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
Revision 2511/26/14 KPS USAR 3-i 3.1 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 3.2 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 Chapter 3: Reactor Fuel Table of Contents Section Title Page
Chapter 3: Reactor Fuel Table of Contents (continued)
Section Title Page Revision 2511/26/14 KPS USAR 3-ii 3.3 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
Revision 2511/26/14 KPS USAR 3-iii Chapter 3: Reactor Fuel List of Tables Table Title Page 3.2-1 Nuclear Design Data..........................................
3.2-13 3.2-2 Deleted....................................................
3.2-14 3.2-3 Deleted....................................................
3.2-15 3.2-4 Deleted....................................................
3.2-16 3.2-5 Deleted....................................................
3.2-17 3.2-6 Deleted....................................................
3.2-18 3.2-7 Core Mechanical Design Parameters.............................
3.2-19 3.2-8 Fuel Assembly and Component Descriptions.......................
3.2-20
Revision 2511/26/14 KPS USAR 3-iv Chapter 3: Reactor Fuel List of Figures Figure Title Page 3.2-1 Deleted...................................................
3.2-22 3.2-2 Deleted...................................................
3.2-23 3.2-3 Deleted...................................................
3.2-24 3.2-4 Typical Rod Cluster Control Assembly..........................
3.2-25 3.2-5 Deleted...................................................
3.2-26 3.2-6 Deleted...................................................
3.2-27 3.2-7 Deleted...................................................
3.2-28 3.2-8 Westinghouse 422V+ and Framatome Fuel Assembly Designs........
3.2-29 3.2-9 Deleted...................................................
3.2-30 3.2-10 Deleted...................................................
3.2-31
Revision 2511/26/14 KPS USAR 3.1-1 CHAPTER 3 REACTOR FUEL The fuel rods are cold-worked partially annealed zirconium alloy tubes containing slightly enriched 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 Control Assembly (RCCA) guide tube thimbles fastened to grids and top and bottom nozzles. The fuel rods are supported and the spacing between fuel rods maintained at several points along their length by spacer grids. Dominion Energy Kewaunee, Inc. (DEK) is no longer authorized under its operating license to place or hold nuclear fuel in the reactor vessel.
The RCCA absorber sections are fabricated of silver-indium-cadmium alloy sealed in stainless steel tubes. The fuel assembly design accommodates the use of additional fixed burnable neutron poison materials including discrete burnable poison rods located in the RCCA guide thimbles and the fuel rod design accommodates the use of integral neutron absorber materials such as Gadolonia (Gd2O3) or ZrB2.
The RCCA drive mechanisms are of the magnetic latch type and are controlled by three magnetic coils. They are so designed that, upon loss of power to the coils, the RCCA is released and falls by gravity to shut down the reactor.
3.1 DESIGN BASES 3.1.1 Performance Objectives Each reload cycle core was designed to yield an approximate 18-month fuel cycle at the rated power level of 1772 MWth. A reload safety evaluation report was prepared for each reload cycle core design to summarize the analyses performed to demonstrate the design would not adversely affect the safety of the plant. In addition, per Technical Specifications, a Core Operating Limits Report (COLR) was prepared. Core operating limits presented in the COLR were determined using analytical methods reviewed and approved by the Nuclear Regulatory Commission (NRC). Operation of the core within the COLR limits ensured that all applicable limits of the safety analyses were met.
The RCCAs provided sufficient control rod worth to shut the reactor down by the shutdown reactivity required by Technical Specifications in the hot condition at any time during the fuel cycle with the most reactive RCCA stuck in the fully withdrawn position.
