ML18122A329
| ML18122A329 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/16/2017 |
| From: | Susquehanna, Talen Energy |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17331A584 | List:
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| References | |
| PLA-7636 | |
| Download: ML18122A329 (291) | |
Text
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-1 4.1
SUMMARY
DESCRIPTION Susquehanna Units 1 and 2 are General Electric BWR/4 Boiling Water Reactors. Each reactor contains 764 fuel assemblies and 185 control rods arranged in an upright cylindrical configuration. Light water acts as both moderator and coolant. The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator and dryer assemblies, and jet pumps. Also included in the reactor assembly are the control rods, control rod drive housings, and the control rod drives. Figure 3.9-3, shows the arrangement of reactor assembly components. Important design and performance characteristics are discussed in Sections 4.2, 4.3 and 4.4. Loading conditions for reactor assembly components are specified in Section 3.9.
4.1.1 REACTOR
VESSEL The reactor vessel design and description are covered in Section 5.3.
4.1.2 REACTOR
INTERNAL COMPONENTS The major reactor internal components are the core (fuel, channels, control rods, and instrumentation), the core support structure (including the shroud, top guide and core plate), the shroud head and steam separator assembly, the steam dryer assembly, the feedwater spargers, the core spray spargers, and the jet pumps. Except for the Zircaloy in the reactor core, these reactor internals are stainless steel or other corrosion resistant alloys. All major internal components of the vessel can be removed except the jet pump diffusers, the jet pump risers, the shroud, the core spray lines, spargers, and the feedwater sparger. The removal of the steam dryers, shroud head and steam separators, fuel assemblies, in-core instrumentation, control rods, orificed fuel supports, and control rod guide tubes, can be accomplished on a routine basis.
4.1.2.1 Reactor Core 4.1.2.1.1 General The design of the boiling water reactor core, including fuel, is based on the proper combination of many design variables and operating experience. These factors contribute to the achievement of high reliability. A number of important features of the boiling water reactor core design are summarized in the following paragraphs:
(1)The BWR core mechanical design is based on conservative application of stress limits, operating experience, and experimental test results. The moderate pressu re level characteristics of a direct cycle reactor (approximately 1050 psia) result in moderate cladding te mperatures and stress levels.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-2 (2)The low coolant saturation temperature, high heat transfer coefficients, and near-neutral water chemistry of the BWR are significant, adva ntageous factors in minimizing Zirca loy temperature and associated temperature-dependent corrosion and hydride buildup.The relatively uniform fuel cladding temperatures throughout the core minimize migration of the hydrides to cold cladding zones and reduce thermal stresses.
(3)The basic thermal and mechanical criteria applied in the design have been proven by irradiation of statistically significant quantities of fuel. The design heat transfer rates an d linear heat generation rates are similar to values proven in fuel assembly irradiation.
(4)The design power distribution used in sizing the core represents a worst expected state of operation.
(5)The AREVA critical power methodology for boiling water reactors (References 4.1-1 2, and 4.1-29) is applied to assure that more than 99.9% of the fuel rods are expected to avoid boiling transition for the most severe abnormal operational transient described in Chapter 15. The possibility of boiling transition o ccurring during normal reactor operation is insignificant.
(6)Because of the large negative moderator density coefficient of reactivity during normal power operation, the BWR has a number of inherent advantages. These are the uses of coolant flow for power maneuvering, the inherent self-flattening of the radial power distribution, the ease of control, the spatial xenon stability, and the abilit y to override xenon, in order to follow load.Boiling water reactors do not have instability problems due to xenon. This has been demonstrated by special tests which have been conducted on operating BWRs in an attempt to force the reactor into xenon instability, and by calculations. No xenon instabilities have ever been observed in the test results. All of these indicators have proven that xenon transients are highly damped in a BWR due to the large negative power coefficient of reactivity (Reference 4.1-1). Important features of the reactor core arrangement are as follows:
(1)The bottom-entry cruciform control rods consist of B 4C in stainless steel tubes (i.e., B 4 C rods) only or a combination of B 4C rods and solid hafnium rods surrounde d by stainless steel. Control Rods are further described in subsections 4.1.3.2 and subs equent sections.(2)The fixed in-core ion chambers provide continuo us power range neutron flux monitoring.
A probe tube in each in-core assembly provides for a traversing ion cham ber for calibration and axial detail. Source and intermediate range monitors are located in-core and are axially retractable. The in-core location of the startup and source r ange instruments provides coverage of the large reactor core and provides an acceptable signal-to-noise ratio and neutron-to-gamma ratio. All in-core instrument leads enter fr om the bottom and the instruments are in service during refueling. In-core instr umentation is further discussed in Subsection 7.7.
1.6.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-3 (3) As shown by experience, the operator, utilizing the in-core flux monitor system, can maintain the desired power distribution within a large core by proper control rod scheduling.
(4) The channels (Zircaloy-2 or Zircaloy-4) provide a fixed flow path for the boiling coolant, serve as a guiding surface for the control rods, and protect the fuel during handling operations.
(5) The core is designed to be subcritical at any time in its operating history with any one control rod fully withdrawn.
(6) The selected control rod pitch provides the ability to finely control the power distribution in the assemblies contained in the reactor core. The pitch also allows ample clearance below the pressure vessel between control rod drive mechanisms for ease of maintenance and removal.
4.1.2.1.2 Core Configuration The reactor core is arranged as an upright circular cylinder containing a large number of fuel cells and is located within the reactor vessel. The coolant flows upward through the core. The Susquehanna SES Units utilize a conventional scatter loading with the lowest reactivity bundles placed in the peripheral region of the core. At periodic refueling intervals, each unit will enter an outage. During this time, fuel assemblies are identified to be discharged, new fuel is loaded and fuel assemblies in the reactor core may be "shuffled" to new locations. Therefore, the core loading patterns are both unit and cycle specific. The core configurations for each unit are discussed in Section 4.3.
4.1.2.1.3 Fuel Assembly Description The fuel assembly is composed of fuel and water rods (or interior water channels), structural components and a fuel channel. The mechanical design of the assembly is described in Section 4.2. The nuclear design of the assembly is described in Section 4.3. Thermal hydraulic design of the assembly is described in Section 4.4.
4.1.2.1.3.1 Fuel Rod A fuel rod consists of UO 2 pellets and a Zircaloy-2 cladding tube. A fuel rod is made by stacking pellets into a Zircaloy-2 cladding tube which is evacuated and back-filled with helium, and sealed by welding Zircaloy end plugs in each end of the tube.
The BWR fuel rod is designed as a pressure vessel. The ASME Boiler and Pressure Vessel Code,Section III, is used as a guide in the mechanical design and stress analysis of the fuel
rod.
The rod is designed to withstand the applied loads, both external and internal. The fuel pellet is sized to provide sufficient volume within the fuel tube to accommodate differential expansion between fuel and clad. Overall fuel rod design is conservative in its accommodation of the mechanisms affecting fuel in a BWR environment. Fuel rod design bases are discussed in more detail in Section 4.2.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-4 4.1.2.1.3.2 Fuel Bundle The fuel bundle has two important design features:
(1)The bundle design places minimum external forces on a fuel rod; each fuel rod is free to expand in the axial direction.
(2)The unique structural design permits the removal and replacem ent, if required, of individual fu el rods.Fuel bundles are designed to meet all the criteria for core performance and to provide ease of handling. Selected fuel rods in each assembly may differ from the others in initial uranium enrichment, burnable poison content, and fuel rod length. The variation in enrichment and burnable poison distribution produces more uniform power production across the fuel assembly, and thus allows a significant reduction in the amount of heat transfer surface required to satisfy the design thermal limitations. The inclusion of part length fuel rods in the assembly improves the two phase pressure drop, enhances the inherent stability of the bundle, and improves the required shutdown margin of the core design. Section 4.2 provides a more detailed description of the mechanical design aspects of the fuel bundles in use at Susquehanna. 4.1.2.1.4 Assembly Support and Control Rod Location All peripheral fuel assemblies are supported by the core plate. Otherwise, individual fuel assemblies in the core rest on fuel support pieces mounted on top of the control rod guide tubes. Each guide tube, with its fuel support piece, bears the weight of four assemblies and is supported by a control rod drive penetration nozzle in the bottom head of the reactor vessel. The core plate provides lateral support and guidance at the top of each control rod guide tube. The top guide, mounted inside the shroud, provides lateral support and guidance for each fuel assembly. The reactivity of the core is controlled by cruciform control rods which occupy alternate spaces between fuel assemblies. The position of each control rod is controlled by independent mechanical hydraulic drive systems. These systems insert and withdraw the control rod from the bottom of the core and can accurately position its associated control rod during normal operation and yet exert approximately ten times the force of gravity to insert the control rod during the scram mode of operation. Bottom entry allows optimum power shaping in the core, ease of refueling, and convenient drive maintenance.
4.1.2.2 Shroud The shroud is a cylindrical, stainless steel structure which surrounds the core and provides a barrier to separate the upward flow through the core from the downward flow in the annulus, and also provides a floodable volume in the unlikely event of an accident which tends to drain the reactor pressure vessel. A flange at the top of the shroud mates with a flange on the shroud head and steam separators. The upper cylindrical wall of the shroud and the shroud head form the core discharge plenum. The jet pump diffusers penetrate the shroud support below the core SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-5 elevation to introduce the coolant to the bottom head volume. The shroud support is designed to support and locate the jet pumps, core support structure, peripheral fuel assemblies and to separate the inlet and outlet flows of the recirculation loops. Mounted inside the upper shroud cylinder in the space between the top of the core and the upper shroud flange are the core spray spargers with spray nozzles for injection of cooling water. The core spray spargers and nozzles do not interfere with the installation or removal of fuel from the core.
4.1.2.3 Shroud Head and Steam Separators The shroud head consists of a flange and dome onto which is welded an array of standpipes, with a steam separator located at the top of each standpipe. The shroud head mounts on the flange at the top of the cylinder and forms the cover of the core discharge plenum region. The joint between the shroud head and shroud flange does not require a gasket or other replacement sealing technique. The fixed axial flow-type steam separators have no moving parts and are made of stainless steel. In each separator, the steam-water mixture rising from the standpipe impinges on vanes which give the mixture a spin to establish a vortex wherein the centrifugal forces separate the steam from the water. Steam leaves the separator at the top and passes into the wet steam plenum below the dryer. The separated water exits from the lower end of the separator and enters the pool that surrounds the standpipes to enter the downcomer annulus. For ease of removal, the shroud head is bolted to the shroud top flange by long shroud head bolts that extend above the separators for easy access during refueling. The shroud head is guided into position on the shroud via guide rods on the inside of the vessel and locating pins located on the shroud head. The objective of the shroud head bolt design is to provide direct access to the bolts during reactor refueling operations with underwater tool manipulation during the removal and installation of the assemblies. 4.1.2.4 Steam Dryer Assembly The steam dryer assembly is mounted in the reactor vessel above the shroud head and forms the top and sides of the wet steam plenum. Vertical guide rods on the inside of the vessel provide alignment for the dryer assembly during installation. The dryer assembly is supported by pads extending from the vessel wall and is prevented from lifting during postulated transients by brackets welded to the reactor vessel top head. Steam from the separators flows upward into the dryer assembly. Moisture is removed by the dryer vanes and flows first through a system of troughs and pipes to the pool surrounding the separators and then into the downcomer annulus between the core shroud and reactor vessel wall. The steam leaving the top of the dryer assembly flows into vessel steam outlet nozzles which are located alongside the steam dryer assembly. The schematics of a typical steam dryer panel are shown in Figures 4.1-2 and 4.1-3. 4.1.3 REACTIVITY CONTROL SYSTEMS 4.1.3.1 Operation SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-6 The control rods perform dual functions of power distribution shaping and reactivity control. Power distribution in the core is controlled during operation of the reactor by manipulation of selected patterns of rods. The rods, which enter from the bottom of the near-cylindrical reactor core, are positioned in such a manner to counter-balance steam voids in the top of the core and which results in significant power flattening.
These groups of control elements, used for power flattening, experience a somewhat higher duty cycle and neutron exposure than the other rods in the control system.
The reactivity control function requires that all rods be available for either reactor "scram" (prompt shutdown) or reactivity regulation. Because of this, the control elements are mechanically designed to withstand the dynamic forces resulting from a scram. They are connected to bottom-mounted, hydraulically actuated drive mechanisms which allow either axial positioning for reactivity regulation or rapid scram insertion.
The design of the rod-to-drive connection permits each blade to be attached or detached from its drive without disturbing the remainder of the control system. The bottom-mounted drives permit the entire control system to be left intact and operable for tests with the reactor vessel
open.
4.1.3.2 Description of Rods
For the original equipment and Duralife D160-C control rods the neutron absorber portion of the control rod is contained in the wings of the cruciform shaped control blades which are inserted in the bypass region between four fuel assemblies. The original equipment control blades contain boron-carbide (B 4C) powder filled stainless steel absorber tubes. Newer generation control blades contain a combination of B 4C filled tubes and solid hafnium rods. The boron-carbide absorber tubes are seal welded with end plugs on either end. Stainless steel balls are used to separate the tubes into individual compartments. The stainless steel balls are held in position by a slight crimp in the tube. The individual tubes act as pressure vessels to contain the helium gas released by the boron-neutron capture reaction. The tubes or rods are held in a cruciform array by a stainless steel sheath extending the full length of the tubes.
A top handle aligns the tubes and provides structural rigidity at the top of the control rod.
Rollers, housed in the handle, provide guidance for control rod insertion and withdrawal. A bottom casting is also used to provide structural rigidity and contains positioning rollers and a parachute-shaped velocity limiter. The handle and lower casting are welded into a single structure by means of a small cruciform post located in the center of the control rod.
Replacement Marathon control rods may use a modified handle assembly that eliminates pins and rollers present in the earlier design.
Marathon control blade wings are made up of an array of square tubes welded together. The tube arrays are welded to center tie rods to form the cruciform blade shape. The square tubes are loaded with either B 4 C or Hafnium. The B 4C is contained in separate capsules to prevent migration within the tubes. The square tubes are sealed at each end to prevent the neutron poisons from washing out into the coolant. The blade handle and velocity limiter are equivalent to previous control blade designs, (Reference 4.1-24). The Marathon Ultra - HD Control Rod design was introduced in U2C18, (Reference 4.1-30).
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-7 Westinghouse CR 99 control rods, introduced in U1C20, are designed similar to the original equipment and newer style Marathon control rod, (Reference 4.1-31). Like the above GE control rods, the Westinghouse CR 99 control rods have a cruciform blade shape, a handle, a B 4C loaded absorber zone and a velocity limiter. Different from the aforementioned GE control rods, the Westinghouse CR 99 control rod has horizontal absorber holes drilled into solid stainless steel wings and uses guide pads (buttons) or no guide pads, rather than upper pins and rollers, to guide control rod motion. Reference 4.1-31 provides additional discussion on the
design of the CR 99 control rod. The control rods can be positioned at 6-in. steps and have a nominal withdrawal and insertion speed of 3 in/sec. The velocity limiter, an engineered safety feature (ESF), is a device which is an integral part of the control rod and protects against the low probability of a rod drop accident. It is designed to limit the free fall velocity and reactivity insertion rate of a control rod so that minimum fuel damage would occur. It is a one-way device, in that control rod scram time is not significantly affected. Control rods are cooled by the core leakage (bypass) flow. The core leakage flow is made up of recirculation flow that leaks through the several leakage flow paths, which are:
(1)The area between fuel channel and fuel assembly nosepiece
- (2)The area between fuel assembly nosepiece and fuel support piece; (3)The area between fuel support piece and core plate; (4)The area between core plate and sh roud; and (5)The bypass flow holes in the fuel assembly nosepiece.
Further details of the control blade design are provided in Section 4.2. 4.1.4 ANALYSIS TECHNIQUES 4.1.4.1 Reactor Internal Components The following computer codes were used for initial design of the reactor internal components. Code descriptions are provided for historical purposes only. Computer codes used for the analysis of the internal components are listed as follows: (1)MASS(2)SNAP (MULTISHELL)(3)GASP (4)NOHEAT (5)FINITE (6)DYSEA (7)SHELL 5(8)HEATER(9)FAP-71(10)CREEP-PL ASTDetailed descriptions of these programs are given in the following sections:
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-8 4.1.4.1.1 MASS (Mechanical Analysis of Space Structure) 4.1.4.1.1.1 Program Description The program, proprietary of the General Electric Company, is an outgrowth of the PAPA (Plate
and Panel Analysis) program originally developed by L. Beitch in the early 1960s. The program is based on the principle of the finite element method. Governing matrix equations are formed in terms of joint displacements using a "stiffness-influence-coefficient" concept originally proposed by L. Beitch (Reference 4.1-2). The program offers curved beam, plate, and shell elements. It can handle mechanical and thermal loads in a static analysis and predict natural frequencies and mode shapes in a dynamic analysis.
4.1.4.1.1.2 Program Version and Computer The GE Nuclear Energy Division is using a past revision of MASS. This revision is identified as revision "0" in the computer production library. The program operates on the Honeywell 6000
computer.
4.1.4.1.1.3 History of Use Since its development in the early 60s, the program has been successfully applied to a wide variety of jet-engine structural problems, many of which involve extremely complex geometries.
The use of the program in the Nuclear Energy Division also started shortly after its development.
4.1.4.1.1.4 Extent of Application
Besides the Jet Engine and Nuclear Energy Divisions, the Missile and Space Division, the Appliance Division, and the Turbine Division of General Electric have also applied the program to a wide range of engineering problems. The Nuclear Energy Division (NED) uses it mainly for piping and reactor internals analyses.
