ML14339A621
| ML14339A621 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 11/24/2014 |
| From: | Dominion Energy Kewaunee |
| To: | Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14339A626 | List: |
| References | |
| 14-572 | |
| Download: ML14339A621 (42) | |
Text
Table of Contents tion Title Page
.1 DESIGN BASES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-1 3.1.1 Performance Objectives. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-1 3.1.2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.1 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.3 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.4 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.5 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.6 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.7 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.2.8 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.3 Safety Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.3.1 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.3.2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.3.3 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2 3.1.3.4 Mechanical Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-2
.2 REACTOR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.1 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.1.1 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.1.2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2.1 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2.2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2.3 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2.4 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2.5 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2.6 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.2.7 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.3 Mechanical Design and Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.3.1 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-2 3.2.3.2 Core Components. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-2 3.2.3.3 Evaluation of Core Components . . . . . . . . . . . . . . . . . . 3.2-8 3.2.3.4 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-10 3.2 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-10
Table of Contents (continued) tion Title Page
.3 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.1 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.2 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.3 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3 References[DELETED] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1
List of Tables le Title Page 1 Nuclear Design Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-13 2 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-14 3 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-15 4 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-16 5 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-17 6 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-18 7 Core Mechanical Design Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-19 8 Fuel Assembly and Component Descriptions. . . . . . . . . . . . . . . . . . . . . . . 3.2-20
List of Figures ure Title Page 1 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-22 2 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-23 3 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-24 4 Typical Rod Cluster Control Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-25 5 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-26 6 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-27 7 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-28 8 Westinghouse 422V+ and Framatome Fuel Assembly Designs. . . . . . . . 3.2-29 9 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-30 10 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-31
The fuel rods are cold-worked partially annealed zirconium alloy tubes containing slightly ched uranium dioxide fuel. All fuel rods are pressurized with helium during fabrication.
The fuel assembly is a canless type with the basic assembly consisting of the Rod Cluster trol Assembly (RCCA) guide tube thimbles fastened to grids and top and bottom nozzles. The rods are supported and the spacing between fuel rods maintained at several points along their th by spacer grids. Dominion Energy Kewaunee, Inc. (DEK) is no longer authorized under its rating license to place or hold nuclear fuel in the reactor vessel.
The RCCA absorber sections are fabricated of silver-indium-cadmium alloy sealed in nless steel tubes. The fuel assembly design accommodates the use of additional fixed burnable tron poison materials including discrete burnable poison rods located in the RCCA guide bles and the fuel rod design accommodates the use of integral neutron absorber materials h as Gadolonia (Gd2O3) or ZrB2.
The RCCA drive mechanisms are of the magnetic latch type and are controlled by three netic coils. They are so designed that, upon loss of power to the coils, the RCCA is released falls by gravity to shut down the reactor.
3.1 DESIGN BASES 1 Performance Objectives Each reload cycle core was designed to yield an approximate 18-month fuel cycle at the d power level of 1772 MWth. A reload safety evaluation report was prepared for each reload le core design to summarize the analyses performed to demonstrate the design would not ersely affect the safety of the plant. In addition, per Technical Specifications, a Core Operating its Report (COLR) was prepared. Core operating limits presented in the COLR were ermined using analytical methods reviewed and approved by the Nuclear Regulatory mmission (NRC). Operation of the core within the COLR limits ensured that all applicable ts of the safety analyses were met.
The RCCAs provided sufficient control rod worth to shut the reactor down by the shutdown tivity required by Technical Specifications in the hot condition at any time during the fuel le with the most reactive RCCA stuck in the fully withdrawn position.
2.1 Deleted 2.2 Deleted 2.3 Deleted 2.4 Deleted 2.5 Deleted 2.6 Deleted 2.7 Deleted 2.8 Deleted 3 Safety Limits 3.1 Deleted 3.2 Deleted 3.3 Deleted 3.4 Mechanical Limits 3.4.1 Deleted 3.4.2 Fuel Assemblies The fuel assemblies were designed to perform satisfactorily throughout their lifetime. The ds, stresses, and strains resulting from the combined effects of flow-induced vibrations, hquakes, reactor pressure, fission-gas pressure, fuel growth, thermal strain, and differential ansion during both steady-state and transient reactor operating conditions were considered in design of the fuel rods and fuel assembly. The assembly is structurally designed to withstand dling and shipping loads prior to irradiation, and to maintain sufficient integrity at the pletion of design burnup to permit safe removal from the core and subsequent handling ng cooldown, shipment and fuel reprocessing or storage.