Revision 2511/26/14 KPS USAR 3.1-2 3.1.2 Deleted 3.1.2.1 Deleted 3.1.2.2 Deleted 3.1.2.3 Deleted 3.1.2.4 Deleted 3.1.2.5 Deleted 3.1.2.6 Deleted 3.1.2.7 Deleted 3.1.2.8 Deleted 3.1.3 Safety Limits 3.1.3.1 Deleted 3.1.3.2 Deleted 3.1.3.3 Deleted 3.1.3.4 Mechanical Limits 3.1.3.4.1 Deleted 3.1.3.4.2 Fuel Assemblies The fuel assemblies were designed to perform satisfactorily throughout their lifetime. The loads, stresses, and strains resulting from the combined effects of flow-induced vibrations, earthquakes, reactor pressure, fission-gas pressure, fuel growth, thermal strain, and differential expansion during both steady-state and transient reactor operating conditions were considered in the design of the fuel rods and fuel assembly. The assembly is structurally designed to withstand handling and shipping loads prior to irradiation, and to maintain sufficient integrity at the completion of design burnup to permit safe removal from the core and subsequent handling during cooldown, shipment and fuel reprocessing or storage.
The fuel rods are supported at several locations along their length within the fuel assemblies by grid assemblies, which are designed to maintain control of the lateral spacing between the rods throughout the design life of the assemblies. The magnitude of the support loads provided by the grids was established to minimize possible fretting without overstressing the cladding at the points of contact between the grids and fuel rods. The grid assemblies also allow axial thermal expansion of the fuel rods without imposing restraint of sufficient magnitude to result in buckling or distortion of the rods. The fuel rod cladding was designed to withstand operating pressure loads without collapse or rupture and to maintain encapsulation of the fuel throughout the design life.
Revision 2511/26/14 KPS USAR 3.1-3 3.1.3.4.3 Rod Cluster Control Assemblies The criteria used for the design of the cladding on the individual absorber rods in the RCCAs are similar to those used for the fuel rod cladding. The stainless steel cladding was designed to be free standing under all operating conditions and will maintain encapsulation of the absorber material throughout the absorber rod design life. Allowance for wear during operation was included in the RCCA cladding thickness.
3.1.3.4.4 Deleted
Revision 2511/26/14 KPS USAR 3.1-4 Intentionally Blank
Revision 2511/26/14 KPS USAR 3.2-1 3.2 REACTOR DESIGN 3.2.1 Deleted 3.2.1.1 Deleted 3.2.1.2 Deleted 3.2.1.2.1 Deleted 3.2.1.2.2 Deleted 3.2.1.2.3 Deleted 3.2.1.2.4 Deleted 3.2.1.2.5 Deleted 3.2.1.2.6 Deleted 3.2.1.2.7 Deleted 3.2.1.2.8 Deleted 3.2.1.2.9 Deleted 3.2.1.2.10 Deleted 3.2.1.2.11 Deleted 3.2.2 Deleted 3.2.2.1 Deleted 3.2.2.2 Deleted 3.2.2.3 Deleted 3.2.2.3.1 Deleted 3.2.2.3.2 Deleted 3.2.2.4 Deleted 3.2.2.5 Deleted 3.2.2.6 Deleted 3.2.2.7 Deleted 3.2.3 Mechanical Design and Evaluation
Revision 2511/26/14 KPS USAR 3.2-2 The fuel is in the form of slightly enriched uranium dioxide ceramic pellets. Pellets in selected fuel rods may contain integral Gadolinia burnable poison or may have a thin zirconium boride (ZrB2) coating. The pellets are stacked to the fuel active height within the tubular cladding, which is plugged and seal-welded at the ends to encapsulate the fuel. All fuel rods are internally pressurized with helium during fabrication.
The control rods, designated as RCCAs, consist of groups of individual absorber rods, which are held together by a spider at the top end and actuated as a group. In the inserted position, the absorber rods fit within hollow guide thimbles in the fuel assemblies. The guide thimbles are an integral part of the fuel assemblies and occupy locations within the regular fuel rod pattern where 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.