4.1.4.1.2 SNAP (MULTISHELL)
4.1.4.1.2.1 Program Description
The SNAP Program, which is also called MULTISHELL, is the General Electric Code which determines the loads, deformations, and stresses of axisymmetric shells of revolution (cylinders, cones, discs, toroids, and rings) for axisymmetric thermal boundary and surface load conditions. Thin shell theory is inherent in the solution of E. Peissner's differential equations for each shell's influence coefficients. Surface loading capability includes pressure, average temperature, and linear-through-wall gradients; the latter two may be linearly varied over the shell meridian. The theoretical limitations of this program are the same as those of classical theory.
4.1.4.1.2.2 Program Version and Computer
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-9 The current version maintained by the General Electric Jet Engine Division at Evandale, Ohio is being used on the Honeywell 6000 computer in GE/NED. 4.1.4.1.2.3 History of Use The initial version of the Shell Analysis Program was completed by the Jet Engine Division in 1961. Since then, a considerable amount of modification and addition has been made to accommodate its broadening area of application. Its application in the Nuclear Energy Division has a history longer than ten years. 4.1.4.1.2.4 Extent of Application The program has been used to analyze jet engine, space vehicle and nuclear reactor components. Because of its efficiency and economy, in addition to reliability, it has been one of the main shell analysis programs in the Nuclear Energy Division of General Electric.
4.1.4.1.3 GASP 4.1.4.1.3.1 Program Description GASP is a finite element program for the stress analysis of axisymmetric or plane two-dimensional geometries. The element representations can be either quadrilateral or triangular. Axisymmetric or plane structural loads can be input at nodal points. Displacements, temperatures, pressure loads, and axial inertia can be accommodated. Effective plastic stress and strain distributions can be calculated using a bilinear stress-strain relationship by means of an iterative convergence procedure. 4.1.4.1.3.2 Program Version and Computer The GE version, originally obtained from the developer, Professor E. L. Wilson, operates on the Honeywell 6000 computer. 4.1.4.1.3.3 History of Use The program was developed by E. L. Wilson in 1965 (Reference 4.1-3). The present version in GE/NED has been in operation since 1967. 4.1.4.1.3.4 Extent of Application The application of GASP in GE/NED is mainly for elastic analysis of axisymmetric and plane structures under thermal and pressure loads. The GE version has been extensively tested and used by engineers in GE.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-10 4.1.4.1.4 NOHEAT 4.1.4.1.4.1 Program Description The NOHEAT program is a two-dimensional and axisymmetric transient nonlinear temperature analysis program. An unconditionally stable numerical integration scheme is combined with an iteration procedure to compute temperature distribution within the body subjected to arbitrary time- and temperature-dependent boundary conditions. This program utilizes the finite element method. Included in the analysis are the three basic forms of heat transfer, conduction, radiation, and convection, as well as internal heat generation. In addition, cooling pipe boundary conditions are also treated. The output includes temperature of all the nodal points for the time instants by the user. The program can handle multitransient temperature input. 4.1.4.1.4.2 Program Version and Computer The current version of the program is an improvement of the program originally developed by I.
Farhoomand and Professor E. L. Wilson of University of California at Berkeley (Reference 4.1-4). The program operates on the Honeywell 6000 computer. 4.1.4.1.4.3 History of Use The program was developed in 1971 and installed in General Electric Honeywell computer by one of its original developers, I. Farhoomand, in 1972. A number of heat transfer problems related to the reactor pedestal have been satisfactorily solved using the program. 4.1.4.1.4.4 Extent of Application The program using finite element formulation is compatible with the finite element stress-analysis computer program GASP. Such compatibility simplified the connection of the two analyses and minimizes human error. 4.1.4.1.5 FINITE 4.1.4.1.5.1 Program Description FINITE is a general-purpose finite element computer program for elastic stress analysis of two-dimensional structural problems including (1) plane stress, (2) plane strain, and (3) axisymmetric structures. It has provision for thermal, mechanical and body force loads. The materials of the structure may be homogeneous or nonhomogeneous and isotropic or orthotopic. The development of the FINITE program is based on the GASP program. (See
Subsection 4.1.4.1.3.) 4.1.4.1.5.2 Program Version and Computer SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-11 The present version of the program at GE/NED was obtained from the developer J. E. McConnelee of GE/Gas Turbine Department in 1969 (Reference 4.1-5). The NED version is used on the Honeywell 6000 computer. 4.1.4.1.5.3 History of Use Since its completion in 1969, the program has been widely used in the Gas Turbine and the Jet Engine Departments of the General Electric Company for the analysis of turbine components. 4.1.4.1.5.4 Extent of Usage The program is used at GE/NED in the analysis of axisymmetric or nearly axisymmetric BWR internals.
4.1.4.1.6 DYSEA 4.1.4.1.6.1 Program Description The DYSEA (Dynamic and Seismic Analysis) program is a GE proprietary program developed specifically for seismic and dynamic analysis of RPV and internals/building system. It calculates the dynamic response of linear structural system by either temporal modal superposition or response spectrum method. Fluid-structure interaction effect in the RPV is taken into account by way of hydrodynamic mass. Program DYSEA was based on program SAPIV with added capability to handle the hydrodynamic mass effect. Structural stiffness and mass matrices are formulated similar to SAPIV. Solution is obtained in time domain by calculating the dynamic response mode by mode. Time integration is performed by using Newmark's b-method. Response spectrum solution is also available as an option. 4.1.4.1.6.2 Program Version and Computer The DYSEA version now operating on the Honeywell 6000 computer of GE, Nuclear Energy Systems Division, was developed at GE by modifying the SAPIV program. Capability was added to handle the hydrodynamic mass effect due to fluid-structure interaction in the reactor. It can handle 3-Dimensional dynamic problem with beam, trusses, and springs. Both acceleration time histories and response spectra may be used as input. 4.1.4.1.6.3 History of Use The DYSEA program wa s developed in the summer of 1976. It has been adopted as a standard production program since 1977 and it has been used extensively in all dynamic and seismic analysis of the RPV and internals/ building system.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-12 4.1.4.1.6.4 Extent of Application The current version of DYSEA has been used in all dynamic and seismic analysis since its development. Results from test problems were found to be in close agreement with those obtained from either verified programs or analytic solutions.
4.1.4.1.7 SHELL 5 4.1.4.1.7.1 Program Description SHELL 5 is a finite shell element program used to analyze smoothly curved thin shell structures with any distribution of elastic material properties, boundary constraints, and mechanical thermal and displacement loading conditions. The basic element is triangular whose membrane displacement fields are linear polynomial functions, and whose bending displacement field is a cubic polynomial function (Reference 4.1-6). Five degrees of freedom (three displacements and two bending rotations) are obtained at each nodal point. Output displacements and stresses are in a local (tangent) surface coordinate system. Due to the approximation of element membrane displacements by linear functions, the in-plane rotation about the surface normal is neglected. Therefore, the only rotations considered are due to bending of the shell cross section and application of the method is not recommended for shell intersection (or discontinuous surface) problems where in-plane rotation can be significant. 4.1.4.1.7.2 Program Version and Computer A copy of the source deck of SHELL 5 is maintained in GE/NED by Y. R. Rashid, one of the originators of the program. SHELL 5 operates on the UNIVAC 1108 computer. 4.1.4.1.7.3 History of Use SHELL 5 is a program developed by Gulf General Atomic Incorporated (Reference 4.1-7) in 1969. The program has been in production status at Gulf General Atomic, General Electric, and at other major computer operating systems since 1970. 4.1.4.1.7.4 Extent of Application SHELL 5 has been used at General Electric to analyze reactor shroud support and torus. Satisfactory results were obtained.
4.1.4.1.8 HEATER 4.1.4.1.8.1 Program Description HEATER is a computer program used in the hydraulic design of feedwater sparger s and their associated delivery header and piping. The program utilizes test data obtained by GE using full SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-13 scale mockups of feedwater spargers combined with a series of models which represent the complex mixing processes obtained in the upper plenum, downcomer, and lower plenum. Mass and energy balances throughout the nuclear steam supply system are modeled in detail (Reference 4.1-8). 4.1.4.1.8.2 Program Version and Computer This program was developed at GE/NED in FORTRAN IV for the Honeywell 6000 computer. 4.1.4.1.8.3 History of Use The program was developed by various individuals in GE/NED beginning in 1970. The present version of the program has been in operation since January 1972. 4.1.4.1.8.4 Extent of Application The program is used in the hydraulic design of the feedwater spargers for each BWR plant, in the evaluation of design modifications, and the evaluation of unusual operational conditions.
4.1.4.1.9 FAP-71 (Fatigue Analysis Program) 4.1.4.1.9.1 Program Description The FAP-71 computer code, or Fatigue Analysis Program, is a stress analysis tool used to aid in performing ASME-III Nuclear Vessel Code structural design calculations. Specifically, FAP-71 is used in determining the primary plus secondary stress range and number of allowable fatigue cycles at points of interest. For structural locations at which the 3Sm (P+Q) ASME Code limit is exceeded, the program can perform either (or both) of two elastic-plastic fatigue life evaluations: 1)the method reported in ASME Pa per 68-PVP-3, 2) the present method documented in Paragraph NB-3228.3 of the 1971 Edition of the ASME Section III Nuclear Vessel Code. The program can accommodate up to 25 transient stress states of as many as 20 struct ural locations.
4.1.4.1.9.2 Program Version and Computer The present version of FAP-71 was completed by L. Young of GE/NED in 1971 (Reference 4.1-9). The program currently is on the NED Honeywell 6000 computer. 4.1.4.1.9.3 History of Use Since its completion in 1971, the program has been applied to several design analyses of GE BWR vessels.
4.1.4.1.9.4 Extent of Use SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-14 The program is used in conjunction with several shell analysis programs in determining the fatigue life of BWR mechanical components subject to thermal transients. 4.1.4.1.10 CREEP/PLAST 4.1.4.1.10.1 Program Description A finite element program is used for the analysis of two-dimensional (plane and axisymmetric) problems under conditions of creep and plasticity. The creep formulation is based on the memory theory of creep in which the constitutive relations are cast in the form of hereditary integrals. The material creep properties are built into the program and they represent annealed 304 stainless steel. Any other creep properties can be included if required. The plasticity treatment is based on kinematic hardening and von Mises yield criterion. The hardening modulus can be constant or a function of strain. 4.1.4.1.10.2 Program Version and Computer The program can be used for elastic-plastic analysis with or without the presence of creep. It can also be used for creep analysis without the presence of instantaneous plasticity. A detailed description of theory is given in Reference 4.1-11. The program is operative on Univac-1108. 4.1.4.1.10.3 History of Use This program was developed by Y. R. Rashid (Ref. 4.1-11) in 1971. It underwent extensive program testing before it was put on production status.
4.1.4.1.10.4 Extent of Application The program is used at GE/NED in the channel cross section mechanical analysis.
4.1.4.2 Fuel Rod Thermal Analysis Fuel Rod Thermal Design Analyses are performed utilizing the classical relationships for heat transfer in cylindrical coordinate geometry with internal heat generation. Steady state fuel rod thermal-mechanical analyses are performed to assure that fuel rod thermal-mechanical limits (e.g., steady state cladding strain and stress, hydrogen absorption, and, corrosion, etc.) are not exceeded. Abnormal operational transients are also evaluated to assure that the damage limit of 1.0% cladding plastic strain is not violated. Fuel rod analyses were performed with RAMPEX and approved versions of RODEX2, RODEX2A, and COLAPX codes. The fuel rod performance characteristics modeled by the RODEX2 and RODEX2A codes are:
-Gas release-Radial thermal conduction and gap conductance
-Free rod volume and gas pressure calculation s
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-15 - Pellet clad interaction (PCI) - Fuel swelling, densification, cracking, and crack healing
- Cladding creep deformation and irradiation induced growth RODEX2 determines the initial conditions for fuel rod power ramping analysis, performed using RAMPEX.
RODEX2A (Reference 4.1-28) determines the steady state strain, internal pressure, fuel cladding temperature, corrosion, hydrogen absorption, fuel temperature, and the fuel rod internal pressure for creep collapse analysis. This computer code is used to determine gap conductance for transient analysis.
Creep collapse analysis is performed using the COLAPX code.
Section 4.2 presents the fuel rod mechanical design and associated methodology.
4.1.4.3 Reactor Systems Dynamics The analysis techniques and computer codes used in reactor systems dynamics are described in section 4 of Reference 4.1-10. Subsection 4.4.4.6 also provides a complete stability analysis for the reactor coolant system.
Channel and core stability analyses were performed by Framatome ANP, Inc. on a fuel design and cycle specific basis using the STAIF code. A description of the methods employed by the STAIF code is provided in Reference 4.1-20. Using the RAMONA 5 code (References 4.1-26 and 27), FANP also performs transient stability analyses in support of the generation of OPRM setpoints.
4.1.4.4 Nuclear Engineering Analysis A brief summary of principal computer codes used in reactor core design and analysis is
provided below.
4.1.4.4.1 CASMO-4
The CASMO-4 computer code (Reference 4.1-25) was developed by STUDSVIK of America to perform steady state modeling of fuel bundles. CASMO-4 uses deterministic transport methods. At the pin cell level it exclusively uses a collision probability method to collapse the energy nuclear data into multi-group data. At the lattice level, it uses a method of characteristics for the neutron equation solution. CASMO-4, as opposed to CASMO-3G, does not need to do pin cell homogenization to perform a 2-D lattice wide transport calculation. The code is used to model each unique fuel lattice in the reactor to calculate few-group cross sections, bundle reactivities and relative fuel rod powers within a fuel bundle. The effects of conditions such as void, control rod presence, moderator temperature, fuel temperature, soluble boron, etc., are included in the model.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-16 4.1.4.4.2 MICROBURN-B2 The MICROBURN-B2 code (Reference 4.1-25) solves a two-group neutron diffusion equation based on an interface current method. It calculates the burnup chain equation for heavy nuclides and burnable poison nuclides, determines the three-dimensional core nodal power distribution, bundle flow, and void distributions. It also determines pin power distributions and thermal margins to technical specification limits. MICROBURN-B2 is used with CASMO-4. 4.1.4.5 Neutron Fluence Calculations Vessel neutron fluence calculations were performed to determine the azimuthal and axial variation of fluence at the vessel inside surface and at 1/4 T depth. The azimuthal and axial results were synthesized to obtain the fluence profile at the vessel inside surface and 1/4 T depth. The calculations also evaluate vessel fluence at power uprate conditions. Sections 4.3 and 5.3 provide additional detail regarding reactor pressure vessel irradiation. 4.1.4.6 Thermal Hydraulic Calculations XCOBRA (References 4.1-21, 4.1-22, and 4.1-23) calculates the steady state thermal hydraulic performance of a BWR. The code determines the flow and local fluid conditions at various axial positions in the core and represents the core as a collection of discrete parallel channels. The only interaction allowed between channels is the equalization of pressure in the inlet and outlet plenums. This is achieved by allowing the core flow to distribute among the various flow channels until the pressure drop in each channel is equalized. Pressure drop in each channel is determined through the application of two-phase pressure drop correlations and various data which hydraulically characterize the fluid channel. At a given axial position in the core, XCOBRA calculates a core-wide distribution of flow, enthalpy, density, quality, void fraction, and mass velocity.
4.
1.5 REFERENCES
4.1-1 Crowther, R. L. "Xenon Considerations in Design of Boiling Water Reactors," APED-5640, June 1968. 4.1-2 Beitch, L., "Shell Structures Solved Numerically by Using a Network of Partial Panels," AIAA Journal, Volume 5, No. 3, March 1967.
4.1-3 E.L. Wilson, "A Digital Computer Program For the Finite Element Analys is of Solids With Non Linear Material Properties," Aerojet General Technical Memo No. 23, Aerojet General, July 1965.
4.1-4 I.Farhoomand and E. L. Wilson, "No n-Linear Heat Transfer Analysis of Axisymmetric Solids," SESM Report SESM71-6, University of California at Berkeley, Berkeley, California, 1971.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-17 4.1-5 J. E. McConnelee, "Finite-Users Manual," General Electric TIS Report DF 69SL206, March 1969.
4.1-6 R. W. Clough and C. P. Johnson, "A Finite Element Approximation For the Analysis of Thin Shells," International Journal Solid Structures, Vol. 4, 1968.
4.1-7 "A Computer Program For the Structural Analysis of Arbitrary Three-Dimensional Thin Shells," Report No. GA-9952, Gulf General Atomic.
4.1-8 Burgess, A. B., "User Guide and Engineering Description of HEATER Computer Program," March 1974.
4.1-9 Young, L. J., "FAP-71 (Fatigue Analysis Program) Computer Code," GE/NED Design Analysis Unit R. A. Report No. 49, January 1972.
4.1-10 Carmichael, L.A. and Scatena, G. J., "Stability and Dynamic Performance of the General Electric Boiling Water Reactor," APED-5652.
4.1-11 Y. R. Rashid, "Theory Report for Creep-Plast Computer Program," GEAP-10546, AEC Research and Development Report, January 1972.
4.1-12 "Advanced Nuclear Fuels Corporation Critical Power Methodology," ANF-524(P)(A), Revision 2, and Supplement 1, Revision 2, November 1990.