The fuel rods are supported at several locations along their length within the fuel assemblies rid assemblies, which are designed to maintain control of the lateral spacing between the rods ughout the design life of the assemblies. The magnitude of the support loads provided by the s was established to minimize possible fretting without overstressing the cladding at the nts of contact between the grids and fuel rods. The grid assemblies also allow axial thermal ansion of the fuel rods without imposing restraint of sufficient magnitude to result in buckling istortion of the rods. The fuel rod cladding was designed to withstand operating pressure loads hout collapse or rupture and to maintain encapsulation of the fuel throughout the design life.
The criteria used for the design of the cladding on the individual absorber rods in the CAs are similar to those used for the fuel rod cladding. The stainless steel cladding was gned to be free standing under all operating conditions and will maintain encapsulation of the orber material throughout the absorber rod design life. Allowance for wear during operation included in the RCCA cladding thickness.
3.4.4 Deleted
Intentionally Blank 1 Deleted 1.1 Deleted 1.2 Deleted 1.2.1 Deleted 1.2.2 Deleted 1.2.3 Deleted 1.2.4 Deleted 1.2.5 Deleted 1.2.6 Deleted 1.2.7 Deleted 1.2.8 Deleted 1.2.9 Deleted 1.2.10 Deleted 1.2.11 Deleted 2 Deleted 2.1 Deleted 2.2 Deleted 2.3 Deleted 2.3.1 Deleted 2.3.2 Deleted 2.4 Deleted 2.5 Deleted 2.6 Deleted 2.7 Deleted 3 Mechanical Design and Evaluation
cted fuel rods may contain integral Gadolinia burnable poison or may have a thin zirconium de (ZrB2) coating. The pellets are stacked to the fuel active height within the tubular cladding, ch is plugged and seal-welded at the ends to encapsulate the fuel. All fuel rods are internally surized with helium during fabrication.
The control rods, designated as RCCAs, consist of groups of individual absorber rods, ch are held together by a spider at the top end and actuated as a group. In the inserted position, absorber rods fit within hollow guide thimbles in the fuel assemblies. The guide thimbles are ntegral part of the fuel assemblies and occupy locations within the regular fuel rod pattern re fuel rods have been deleted. Figure 3.2-4 shows a typical rod cluster control assembly.
A summary of nuclear design data is presented in Table 3.2-1.
A listing of the mechanical design parameters for the RCCAs is given in Table 3.2-7.
3.1 Deleted 3.1.1 Deleted 3.1.2 Deleted 3.1.3 Deleted 3.1.4 Deleted 3.1.5 Deleted 3.2 Core Components The final reload core design consisted of a full core of Westinghouse fuel assemblies of the V+ fuel design. Transition to the full use of the 422V+ fuel design began with Cycle 26 and first full core of 422V+ fuel assemblies was Cycle 28.
Selected mechanical design parameter values for FRA/ANP Heavy and Westinghouse V+ fuel designs are presented in Table 3.2-8.
The fuel rods are arranged in a square array with 14-rod locations per side and a nominal terline-to-centerline pitch of 0.556 inch between rods. Of the total possible 196-rod locations assembly, 16 are occupied by guide thimbles for the RCC rods or discrete burnable poison s and one by a tube for in-core instrumentation. The remaining 179 locations contain fuel rods.
ddition to fuel rods, a fuel assembly is composed of a top nozzle, a bottom nozzle, 7 spacer assemblies, 16 absorber rod guide thimbles, and 1 instrumentation thimble.
prise the basic structural fuel assembly skeleton. The top and bottom ends of the guide bles are fastened to the top and bottom nozzles, respectively. The grid assemblies, in turn, are ened to the guide thimbles at each location along the height of the fuel assembly at which ral support for the fuel rods is required. Within this skeletal frame-work the fuel rods are tained and supported and the rod-to-rod centerline spacing is maintained along the assembly.
