirement o.
p.
FPC Calculation S99-0166, R0 " Analysis of Spent Fuel Pool Structure for New Racks".
\\
Codes, Standards, Specifications and Other References August 1999 a%4sanoitdocit>-090199 7-4 Revision 0
I I
- U. S. Nuclear Regulatory Commission Attachment I 3F0999-07 Page 1 of1 LIST OF COMMITMENTS The following table identifies those actions committed to by the Florida Power Corporation in License Amendment Request (LAR) #239. Any other actions discussed in the submittal represent intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing of any questions regarding this document or any associated regulatory commitments.
COMMITMENT DUE DATE Implement administrative controls to maintain both the current and proposed Improved Technical Specification (ITS) and ITS Bases related to spent fuel storage in the controlled copies of CR-3 ITS during the rack replacement process. Remove the current ITS and associated ITS Bases after existing racks have been replaced with the new racks.
[ Attachment A, page 6 of 15, Interim Rack Conficurations and Implementation of Revised ITS]
January 15,2001
)
FPC-qualified operators will make the fuel movements required for the rack replacement project.
[ Attachment A, page 7 of 15, Fuel and Rack Handline, and Attachment D, Section 1.4.1]
January 15,2001 All load hanilling operations for the Spent Fuel Pool B rack replacement project will be conducted in accordance with FPC procedures and practices -
that implement the criteria of NUREG-0612.
[ Attachment A, page 7 of 15, Fuel and Rack Handline, and Attachment D.
Section 1.4]
January 15,2001 After removing the existing racks the Spent Fuel Pool B will be inspected and' debris removed prior to installing any replacement racks.
[ Attachment D, Section 1.4.2.2]
February 1,2001 Following installation of the racks a drag test is performed using a fuel assembly mockup to verify that a fuel assembly can be inserted into the fuel locations.
[ Attachment D, Section 1.4.2.2]
June 1,2001 I
k i
i
FLORIDA POWE.R CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT D 1
ENHANCED SPENT FUEL STORAGE PROJECT ENGINEERING INPUT (PROPRIETARY)
I LICENSE AMENDMENT REQUEST No. 239
)
I