4.1-13 Deleted 4.1-14 Deleted
4.1-15 Deleted
4.1-16 Deleted
4.1-17 Deleted
4.1-18 Deleted
4.1-19 Deleted
4.1-20 "STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain", EMF-CC-074(P)(A) Volumes 1 and 2, July 1994, and Volume 4, August
2000, Siemens Power Corporation, Richland WA 99352.
4.1-21 "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description", XN-NF-80-19(P)(A) Volume 3, Revision 2, Siemens Power Corporation, January 1987.
4.1-22 "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies", XN-NF-79-59(P)(A), November 1979.
4.1-23 "XCOBRA Users Manual", EMF-CC-43, Rev. 3, December 1995.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-18 4.1-24 NEDE-31758P-A, "GE Marathon Control Rod Assembly" GE Nuclear Energy, October 1991. 4.1-25 EMF-2158 (P) (A) "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, " October 1999. 4.1-26 NEDO-32465-A "Reactor Stability Detect and Suppress Solution Licensing Basis Methodology for Reload Applications". 4.1-27 BAW-10255(P), "Cycle Specific Divom Methodology Using the ROMONA5-FA Code," Framatome ANP, Inc. September 2004. 4.1-28 XN-NF-85-74(P)(A). "RODEX2A(BWR) Fuel Thermal-Mechanical Evaluation Model", Exxon-Nuclear Company, Inc. February 1998. 4.1-29 EMF-2209 (P) (A), "SPCB SPCB Critical power Correlation," Framatome ANP, September 2003. 4.1-30 NEDE-33284 Supplement 1 P-A, Revision 1 March 2012, Licensing Topical Report, "Marathon - Ultra Control Rod Assembly" 4.1-31 WCAP-16182-P-A, Revision 1 October 2009, "Westinghouse BWR Control Rod CR 99 Licensing Report - Update to Mechanical Design Limits" THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.1-1, Rev. 54 AutoCAD Figure 4_1_1.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.1-4, Rev. 54 AutoCAD Figure 4_1_4.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.1-5, Rev. 54 AutoCAD Figure 4_1_5.doc THIS FIGURE HAS BEEN RENUMBERED TO 4.2-14 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure renumbered from 4.1-6 to 4.2-14 FIGURE 4.1-6, Rev. 54 AutoCAD Figure 4_1_6.doc THIS FIGURE HAS BEEN RENUMBERED TO 4.2-20 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure renumbered from 4.1-7 to 4.2-20 FIGURE 4.1-7, Rev. 54 AutoCAD Figure 4_1_7.doc THIS FIGURE HAS BEEN RENUMBERED TO 4.2-21 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure renumbered from 4.1-8 to 4.2-21 FIGURE 4.1-8, Rev. 54 AutoCAD Figure 4_1_8.doc Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-1 4.2 FUEL SYSTEM DESIGN The fuel system includes the fuel assembly (channeled fuel bundle) and the portion of the control
rod assembly which extends above the coupling mechanism on the control rod drive. The following
sections discuss the thermal/mechanical design bases, design descriptions, and design evaluations
for the fuel system components. Nuclear design is described in Section 4.3 and thermal hydraulic
design is described in Section 4.4.
4.2.1 Design
Bases
4.2.1.1 Fuel Assembly
The core designs described in Section 4.3 contain one fuel assembly design. The FANP
ATRIUM TM-10 is the primary fuel type loaded into the core. Occasionally, Lead Use Assemblies (LUAs) are also used in the core to provide operating experience with alternative fuel designs.
When used, LUAs are loaded in non-limiting locations in the core.
FANP ATRIUM TM-10 Fuel The mechanical design for the ATRIUM-10 TM assembly is based on compliance with generic mechanical design criteria established by FANP and approved by the NRC in Reference 4.2.6-10.
In accordance with the requirements of the approved mechanical design criteria, the ATRIUM TM-10 mechanical analyses were performed to provide the following assurances.
- 1) The fuel assembly shall not fail as a result of normal operation and AOO's.
- 2) Damage to fuel assemblies shall never prevent control rod insertion when required.
- 3) The number of fuel rod failures is not underestimated for postulated accidents.
- 4) Fuel coolability shall always be maintained.
Mechanical design analyses have been performed to evaluate the cladding stress and strain limits, fretting wear, oxidation, hydriding and crud buildup, fuel rod bowing, differential fuel rod growth, internal hydriding, cladding collapse, and cladding and fuel pellet overheat. The RODEX2, RODEX2A, RAMPEX, and COLAPX codes were used in the mechanical design analyses.
Mechanical analyses have also been performed to evaluate the ATRIUM TM-10 fuel design for Seismic/LOCA loads and for normal shipping and handling. In addition, FANP ATRIUM TM-10 fuel has been evaluated for power uprate conditions.
Results of these analyses and evaluations are discussed in Section 4.2.3.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-2 Lead Use Assemblies (LUA)
Occasionally, LUAs are loaded into the reactor core. The core loading of the LUAs will be such that
the assemblies are not loaded into thermally limiting locations. Mechanical design analyses are
performed for the LUA to evaluate fuel design parameters similar to that provided for the reload fuel.
4.2.1.2 Original Equipment Control Rod Assembly
The design bases for the original equipment control rod assembly are presented in Reference
4.2.6-3. The End-of-Life evaluation for the original equipment control rod assembly was modified in
accordance with PPL's response to IE Bulletin 79-26, Rev. 1. PPL has committed to replacing these
control rod assemblies prior to exceeding a limit of 34% B 10 depletion averaged over the upper one-fourth of the control rod assembly.
4.2.1.3 GE Duralife 160C Control Rod Assembly
The Duralife 160C control rod has been evaluated to assure it has adequate structural margin
under loading due to handling, and normal, emergency, and faulted operating modes. The loads
evaluated include those due to normal operating transients (scram and jogging), pressure
differentials, thermal gradients, seismic deflection, irradiation growth, and all other lateral and
vertical loads expected for each condition. The Duralife 160C control rod assembly design bases
have been reviewed and approved by the NRC (References 4.2.6-4 and 4.2.6-5).
4.2.1.4 GE Marathon Control Rod Assembly
The Marathon control rod has been evaluated to assure it has adequate structural margin under
loading due to handling, and normal, emergency, and faulted operating modes. The loads
evaluated include those due to normal operating transients (scram and jogging), pressure
differentials, thermal gradients, seismic deflection, irradiation growth, and all other lateral and
vertical loads expected for each condition. The Marathon control rod assembly design bases have
been reviewed and approved by the NRC (References 4.2.6-12 and 4.2.6-16).
4.2.1.5 Westinghouse CR 99 Control Rod Assembly
The Westinghouse CR 99 control rod has been evaluated to assure it has adequate structural
margin under loading due to handling, and normal, emergency, and faulted operating modes. The
loads evaluated include those due to normal operating transients (scram and jogging), pressure
differentials, thermal gradients, seismic deflection, irradiation growth, and all other lateral and
vertical loads expected for each condition. The Westinghouse CR 99 control rod assembly design
bases have been reviewed and approved by the NRC (Reference 4.2.6-17).
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-3
4.2.2 General
Design Description A summary of fuel design characteristics is provided in Table 4.2-14 for fuel designs loaded in the
core at SSES.
4.2.2.1 Core Cell
A core cell consists of a control rod assembly and the four fuel assemblies which immediately
surround the control rod. Figure 4.2-15 provides nominal dimensions for a core cell loaded with
FANP ATRIUM TM-10 fuel and a Duralife 160C control rod. Figures 4.2-15A and 4.2.15B provide nominal dimensions for a core cell loaded with FANP ATRIUM TM-10 fuel and a Marathon C+ control rod and a Marathon Ultra - HD control rod, respectively. Figure 4.2-15C provides the nominal
dimensions for a core cell loaded with FANP ATRIUM TM-10 fuel and a Westinghouse CR 99 control rod. These figures illustrate the general layout of a core cell while providing nominal dimensions for
fuel and control rods. A core cell may contain multiple fuel types, regardless of control rod type
utilized.
Each core cell is associated with a four-lobed fuel support piece. Around the outer edge of the
core, certain fuel assemblies are not immediately adjacent to a control rod and are supported by
individual peripheral fuel support pieces.
4.2.2.2 Fuel Bundle
FANP ATRIUM TM-10 Fuel An FANP ATRIUM TM-10 fuel bundle contains 83 full length and 8 part length fuel rods. The 8 part length fuel rods are provided to decrease the two phase pressure loss in the top of the bundle
thereby providing fuel bundle design that is more stable. A central water channel, which displaces
a 3x3 array of fuel rods near the center of the bundle, provides additional moderation within the
bundle thereby enhancing fuel utilization. The water channel is also used to fix the spacer locations
within the fuel bundle and serves as the main structural member connecting the upper and lower tie
plates. A total of 8 spacers are used to maintain fuel rod spacing. Reference 4.2.6-11 provides
detailed discussion of the various components of the ATRIUM TM-10 fuel bundle. Nominal dimensions for the ATRIUM TM-10 fuel bundle are provided in Figure 4.2-15. A schematic of an ATRIUM TM-10 fuel bundle is shown in Figure 4.2-17.
4.2.2.3 Fuel Assembly
A fuel assembly is a fuel bundle including the surrounding fuel channel. The fuel assemblies are
arranged in the reactor core to approximate a right circular cylinder inside the core shroud. Each
fuel assembly is supported by a fuel support piece and the top guide.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-4 The fuel channel enclosing the FANP fuel bundles is fabricated from Zircaloy-4 or Zircaloy-2 and performs these functions: the channel separates flow inside the bundle from the bypass flow
between channels; the channel guides the control rod and provides a bearing surface for it; the
channel provides rigidity for the fuel bundle. A channel fastener attaches the fuel channel to the
fuel bundle using a threaded post on the upper tie plate. Once the channel fastener has been
installed and the fuel assembly has been positioned in the core, the spring on the channel fastener
helps to hold the assembly against the core grid. A schematic of a fuel channel is shown in Figure
4.2-18. The fuel channel for U2C12 has a slightly different design and is shown in Figure 4.2-18-1.
The fuel channel for U1C14 reload fuel and subsequent reloads for Units 1 and 2 has been
changed to a 100 mill wall thickness and is shown in Figure 4.2-18-2, (Reference 4.2.6-9).
Co-resident fuel assemblies use the 80-mil fuel channel design. The difference between the 80-mil
and 100-mil channel is the thickness of the channel wall. Because the inside of the channel is the
same for the two designs, the top 20.1" of the 100-mil channel outer sides that face the upper guide
is reduced to 80-mil. This wall thickness will then allow for the bundle to be in the same lateral
position as the 80-mil channel. To maintain compatibility with an adjacent 80-mil channel fastener
in a control cell, there are also slight modifications to the 100-mil channel fastener to accommodate
the 100-mil channel. Starting with U1C15 the fuel channel mechanical design is based on
Reference 4.2.6-8.
The Advanced Fuel Channel (AFC) has been introduced on fuel for Units 1 and 2 beginning with
the reload for U1C20 and is shown in Figure 4.2-18-3. Co-resident fuel assemblies use the
standard 100-mil fuel channel design. The AFC dimensions are the same as the standard 100-mil
fuel channels with the difference that the AFC has thinned side-walls in the active core region. The
AFC use similar channel fasteners as the standard 100-mil fuel channels. Details regarding the
AFC are available in Reference 4.2.6-8.
U1C14 reload fuel uses the Framatome-ANP FUELGUARD lower tie plate design. The
FUELGUARD lower tie plate has the same outer envelope dimensions as the small hole lower tie
plate, including the same seal/finger springs. The basic difference between the two types of lower
tie plates is the grid. The FUELGUARD grid is constructed from wavy plates with support bars for
the fuel pins to sit on. The small hole design has machined holes to form a grid for the coolant to
flow through and the fuel pins to sit on. Framatome ANP performed flow tests to demonstrate that
the FUELGUARD lower tie plate is hydraulically compatibility with the small hole lower tie plate.
Both lower tie plates basically have the same loss coefficient. Subsequent reloads for Units 1 and 2
will use the Framatome-ANP FUELGUARD lower tie plate design.
A design change of the upper locking mechanism was incorporated into the ATRIUM-10 fuel
assemblies commencing with the Unit 2 Cycle 16 core loading. This design change is referred to
as the "Harmonized Advanced Load Chain" or HALC and replaces the previous load chain design.
The HALC was made to improve manufacturing reliability and improve the ability to remove the
upper tie plate of the fuel assembly. The design change does not affect the thermal, hydraulic or
mechanical performance of the fuel assembly. Subsequent reloads for Units 1 and 2 will use the
HALC design.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-5 Proper assembly orientation in the core is verified by visual inspection and is assured by verification procedures during core loading. Five visual indications of proper fuel assembly orientation exist.
These indications are:
- 1) The channel fastener assemblies are located at one corner of each fuel assembly adjacent to the center of the control rod.
- 2) The orientation boss on the fuel assembly handle points toward the adjacent control rod.
- 3) The channel spacing buttons are adjacent to the control rod passage area.
- 4) The assembly identification numbers which are located on the fuel assembly handles are all readable from the direction of the center of the cell.
- 5) There is cell-to-cell replication.
Proper assembly orientation in the core is shown in Figure 4.2-19.
4.2.2.4 Reactivity Control Assembly
4.2.2.4.1 Original Equipment Control Rod Assembly
The control rod consists of a sheathed cruciform array of commercial grade stainless steel tubes
filled with B 4 C powder. The main structural member of a control rod is made of Type-304 stainless steel and consists of a top handle, a bottom casting with a velocity limiter and control rod drive
coupling, a vertical cruciform center post, and four U-shaped absorber tube sheaths. Control rods
are cooled by core bypass flow.
Stellite rollers with Haynes Alloy 25 pins located at the top and bottom of the control rod help guide
the control rod as it is inserted and withdrawn from the core.
The control rod velocity limiter consists of cast austenitic stainless steel (Grade CF-8) and is an
integral part of the control rod bottom casting. The velocity limiter protects against high reactivity
insertion rate by limiting the control rod velocity in the event of a control rod drop accident.
Reference 4.2.6-3 provides the general design characteristics of the original equipment control rod
assembly. A diagram of this control rod is provided in Figure 4.2-20.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-6 4.2.2.4.2 Duralife 160C Control Rod Assembly The main differences between the Duralife 160C control rods and the original equipment control
rods are:
the Duralife 160C control rods utilize three solid hafnium rods at each edge of the cruciform to replace the three B 4 C rods that are most susceptible to cracking and to increase control rod life; the Duralife 160C control rods utilize improved B 4 C tube material (i.e. high purity type 304 stainless steel vs. commercial purity stainless steel) to eliminate cracking in the remaining
B 4 C rods during the lifetime of the control rod; the Duralife 160C control rods use GE's crevice-free structure design, which includes additional B 4 C tubes in place of the stiffeners, an increased sheath thickness, a full length weld to attach the handle and velocity limiter, and additional coolant holes at the top and
bottom of the sheath; the Duralife 160C control rods utilize low cobalt-bearing pin and roller materials in place of stellite which was previously used (PH13-8 Mo for pins, Inconel X750 for rollers);
the Duralife 160C control rod handles are longer by approximately 3.1 inches. The extended handle provides lateral support against the top grid and facilitates fuel moves
within the reactor vessel during refueling outages; the Duralife 160C control rods are roughly 10% to 15% heavier (depending on the velocity limiter design) as a result of the design changes described above; and the Duralife 160C control rod velocity limiter material is cast austenitic stainless steel grade CF-3.
References 4.2.6-4 and 4.2.6-5 provide additional discussion on the design of the various Duralife
control rod assemblies, including the features of the Duralife 160C inserted in the SSES Units. A
cross section of the Duralife 160C control rod is shown in Figure 4.2-15, and a diagram of this
control rod is provided in Figure 4.2-21.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-7 4.2.2.4.3 GE Marathon Control Rod Assembly The main difference between the Marathon control rod and the Duralife design control rods are:
the absorber tube and sheath arrangement of the Duralife designs is replaced with an array of square tubes resulting in reduced weight and increased absorber volume; the full length center tie rod is replaced with a segmented tie rod which also reduces weight.
References 4.2.6-12 and 4.2.6-16 provide additional discussion on the designs of the Marathon
control rod assemblies. A cross section of the Marathon C+ control rod is shown in Figure 4.2-15A.
A cross section of the Marathon Ultra HD Control Rod is shown in Figure 4.2-15B. A diagram of the
control rod is provided in Figure 4.2-22.
Replacement Marathon control rods may use a modified handle assembly that eliminates pins and
rollers present in the earlier design. Figure 4.2-22 is applicable to the rollerless design; the only
difference is that the small circles depicting the rollers would be removed from the diagram for the
rollerless design. The overall shape and dimensions of the upper handle remains unchanged.
4.2.2.4.4 Westinghouse CR 99 Control Rod Assembly
The Westinghouse BWR CR 99 control rod design is comparable to that of the GE design. Both
the Westinghouse and GE control blade designs have a coupling socket, velocity limiter, coupling
release handle, 143-inch active absorber zone, B 4 C as their main absorber material and a handle.