Bottom Nozzle The bottom nozzle is a square pedestal structure, which controls the coolant flow ribution to the fuel assembly and functions as the bottom structural element of the fuel mbly. The nozzle, which is square in cross-section, is fabricated from 304 stainless steel parts sisting of a perforated plate, four angle legs, and four pads or feet. The legs form a plenum ce for the inlet coolant to the fuel assembly. The perforated plate serves as the bottom end port for the fuel rods. The bottom support surface for the fuel assembly is formed under the um space by the four pads, which are attached to the corner angles.
The guide thimbles, which carry axial loads imposed on the assembly, are fastened to the om nozzle plate. These loads as well as the weight of the assembly are distributed through the zle to the lower core support plate. Indexing and positioning of the fuel assembly in the core is trolled through two holes in diagonally opposite pads, which mate with locating pins in the er core plate. Lateral loads imposed on the fuel assembly are also transferred to the core port structures through the locating-pins.
Top Nozzle The top nozzle is a box-like structure, which functions as the fuel assembly upper structural ment and forms a plenum space where the heated reactor coolant was directed toward the flow es in the upper core plate. The top nozzle is made of stainless steel with four Inconel d-down springs attached.
The top nozzle adaptor plate is provided with perforations to provide for coolant flow and vides a means of evenly distributing any axial loads imposed on the fuel assemblies among the de thimbles.
The springs are fastened in pairs to the top nozzle at the two corners where alignment holes not used and radiate out from the corners parallel to the sides of the nozzle. Fastening of each of springs is accomplished with a clamp, which fits over the ends of the springs and bolts ch pass through the clamp and spring, and thread into the top nozzle. At assembly, the spring unting bolts are torqued sufficiently to pre-load against the maximum spring load and then wired to the clamp, which is counter-bored to receive the bolt head.
The top nozzle is also designed to accommodate other types of insert components that are ently not used in KPS core design including thimble plugging devices, source assemblies, and rete burnable poison rod assemblies.
Guide Thimbles The guide thimbles in the fuel assembly provide guided channels for the absorber rods ing insertion and withdrawal of the RCCAs. They are fabricated from a single piece of aloy-4 or Zirlo tubing, which is drawn to two different diameters. The larger inside diameter he top provides a relatively large annular area for rapid insertion during a reactor trip and to ommodate a small amount of upward cooling flow during normal operation. The bottom ion of the guide thimble is of reduced diameter to produce a dashpot action when the absorber s near the end of travel in the guide thimbles during a reactor trip. The transition zone at the hpot section is conical in shape so that there are no rapid changes in diameter in the tube.
Flow holes are provided just above the transition of the two diameters to permit the entrance ooling water during normal operation, and to accommodate the outflow of water from the hpot during reactor trip.
The dashpot is closed at the bottom by means of a welded end plug. The end plug is ened to the bottom nozzle during fuel assembly fabrication. Flow holes are provided in the plugs to permit entrance of cooling water during normal operation and to regulate dashpot on during control rod trip.
Grids The spacer grid assemblies consist of individual slotted straps, which are assembled and rlocked in an egg-crate type arrangement. The straps are permanently joined at their points ntersection.
Two types of grid assemblies are used in the Westinghouse 422V+ fuel assemblies. Inconel grids are utilized for their corrosion resistance and high strength properties to hold the fuel s tightly. The inner straps on the Inconel grids do not contain mixing vanes. Mid-grids are ctural grids with mixing vanes on the upper edges of the inner grid straps. These grids are used he high heat region of the fuel assemblies to promote mixing of the coolant. ZIRLO' is zed as the material for the mid-grids primarily for its low-neutron absorption cross section.
outside straps on all grids on the 422V+ fuel assemblies contain mixing vanes on their upper es and small tabs projecting downwards from the lower edge, to aid in guiding the grids and assemblies past projecting surfaces during handling or core loading and unloading. Each grid in both grid types includes two springs and four dimples that serve as contact points with the
erence 14.
Fuel Rods The fuel rods consist of uranium dioxide ceramic pellets in slightly cold-worked and ially annealed Zircaloy-4 or ZIRLO' tubing. The tubing is plugged and seal-welded at the s to encapsulate the fuel. Sufficient void volume and clearances are provided within the rod to ommodate fission gases released from the fuel, differential thermal expansion between the ding and the fuel, and fuel swelling due to accumulated fission products without overstressing he cladding or seal welds. Shifting of the fuel within the cladding is prevented during handling hipping prior to core loading by a stainless steel helical compression spring.