3.2.3.1 Deleted 3.2.3.1.1 Deleted 3.2.3.1.2 Deleted 3.2.3.1.3 Deleted 3.2.3.1.4 Deleted 3.2.3.1.5 Deleted 3.2.3.2 Core Components The final reload core design consisted of a full core of Westinghouse fuel assemblies of the 422V+ fuel design. Transition to the full use of the 422V+ fuel design began with Cycle 26 and the first full core of 422V+ fuel assemblies was Cycle 28.
Selected mechanical design parameter values for FRA/ANP Heavy and Westinghouse 422V+ 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 centerline-to-centerline pitch of 0.556 inch between rods. Of the total possible 196-rod locations per assembly, 16 are occupied by guide thimbles for the RCC rods or discrete burnable poison rods and one by a tube for in-core instrumentation. The remaining 179 locations contain fuel rods.
In addition to fuel rods, a fuel assembly is composed of a top nozzle, a bottom nozzle, 7 spacer grid assemblies, 16 absorber rod guide thimbles, and 1 instrumentation thimble.
Revision 2511/26/14 KPS USAR 3.2-3 The guide thimbles, in conjunction with the grid assemblies and the top and bottom nozzles, comprise the basic structural fuel assembly skeleton. The top and bottom ends of the guide thimbles are fastened to the top and bottom nozzles, respectively. The grid assemblies, in turn, are fastened to the guide thimbles at each location along the height of the fuel assembly at which lateral support for the fuel rods is required. Within this skeletal frame-work the fuel rods are contained 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 distribution to the fuel assembly and functions as the bottom structural element of the fuel assembly. The nozzle, which is square in cross-section, is fabricated from 304 stainless steel parts consisting of a perforated plate, four angle legs, and four pads or feet. The legs form a plenum space for the inlet coolant to the fuel assembly. The perforated plate serves as the bottom end support for the fuel rods. The bottom support surface for the fuel assembly is formed under the plenum 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 bottom nozzle plate. These loads as well as the weight of the assembly are distributed through the nozzle to the lower core support plate. Indexing and positioning of the fuel assembly in the core is controlled through two holes in diagonally opposite pads, which mate with locating pins in the lower core plate. Lateral loads imposed on the fuel assembly are also transferred to the core support structures through the locating-pins.
Top Nozzle The top nozzle is a box-like structure, which functions as the fuel assembly upper structural element and forms a plenum space where the heated reactor coolant was directed toward the flow holes in the upper core plate. The top nozzle is made of stainless steel with four Inconel hold-down springs attached.
The top nozzle adaptor plate is provided with perforations to provide for coolant flow and provides a means of evenly distributing any axial loads imposed on the fuel assemblies among the guide thimbles.
The springs are fastened in pairs to the top nozzle at the two corners where alignment holes are not used and radiate out from the corners parallel to the sides of the nozzle. Fastening of each pair of springs is accomplished with a clamp, which fits over the ends of the springs and bolts which pass through the clamp and spring, and thread into the top nozzle. At assembly, the spring mounting bolts are torqued sufficiently to pre-load against the maximum spring load and then lock wired to the clamp, which is counter-bored to receive the bolt head.
Revision 2511/26/14 KPS USAR 3.2-4 The top nozzle is also designed to accommodate other types of insert components that are currently not used in KPS core design including thimble plugging devices, source assemblies, and discrete burnable poison rod assemblies.
Guide Thimbles The guide thimbles in the fuel assembly provide guided channels for the absorber rods during insertion and withdrawal of the RCCAs. They are fabricated from a single piece of Zircaloy-4 or Zirlo tubing, which is drawn to two different diameters. The larger inside diameter at the top provides a relatively large annular area for rapid insertion during a reactor trip and to accommodate a small amount of upward cooling flow during normal operation. The bottom portion of the guide thimble is of reduced diameter to produce a dashpot action when the absorber rods near the end of travel in the guide thimbles during a reactor trip. The transition zone at the dashpot 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 of cooling water during normal operation, and to accommodate the outflow of water from the dashpot during reactor trip.