The overall length of the Westinghouse CR 99 control rod is the same as the OE GE control rod
design at 173-inches. The main difference in the design between the Westinghouse and GE
control rod is that the Westinghouse CR 99 control blade has horizontal absorber holes drilled in
solid stainless steel wings. Reference 4.2.6-17 provides additional discussion on the design of the
CR 99 control rod. A diagram of the CR 99 control rod is provided in Figure 4.2-23.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-8
4.2.3 Design
Evaluations 4.2.3.1 Fuel Design Evaluations
FANP ATRIUM TM-10 Fuel For the FANP ATRIUM TM-10 fuel, the design is such that adequate margins to fuel mechanical design limits (e.g., centerline melting temperature, transient strain, etc.) are assured for all
anticipated operational occurrences (AOOs) throughout the life of the fuel as demonstrated by the
fuel mechanical design analyses (References 4.2.6-10 and 4.2.6-11), provided that the steady state
fuel rod power remains within the power history assumed in the analyses. The design steady state
power history for the FANP ATRIUM TM-10 fuel is shown in Reference 4.2.6-11 and is incorporated into the Unit/Cycle specific Core Operating Limits Report (COLR) as an operating limit. The
operating limit may be in terms of planar or pellet exposure. ARTS has been implemented for Unit
1 and Unit 2. The COLR has a flow dependent LHGR multiplier and a power dependent multiplier, which are used to adjust the LHGR limit at off-rated conditions to assure that design limits are not
exceeded. The mechanical analyses support a maximum assembly average exposure of 49,400
MWD/MTU for fresh ATRIUM TM-10 fuel loaded in Cycles 12 and 13 on Unit 1 and Cycles 10 and 11 on Unit 2. Commencing with Unit 1 Cycle 14 and Unit 2 Cycle 12, the mechanical analyses for
fresh ATRIUM TM-10 assemblies support a maximum assembly average exposure of 54,000 MWd/MTU.
For U2C14, ARTS has been implemented. With ARTS the need for the FRTP/MFLPD adjustment
factor has been eliminated from the U2C14 Technical Specifications and the COLR. For U2C14, the COLR has a flow dependent LHGR multiplier and a power dependent multiplier, which are used
to adjust the LHGR limit at off-rated conditions.
FANP has evaluated the performance of ATRIUM TM-10 fuel assemblies under Susquehanna Seismic LOCA conditions. For this evaluation, maximum loads and/or stresses were calculated for
the fuel components under an acceleration load equivalent to a maximum dynamic load which
bounds the allowable bending moment in BWR/4 80 mil and 100 mil fuel channels, (Reference
4.2.6-9). The large margin that resulted from these analyses shows that the ATRIUM TM-10 fuel assembly with an 80 mil or 100 mil fuel channel demonstrates adequate structural integrity in the
Susquehanna Units under Seismic LOCA conditions. With regard to assembly liftoff, the net force
for the ATRIUM TM-10 fuel assembly was found to be downwards. Starting with U1C15 the fuel channel mechanical design is based on Reference 4.2.6-8.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-9 4.2.3.2 Results of Control Rod Assembly Design Evaluations 4.2.3.2.1 Original Equipment Control Rod Assembly
The original equipment control rod assembly design evaluations are discussed in Reference
4.2.6-3. Subsequent to the completion of the above referenced evaluations, a new failure
mechanism was identified for the original equipment control rod assembly. IE Bulletin 79-26 Rev-1
discusses this failure mechanism and recommends a reduction in the end-of-life criteria for the
original equipment control rod assembly. PPL committed to replacing the original equipment control
rod assemblies in accordance with this IE bulletin.
4.2.3.2.2 Duralife 160C Control Rod Assembly
The Duralife 160C control rod stresses, strains, and cumulative fatigue have been evaluated and
result in an acceptable margin to safety. The control rod insertion capability has been evaluated
and found to be acceptable during all modes of plant operation within the limits of plant analyses.
The Duralife 160C control rod coupling mechanism is equivalent to the original equipment coupling
mechanism, and is therefore fully compatible with the existing control rod drives in the plant. In
addition, the materials used in the Duralife 160C are compatible with the reactor environment. The
Duralife 160C control rods are roughly 10% to 15% heavier than the original equipment control rod
assembly, depending on the velocity limiter design utilized. The impact of the increased weight of
the control rods on the seismic and hydrodynamic load evaluation of the reactor vessel and
internals has been evaluated and found to be negligible.
With the exception of the crevice-free structure and the extended handle, the Duralife 160C control
rod is equivalent to the NRC approved Hybrid I Control Rod Assembly (Reference 4.2.6-4). The
mechanical aspects of the crevice-free structure were approved by the NRC for all control rod
designs in Reference 4.2.6-5. A neutronics evaluation of the crevice-free structure for the Duralife
160C design was performed by GE using the same NRC approved nuclear interchangeability
evaluation methodology as described in Reference 4.2.6-4. These calculations were performed for
the original equipment control rods and the Duralife 160C control rods assuming an infinite array of
FANP 9x9-2 fuel. The Duralife 160C control rod has a slightly higher worth than the original
equipment design, but the increase in worth is within the criterion for nuclear interchangeability.
The increase in rod worth has been taken into account in the appropriate reload analyses.
In Reference 4.2.6-4, the NRC approved the Hybrid I (Duralife 160C) control rod which weighs less
than the D lattice control rod. The basis of the Control Rod Drop Accident analysis continues to be
conservative with respect to control rod drop speed since the Duralife 160C control rod (including the extended handle, crevice free structure, and heavier velocity limiter) weighs less than the D
lattice control rod, and the heavier D lattice control rod drop speed is used in the
analysis. In addition, GE performed scram time analyses and determined that the Duralife 160C
control rod scram times are not significantly different than the original equipment control rod scram
times.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-10 Also, the scram speeds are monitored in the plant to assure compliance with safety analysis assumptions and technical specification limits.
IE Bulletin 79-26, Rev. 1 was issued to address B 4 C rod cracking and subsequent loss of boron in GE original equipment control rods. The Duralife 160C control rod design contains solid hafnium
absorber rods in locations where B 4 C tubes have historically failed. The remaining B 4 C rods are manufactured with an improved tubing material (high purity stainless steel vs. commercial purity
stainless steel), thus, boron loss due to cracking is not expected to occur.
Due to the control rod design, IE Bulletin 79-26, Rev. 1 does not apply to Duralife 160C control
rods. However, PPL plans to continue tracking the depletion of each control rod and discharge any
control rod prior to a ten percent loss in reactivity worth.
4.2.3.2.3 GE Marathon Control Rod Assembly
The form, fit and function of the Marathon control rod design are equivalent to the original
equipment control rods used at Susquehanna. Reference 4.2.6-12 documents NRC acceptance of
the GE Marathon control rod mechanical design.
The control rod stresses, strains, and cumulative fatigue were evaluated by GE Nuclear and result
in acceptable margins to safety. The control rod insertion capability was evaluated and found to be
acceptable during all modes of plant operation within the limits of plant analyses. In addition, the
coupling mechanism is fully compatible with the existing control rod drives in the plant. The
materials used in the Marathon control rods were also evaluated and are compatible with the
reactor environment. The Marathon control rods are approximately the same weight as the original
equipment control rods and, therefore, there is no impact on the seismic and hydrodynamic load
evaluation for the reactor vessel and internals. With lighter weight than the D160 control rods and
envelope dimensions less than or the same as the original equipment, the Marathon design is
compatible with existing NSSS hardware and there is no change in scram performance or drop
time.
Neutronics evaluations of the Marathon control rods by GE Nuclear using the methodologies
described in Reference 4.2.6-12 indicate the C lattice Marathon design for Susquehanna slightly
exceeds the +5% beginning-of-life reactivity worth constraint relative to the original equipment all
B 4 C design. Therefore, the effect of the increased reactivity worth on plant analyses had to be considered. The increased reactivity worth was found to not adversely impact normal operation, and is considered in the analysis of abnormal operational occurrences, infrequent events, or
accidents. The Marathon Ultra - HD control rod design satisfies the +5% beginning-of-life reactivity
worth constraint, (Reference 4.2.6-16).
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-11 The Marathon control rods used improved materials and contain significant design improvements to eliminate cracking and the associated loss of Boron experienced by the original equipment. GE
defines the end of life as a 10% reduction in cold reactivity worth in any 1/4 axial segment relative to
the initial undepleted state of the original equipment control rods. PPL will track the depletion of the
Marathon control rods and discharge any control rod prior to reaching the defined end of life, or
provide technical justification for its continued use.
4.2.3.2.4 Westinghouse CR 99 Control Rod Assembly
The form, fit and function of the Westinghouse CR 99 control rod design are equivalent to the
OE control rods used at Susquehanna. Reference 4.2.6-17 documents NRC acceptance of the
CR 99 control rod mechanical design.
The CR 99 control rod stresses, strains, and cumulative fatigue were evaluated by Westinghouse
and result in acceptable margins to safety. The CR 99 insertion capability was evaluated and found
to be acceptable during all modes of plant operation within the limits of plant analyses. The CR 99 coupling mechanism is fully compatible with the existing control rod drives in the plant. The
materials used in the CR 99 control rods were also evaluated and are compatible with the reactor
internals and the reactor environment. The CR 99 control rods are similar in nominal weight of the
OE control rods and, therefore, there is no impact on the seismic and hydrodynamic load evaluation
for the reactor vessel and internals. Scram speeds and settling times in the reactor are not
adversely affected by the CR 99 control rods. The CR 99 velocity limiter design is identical to the
design of the OE control rods and meets the assumption for the control rod drop accident.
The total worth of the CR 99 control rod is within +/-5% of the OE control rod. There is no
negative impact to shutdown margin and minimal impact on LPRM detector indications. The
nuclear end of life criteria is maintained as 10% reactivity worth decrease relative to the OE
control rod (Reference 4.2.6-17).
The CR 99 use of an improved high density absorber material, which is less sensitive to both
powder densification and absorber swelling due to neutron absorption reactions, minimizes the
possibility of absorber swelling causing contact with the surrounding stainless steel and
contributing stress. The CR 99 use of AISI 316L stainless steel, with its better resistance to fast
neutron irradiation assisted stress corrosion cracking (IASCC), also reduces the potential for
control blade cracking. SSES will track the depletion of the CR 99 control rod and discharge
any control rods prior to reaching the defined end of life, or provide technical justification for its
continued use.
4.2.4 Testing
and Inspection
4.2.4.1 Fuel Hardware and Assembly
Framatome - ANP, Inc (FANP) has developed Quality Control Standards for manufacturing, testing, and inspection of FANP components and fuel bundles. Details regarding FANP
manufacturing, testing, and inspection processes are available in Reference 4.2.6-6.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-12 On-site inspection of all new fuel bundles, fuel channels, and control rods is performed prior to installation into a Sus quehanna Unit. These inspections are controlled by plant procedures. The procedures were developed bas ed on guidelines provided by the fuel, channel, and control rod suppliers for rec eipt inspection and include acceptance criteria which are verified for each fuel bundle, fuel channel, or control rod.
The fuel channel management practices in place at Susquehanna are consistent with the recommendations contained in GE SIL 320, 'Recommendations for Mitigation of the Effects of Fuel Channel Bowing'. In addition, PPL will only use fuel channels for only one fuel bundle lifetime and will not reuse them. The fuel c hannel management practices are continuously reviewed against plant operation and industry practices.
4.2.4.2 Enrichment, Burnable Poison, and Absorber Rod Concentrations FANP has established adequate measures, in accordance with Reference 4.2.6-6, to assure that nuclear materials of v arying enrichment and form are pos itively identified and physically segregated as required to assure no inadvertent intermixing of enrichment forms. These measures include, as appropriate, identification of storage and processing containers, gamma sc an verific ation of powder, nuclear rod assay, analy tical ex aminations, in-process inspections, cleanouts of processing equipment between enrichments, administrative controls on the handling of materials, and audits of processing and product.
FANP fuel pellets are manufactured in accordance with approved procedures and are controlled by Product Design Specifications which define the allowable concentration tolerances and confidence levels required to verify enrichment and burnable poison concentrations.
General Electric (GE) supplied the original equipment control rods for both Sus quehanna Units and is the supplier of the Duralife-160C and Marathon replacement control rods. The absorber materials, boron carbide and hafnium, are c ertified by GE to meet GE Material Specific ations. The isotopic B 10 content and boron content is verified for eac h powder lot received by General Electric.
All boron c arbide absorber rod assemblies are s ubjected to a leakage test to insure abs orber rod integrity. GE performs analysis of hafnium absorber rod lots to insure chemical composition is in c onformance with GE Material Specifications.
Wes tinghouse supplied the CR 99 control rods for use at both Susquehanna Units. The boron c arbide absorber material is certified by Westinghouse to meet Westinghouse Material Specifications. The isotopic B 10 content and boron content is verified for each powder lot receiv ed by Westinghous
- e. Eac h CR 99 control rod blade is leak tested with helium per the Westinghouse Materials Specifications.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-13 4.2.4.3 Surveillance, Inspection, and Testing PPL has a fuel reliability program that includes fuel performance monitoring and fuel failure
response. On-line fuel performance monitoring is conducted to determine whether there is a fuel
failure and may include evaluation of the general location of the failed assembly, the number of fuel
assemblies suspected, when the failure occurred, and the approximate exposure of the failed
assembly. Determination of this information prior to refueling allows preparation for changes in the
following cycle's core design. In addition, control rod sequence exchanges and full power control
rod patterns can be developed to minimize the offgas release from the failed rod(s) and stress on
the suspect assembly during power maneuvering.
On-line fuel performance monitoring is performed at the Susquehanna station by periodic
evaluation of pretreatment offgas activity and/or reactor coolant samples. Verification of failed fuel
is made by periodic evaluation of the pretreatment xenon and krypton offgas activity and reactor
water cleanup system iodine and cesium activity. The general location of the failed fuel assembly
is, typically, identified by control rod motion testing and monitoring of the pretreatment offgas
activity. Identification of the exact assembly may be performed by sipping or ultrasonic testing (UT)
of the suspect assemblies.
Post-irradiation fuel failure evaluations are, typically, performed to determine the exact fuel rod
location within the assembly and the root cause of a fuel rod failure. The exact location of the failed
rod may be determined by UT or eddy current testing. Root cause evaluations may include review
of manufacturing and inspection records, visual examination of the failure location, and destructive
examination of the failed fuel rod.
Post-irradiation inspection programs have been developed by FANP to evaluate fuel design
performance. Reference 4.2.6-11 discusses the FANP inspection and surveillance program for
irradiated ATRIUM TM-10 fuel.
4.2.5 Operating
and Developmental Experience
FANP ATRIUM TM-10 fuel has been utilized at SSES beginning with Unit 1 Cycle 11 and Unit 2 Cycle 9.
PPL continually tracks the performance of all fuel in the Susquehanna Units in an effort to identify
indications of potential fuel rod failures.
Prior to the implementation of a mechanical fuel design into either Susquehanna Unit, that
introduces features not currently in other operating plants, a plan will be developed to evaluate the
performance of this fuel design in the Susquehanna Units. This plan may include pre and post irradiation fuel assembly characterization, visual inspection, power maneuvering evaluations, fuel
clad corrosion evaluations, and UT inspections.
PPL occasionally participates in Lead Use Assembly programs. These programs allow the
company to evaluate and gain operating experience with new fuel designs.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-14
4.2.6 References
4.2.6-1 Deleted 4.2.6-2 Deleted 4.2.6-3 "BWR/4 and BWR/5 Fuel Design," NEDE-20944(P), General Electric Company, October 1976, and Letter from Olan D. Parr (NRC) to Dr. G. G. Sherwood (GE), "Review of General Electric Topical Report NEDE-20944-P, BWR/4 and BWR/5
Fuel Design (NEDO-20944 Non-Proprietary Version)", September 30, 1977. 4.2.6-4 "Safety Evaluation of the General Electric Hybrid I Control Rod Assembly for the BWR 4/5 C Lattice, "NEDE-22290-A, General Electric Company, September 1983, and Supplement 1, General Electric Company, July 1985. 4.2.6-5 "Safety Evaluation of the General Electric Duralife 230 Control Rod Assembly, "
NEDE-22290-A, Supplement 3, General Electric Company, May 1988. 4.2.6-6 "Nuclear Fuel Business Group Quality Management Manual, "NFQM, Rev. 0, Framatome-ANP,U.S. Version, June 2002.
4.2.6-7 Deleted 4.2.6-8 "Mechanical Design for BWR Fuel Channels", EMF-93-177 (P) (A) Rev. 1, August 2005. 4.2.6-9 "Mechanical Design for BWR Fuel Channels," EMF-93-177 (P)(A) and Supplement 1, Siemens Power Corporation, August 1995. 4.2.6-10 "Generic Mechanical Design Criteria for BWR Fuel Designs," ANF-89-98(P)(A), Rev.
1 and Rev. 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995. 4.2.6-11 "Mechanical Design Evaluation for Siemens Power Corporation ATRIUM TM BWR Reload Fuel, " EMF-95-52(P), Rev. 2, Siemens Power Corporation -
Nuclear Division, December 1998. 4.2.6-12 "GE Marathon Control Rod Assembly, " NEDE-31758P-A, GE Nuclear Energy, October 1991.
4.2.6-13 Deleted
4.2.6-14 Deleted
4.2.6-15 Deleted Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-15 4.2.6-16 NEDE - 33284 Supplement 1 P-A, Revision 1 March 2012, Licensing Topical Report, "Marathon Ultra Control Rod Assembly". 4.2.6-17 WCAP-16182-P-A, Revision 1 October 2009, "Westinghouse BWR Control Rod CR 99 Licensing Report - Update to Mechanical Design Limits".