All fuel rods are internally pressurized with helium during fabrication. The fuel rod void ce is sized to ensure adherence to the pressure criteria. The rod internal pressure is evaluated the limiting fuel rod, assuming a conservative operating history. The evaluation is based on ected operating conditions at the peak steady-state power, and also considers the fission gas ase from normal operating transients. The model used to predict the quantity of fission gas in gap is based on an extensive comparison with both published and proprietary data covering a ety of conditions. The internal pressure of the lead rod in the reactor is limited to a value that s not cause the diametral gap to increase due to outward cladding creep during steady state ration, and does not cause extensive DNB propagation to occur.
The fuel pellets in the central region of the fuel rod are right circular cylinders consisting of htly enriched uranium-dioxide powder, which has been compacted by cold-pressing and then ered to the required density. The edges of the pellets are chamfered slightly, and the ends of pellets are dished to accommodate axial expansion at the center of the fuel. The pellets in fuel s for selected fuel assemblies may have an integral burnable absorber dispersed in the pellet dolinia) or may have a thin zirconium diboride (ZrB2) coating on the outside surface of the et (IFBA).
The Westinghouse 422V+ fuel rods are sized to provide sufficient volume to accommodate ion gas release at extended burnup. Features of the 422V+ rods that help provide volume for accommodation are the selected pellet stack height and the use of annular blanket pellets at top and bottom of the fuel stack. The bottom end-plug of the fuel rod has an internal grip ure to facilitate rod loading and to provide appropriate lead-in for the removable top nozzle nstitution feature. The 422V+ fuel rod also may have a zirconium oxide coating at the bottom of the fuel rod to provide additional protection against debris fretting wear.
The 422V+ fuel contains mid-enriched (mid-enriched means greater than natural uranium less than the enrichment of the solid pellet used for the majority of the fuel rod) annular pellets xial blankets. This fuel rod design provides sufficient peaking factor and shutdown margin for core designs. Annular pellets provide additional plenum volume for fission gas releases while
ar heat rate which the annular pellets in axial blankets experience, and the modest fraction of fuel volume which they occupy, assures that their use does not have any significant effect on limiting fuel temperature or adverse effect on rod internal pressure.
The IFBA coated fuel pellets are identical to the enriched uranium dioxide pellets except for addition of a thin zirconium diboride (ZrB2) coating on the pellet cylindrical surface. Coated ets occupy the central portion of the fuel stack; the annular blanket pellets have no IFBA ting. The number and pattern of IFBA rods within an assembly may vary depending on cific application. The ends of the enriched coated pellets are dished to allow greater axial ansion at the pellet centerline and to increase the void volume for fission gas release.
Although the initial helium backfill pressure is lower in IFBA fuel rods than in non-IFBA s, IFBA rods operate with greater internal pressure (for a given irradiation) for most of their rating life because neutron absorption in the ZrB2 coating creates helium gas. Details of the A design are given in References 14 and 15.
Fuel rods containing IFBA coated pellets are identical in all external features to non-IFBA rods. All physical changes for IFBA are internal to the fuel rod cladding. Fuel rod design eria for all fuel rods, including the IFBA fuel rods, are verified as part of each reload design.
Rod Cluster Control Assemblies (RCCAs)
The control rods or RCCAs each consist of a group of individual absorber rods fastened at top end to a common hub or spider assembly. An example of one of these assemblies is shown igure 3.2-4.
The absorber material used in the control rods is a silver-indium-cadmium alloy, which is entially black to thermal neutrons and has sufficient additional resonance absorption to ificantly increase its worth. The absorber material in the form of rods is sealed in stainless l to prevent the rods from coming in direct contact with the coolant. In the current RCCAs the orber rodlet diameter in the lower twelve inches of the absorber rods is reduced to decrease ding strain caused by irradiation - induced swelling of the absorber.