The dashpot is closed at the bottom by means of a welded end plug. The end plug is fastened to the bottom nozzle during fuel assembly fabrication. Flow holes are provided in the endplugs to permit entrance of cooling water during normal operation and to regulate dashpot action during control rod trip.
Grids The spacer grid assemblies consist of individual slotted straps, which are assembled and interlocked in an egg-crate type arrangement. The straps are permanently joined at their points of intersection.
Two types of grid assemblies are used in the Westinghouse 422V+ fuel assemblies. Inconel end grids are utilized for their corrosion resistance and high strength properties to hold the fuel rods tightly. The inner straps on the Inconel grids do not contain mixing vanes. Mid-grids are structural grids with mixing vanes on the upper edges of the inner grid straps. These grids are used in the high heat region of the fuel assemblies to promote mixing of the coolant. ZIRLO' is utilized as the material for the mid-grids primarily for its low-neutron absorption cross section.
The outside straps on all grids on the 422V+ fuel assemblies contain mixing vanes on their upper edges and small tabs projecting downwards from the lower edge, to aid in guiding the grids and fuel assemblies past projecting surfaces during handling or core loading and unloading. Each grid cell in both grid types includes two springs and four dimples that serve as contact points with the
Revision 2511/26/14 KPS USAR 3.2-5 fuel rod. The 422V+ grid assembly design bases and evaluation are given in Section 2.3 of Reference 14.
Fuel Rods The fuel rods consist of uranium dioxide ceramic pellets in slightly cold-worked and partially annealed Zircaloy-4 or ZIRLO' tubing. The tubing is plugged and seal-welded at the ends to encapsulate the fuel. Sufficient void volume and clearances are provided within the rod to accommodate fission gases released from the fuel, differential thermal expansion between the cladding and the fuel, and fuel swelling due to accumulated fission products without overstressing of the cladding or seal welds. Shifting of the fuel within the cladding is prevented during handling or shipping 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 space is sized to ensure adherence to the pressure criteria. The rod internal pressure is evaluated for the limiting fuel rod, assuming a conservative operating history. The evaluation is based on expected operating conditions at the peak steady-state power, and also considers the fission gas release from normal operating transients. The model used to predict the quantity of fission gas in the gap is based on an extensive comparison with both published and proprietary data covering a variety of conditions. The internal pressure of the lead rod in the reactor is limited to a value that does not cause the diametral gap to increase due to outward cladding creep during steady state operation, 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 slightly enriched uranium-dioxide powder, which has been compacted by cold-pressing and then sintered to the required density. The edges of the pellets are chamfered slightly, and the ends of the pellets are dished to accommodate axial expansion at the center of the fuel. The pellets in fuel rods for selected fuel assemblies may have an integral burnable absorber dispersed in the pellet (gadolinia) or may have a thin zirconium diboride (ZrB2) coating on the outside surface of the pellet (IFBA).
The Westinghouse 422V+ fuel rods are sized to provide sufficient volume to accommodate fission gas release at extended burnup. Features of the 422V+ rods that help provide volume for gas accommodation are the selected pellet stack height and the use of annular blanket pellets at the top and bottom of the fuel stack. The bottom end-plug of the fuel rod has an internal grip feature to facilitate rod loading and to provide appropriate lead-in for the removable top nozzle reconstitution feature. The 422V+ fuel rod also may have a zirconium oxide coating at the bottom end 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 but less than the enrichment of the solid pellet used for the majority of the fuel rod) annular pellets in axial blankets. This fuel rod design provides sufficient peaking factor and shutdown margin for the core designs. Annular pellets provide additional plenum volume for fission gas releases while
Revision 2511/26/14 KPS USAR 3.2-6 axial blankets reduce neutron leakage and improve fuel utilization. The relatively low range of linear heat rate which the annular pellets in axial blankets experience, and the modest fraction of the fuel volume which they occupy, assures that their use does not have any significant effect on the 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 the addition of a thin zirconium diboride (ZrB2) coating on the pellet cylindrical surface. Coated pellets occupy the central portion of the fuel stack; the annular blanket pellets have no IFBA coating. The number and pattern of IFBA rods within an assembly may vary depending on specific application. The ends of the enriched coated pellets are dished to allow greater axial expansion 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 rods, IFBA rods operate with greater internal pressure (for a given irradiation) for most of their operating life because neutron absorption in the ZrB2 coating creates helium gas. Details of the IFBA design are given in References 14 and 15.