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-1, Rev. 54 AutoCAD Figure 4_2_1.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-2, Rev. 54 AutoCAD Figure 4_2_2.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-3, Rev. 54 AutoCAD Figure 4_2_3.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-4, Rev. 54 AutoCAD Figure 4_2_4.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-5, Rev. 54 AutoCAD Figure 4_2_5.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6, Rev. 54 AutoCAD Figure 4_2_6.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-7, Rev. 54 AutoCAD Figure 4_2_7.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8, Rev. 54 AutoCAD Figure 4_2_8.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-9, Rev. 54 AutoCAD Figure 4_2_9.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-10, Rev. 54 AutoCAD Figure 4_2_10.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-11, Rev. 54 AutoCAD Figure 4_2_11.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-12, Rev. 54 AutoCAD Figure 4_2_12.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-13, Rev. 54 AutoCAD Figure 4_2_13.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-14, Rev. 55 AutoCAD Figure 4_2_14.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-16, Rev. 55 AutoCAD Figure 4_2_16.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6a, Rev. 48 AutoCAD Figure 4_2_6a.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6b, Rev. 48 AutoCAD Figure 4_2_6b.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6c, Rev. 48 AutoCAD Figure 4_2_6c.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8a, Rev. 54 AutoCAD Figure 4_2_8a.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8b, Rev. 54 AutoCAD Figure 4_2_8b.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8c, Rev. 54 AutoCAD Figure 4_2_8c.doc
For Information Only SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-1 4.3 NUCLEAR DESIGN The nuclear design of the initial cores for Susquehanna is described in References 4.3-1, 4.3-2, and 4.3-3. This section incorporates much of the general nuclear design information in Reference 4.3-1 and presents detailed design information for reload cores. 4.3.1 Design Bases Nuclear design bases fall into two categories: safety design bases and core performance design bases. Safety design bases are required by the General Design Criteria to ensure safe operation of the core. Core performance design bases are required to meet power production objectives. 4.3.1.1 Safety Design Bases Safety design bases protect the nuclear fuel from damage which would result in a release of radioactivity, representing an undue risk to the health and safety of the public. Safety design bases are listed below.
1)The core shall be capable of being rendered subcritical at any time or core condition wi th the highest worth control rod fully withdrawn.
2)The void coefficient shall be negative over the entire operating range.
3)Technical specification limits on Linear Heat Generation Rate (LHGR), Minimum CriticalPower Ratio (MCPR), and the Average Planar Linear Heat Generation Rate (APLHGR), shall not be exceeded during steady state operation.
4)The nuclear characteristics of the design shall not exhibit any tendency toward divergen t operation.5)Reload fuel lattice enrichments shall be such that the nuclear design bases are met for the new fuel storage racks (section 9.1.1.1.1.2) and spent fuel storage (section 9.1.2.1.1.2).4.3.1.2 Plant Performance Design Bases 1)The core design shall have adequate excess reactivity to reach the desired cycle length
.2)The core design shall be capable of operating without exceeding technical specification lim its.3)The core and fuel design and the reactivity control system shall allow continuous, stableregulation of reactivity
.4)The core and fuel design shall have adequate reactivity feedback to facilitate normaloperation.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-2 4.3.2 Description A general description of BWR nuclear characteristics is provided in Reference 4.3-1. A summary of reactor core characteristics for Susquehanna is listed in Table 4.3-1. 4.3.2.1 Nuclear Design Description The nuclear design of Susquehanna is both unit and cycle specific. A detailed description of the initial core nuclear design is available in Reference 4.3-1. Susquehanna Steam Electric Station Units 1 and 2 operate at power conditions in Table 4.3-1 with increased core flow. Fuel bundle and core reload designs have been developed using NRC approved methods. 4.3.2.1.1 Core Composition The core contains 764 fuel assemblies arranged in a conventional scatter loaded pattern. Typically, the lowest reactivity fuel assemblies are placed in the peripheral region of the core. SSES uses FANP ATRIUM TM-10 fuel as the primary reload fuel mechanical design. In addition, a limited number of Lead Use Assemblies (LUAs) may be loaded to evaluate new fuel designs. Detailed core compositions are presented in Tables 4.3-2 and 4.3-3 for Units 1 and 2, respectively. The core loading patterns for both units are shown in Figures 4.3-1 and 4.3-2. 4.3.2.1.2 Fuel Bundle Nuclear Design Reference 4.3-1 describes the first core bundle designs and related fuel nuclear properties. Reload fuel bundle design descriptions are presented below. The burnup dependent behavior of certain nuclear properties is primarily a function of enrichment and does not change significantly with bundle mechanical design. These characteristics include Uranium depletion and Plutonium buildup, fission fraction, delayed neutron fraction, and neutron lifetime. Figures 4.3-3 through 4.3-7 show the typical response of these characteristics with burnup for an enriched reload fuel bundle lattice. Several FANP ATRIUM TM-10 bundle designs are in use at SSES. Each design may have a unique axial enrichment distribution, radial enrichment distribution, or burnable absorber loading. Figure series 4.3-8 shows the nominal axial zoning for the various fuel bundles used in the reload cores. Figure series 4.3-9 shows the nominal radial enrichment distributions for the various lattice types used in the fuel bundles. Table 4.3-6 lists the fuel types currently used and the associated 4.3-8 series figure numbers. Table 4.3-6 also lists the lattice numbers used by each fuel type. Table 4.3-7 lists the lattice currently used and the associated 4.3-9 series figure numbers. 4.3.2.2 Power Distributions This section presents typical power distributions for SSES reload cores. Typical local, core radial, and core axial power distributions for the initial core are described in Reference 4.3-1. The core is designed such that the resulting power distributions meet the thermal limits identified in the plant Technical Specifications. The primary criteria for thermal limits are the Maximum Linear Heat Generation Rate (MLHGR) and the Minimum Critical Power Ratio (MCPR). In addition, a Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit is applied to the plant.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-3 Each of these parameters is a function of the core 3-D power distribution and the local rod-to-rod power distribution. Design calculations are performed to ensure that the core meets thermal limits and to demonstrate that the power distributions comply with the cycle design envelope.
The local peaking factor is defined as the ratio of the power density in the highest power rod in a lattice to the average power density in the lattice. Local effects on Critical Power Ratio are characterized by F-effective. Both the local peaking factor and F-effective have associated target values which typically satisfy the design envelope. Gross power peaking in the core is defined as the ratio of the maximum power density in any axial segment of any bundle in the core to the average core power density. Design allowances are included in the design stage to ensure that thermal limits are met. During plant operation, the power distributions are measured by the in-core instrumentation system and thermal margins are calculated by the core monitoring system.
4.3.2.2.1 Local Power Distribution
The local rod-to-rod power distribution and associated F-effective distribution are a direct function of the lattice enrichment distribution. Near the outside of the lattice where thermal flux peaks due to interbundle water gaps, low enrichment fuel rods are utilized to reduce power peaking. Closer to the center of the bundle, higher enrichment rods are used to increase power peaking and flatten the bundle power distribution. In addition, the water rods (or water channels) in the center of the lattice increase thermal flux and cause more power to be produced in the center of the lattice. The combination of enrichment and water channels results in a relatively flat power distribution.
To control bundle reactivity, Gd 2 O 3 is utilized as a burnable absorber. Power is suppressed in gadolinia bearing fuel rods early in bundle life. As gadolinia is depleted, power in these rods initially increases, then decreases as fuel is depleted. Local power distributions are calculated using licensed methodology described in Section 4.3.3.
Figure 4.3-11-1 shows bundle reactivity (k) as a function of void fraction and burnup for an FANP ATRIUM TM-10 fuel assembly dominant lattice. At low exposure, reactivity is higher for lower void fractions. As exposure increases the curves cross, largely due to the effect of void history and the increase in plutonium buildup.
Figures 4.3-11-2 to 4.3-11-4 show typical unrodded local power distributions for an FANP ATRIUM TM-10 fuel assembly dominant lattice as a function of burnup with a constant void fraction. Figures 4.3-11-2, 4.3-11-5, and 4.3-11-6 show typical unrodded local power distributions for a fresh ATRIUM TM-10 fuel dominant lattice as a function of void fraction at BOC. Figure 4.3-11-7 shows the typical response of the unrodded maximum local peaking factor as a function of void fraction and burnup.
4.3.2.2.2 Radial Power Distribution The integrated bundle power, commonly referred to as the radial power, is the primary factor for determining MCPR. At rated conditions the MCPR is directly proportional to the radial power. The radial power distribution is a function of the control rod pattern in the core, the fuel bundle type and loading pattern, and void distribution. Radial power is calculated using the licensed methodology described in Section 4.3.3.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-4 4.3.2.2.3 Axial Power Distribution Axial power distributions in a BWR are a function of control rod position, steam voids, axial gadolinia distribution, and the exposure distribution. Voids tend to skew power toward the bottom of the core; bottom entry control rods reduce the power in the bottom of the core; and the axial gadolinia distribution assists in flattening the power in the bottom of the core. Since the void distribution is primarily determined by the power shape, the two means available for axial power shape optimization are the control rods and gadolinia. Typically, the core axial power shape is bottom peaked at BOC and becomes top peaked at EOC. Axial power shapes are calculated using the licensed methodology described in Section 4.3.3. 4.3.2.2.4 Power Distribution Measurements Power distribution measurement methodology and measurement uncertainties are described in References and 4.3-10 and 4.3.13. 4.3.2.2.5 Power Distribution Accuracy The accuracy of calculated power distributions is discussed in References 4.3-10 and 4.3-13. 4.3.2.3 Reactivity Coefficients Reactivity coefficients are differential changes in reactivity produced by differential changes in core conditions. These coefficients are useful in calculating the response of the core to varying plant conditions. The initial condition of the core and the postulated initiating event determine which of the coefficients are significant in evaluating core response. The dynamic behavior of BWRs over all operating states can be characterized by three reactivity coefficients. These coefficients are the Doppler coefficient, the moderator temperature coefficient, and the void coefficient. The Power coefficient is also associated with a BWR; however, this coefficient is the combination of the Doppler and void coefficients in the operating range. Reactivity coefficients are calculated using the licensed methods described in Section 4.3.3. 4.3.2.3.1 Void Coefficient The most important reactivity coefficient in a BWR is the void coefficient. The void coefficient must be large enough to prevent power oscillation due to spatial xenon changes, but it must be small enough that pressurization transients do not limit plant operation. The void coefficient inherently flattens the radial power distribution during normal operation and provides enhanced reactor control through the void feedback mechanism. The overall void coefficient is always negative over the complete operating range since the BWR design is typically undermoderated. Void formation changes reactivity by reducing the amount of water available for neutron moderation, thus increasing neutron leakage. Typical values for the void coefficient are listed in Table 4.3-4.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-5 4.3.2.3.2 Moderator Temperature Coefficient The moderator temperature coefficient (MTC) is the least important of the reactivity coefficients since its effect is limited to a very small portion of the reactor operating range. Once the reactor reaches the power production range, boiling begins and the MTC remains essentially constant.
Like the void coefficient, the moderator coefficient is associated with the amount of neutron moderation in the water. The MTC is negative during power operation; however, under cold
conditions beginning soon after BOC, the MTC may become slightly positive. The range of values of MTCs in reload lattices does not include any that are significant from a safety point of view. Typical values for the MTC are listed in Table 4.3-4. The small magnitude of this coefficient, relative to that associated with steam voids, combined with the long time-constant associated with heat transfer from fuel to coolant, makes the reactivity contribution of a change in moderator temperature insignificant during rapid transients. 4.3.2.3.3 Doppler Temperature Coefficient The Doppler Temperature coefficient (DTC) is the change in reactivity due to a change in fuel temperature. This change in reactivity occurs due to the broadening of the fuel resonance absorption cross sections as temperature increases. The DTC is primarily a measure of the Doppler broadening of U238 and Pu240 resonance absorption peaks. An increase in fuel temperature increases the effective resonance absorption cross section of the fuel and produces a corresponding reduction in reactivity. The Doppler coefficient changes as a function of core life representing the combined effects of fuel temperature reduction with burnup and the buildup of Pu240. Typical values for the Doppler coefficient are listed in Table 4.3-4. 4.3.2.3.4 Power Coefficient The power coefficient is determined from the composite of all the significant individual sources of reactivity change associated with a differential change in reactor power. This coefficient assumes constant xenon. Typical values for the power coefficient may be obtained from Reference 4.3-1 for the initial cores. 4.3.2.4 Control Requirements The core and fuel design in conjunction with the reactivity control system provide a stable system for BWRs. The control rod system is designed to provide adequate control of the maximum excess reactivity anticipated during the equilibrium fuel cycle operation. Since fuel reactivity is a maximum and control rod worth is a minimum at ambient temperature, the shutdown capability is evaluated assuming a cold, xenon free core. The safety design basis requires that the core, in its maximum reactivity condition, shall be subcritical with all control rods inserted except with the highest worth rod completely withdrawn. This limit allows control rod testing at any time in core life and assures that the reactor can be made subcritical by control rods alone. The typical behavior of hot excess reactivity as a function of cycle exposure for SSES Units 1 and 2 is shown in Figure 4.3-12.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-6 4.3.2.4.1 Shutdown Reactivity Core Shutdown Margin calculations are performed to assess whether the basic criterion for reactivity control is met by the reload design. This criterion requires that the core, under cold, no xenon conditions, must be subcritical with the highest worth control rod fully withdrawn and all other rods fully inserted. SSES Technical Requirements Manual requires a shutdown margin of at least 0.38% k/k. The shutdown margin requirement is based on the uncertainties associated with the statistical variance of cold criticality calculations at a given exposure, plus a manufacturing uncertainty. The manufacturing uncertainty results from the fact that the calculated highest worth control rod may not be the highest worth rod in reality due to the stackup of manufacturing tolerances in a control cell.
Core Shutdown margin is very dependent on bundle and core designs and is a function of core exposure. Gadolinia loading, enrichment loading, and core loading all significantly affect core and local cell reactivity as a function of exposure. As a result, shutdown margin must be evaluated throughout the expected cycle operation to assure adequate margin to Technical Specification requirements. For design purposes, an additional uncertainty is added to the Technical Specification value to account for prediction uncertainties.
Shutdown margin is calculated as a function of cycle exposure in the following manner:
%100*)E (bias)E (k)E (bias)E (k 1)E (SDM eff eff where; SDM(E) = core shutdown margin (%k/k) at cycle exposure E, K eff(E) = core k-effective at cycle exposure (E) with all rods in except the strongest worth rod (68F with no xenon),
Bias(E) = core k-effective bias for cold core simulation model at cycle exposure (E). The bias equals the target cold core simulation model critical k-effective minus 1.0.
The Cycle R value is determined from the evaluation of shutdown margin as a function of cycle exposure. The R value is used to determine shutdown margin testing requirements, and it is defined as the difference between the calculated beginning of cycle shutdown margin minus the calculated minimum shutdown margin during the cycle, where shutdown margin is a positive number. The value of R must be either positive or zero and must be determined for each fuel loading cycle.
Typical behavior of shutdown margin as a function of cycle exposure for SSES Units 1 and 2 is shown in Figure 4.3-13.