The spider assembly is in the form of a center hub with radial vanes containing cylindrical ers from which the absorber rods are suspended. Handling detents and detents for connection he drive shaft are machined into the upper end of the hub. A spring pack is assembled into a t integral to the bottom of the hub to stop the RCCA and absorb the impact energy at the end trip insertion. The radial vanes are joined to the hub, and the fingers are joined to the vanes by ace-brazing. A centerpost, which holds the spring pack and its retainer is threaded into the within the skirt and welded to prevent loosening in service. All components of the spider
y and the retainer, which is of 17-4 pH material.
The absorber rods are securely fastened to the spider. The rods are first threaded into the er fingers and then pinned to maintain joint tightness, after which the pins are welded in
- e. The end plug below the pin position is designed with a reduced section to permit flexing of rods to correct for small operating or assembly misalignments.
The silver-indium-cadmium rods are contained within cold-worked stainless steel tubing, ch is sealed at the bottom and the top by welded end plugs. Sufficient diametral and end rances are provided to accommodate relative thermal expansions and to limit the internal sure to acceptable levels. The current control rods use cladding tubes that are hard chrome ed to increase wear resistance. The ends of the cladding tubes are not plated to preclude tamination of the end plug welds during fabrication.
The bottom plugs are made bullet-nosed to reduce the hydraulic drag during a reactor trip to guide smoothly into the dashpot section of the fuel assembly guide thimbles. The upper g is threaded for assembly to the spider and has a reduced end section to make the joint more ible.
Stainless steel clad silver-indium-cadmium alloy absorber rods are resistant to radiation and mal damage thereby ensuring their effectiveness under all operating conditions.
Neutron Source Assemblies Two primary-secondary neutron source assemblies were utilized in the cores and were sequently replaced during reload Cycles 4 and 5 with a pair of secondary sources. The initial rce assemblies each contained one combination primary-secondary source rod and three ondary source rods. Neutron source assemblies were removed from the core in 1992 and have been used in subsequent cycles.
Plugging Devices When necessary to limit bypass flow through the guide thimbles in fuel assemblies that do contain control rods, source assemblies, or burnable poison rods, the fuel assemblies at those tions may be fitted with plugging devices. The plugging devices consist of a flat plate with rt rods suspended from the bottom surface and a spring pack assembly attached to the top ace. The plugging devices fit within the fuel assembly top nozzles. The thimble plug plate s on the top nozzle adaptor plate and the short rods project into the upper ends of the thimble es to reduce the bypass flow area. The spring pack is compressed by the upper core plate when upper internals package is lowered into place.
All components in the plugging device, except for the springs, are constructed from type stainless steel. The springs are wound from an age-hardenable nickel-base alloy to obtain
e not been used in subsequent cycles.
Discrete Burnable Poison Rods Discrete burnable poison rods were employed during the first core cycle and are odically used as required by the reload core design. The burnable poison rods are statically pended and positioned in the guide thimbles of fuel assemblies at nonrodded core locations.
poison rods used at KPS were of two types, the standard design and the wet annular burnable orber (WABA) design. The WABA poison rods were removed from the reactor after one fuel le and have not been used in subsequent cycles.
The standard design burnable poison rods consist of borosilicate glass tubes contained hin type-304 stainless steel tubular cladding, which is plugged and seal-welded at the ends to apsulate the glass. The glass is also supported along the length of its inside diameter by a
-wall type-304 stainless steel tubular inner liner.
The standard design burnable poison rods are designed in accordance with the standard fuel design criteria. The cladding is free-standing at reactor operating pressures and temperatures sufficient cold void volume is provided within the rods to limit internal pressures to less than reactor operating pressure assuming total release of all helium generated in the glass as a lt of the B10 (n, a) reaction. The void volume required for the helium is obtained through the of glass in tubular form, which provides a central void along the length of the rods. The lting clad stresses at temperature and pressure are given in WCAP 7113 (Reference 26).
3.3 Evaluation of Core Components Evaluation of the capability of fuel assemblies of the Westinghouse 422V+ fuel design to t mechanical design criteria at a rated power level of 1772 MWth was first performed prior to use of the 422V+ fuel design in Cycle 26 and documented in the Reload Transition Safety ort (Reference 24). The analyses reported in the Reload Transition Safety Report (RTSR) luated reload cycle core designs consisting of both FRA/ANP and Westinghouse 422V+ fuel gns (mixed cores) and reload cycle core designs consisting only of Westinghouse 422V+ fuel gns.