Fuel rods containing IFBA coated pellets are identical in all external features to non-IFBA fuel rods. All physical changes for IFBA are internal to the fuel rod cladding. Fuel rod design criteria 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 the top end to a common hub or spider assembly. An example of one of these assemblies is shown in Figure 3.2-4.
The absorber material used in the control rods is a silver-indium-cadmium alloy, which is essentially black to thermal neutrons and has sufficient additional resonance absorption to significantly increase its worth. The absorber material in the form of rods is sealed in stainless steel to prevent the rods from coming in direct contact with the coolant. In the current RCCAs the absorber rodlet diameter in the lower twelve inches of the absorber rods is reduced to decrease cladding 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 fingers from which the absorber rods are suspended. Handling detents and detents for connection to the drive shaft are machined into the upper end of the hub. A spring pack is assembled into a skirt integral to the bottom of the hub to stop the RCCA and absorb the impact energy at the end of a trip insertion. The radial vanes are joined to the hub, and the fingers are joined to the vanes by furnace-brazing. A centerpost, which holds the spring pack and its retainer is threaded into the hub within the skirt and welded to prevent loosening in service. All components of the spider
Revision 2511/26/14 KPS USAR 3.2-7 assembly are made from Type 304 stainless steel except for the springs, which are Inconel X-750 alloy 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 spider fingers and then pinned to maintain joint tightness, after which the pins are welded in place. The end plug below the pin position is designed with a reduced section to permit flexing of the rods to correct for small operating or assembly misalignments.
The silver-indium-cadmium rods are contained within cold-worked stainless steel tubing, which is sealed at the bottom and the top by welded end plugs. Sufficient diametral and end clearances are provided to accommodate relative thermal expansions and to limit the internal pressure to acceptable levels. The current control rods use cladding tubes that are hard chrome plated to increase wear resistance. The ends of the cladding tubes are not plated to preclude contamination of the end plug welds during fabrication.
The bottom plugs are made bullet-nosed to reduce the hydraulic drag during a reactor trip and to guide smoothly into the dashpot section of the fuel assembly guide thimbles. The upper plug is threaded for assembly to the spider and has a reduced end section to make the joint more flexible.
Stainless steel clad silver-indium-cadmium alloy absorber rods are resistant to radiation and thermal 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 subsequently replaced during reload Cycles 4 and 5 with a pair of secondary sources. The initial source assemblies each contained one combination primary-secondary source rod and three secondary source rods. Neutron source assemblies were removed from the core in 1992 and have not been used in subsequent cycles.
Plugging Devices When necessary to limit bypass flow through the guide thimbles in fuel assemblies that do not contain control rods, source assemblies, or burnable poison rods, the fuel assemblies at those locations may be fitted with plugging devices. The plugging devices consist of a flat plate with short rods suspended from the bottom surface and a spring pack assembly attached to the top surface. The plugging devices fit within the fuel assembly top nozzles. The thimble plug plate rests on the top nozzle adaptor plate and the short rods project into the upper ends of the thimble tubes to reduce the bypass flow area. The spring pack is compressed by the upper core plate when the upper internals package is lowered into place.
All components in the plugging device, except for the springs, are constructed from type 304 stainless steel. The springs are wound from an age-hardenable nickel-base alloy to obtain
Revision 2511/26/14 KPS USAR 3.2-8 higher strength. Thimble plugging devices were removed from the reactor in Cycle 14 (1988) and have not been used in subsequent cycles.