A description of the methods used to calculate shutdown margin is provided in Section 4.3.3.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-7 4.3.2.4.2 Reactivity Variations Reference 4.3-1 provides a general discussion of reactivity variations in a BWR/4. The reference provides tables showing typical k-effective values for various power levels, control fractions, and Xenon concentrations. From this data, the general reactivity effect of changing a single core variable can be determined. 4.3.2.5 Control Rod Patterns and Reactivity Worths 4.3.2.5.1 RWM Range Below the low power setpoint, control rod patterns follow prescribed withdrawal and insertion sequences restricted by the Rod Worth Minimizer (RWM). The sequences are established to assure that the maximum insequence control rod or rod segment reactivity worth would not be sufficient to result in a deposited fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal or insertion. Further discussion of the RWM and control rod sequence limitations is provided in Section 15.4.9 (Control Rod Drop Accident). 4.3.2.5.2 Operating Range In the power range, above the low power setpoint, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak fuel enthalpy of 280 cal/gm. Therefore, restrictions on control rod patterns are not required to minimize control rod worths. During power operation the control rod patterns are selected based on the measured core power distributions. For reload design purposes, optimized control rod patterns are selected for the cycle depletion. The series of design control rod patterns form the Cycle Step Out. Control rod sequence identification (A2, B2, A1, B1) is defined in Reference 4.3-1. 4.3.2.5.3 SCRAM Reactivity The reactor protection system (RPS) is capable of shutting down the reactor by initiating a SCRAM. The RPS and the control rod drive (CRD) system act quickly enough to prevent the initiating event from driving the fuel beyond transient limits. During a SCRAM from operating conditions, the control rod worth, reactor power, delayed neutron fraction, and void distributions must be properly accounted for as a function of time. The methodology used to account for these variables and determine SCRAM reactivity is described in Section 4.3.3. 4.3.2.6 Criticality of Reactor During Refueling Criticality of fuel assemblies in the core during refueling is avoided by assuring that the Technical Specification shutdown margin requirement is met. For core shuffles, a shutdo wn margin design criterion is defined to account for prediction uncertainties. This criterion helps determine the acceptability of a fuel move for meeting the Technical Specification limit. A description of the methods used to calculate shutdown margin is provided in Section 4.3.3.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-8 4.3.2.7 Stability Boiling Water Reactors do not have instability problems due to Xenon. Xenon transients are highly damped in a BWR due to the large negative power coefficient. References 4.3-1 and 4.3-3 provide additional discussion of Xenon instability. Thermal hydraulic stability is discussed in detail in Section 4.4. 4.3.2.8 Vessel Irradiations The RAMA Fluence Methodology (Reference 4.3-14) is used to evaluate the Reactor Pressure Vessel (RPV) fluence for both units. This methodology has been reviewed and approved by the NRC for RPV fluence evaluations (Reference 4.3-15) and is consistent with applicable regulatory guidance (Reference 4.3-16). Detailed descriptions of the calculations for each unit are provided in References 4.3-7 and 4.3-8. The fast fluence evaluations are based on the RAMA Code Methodology. The Methodology includes a transport code, model builder codes, a fluence calculator code, an uncertainty methodology, and a nuclear data library. The transport code, fluence calculator, and nuclear data library are the primary software components for calculating the neutron flux and fluence. The transport code uses a deterministic, three-dimensional, multigroup nuclear particle transport theory to perform the neutron flux calculations. The transport code couples the nuclear transport method with a general geometry modeling capability to provide a flexible and accurate tool for calculating fluxes in light water reactors. The fluence calculator uses reactor operating history information with isotopic production and decay data to estimate activation and fluence in the reactor components over the operating life of the reactor. The nuclear data
library contains nuclear cross-section data and response functions that are needed in the flux, fluence, and reaction rate calculations. The cross sections and response functions are based on the BUGLE-96 nuclear data library. Fluxes are calculated at the inner vessel surface, at 1/4 T and 3/4 T depths. The RAMA methodology calculates RPV fluence and uncertainty at all locations in the RPV in the active core region in accordance with applicable regulatory guidance. The results from the vessel fluence coupon analyses are solely used to support the methodology uncertainty analysis. The RAMA methodology directly calculates the fluence at all RPV locations in the active core region. Therefore, lead factors, which were historically used to extrapolate the measured fluence at the coupon locations to the RPV 1/4 T depths are no longer used or calculated. Previous fluence calculations were performed using the DORT computer code, which is described in Section 4.1. The RAMA Fluence Methodology will continue to be used to calculate the fluence for both units and is described in BWRVIP-114 (Reference 4.3-14). The analytical model for (R, ) geometry is shown in Figure 4.3-14. The model consists of an inner and outer core region, the shroud, water regions inside and outside the shroud, jet pump components, the vessel wall, inner and outer Cavity, Mirror Insulation and the Biological Shield.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-9 Neutron fluence was determined based on actual and expected operating history for each unit. This included the effects of several power uprates that have occurred during the operating history. Final end of life RPV fluence is calculated for both units at 32 EFPY at the RPV [both inner diameter (ID) and 1/4 T (1/4 of the distance from the inside diameter to the outside diameter)] based on actual and expected operating history. Details on the power history assumed in the fluence analysis are provided in footnotes to the data in Table 4.3-5. Table 4.3-5 lists the 32 EFPY maximum fast fluence results and also provides historical results from the original analyses for comparison. 4.3.3 Analytical Methods Reload design for SSES Units 1 and 2 is performed using NRC approved methodology. The approved methods used for nuclear design are fully described in Reference 4.3-13. A summary description of several nuclear design codes is provided in Section 4.1. Reference 4.3-1 describes the methods used for initial core nuclear design. 4.3.4 Changes Reference 4.3-1 lists several changes made to the initial reactor nuclear design. Reload core nuclear designs incorporate the following significant changes. Unit 2, Cycle 9 and Unit 1, Cycle 11 were the first cores to utilize the FANP ATRIUM TM-10 fuel design at SSES. ATRIUM TM-10 has a 10x10 lattice which is significantly different from the 9x9 lattice utilized in previous cycles. Nuclear characteristics of ATRIUM TM-10 fuel are discussed in Section 4.3. Mechanical design of ATRIUM TM-10 fuel is discussed in Section 4.2. The CASMO-3G lattice physics code was first used to support the U1C10 reload design. Unit 2, Cycle 9 was designed for a 24 month cycle. This cycle length represents a change from the 18 month cycle used for previous core designs. The effects of a 24 month cycle on the U2C9 reload were evaluated in Reference 4.3-11. Unit 1, Cycle 11 was designed for a 24 month cycle. This cycle length represents a change from the 18 month cycle used for previous core designs. The effects of a 24 month cycle on the U1C11 reload were evaluated in Reference 4.3-12. The CASMO-4/MICROBURN-B2 code system was first used to support the U1C14 reload design. A summary description of CASMO-4 and MICROBURN-B2 is provided in Section 4.1. Unit 1 Cycle 14 was the first cycle to utilize 100 mil fuel channels and the Framatome-ANP FUELGUARD Lower Tie Plate design. The 100 mil fuel channel and FUELGUARD Lower Tie Plate are described in Section 4.2. Unit 1 Cycle 20 was the first cycle to utilize the Advanced Fuel Channel (AFC). The AFC is described in Section 4.2.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-10 4.3.5 References 4.3-1 "BWR/4 and BWR/5 Fuel Design", NEDE-20944(P), General Electric Company, October 1976. 4.3-2 "BWR/4 and BWR/5 Fuel Design", Amendment 1 NEDE-20944-1(P), General Electric Company, January 1977. 4.3-3 Letter from Olan. D. Parr (NRC) to Dr. G. G. Sherwood (GE), "Review of General Electric Topical Report NEDE-20944-P, BWR/4 and BWR/5 Fuel Design (NEDO-20944 Non-Proprietary Version)", September 30, 1977.
4.3-4D e leted 4.3-5D e leted 4.3-6D e leted 4.3-7 "Susquehanna Unit 1 Reactor Pressure Vessel Fluence Evaluation, "PPL-FLU-002-R-002, Rev. 1, TransWare Enterprises, Inc., October 2005. 4.3-8 "Susquehanna Unit 2 Reactor Pressure Vessel Fluence Evaluation,"PPL-FLU-002-R-001, Rev. 0, TransWare Enterprises, Inc., May 2005.4.3-9 "Power Uprate Engineering Report for Susquehanna Steam Electric Station Units 1 and 2", NEDC-32161P, GE Nuclear Energy, December 1993. 4.3-10 "Advanced Nuclear Fuels Methodology For Boiling Water Reactors", XN-NF-80-19 (P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990. 4.3-11 "Susquehanna SES Unit 2 Cycle 9 Reload Summary Report", PL-NF-97-003, Rev. 1, PP&L, September 1997. 4.3-12 "Susquehanna SES Unit 1 Cycle 11 Reload Summary Report", PL-NF-98-002, Rev. 1, PP&L, Inc., July 1998. 4.3-13 EMF-2158 (P) (A), Siemens Power Corporation Methodology For Boiling Water Reactors Evaluation and Validation of Casmo-4/Microburn-B2", October 1999. 4.3-14 "BWR vessel and Internals Project RAMA Fluence Methodology Manual," BWRVIP-114, May 2003.
4.3-15 Safety Evaluation of proprietary EPRI Reports, "BWR Vessel And Internals Project, RAMA Fluence Methodology Manual (BWRVIP-114)," "RAMA Fluence Methodology Benchmark Manual - Evaluation of Regulatory Guide 1.190 Benchmark Problems (BWRVIP-115)," "RAMA Fluence Methodology - Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5 (BWRVIP-117)," and "RAMA Fluence Methodology Procedures Manual (BWRVIP-121)," and "Hope Creek Flux Wire Dosimeter Activation Evaluation for SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-11 Cycle 1 (TWE-PSE-001-R-001)" (TAC No. MB9765), BWRVIP 2005-208B, William H. Bateman, NRC to Bill Eaton, BWRVIP Chairman, May 13, 2005.
4.3-16 "Calculation and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Guide 1.190, March 2001.
SSES-FSAR Table Rev. 56 FSAR Rev. 65 Page 1 of 1 TABLE 4.3-1 REACTOR CORE CHARACTERISTICS Reactor Type/Configuration BWR-4/2 Loop Jet Pump Recirculation System, C-Lattice Rated Thermal Power, Unit 1 3952 MWt Rated Thermal Power, Unit 2 3952 Mwt Number of Fuel Assemblies 764 Number of Control Rods 185 Number of Traversing In-core Probe Locations 43 Active Core Height, ft 12.45 Control Rod Pitch, inches 12.0 Fuel Assembly Pitch, inches
6.0 AutoCAD
Figure Fsar 4_3_2.dwg FIGURE 4.3-2, Rev 63 UNIT 2 CORE LOADING MAP SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-8, Rev. 54 AutoCAD Figure 4_3_8.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9, Rev. 54 AutoCAD Figure 4_3_9.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10, Rev. 54 AutoCAD Figure 4_3_10.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-11, Rev. 54 AutoCAD Figure 4_3_11.doc
SUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 AND 2FINAL SAFETY ANALYSIS REPORTFIGURE 4.3-14, Rev. 55VESSEL FLUENCE (R.0) MODEL FORAZIMUTHAL FLUX DISTRIBUTIONAuto Cad: Figure Fsar 4_3_14.dwg THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-15, Rev. 55 AutoCAD Figure 4_3_15.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-16, Rev. 55 AutoCAD Figure 4_3_16.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-17, Rev. 54 AutoCAD Figure 4_3_17.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-18, Rev. 54 AutoCAD Figure 4_3_18.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-19, Rev. 54 AutoCAD Figure 4_3_19.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-20, Rev. 54 AutoCAD Figure 4_3_20.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-21, Rev. 54 AutoCAD Figure 4_3_21.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-22, Rev. 54 AutoCAD Figure 4_3_22.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-23, Rev. 54 AutoCAD Figure 4_3_23.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-24, Rev. 54 AutoCAD Figure 4_3_24.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-25, Rev. 54 AutoCAD Figure 4_3_25.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-26, Rev. 54 AutoCAD Figure 4_3_26.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-27, Rev. 54 AutoCAD Figure 4_3_27.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-28, Rev. 54 AutoCAD Figure 4_3_28.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-29, Rev. 54 AutoCAD Figure 4_3_29.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-30, Rev. 54 AutoCAD Figure 4_3_30.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-8-2, Rev. 56 AutoCAD Figure 4_3_8_2.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-8-9, Rev. 3 AutoCAD Figure 4_3_8_9.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-1, Rev. 55 AutoCAD Figure 4_3_9_1.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-5, Rev. 57 AutoCAD Figure 4_3_9_5.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-6, Rev. 55 AutoCAD Figure 4_3_9_6.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-7, Rev. 55 AutoCAD Figure 4_3_9_7.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-1, Rev. 55 AutoCAD Figure 4_3_10_1.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-2, Rev. 55 AutoCAD Figure 4_3_10_2.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-5, Rev. 55 AutoCAD Figure 4_3_10_5.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-6, Rev. 55 AutoCAD Figure 4_3_10_6.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-7, Rev. 55 AutoCAD Figure 4_3_10_7.doc
AutoCAD: Figure Fsar 4_3_8_32.dwgFIGURE 4.3-8-32, Rev 0ATRIUM -10 FUEL AXIAL ENRICHMENT(NOMINAL)PPL ASSEMBLY TYPE 79SUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 & 2FINAL SAFETY ANALYSIS REPORT TM AutoCAD: Figure Fsar 4_3_8_33.dwgFIGURE 4.3-8-33, Rev 0ATRIUM -10 FUEL AXIAL ENRICHMENT(NOMINAL)PPL ASSEMBLY TYPE 80SUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 & 2FINAL SAFETY ANALYSIS REPORT TM
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-10, Rev. 55 AutoCAD Figure 4_3_9_10.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-12, Rev. 55 AutoCAD Figure 4_3_9_12.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-13, Rev. 55 AutoCAD Figure 4_3_9_13.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-14, Rev. 56 AutoCAD Figure 4_3_9_14.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-15, Rev. 56 AutoCAD Figure 4_3_9_15.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-16, Rev. 56 AutoCAD Figure 4_3_9_16.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-17, Rev. 3 AutoCAD Figure 4_3_9_17.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-18, Rev. 3 AutoCAD Figure 4_3_9_18.doc
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-1 4.4 THERMAL AND HYDRAULIC DESIGN This section addresses the original plant thermal hydraulic design (number of assemblies, core power and flow, etc), the compatibility of co-resident fuel designs and the relative stability of reload
cores. 4.4.1 DESIGN BASES 4.4.1.1 Safety Design Bases Thermal-hydraulic design of the core shall establish:
(1)Actuation limits for the devices of the nuclear safety systems such that no fuel damag e occurs as a result of moderate frequency transient events. For example, the Minimu m Critical Power Ratio (MCPR) operating limit is specified such that at least 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during the mostsevere moderate (Per Regulatory Guide 1.70 Revision 2) frequency transient even ts.(2)The thermal-hydraulic safety limits for use in evaluating the safety margin relating theconsequences of fuel barrier failure to public safety
.(3)That the nuclear system exhibits no inherent tendency toward divergent or limit cyc le oscillations which would compromise the integrity of the fuel or nuclear system proces s barri er.4.4.1.2 Power Generation Design Bases The thermal-hydraulic design of the core shall provide the following operational characteristics:
(1)The ability to achieve rated core power output throughout the design life of the fuel witho ut sustaining premature fuel failure.
(2)Flexibility to adjust core output over the range of plant load and load maneuveringrequirements in a stable, predictable manner without sustaining fuel dama ge.4.4.1.3 Requirements for Steady-State Conditions Steady-State Limits For purposes of maintaining adequate thermal margin during normal steady-state operation, the minimum critical power ratio must not be less than the required MCPR operating limit, and the maximum linear heat generation rate (LHGR) must be maintained below the LHGR limit. This does not specify the operating power nor does it specify peaking factors. These parameters are determined subject to a number of constraints including the thermal limits given previously. The core and fuel design basis for steady-state operation, i.e., MCPR and LHGR limits, have been defined to provide margin between the steady-state operating conditions and any fuel damage SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-2 condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time in life.
Steady-state limits also exist on the maximum average planar linear heat generation rate (MAPLHGR). The MAPLHGR limits protect against violation of the ECCS acceptance criteria during a Loss of Coolant Accident and are derived from the LOCA analyses described in Section 6.3.
4.4.1.4 Requirements for Transient Conditions Transient Limits
The transient thermal limits are established such that no fuel damage is expected to occur during the most severe moderate frequency transient event. Fuel damage is defined as perforation of the cladding that permits release of fission products. Mechanisms that cause fuel damage in reactor transients are:
(1) Severe overheating of fuel cladding caused by inadequate cooling, and (2) Fracture of the fuel cladding caused by relative expansion of the uranium dioxide pellet inside the fuel cladding.
For design purposes, the transient limit requirement relating to cladding overheating is met if at least 99.9 percent of the fuel rods in the core do not experience boiling transition during any moderate frequency transient event. No fuel damage would be expected to occur even if a fuel rod actually experienced a boiling transition.
A value of 1 percent plastic strain of Zircaloy cladding is conservatively defined as the limit below which fuel damage from overstraining the fuel cladding is not expected to occur. The linear heat generation rate required to cause this amount of cladding strain depends on the fuel type and burnup. The linear heat generation rates are discussed on a fuel type specific basis in Section 4.2.3.
4.4.1.5 Summary of Design Bases
In summary, the steady-state operating limits have been established to assure that the design basis is satisfied for the most severe moderate frequency transient event. There is no steady-state design overpower basis. An overpower which occurs during an incident of a moderate frequency transient event must meet the plant transient MCPR limit and 1% plastic strain limit. Demonstration that the transient limits are not exceeded is sufficient to conclude that the design basis is satisfied.
The MCPR, MAPLHGR, and LHGR limits are sufficiently general so that no other limits need to be stated. For example, cladding surface temperatures will always be maintained within 10 to 15 oF of the coolant temperature as long as the boiling process is in the nucleate regime. The cladding and fuel bundle integrity criterion is assured as long as MCPR, MAPLHGR, and LHGR limits are met. There are no additional design criteria on coolant void fraction, core coolant flow-velocities, or flow distribution, nor are they needed. The coolant flow velocities and void fraction become constraints upon the mechanical and physics design of reactor components and are partially constrained by stability and control requirements.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-3 4.
4.2 DESCRIPTION
OF THERMAL-HYDRAULIC DESIGN OF THE REACTOR CORE 4.4.2.1 Summary Comparison
An evaluation of plant performance from a thermal and hydraulic standpoint is provided in Subsection 4.4.3.
A tabulation of thermal and hydraulic parameters of the core is given in Table 4.4-1.
4.4.2.2 Critical Power Ratio
There are three different types of boiling heat transfer to water in a forced convection system:
nucleate boiling, transition boiling, and film boiling. Nucleate boiling, at lower heat transfer rates, is an extremely efficient mode of heat transfer, allowing large quantities of heat to be transferred with a very small temperature rise at the heated wall. As heat transfer rate is increased the boiling heat transfer surface alternates between film and nucleate boiling, leading to fluctuations in heated wall temperatures. The point of departure from the nucleate boiling region into the transition boiling region is called the boiling transition. Transition boiling begins at the critical power and is characterized by fluctuations in cladding surface temperature. Film boiling occurs at the highest heat transfer rates; it begins as transition boiling comes to an end. Film boiling heat transfer is characterized by stable wall temperatures which are higher than those experienced during nucleate boiling.
4.4.2.2.1 Boiling Correlations The occurrence of boiling transition is a function of the fluid enthalpy, mass flow rate, pressure, flow geometry and assembly power distribution. Framatome ANP, Inc. (FANP) has conducted extensive experimental investigations of these parameters. These parametric studies encompass the entire design range of these variables. The SPCB critical power correlation, Reference 4.4-58, is used for ATRIUMŽ-10 fuel. This correlation is based on accurate test data of full-scale prototypic simulations of reactor fuel assemblies operating under conditions typical of those in actual reactor designs. The correlation is a "best fit" to the data and is used together with a statistical analysis to assure adequate reactor thermal margins (Reference 4.4-42).