The reload design methodology included fuel rod mechanical design analyses and luations to ensure that the reload fuel continues to meet applicable mechanical design and ty criteria. These fuel rod design evaluations were performed using NRC approved models ferences 9, 14, and 33) and the NRC approved design methods (References 27, 28, and 29) to onstrate that all of the fuel design criteria (Reference 27) were satisfied throughout the rating life of the fuel.
Fuel rod evaluation involves mechanical design analyses and evaluations to ensure that the meets applicable mechanical design and safety criteria.
The integrity of fuel rod cladding is important for retaining fission products or fuel material is directly related to cladding stress and strain under normal operating and overpower ditions. For most of the fuel rod life the actual cladding stresses and strains are considerably w the design limits. Thus, significant margin exists between actual operating conditions and fuel damage limits. In the event of cladding defects, the high resistance of uranium dioxide pellets to attack by hot water protects against fuel deterioration or decrease in fuel integrity.
consequences of a breach of cladding are greatly reduced by the ability of uranium dioxide to in fission products including those that are gaseous or highly volatile.
The following fuel mechanical design criteria and specified acceptable fuel design limits confirmed to be met for the fuel in the reload core design.
Rod Internal Pressure The internal pressure of the lead fuel rod in the reactor is limited to a value below that ch could cause the diametral gap to increase due to outward clad creep during steady state ration or extensive DNB propagation to occur. The rod internal pressure for the KPS 422V+
rods is evaluated on a case by case basis by modeling the gas inventories, gas temperature rod internal volumes for the projected operating conditions throughout the rods' life. The lting rod internal pressure is then compared to the design limit.
Fuel Rod Axial Growth This criterion assures that sufficient axial space exists to accommodate the maximum ected fuel rod growth without degradation of the assembly function. Fuel rods are designed h adequate clearance between the fuel rod and the top and bottom nozzles to accommodate the erences in the growth of fuel rods and the growth of the fuel assembly to preclude interference ween these members.
3.3.3 Deleted 3.4 Deleted 3.4.1 Deleted 3.4.2 Deleted 3.4.3 Deleted 3.4.4 Deleted 3.4.4.1 Deleted 3.4.4.2 Deleted 3.4.4.3 Deleted 3.4.5 Deleted 3.2 References Deleted Deleted Nguyen, T. Q., et al., Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, WCAP-11596-P-A, June 1988.
Liu, Y. S., et al., ANC: A Westinghouse Advanced Nodal Computer Code, WCAP 10965 P A, September 1986.
Deleted Deleted Reactor Test Program, Kewaunee Power Station, (Revision 9, September 2006).
Deleted Weiner, R. A., et al., Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations, WCAP-10851-P-A, August 1988.
Deleted Deleted Deleted Deleted
WCAP-12610-P-A, April 1995.
Davidson, S.L. and Kramer, W.R., Reference Core Report Vantage 5 Fuel Assembly, WCAP-10444-P-A, September 1985.
Deleted Deleted Wisconsin Public Service Corporation, Kewaunee Nuclear Power Plant, Topical Report WPSRSEM-NP-A entitled, Reload Safety Evaluation Methods for Application to Kewaunee, Revision 3, October 2001.
Letter from G. T. Bischof (DEK) to NRC, Implementation of the Dominion Statistical DNBR Methodology with VIPRE-D/WRB-1 at Kewaunee Power Station, dated May 4, 2007.
Deleted Deleted Deleted Deleted Sisk, R.B., Reload Transition Safety Report for the Kewaunee Nuclear Power Plant, July 2002.
Deleted WCAP-7113, Use of Burnable Poison Rods in Westinghouse Pressurized Water Reactors, October 1967.
Davidson, S. L., (Ed.), et al., Westinghouse Fuel Criteria Evaluation Process, WCAP-12488-A, October 1994 and Addendum 1 to WCAP-12488-A Revision to Design Criteria, WCAP-12488-P-A, Addendum 1, Revision 1, January 2002.
Davidson, S. L. (Ed.) et al., Extended Burnup Evaluation of Westinghouse Fuel, WCAP-10125-P-A (Proprietary), December 1985 and Slagle, W. H., Revisions to Design Criteria, WCAP-10125-P-A, Addendum 1-A, May 2003.