Discrete Burnable Poison Rods Discrete burnable poison rods were employed during the first core cycle and are periodically used as required by the reload core design. The burnable poison rods are statically suspended and positioned in the guide thimbles of fuel assemblies at nonrodded core locations.
The poison rods used at KPS were of two types, the standard design and the wet annular burnable absorber (WABA) design. The WABA poison rods were removed from the reactor after one fuel cycle and have not been used in subsequent cycles.
The standard design burnable poison rods consist of borosilicate glass tubes contained within type-304 stainless steel tubular cladding, which is plugged and seal-welded at the ends to encapsulate the glass. The glass is also supported along the length of its inside diameter by a thin-wall type-304 stainless steel tubular inner liner.
The standard design burnable poison rods are designed in accordance with the standard fuel rod design criteria. The cladding is free-standing at reactor operating pressures and temperatures and sufficient cold void volume is provided within the rods to limit internal pressures to less than the reactor operating pressure assuming total release of all helium generated in the glass as a result of the B10 (n, a) reaction. The void volume required for the helium is obtained through the use of glass in tubular form, which provides a central void along the length of the rods. The resulting clad stresses at temperature and pressure are given in WCAP 7113 (Reference 26).
3.2.3.3 Evaluation of Core Components Evaluation of the capability of fuel assemblies of the Westinghouse 422V+ fuel design to meet mechanical design criteria at a rated power level of 1772 MWth was first performed prior to the use of the 422V+ fuel design in Cycle 26 and documented in the Reload Transition Safety Report (Reference 24). The analyses reported in the Reload Transition Safety Report (RTSR) evaluated reload cycle core designs consisting of both FRA/ANP and Westinghouse 422V+ fuel designs (mixed cores) and reload cycle core designs consisting only of Westinghouse 422V+ fuel designs.
The reload design methodology included fuel rod mechanical design analyses and evaluations to ensure that the reload fuel continues to meet applicable mechanical design and safety criteria. These fuel rod design evaluations were performed using NRC approved models (References 9, 14, and 33) and the NRC approved design methods (References 27, 28, and 29) to demonstrate that all of the fuel design criteria (Reference 27) were satisfied throughout the operating life of the fuel.
Revision 2511/26/14 KPS USAR 3.2-9 3.2.3.3.1 Fuel Rod Evaluation Fuel rod evaluation involves mechanical design analyses and evaluations to ensure that the fuel meets applicable mechanical design and safety criteria.
The integrity of fuel rod cladding is important for retaining fission products or fuel material and is directly related to cladding stress and strain under normal operating and overpower conditions. For most of the fuel rod life the actual cladding stresses and strains are considerably below the design limits. Thus, significant margin exists between actual operating conditions and the fuel damage limits. In the event of cladding defects, the high resistance of uranium dioxide fuel pellets to attack by hot water protects against fuel deterioration or decrease in fuel integrity.
The consequences of a breach of cladding are greatly reduced by the ability of uranium dioxide to retain fission products including those that are gaseous or highly volatile.
The following fuel mechanical design criteria and specified acceptable fuel design limits are 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 which could cause the diametral gap to increase due to outward clad creep during steady state operation or extensive DNB propagation to occur. The rod internal pressure for the KPS 422V+
fuel rods is evaluated on a case by case basis by modeling the gas inventories, gas temperature and rod internal volumes for the projected operating conditions throughout the rods' life. The resulting 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 expected fuel rod growth without degradation of the assembly function. Fuel rods are designed with adequate clearance between the fuel rod and the top and bottom nozzles to accommodate the differences in the growth of fuel rods and the growth of the fuel assembly to preclude interference between these members.