The figure of merit used for reactor design and operation is the Critical Power Ratio (CPR). This is defined as the ratio of the bundle power at which boiling transition occurs to the bundle power at the reactor condition of interest (i.e., the ratio of critical bundle power to operating bundle power).
In this definition, the critical power is determined at the same mass flux, inlet temperature, and pressure which exist at the specified reactor condition.
4.4.2.3 Thermal Operating Limits
The limiting constraints in the design of the reactor core are stated in terms of the MCPR, MAPLHGR, and LHGR limits. The design philosophy used to assure that these limits are met involves the selection of one or more power distributions which are more limiting than expected operating conditions and subsequent verification that under these more stringent conditions, the design limits are met. Therefore, the "design power distributions" represent extreme conditions of power. Use of these power distributions in the analyses is a fair and stringent test of the operability SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-4 of the reactor as designed to comply with the foregoing limits. Expected operating conditions are less severe than those represented by the design power distributions which give the MCPR, MAPLHGR and LHGR limits.
However, it must be established that operation with a less severe power distribution is not a necessary condition for the safety of the reactor. Because there are an infinite number of operating reactor states which can exist (with variations in rod patterns, time in cycle, power level, distribution, flow etc.) which are within the design constraints, it is not possible to determine them all. However, constant monitoring of operating conditions using the available plant measurements can ensure compliance with design objectives.
4.4.2.3.1 Design Power Distribution Thermal design of the reactor--including the selection of the core size and effective heat transfer area, the design steam quality, the total recirculation flow, the inlet subcooling, and the specification of internal flow distribution -- was performed by the NSSS vendor and is based on the concept and application of a design power distribution. The design power distribution was an appropriately conservative representation of the most limiting thermal operating state at rated conditions and included design allowances for the combined effects (on the fuel rod, and the fuel assembly heat flux and temperature) of the gross and local steady-state power density distributions and adjustments of the control rods.
4.4.2.3.2 Design Linear Heat Generation Rates
The maximum and core average linear heat generation rates are shown in Table 4.4-1. The maximum linear heat generation rate at any location is the average linear heat generation rate at a given axial location multiplied by the total peaking factor of that location.
Fuel type specific LHGR limits and MAPLHGR limits are provided in the Core Operating Limits Report for each unit (see FSAR section 16.3, Technical Requirements Manuals).
4.4.2.4 Void Fraction Distribution Typical core average and maximum exit void fractions in the core at rated condition are given in Table 4.4-2. The axial distribution of core void fractions for the average radial channel and the maximum radial channel (end of node value) are also given in Table 4.4-2. Similar distributions for steam quality are provided in Table 4.4-3. The core average axial power distributions used to produce these tables are given in Table 4.4-2a.
4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern
Correct distribution of core coolant flow among the fuel assemblies is accomplished by the use of an accurately calibrated fixed orifice at the inlet of each fuel assembly. The orifices are located in the fuel support piece. They serve to control the flow distribution and, hence, the coolant conditions within prescribed bounds throughout the design range of core operation.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-5 The sizing and design of the orifices ensure stable flow in each fuel assembly during all phases of operation at normal operating conditions.
The core is divided into two orificed flow zones. The outer zone is a narrow, reduced-power region around the periphery of the core. The inner zone consists of the core center region. No other control of flow and steam distribution, other than that incidentally supplied by adjusting the power distribution with the control rods, is used or needed. The orifices can be changed during refueling, if necessary.
Design core flow distribution calculations were performed by the NSSS vendor using a design power distribution which consists of a hot and average powered assembly in each of the two orifice zones. The design bundle power and resulting relative flow distribution are given in Table 4.4-4.
The flow distribution to the fuel assemblies is calculated on the assumption that the pressure drop across all fuel assemblies is the same. This assumption has been confirmed by measuring the flow distribution in a modern boiling water reactor as reported in References 4.4-2 and 4.4-36.
There is reasonable assurance, therefore, that the calculated flow distribution throughout the core is in close agreement with the actual flow distribution of an operating reactor.
The use of the design power distribution discussed previously ensures the orificing chosen covers the range of normal operation. The expected shifts in power production during core life are less severe and are bounded by the design power distribution.
4.4.2.6 Core Pressure Drop and Hydraulic Loads The pressure drop across various core components under the steady state design conditions is included in Table 4.4-1. Analyses for the most limiting conditions, the recirculation line break and the steam line break are reported in Chapter 15.
The components of bundle pressure drop considered are friction, local elevation and acceleration.
Reference 4.4-43 presents the methodology and constitutive relationships used by FANP for the calculation of pressure drop in BWR fuel assemblies. These are implemented in the XCOBRA computer code which is used to perform steady state thermal-hydraulic analyses, Reference 4.4-49. The thermal hydraulic loads on the fuel rods during steady-state operation, transient, and accident conditions are negligible, primarily because of the channel confinement, thereby resulting in small cross flow between rods (i.e., essentially constant pressure at any given elevation in the fuel bundle). The loads (i.e. horizontal) across the control blades are minimal or negligible primarily due to the flat interchannel velocity profile as given in Reference 4.4-13.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-6 4.4.2.6.1 Friction Pressure Drop Friction pressure drop is calculated using the model relation 2 TPFchH 2 f fL 2g w = P A D 2 where P f = friction pressure drop, psi w = mass flow rate, g = acceleration of gravity, = water density, D H = channel hydraulic diameter, A ch = channel flow area, L = length, f = friction factor, and 2TPF = two-phase friction multiplier This basic model is similar to that used throughout the nuclear power industry. The formulation for the two-phase multiplier used by FANP is the correlation determined by Jones, Reference 4.4-43, which represents a mass velocity correction to the Martinelli-Nelson correlation, Reference 4.4-3.
Significant amounts of friction pressure drop data in multirod geometries representative of modern BWR plant fuel bundles have been taken and both the friction factor and two-phase multipliers have been correlated on a best-fit basis using the above pressure drop formulation.
4.4.2.6.2 Local Pressure Drop
The local pressure drop is defined as the irreversible pressure loss associated with an area change such as the orifice, lower tie plates, and spacers of a fuel assembly.
The general local pressure drop model is similar to the friction pressure drop and is:
L 2 2 2 TPL 2 P = w 2g K A where P L = local pressure drop, psi, K = local pressure drop loss coefficient, A 2 = reference area for local loss coefficient, and 2 TPL = two-phase local multiplier, and w and g are defined the same as for friction. This basic model is similar to that used throughout the nuclear power industry. The two-phase multiplier used by FANP is given by the ratio of the saturated water and two-phase mixture densities. Tests are performed in both single and two-phase flow to arrive at best-fit design values for spacer and upper tie plate pressure drop. The range of test variables is specified to include the range of interest to boiling water reactors.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-7 New data are taken whenever there is a significant design change to ensure the most applicable methods are in use at all times. For ATRIUM-10 fuel, a simple multiplier based on local quality is also used to calculate the spacer pressure drop for the two-phase conditions, Reference 4.4-45.
4.4.2.6.3 Elevation Pressure Drop
The elevation pressure drop is based on the well-known relation
E P gL=
=+fg ()1 where PE = elevation pressure drop, psi L = incremental length
= average water density
= average void fraction over the length L f, g = saturated water and vapor density, respectively. g = acceleration of gravity
4.4.2.6.4 Acceleration Pressure Drop A reversible pressure change occurs when an area change is encountered, and an irreversible loss occurs when the fluid is accelerated through the boiling process. The basic formulation for the reversible pressure change resulting from a flow area change is given by:
P W gA A A ACC==();1 2 2 2 2 2 2 1 where PACC = acceleration pressure drop, A 2 = final flow area, A 1 = initial flow area, and other terms are as previously defined. The basic formulation for the acceleration pressure change due to density change is:
P W gAACC ch M out M in=2 2 11 where
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-8
()1 1 1 2 2 1Mg x x=+() M = momentum density, x = steam quality, 1 = saturated liquid density and other terms are as previously defined. The total acceleration pressure drop in boiling water reactors is on the order of a few percent of the total pressure drop.
4.4.2.7 Correlation and Physical Data
Substantial amounts of physical data support the pressure drop and thermal hydraulic loads discussed in Subsection 4.4.2.6. Correlations have been developed to fit these data to the formulations discussed.
4.4.2.7.1 Pressure Drop Correlations
Pressure drop data in multirod geometries representative of modern BWR plant fuel bundles has been correlated to the friction factor and two-phase multipliers on a best fit basis using the pressure drop formulations reported in Subsections 4.4.2.6.1 and 4.4.2.6.2 FANP's pressure drop methodology is described in Reference 4.4-43.
New data are taken whenever there is a significant design change. Applicability of the pressure drop correlations is confirmed by full scale prototype flow tests. Pressure drop tests for the FANP ATRIUM-10 fuel designs is reported in Reference 4.4-45. The range of tests variables is specified to include the range of interest to boiling water reactors.
4.4.2.7.2 Void Fraction Correlation
The void fraction is determined by a Zuber-Findlay model with constitutive relations as supplied by Ohkawa and Lahey, Reference 4.4-43.
4.4.2.7.3 Heat Transfer Correlation
The Jens-Lottes (Reference 4.4-5) wall superheat equation is used in fuel design to determine the cladding-to-coolant heat transfer coefficients for nucleate boiling.
4.4.2.8 Thermal Effects of Operational Transients
The evaluation of the core's capability to withstand the thermal effects resulting from anticipated operational transients is covered in Chapter 15.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-9 4.4.2.9 Uncertainties in Estimates Uncertainties in thermal-hydraulic parameters are considered in the statistical analysis which is performed to establish the fuel cladding integrity safety MCPR limit such that at least 99.9% of the fuel rods in the core are expected not to experience boiling transition during any moderate frequency transient event. The statistical model and analytical procedure are described in Reference 4.4-42.
The MCPR safety limit is determined by a statistical convolution of the uncertainties associated with the calculation of thermal margin. Some uncertainties are fuel related and others are characteristics of the reactor system. Examples of fuel related uncertainties are those introduced by the critical power correlation, the calculation of core wide power peaking and the calculation of core wide flow distribution. Examples of uncertainties which are characteristics of the reactor system are the measurement uncertainties associated with reactor pressure, total core flow, feedwater flow and feedwater temperature. The uncertainties which are considered are shown in Table 4.4-6.
4.4.2.10 Flux Tilt Considerations
The inherent design characteristics of the BWR are particularly well suited to handle perturbations due to flux tilt. The stabilizing nature of the moderator void coefficient effectively damps oscillations in the power distribution. In addition to this damping, the incore instrumentation system and the associated on-line computer provide the operator with prompt and reliable power distribution information. Thus, the operator can readily use control rods or other means to effectively limit the undesirable effects of flux tilting. Because of these features and capabilities, it is not necessary to allocate a specific peaking factor margin to account for flux tilt. If for some reason, the power distribution could not be maintained within normal limits using control rods, then the operating power limits would have to be reduced as prescribed in the Plant Technical Specifications. The power distributions will be maintained such that the operating limits given in the Core Operating Limits Report will not be exceeded.
4.4.2.11 Crud Deposition
In general, the CPR is not affected as crud accumulates on fuel rods (References 4.4-34 and 4.4-35). Therefore, no modifications to the critical power correlation are made to account for crud deposition. The effect of crud deposition on pressure drop and flow is to increase the pressure drop and decrease the flow. An increase in crud deposition for high exposure assemblies would tend to reduce the flow in these assemblies and increase the flow in low exposure, CPR limiting assemblies. No credit is taken for the increase in CPR margin due to crud deposition.
The effects of crud deposition are included in thermal and rod internal pressure calculations, Reference 4.4-47.
4.
4.3 DESCRIPTION
OF THE THERMAL AND HYDRAULIC DESIGN OF THE REACTOR COOLANT SYSTEM The thermal and hydraulic design of the reactor coolant system is described in this subsection.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-10 4.4.3.1 Plant Configuration Data 4.4.3.1.1 Reactor Coolant System Configuration The reactor coolant system is described in Section 5.4 and shown in isometric perspective in Figure 5.4-1. The piping sizes, fittings and valves are listed in Table 5.1-1.
4.4.3.1.2 Reactor Coolant System Thermal Hydraulic Data Table 5.1-1 provides design temperatures, pressures and flow rates for the reactor coolant system and its components.
4.4.3.1.3 Reactor Coolant System Geometric Data
Coolant volumes of regions and components within the reactor vessel are shown in Figure 5.1-2.
Table 4.4-8 provides the flow path length, height and liquid level, minimum elevations, and minimum flow areas for each major flow path volume within the reactor vessel and recirculation loops of the reactor coolant systems.
Table 4.4-9 provides the lengths and sizes of all safety injection lines to the reactor coolant system.
4.4.3.2 Operating Restrictions on Pumps
Expected recirculation pump performance curves are shown in Figure 5.4-3. These curves are valid for all conditions with a normal operating range varying from approximately 20% to 115% of rated pump flow.
The pump characteristics including considerations of NPSH requirements are the same for the conditions of two pump and one pump operation as described in Subsection 5.4.1. Subsection 4.4.3.3 gives the limits imposed on the recirculation pumps by cavitation, pump loads, bearing design, flow starvation, and pump speed.
4.4.3.3 Power-Flow Operating Map
4.4.3.3.1 Limits for Normal Operation
A boiling water reactor must operate with certain restrictions because of pump net positive suction head (NPSH), overall plant control characteristics, core thermal power limits, etc. A representation of a simplified power-flow map for the power range of operation is shown in Figure 4.4-5. The actual power-flow maps for Units 1 and 2 are found in the respective COLR, FSAR Section 16.3.
The nuclear system equipment, nuclear instrumentation, and the reactor protection system, in conjunction with operating procedures, maintain operations within the area of this map for normal operating conditions. The boundaries on this map are as follows:
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-11 Natural Circulation Line: The operating state of the reactor moves along this line for the normal control rod withdrawal sequence in the absence of recirculation pump operation.
30 Percent Recirculation Pump Constant Speed Line: Startup operations of the plant are normally carried out with the recirculation pumps operating at approximately 30 percent speed. The operating state for the reactor follows this line for the normal control rod withdrawal sequence.
Rated Flow Control Line: The rated flow control line (100% rod line) passes through 100 percent power at 108 Mlb/hr flow. The operating state for the reactor follows this line for recirculation flow changes with a fixed control rod pattern. The line is based on full power constant xenon concentration.
Cavitation Protection Line: This line (minimum power line) results from the recirculation pump and jet pump NPSH requirements. The recirculation pumps are automatically switched to 30 percent speed when the feedwater flow drops below a preset value.
Note that an actual power-flow map will contain stability related regions. The actual Unit 1 and Unit 2 power-flow maps are included in their respective COLR, FSAR Section16.3.
4.4.3.3.1.1 Performance Characteristics Other Power Flow Operating Map performance characteristics are:
Recirculation Pump Constant Speed Line: This line shows the change in flow associated with power changes while maintaining constant recirculation pump speed.
Constant Rod Lines: These lines show the change in power associated with flow changes while maintaining constant control rod position (e.g. 80% rod line).
4.4.3.3.2 Regions of the Power Flow Map For normal operating conditions, the nuclear system equipment, nuclear instrumentation, and the reactor protection system, in conjunction with operating procedures, maintain operation outside the exclusion areas of the power flow map. Main regions of the map are discussed below to clarify operational capabilities.
Region A - This is the transition region between natural circulation operation and 30% pump speed operation. Operation at less than 30% pump speed with two recirculation loops results in flow instabilities (causing flow induced vibrations), therefore the recirculation pumps are not continually operated below 30% pump speed. Normal startup is along the 30% pump speed boundary of this region.
Region B - This region represents the normal operating zone of the map where power changes can be made either by control rod movement or by core flow changes by changing recirculation pump drive speed.
Region C - This is the low power area of the map where cavitation can be expected in the recirculation pumps and in the jet pumps. Operation within this region is precluded by system interlocks which runback the recirculation pumps to 30% speed whenever feedwater flow is less than a preset value (typically 20% of rated).
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-12 4.4.3.4 Temperature-Power Operating Map (PWR)
Not Applicable.
4.4.3.5 Load Following Characteristics
The following simple description of boiling water reactor operation with recirculation flow control summarizes the principal modes of normal power range operation. Assuming the plant to be initially hot with the reactor critical, full power operation can be approached by initially moving along the two pump 30% speed line until power is at least above the minimum power line (cavitation interlock) of Region C (see Figure 4.4-5). Note, other low power restrictions may apply as a result of cycle specific transient analyses. This initial sequence may be achieved with control rod withdrawal and manual, individual recirculation pump control. Individual pump startup procedures are provided which achieve 30 percent of full pump speed in each loop. Power, steam flow, and feedwater flow are increased as control rods are manually withdrawn. An interlock prevents low power-high recirculation flow combinations which create recirculation pump and jet pump NPSH problems.
Reactor power increases as the operating state moves to the right on Figure 4.4-5 as the operator manually increases recirculation flow in each loop. Eventually, the operator can switch to simultaneous recirculation pump control. Thermal output can then be increased by either control rod withdrawal or recirculation flow increase. Both combinations are required to achieve full power.
The operating map is shown in Figure 4.4-5 with the designated flow control range expected.