Kersting, P. J., et al., Assessment of Clad Flattening and Densification Power Spike Factor Elimination in Westinghouse Nuclear Fuel, WCAP-13589-A, March 1995.
Deleted Deleted Deleted
(PAD 4.0),WCAP-15063-P-A with errata, July 2000.
Blanchard, A., and D.N. Katz, Solid State Rod Control System, Full Length, WCAP-7778, January 1973.
Deleted
NUCLEAR DESIGN DATA Structural Characteristics el Weight (U) (kg), per assembly (approximate) 395 estinghouse 422V+)
tal Assembly Weight (lb), (approximate) 1266 estinghouse 422V+)
tive Full Length (in.) (Westinghouse 422V+) 143.25 mber of UO2 Rods per Assembly 179 Performance Characteristics el Enrichment (weight %) Typical Cycle 4.0 to 4.95 Control Characteristics d Cluster Control Assemblies (RCCAs) Material 5% Cd; 15% In; 80% Ag egral Burnable Poison Material (IFBA) Boron 10 - ZrB2 egral Burnable Poison Material (Gadolinia) Gd2O3 BA B10 Loading (gm/in) 0.02213 dolinia Weight Percent 2, 4, 6 or 8
Deleted Deleted Deleted Deleted Deleted CORE MECHANICAL DESIGN PARAMETERS d Cluster Control Assemblies Neutron Absorber 5% Cd, 15% In, 80% Ag Cladding Material Type 304 SS - Cold Worked Clad Thickness, in. 0.019 Number of Clusters:
Full Length 29 Number of Control Rods per Cluster 16 Weight in 60°F water: 114 Length of Control Rods, in. 150.58 (insertion length)
Length of Absorber Section, in. 142.01
FUEL ASSEMBLY AND COMPONENT DESCRIPTIONS Design Characteristics FRA-ANP mponent Heavy Fuel Westinghouse 422V+
el Assembly Array 14x14 14x14 Pitch (Assy) 7.803 in. 7.803 in.
Pitch (Rod) 0.556 in. 0.556 in.
Length 159.7 in. 159.8 in.
Distance between tie plates 153.6 in. 153.9 in.
No. fuel rods 179 179 No. guide tubes 16 16 No. instrument tubes 1 1 No. spacer grids 7 7 Fuel (U) weight per assy (approx.) 406 kg 403 kg Total assy. weight (approx.) 1283 lb 1266 lb el Rods Total Length 152.07 in. 152.56 in.
Active Fuel Length 144.0 in. 143.25 in.
Fuel/Clad Diametral Gap 0.0070 in. 0.0075 in.
Annular Axial Blanket Length N/A 6.0 in.
el Pellet 0.3670 in. 0.3659 in.
Diameter Theoretical Density 95.35% 96.56%
Material sintered UO2 sintered UO2 nular Axial Blanket Pellet Diameters 0.1830 in.
I.D. N/A 0.3659 in.
O.D. N/A adding O.D. 0.424 in. 0.422 in.
I.D. 0.374 in. 0.3734 in.
Material Zr-4 ZIRLO
FUEL ASSEMBLY AND COMPONENT DESCRIPTIONS Design Characteristics FRA-ANP mponent Heavy Fuel Westinghouse 422V+
acer Grids (Top - Bottom/Mid)
Height 2.25 in./1.75 in. 1.904 in./2.672 in Outer Dimension 7.763 in./7.761 in. 7.759 in./7.756in.
Material Zr-4/ Inconel Inconel/ZIRLO ide Tube O.D. (above dashpot) 0.541 in. 0.526 in.
I.D. (above dashpot) 0.507 in. 0.492 in.
O.D. (below dashpot) 0.481 in. 0.4815 in.
I.D. (below dashpot) 0.447 in. 0.4465 in.
aterial Zr-4 ZIRLO
Deleted Deleted Deleted TYPICAL ROD CLUSTER CONTROL ASSEMBLY Deleted Deleted Deleted WESTINGHOUSE 422V+ AND FRAMATOME FUEL ASSEMBLY DESIGNS Deleted Deleted Intentionally Blank 1 Deleted 2 Deleted 3 Deleted 3.3 References[Deleted]
Deleted Deleted
Intentionally Blank