Revision 2511/26/14 KPS USAR 3.2-10 3.2.3.3.2 Deleted 3.2.3.3.3 Deleted 3.2.3.4 Deleted 3.2.3.4.1 Deleted 3.2.3.4.2 Deleted 3.2.3.4.3 Deleted 3.2.3.4.4 Deleted 3.2.3.4.4.1 Deleted 3.2.3.4.4.2 Deleted 3.2.3.4.4.3 Deleted 3.2.3.4.5 Deleted 3.2 References
- 1. Deleted
- 2. Deleted
- 3. 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.
- 4. Liu, Y. S., et al., ANC: A Westinghouse Advanced Nodal Computer Code, WCAP 10965 P A, September 1986.
- 5. Deleted
- 6. Deleted
- 7. Reactor Test Program, Kewaunee Power Station, (Revision 9, September 2006).
- 8. Deleted
- 9. Weiner, R. A., et al., Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations, WCAP-10851-P-A, August 1988.
- 10. Deleted
- 11. Deleted
- 12. Deleted
- 13. Deleted
Revision 2511/26/14 KPS USAR 3.2-11
- 14. Davidson, S.L. and Ryan, T.L. (Eds.), Vantage & Fuel Assembly Reference Core Report, WCAP-12610-P-A, April 1995.
- 15. Davidson, S.L. and Kramer, W.R., Reference Core Report Vantage 5 Fuel Assembly, WCAP-10444-P-A, September 1985.
- 16. Deleted
- 17. Deleted
- 18. 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.
- 19. 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.
- 20. Deleted
- 21. Deleted
- 22. Deleted
- 23. Deleted
- 24. Sisk, R.B., Reload Transition Safety Report for the Kewaunee Nuclear Power Plant, July 2002.
- 25. Deleted
- 26. WCAP-7113, Use of Burnable Poison Rods in Westinghouse Pressurized Water Reactors, October 1967.
- 27. 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.
- 28. 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.
- 29. Kersting, P. J., et al., Assessment of Clad Flattening and Densification Power Spike Factor Elimination in Westinghouse Nuclear Fuel, WCAP-13589-A, March 1995.
- 30. Deleted
- 31. Deleted
- 32. Deleted
Revision 2511/26/14 KPS USAR 3.2-12
- 33. Foster, J. P., Sidener, S, Westinghouse Improved Performance Analysis and Design Model (PAD 4.0),WCAP-15063-P-A with errata, July 2000.
- 34. Blanchard, A., and D.N. Katz, Solid State Rod Control System, Full Length, WCAP-7778, January 1973.
- 35. Deleted
Revision 2511/26/14 KPS USAR 3.2-13 Table 3.2-1 NUCLEAR DESIGN DATA Structural Characteristics Fuel Weight (U) (kg), per assembly (approximate)
(Westinghouse 422V+)
395 Total Assembly Weight (lb), (approximate)
(Westinghouse 422V+)
1266 Active Full Length (in.) (Westinghouse 422V+)
143.25 Number of UO2 Rods per Assembly 179 Performance Characteristics Fuel Enrichment (weight %) Typical Cycle 4.0 to 4.95 Control Characteristics Rod Cluster Control Assemblies (RCCAs) Material 5% Cd; 15% In; 80% Ag Integral Burnable Poison Material (IFBA)
Boron 10 - ZrB2 Integral Burnable Poison Material (Gadolinia)
Gd2O3 IFBA B10 Loading (gm/in) 0.02213 Gadolinia Weight Percent 2, 4, 6 or 8
Revision 2511/26/14 KPS USAR 3.2-14 Table 3.2-2 Deleted
Revision 2511/26/14 KPS USAR 3.2-15 Table 3.2-3 Deleted
Revision 2511/26/14 KPS USAR 3.2-16 Table 3.2-4 Deleted
Revision 2511/26/14 KPS USAR 3.2-17 Table 3.2-5 Deleted
Revision 2511/26/14 KPS USAR 3.2-18 Table 3.2-6 Deleted
Revision 2511/26/14 KPS USAR 3.2-19 Table 3.2-7 CORE MECHANICAL DESIGN PARAMETERS Rod 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
Revision 2511/26/14 KPS USAR 3.2-20 Table 3.2-8 FUEL ASSEMBLY AND COMPONENT DESCRIPTIONS Component Design Characteristics FRA-ANP Heavy Fuel Westinghouse 422V+
Fuel Assembly Array Pitch (Assy)
Pitch (Rod)
Length Distance between tie plates No. fuel rods No. guide tubes No. instrument tubes No. spacer grids Fuel (U) weight per assy (approx.)
Total assy. weight (approx.)
14x14 7.803 in.
0.556 in.
159.7 in.
153.6 in.
179 16 1
7 406 kg 1283 lb 14x14 7.803 in.
0.556 in.
159.8 in.
153.9 in.
179 16 1
7 403 kg 1266 lb Fuel Rods Total Length Active Fuel Length Fuel/Clad Diametral Gap Annular Axial Blanket Length 152.07 in.
144.0 in.
0.0070 in.
N/A 152.56 in.
143.25 in.
0.0075 in.
6.0 in.
Fuel Pellet Diameter Theoretical Density Material 0.3670 in.
95.35%
sintered UO2 0.3659 in.
96.56%
sintered UO2 Annular Axial Blanket Pellet Diameters I.D.
O.D.
N/A N/A 0.1830 in.
0.3659 in.
Cladding O.D.
I.D.
Material 0.424 in.
0.374 in.
Zr-4 0.422 in.
0.3734 in.
Revision 2511/26/14 KPS USAR 3.2-21 Spacer Grids (Top - Bottom/Mid)
Height Outer Dimension Material 2.25 in./1.75 in.
7.763 in./7.761 in.
Zr-4/ Inconel 1.904 in./2.672 in 7.759 in./7.756in.
Inconel/ZIRLO Guide Tube O.D. (above dashpot)
I.D. (above dashpot)
O.D. (below dashpot)
I.D. (below dashpot)
Material 0.541 in.
0.507 in.
0.481 in.
0.447 in.
Zr-4 0.526 in.
0.492 in.
0.4815 in.
0.4465 in.
ZIRLO Table 3.2-8 (continued)
FUEL ASSEMBLY AND COMPONENT DESCRIPTIONS Component Design Characteristics FRA-ANP Heavy Fuel Westinghouse 422V+
Revision 2511/26/14 KPS USAR 3.2-22 Figure 3.2-1 Deleted
Revision 2511/26/14 KPS USAR 3.2-23 Figure 3.2-2 Deleted
Revision 2511/26/14 KPS USAR 3.2-24 Figure 3.2-3 Deleted
Revision 2511/26/14 KPS USAR 3.2-25 Figure 3.2-4 TYPICAL ROD CLUSTER CONTROL ASSEMBLY
Revision 2511/26/14 KPS USAR 3.2-26 Figure 3.2-5 Deleted
Revision 2511/26/14 KPS USAR 3.2-27 Figure 3.2-6 Deleted
Revision 2511/26/14 KPS USAR 3.2-28 Figure 3.2-7 Deleted
Revision 2511/26/14 KPS USAR 3.2-29 Figure 3.2-8 WESTINGHOUSE 422V+ AND FRAMATOME FUEL ASSEMBLY DESIGNS
Revision 2511/26/14 KPS USAR 3.2-30 Figure 3.2-9 Deleted
Revision 2511/26/14 KPS USAR 3.2-31 Figure 3.2-10 Deleted
Revision 2511/26/14 KPS USAR 3.2-32 Intentionally Blank
Revision 2511/26/14 KPS USAR 3.3-1 3.3 Deleted 3.3.1 Deleted 3.3.2 Deleted 3.3.3 Deleted 3.3 References[Deleted]
- 1. Deleted
- 2. Deleted