The curve labeled "100% Xe Rod Line" (i.e., the "Rated Flow Control Line") represents a typical steady state power flow characteristic for a fixed rod pattern. It is affected by xenon, core leakage flow assumptions, and reactor vessel pressure variations.
Normal power range operation is along the "Rated Flow Control Line", below the APRM Rod Block Trip Setpoint, and below 100% rated power.
The large negative operating reactivity and power coefficients, which are inherent in the boiling water reactor, provide important advantages as follows:
(1) Good load following with well damped behavior and little undershoot or overshoot in the heat transfer response.
(2) Load following with recirculation flow control.
(3) Strong damping of spatial power disturbances.
The reactor power level can be controlled by flow control over approximately 35 percent of the power level on the rated rod line. Load following is accomplished by varying the recirculation flow to the reactor. This method of power level control takes advantage of the reactor negative void coefficient. To increase reactor power, it is necessary to increase the recirculation flow rate which sweeps some of the voids from the moderator, causing an increase in core reactivity. As the reactor power increases, more steam is formed and the reactor stabilizes at a new power level with the transient excess reactivity balanced by the new void formation. No control rods are moved to accomplish this power level change. Conversely, when a power reduction is required, it is necessary only to reduce the recirculation flow rate. When this is done, more voids in the SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-13 moderator automatically decrease the reactor power level to that commensurate with the new recirculation flow rate. Again, no control rods are moved to accomplish the power reduction.
Varying the recirculation flow rate (flow control) is more advantageous, relative to load-following, than using control rod positioning. Flow variations perturb the reactor uniformly in the horizontal planes and ensure a flatter power distribution and reduced transient allowances. As flow is varied, the power and void distributions remain approximately constant at the steady state end points for a wide range of flow variations. After adjusting the power distribution by positioning the control rods at a reduced power and flow, and taking into account any effects due to Xe variations, the operator can then bring the reactor to rated conditions by increasing flow, with the assurance that the power distribution will remain approximately constant. Section 7.7 describes how recirculation flow is
varied.
4.4.3.6 Thermal and Hydraulic Characteristics Summary Table The thermal hydraulic characteristics are provided in Table 4.4-1 for the core and tables of Sections 5.1 and 5.4 for other portions of the reactor coolant system.
4.4.4 EVALUATION
The design basis employed for the thermal and hydraulic characteristics incorporated in the core design, in conjunction with the plant equipment characteristics, nuclear instrumentation, and the reactor protection system, is to require that no fuel damage occur during normal operation or during abnormal operation transients. Demonstration that the applicable thermal-hydraulic limits are not exceeded is given by analyses.
4.4.4.1 Critical Power
The SPCB critical power correlation is utilized in thermal-hydraulic evaluations. This correlation is discussed in more detail in Subsection 4.4.2.2.1.
4.4.4.2 Core Hydraulics
Core hydraulic models and correlations are discussed in Subsections 4.4.2.6, 4.4.2.7, and 4.4.4.5.
4.4.4.3 Influence of Power Distributions
The influence of power distributions on the thermal-hydraulic design is discussed in Reference 4.4-1, Appendix V for the initial core. The influence of power distribution is included in the cycle specific licensing calculations.
4.4.4.4 Core Thermal Response The thermal response of the core for accidents and expected transient conditions is discussed in Chapter 15.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-14 4.4.4.5 Analytical Methods The analytical methods, thermodynamic data, and hydrodynamic data used in determining the thermal and hydraulic characteristics of the core are similar to those used throughout the nuclear power industry.
Core thermal-hydraulic analyses are performed with the aid of a digital computer program. This program models the reactor core through a hydraulic description of orifices, lower tie plates, fuel rods, fuel rod spacers, upper tie plates, fuel channel, and the core bypass flow paths.
4.4.4.5.1 Reactor Model
The reactor model includes a hydraulic representation of the orifice, lower tie plate, fuel rods, water rods or inner water channel, spacers, upper tie plate and the fuel channel.
The code can handle a number of fuel channel types and bypass flow paths. Usually there is one fuel assembly representing each of the "hot" fuel types. The average types then make up the balance of the core.
The computer program iterates on flow through each flow path (fuel assemblies and bypass paths) until the total differential pressure (plenum to plenum) across each path is equal, and the sum of the flows through each path equals the total core flow.
For the initial core, orificing was selected to optimize the core flow distribution between orifice regions as discussed in Subsection 4.4.2.5. The core design pressure is determined from the required turbine throttle pressure, the steam line pressure drop, steam dryer pressure drop, and the steam separator pressure drop. The core inlet enthalpy is determined from the reactor and turbine heat balances. The required core flow is then determined by applying the procedures of this section and specifications such that the applicable thermal limits are satisfied. The results of applying these methods and specifications are:
(1) Flow for each bundle type, (2) Flow for each bypass path, (3) Core pressure drop, (4) Fluid property axial distribution for each bundle type, and (5) CPR calculations for each bundle type.
For reload cores, the appropriate orificing, core flow and system pressure are used as model input.
The same type of calculations that were used for the initial core are performed to calculate the parameters stated in (1)-(5) above.
4.4.4.5.2 System Flow Balances
The basic assumption used by the code in performing the hydraulic analysis is that the flow entering the core will divide itself between the fuel bundles and the bypass flow paths such that each assembly and bypass flow path experience the same pressure drop. The bypass flow paths considered are described in Table 4.4-7 and shown in Figure 4.4-1. Due to the large flow area, the pressure drop in the bypass region above the core plate is essentially all elevation head. Thus, the SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-15 sum of the core plate differential pressure and the bypass region elevation head is equal to the core differential pressure.
The total core flow less the control rod cooling flow enters the lower plenum through the jet pumps. A fraction of this passes through the various bypass paths. The remainder passes through the orifices in the fuel support (experiencing a pressure loss) where more flow is lost through the fit-up between the fuel support and the lower tie plate and also through the lower tie plate holes into the bypass region. The majority of the flow continues through the lower tie plate (experiencing a pressure loss) where some flow is lost through the flow path defined by the fuel channel and lower tie plate, and restricted by the finger springs, into the bypass region.
Full-scale tests have been performed to establish the flow coefficients for the major flow paths (Reference 4.4-14). The results of these tests were used to support the initial core design. These tests simulate actual plant configurations which have several parallel flow paths and therefore the flow coefficients for the individual paths could not be separated. However, analytical models of the individual flow paths were developed as an independent check of the tests. The models were derived for actual BWR design dimensions and considered the effects of dimensional variations.
These models predicted the test results when the "as built" dimensions were applied. When using these models for hydraulic design calculations, nominal drawing dimensions were used. This is done to yield the most accurate prediction of the expected bypass flow. With the large number of components in a typical BWR core, deviations from the nominal dimensions will tend to statistically cancel, resulting in a total bypass flow best represented by that calculated using nominal dimensions.
The bypass and active channel path loss coefficients are based on test data or analytical models. Use of these coefficients produces an accurate prediction of flow through the various flow paths.
The balance of the flow enters the fuel bundle from the lower tie plate and passes through either the fuel rod channel spaces or into a non-fueled water rod or water channel, depending on fuel type. This water rod or water channel flow, remixes with the active coolant channel flow below the upper tie plate. The uncertainties associated with the calculation of total core flow and assembly flow are considered in the MCPR safety limit calculation, Subsection 4.4.2.9.
4.4.4.5.3 System Heat Balances Within the fuel assembly, heat balances on the active coolant are performed nodally. Fluid properties are expressed as the bundle average at the particular node of interest. In evaluating fluid properties, a constant pressure model is used. The core power is divided into two parts: an active coolant power and a bypass flow power. The bypass flow is heated by neutron-slowing down and gamma heating transferred to the bypass flow from structures and control elements which are themselves heated by gamma absorption and by the (n, a) reaction in the control material. The fraction of total reactor power deposited in the bypass region is very nearly 2%. A similar phenomenon occurs within the fuel bundle relative to the active coolant and the water rod or inner water channel flows. The net effect is that approximately 96% of the core power is conducted through the fuel cladding and appears as heat flux.
In design analyses the power is allocated to the individual fuel bundles using a relative power factor. The power distribution along the length of the fuel bundle is specified with axial power factors which distribute the bundle's power among the axial nodes. A nodal local peaking factor is used to establish the peak heat flux at each nodal location.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-16 The relative (radial) and axial power distributions, when used with the bundle flow, determine the axial coolant property distribution resulting in sufficient information to calculate the pressure drop components within each fuel type. Once the equal pressure drop criterion has been satisfied, the critical bundle power is determined by an iterative process for each fuel type.
4.4.4.6 Thermal-Hydraulic Stability Analysis
4.4.4.6.1 Introduction
There are many definitions of stability, but for feedback processes and control systems it can be defined as follows: A system is stable if, following a disturbance, the transient settles to a steady, noncyclic state.
A system may also be acceptably safe even if oscillatory, provided that any limit cycle of the oscillations is less than a prescribed magnitude. Instability then, is either a continual departure from a final steady-state value or a greater-than-prescribed limit cycle about the final steady-state
value.
The mechanism for instability can be explained in terms of frequency response. Consider a sinusoidal input to a feedback control system which, for the moment, has the feedback disconnected. If there were no time lags or delays between input and output, the output would be in phase with the input. Connecting the output so as to subtract from the input (negative feedback or 180 o out-of-phase connection) would result in stable closed loop operation. However, natural laws can cause phase shift between output and input and should the phase shift reach 180 degrees, the feedback signal would be reinforcing the input signal rather than subtracting from it. If the feedback signal were equal to or larger than the input signal (loop gain equal to one or greater),
the input signal could be disconnected and the system would contin ue to oscillate.
If the feedback signal were less than the input signal (loop gains less than one), the oscillations would die out.
The design of the BWR is based on the premise that power oscillations can be readily detected and suppressed.
4.4.4.6.2 Description Three types of stability considered in the design of boiling water reactors are: (1) reactor core (reactivity) stability, (2) channel hydrodynamic stability, and (3) total system stability. Reactivity feedback instability of the reactor core could drive the reactor into power oscillations. Hydrodynamic channel instability could impede heat transfer to the moderator and drive the reactor into power oscillations. The total system stability considers control system dynamics combined with basic process dynamics. A stable system is analytically demonstrated if no inherent limit cycle or divergent oscillation develops within the system as a result of calculated step disturbances of any critical variable, such as steam flow, pressure, neutron flux, and recirculation flow.
The criteria to be considered are stated in terms of two compatible parameters. First is the decay ratio x 2/x 0, designated as the ratio of the magnitude of the second overshoot to the first overshoot resulting from a step perturbation. A plot of the decay ratio is a graphic representation of the physical responsiveness of the system, which is readily evaluated in a time-domain analysis. Second is the damping coefficient n, the definition of which corresponds to the pole pair closest to the j, axis in the s-plane for the system closed loop transfer function. This parameter also applies SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-17 to the frequency-domain interpretation. The damping coefficient is related to the decay ratio as shown in Figure 4.4-2.
4.4.4.6.3 Stability Criteria
The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable design limits are not possible or can be reliably and readily detected and suppressed.
The assurance that the total plant is stable and, therefore, has significant safety margin shall be demonstrated analytically when the decay ratio, x 2/x 0, is less than 1.0 or, equivalently, when the damping coefficient, n, is greater than zero for each type of stability discussed. Special attention is given to differentiate between inherent system limit cycles and small, acceptable limit cycles that are always present, even in the most stable reactors. The latter are caused by physical nonlinearities (deadband, striction, etc.) in real control systems and are not representative of inherent hydrodynamic or reactivity instabilities in the reactor. The ultimate performance limit criteria for the three types of dynamic performance are summarized below in terms of decay ratio and damping coefficient:
Channel hydrodynamic stability x 2/x 0 < 1, n > 0 Reactor core (reactivity) stability x 2/x 0 < 1, n > 0 Total system stability x 2/x 0 < 1, n > 0 These criteria shall be satisfied for all attainable conditions of the reactor that may be encountered in the course of plant operation. For stability purposes the most severe core power and core flow conditions to which these criteria will be applied correspond to the highest attainable rodline intersection with natural circulation flow.
New FANP fuel designs are designed to exhibit channel decay ratio characteristics equivalent to existing FANP fuel designs. Evaluation of the effect of all fuel designs present in the core on the core stability is currently made on a cycle specific basis. In support of these evaluations, FANP uses the STAIF computer code for stability calculations, Reference 4.4-48. SSES has implemented Option 3 (osc illation power range monitor system) for the long term stability solution.
4.4.4.6.4 Mathematical Model
For the initial core, the mathematical model representing the core examines the linearized reactivity response of a reactor system with density-dependent reactivity feedback caused by boiling. The core model (References 4.4-27 through 4.4-32), shown in block diagram form in Figure 4.4-3, solves the dynamic equations that represent the reactor core in the frequency domain.
The plant model considers the entire reactor system, neutronics, heat transfer, hydraulics, and the basic processes, as well as associated control systems such as the flow controller, pressure regulator, feedwater controller, etc. Although, the control systems may be stable when analyzed individually, final control system settings must be made in conjunction with the operating reactor so that the entire system is stable. The plant model yields results that are essentially equivalent to SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-18 those achieved with the core model and allows the addition of the controllers, which have adjustable features permitting the attainment of the desired performance.
The plant model solves the dynamic equations that present the BWR system in the time domain. The variables, such as steam flow and pressure, are represented as a function of time. The extensiveness of this model (Reference 4.4-10, which describes the version of the code used for Susquehanna system stability calculations) is shown in block diagram form in Figure 4.4-3. Many of the blocks are extensive systems in themselves.
For reload cores, the continued applicability of the exclusion region that has been established to assure thermal-hydraulic stability is demonstrated or the exclusion region is redefined. Stability calculations, when required are performed using the STAIF computer code (Reference 4.4-48).
4.4.4.6.5 Analytical Confirmation References 4.4-37 and 4.4-48 provide a description of the analytical methods used by GE and FANP as well as model qualification through comparison with test data.
4.4.4.6.6 Analysis Results
Using actual design parameters, the responses of important nuclear system variables for the first core to step disturbances were calculated for three different power/flow conditions. Figures 4.4-7A, 4.4-7B, and 4.4-7C show the responses at 51.5% power and natural circulation. Figures 4.4-8A, 4.4-8B, and 4.4-8C show the responses at rated power/flow conditions. Figures 4.4-9A, 4.4-9B, and 4.4-9C show the responses at the lower end of the automatic power-flow control path. For all of these cases the responses met the stability criterion.
For reload cores, a confirmatory analysis is performed to demonstrate the continued applicability of the core stability regions identified in the COLR. The analysis is based on comparison of core stability performance to previously analyzed cycles. A stability code is used to calculate the variations in decay ratio from cycle to cycle for operating conditions at representative state points near the stability exclusion region.
4.4.5 TESTING
AND VERIFICATION The testing and verification techniques to be used to assure that the planned thermal and hydraulic design characteristics of the core have been provided, and will remain within required limits throughout core lifetime, are discussed in Chapter 14. A summary is as follows:
(1) Preoperational Testing
Tests are performed during the preoperational test program to confirm that construction is complete and that all process and safety equipment is operational. Baseline data are taken to assist in the evaluation of subsequent tests. Heat balance instrumentation, jet pump flow and core temperature instrumentation, is calibrated and set points verified.
(2) Initial Start-Up
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-19 Hot functional tests are conducted with the reactor between 5 and 10% power. Core performance is monitored continuously to assure that the reactor is operating within allowable limits (e.g., peaking factors, linear heat generation rate, etc.) and is evaluated periodically to verify the core expected and actual performance margins.
4.4.6 INSTRUMENTATION
REQUIREMENTS
The reactor vessel instrumentation monitors the key reactor vessel operating parameters during planned operations. This ensures sufficient control of the parameters. The following reactor vessel sensors are discussed in Subsection 7.7.1.1.
(1) Reactor Vessel Temperature (2) Reactor Vessel Water Level (3) Reactor Vessel Coolant Flow Rates and Differential Pressures (4) Reactor Vessel Internal Pressure (5) Neutron Monitoring System
4.
4.7 REFERENCES
4.4-1 General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Company, January 1977, (NEDO-10958A).
4.4-2 Core Flow Distribution in a Modern Boiling Water Reactor as Measured in Monticello, August 1976, (NEDO-10722A).
4.4-3 R.C. Martinelli and D. E. Nelson, "Prediction of Pressure Drops During Forced Convection Boiling of Water," ASME Trans., 70, pp 695-702, 1948.
4.4-4 Deleted
4.4-5 Jens, W. H., and Lottes, P.A., Analysis of Heat Transfer, Burnout, Pressure Drop, and Density Data for High Pressure Water, USAEC Report-4627, 1972.
4.4-6 Deleted
4.4-7 Deleted
4.4-8 Deleted
4.4-9 Deleted
4.4-10 Analytical Methods of Plant Transient Evaluations for General Electric Boiling Water Reactor, General Electric Company, BWR Systems Department, February 1973, (NEDO-10802).
4.4-11 Deleted
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-20 4.4-12 Deleted
4.4-13 Peach Bottom Atomic Power Station Units 2 and 3, Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibration, General Electric Co.,
NEDO-20994, September, 1975.
4.4-14 "Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration," NEDE-21156, Class III, January 1976.
4.4-15 Deleted
4.4-16 Deleted
4.4-17 Deleted 4.4-18 Deleted
4.4-19 Deleted
4.4-20 Deleted
4.4-21 Deleted
4.4-22 Deleted
4.4-23 Deleted
4.4-24 Deleted
4.4-25 Deleted 4.4-26 Deleted
4.4-27 KAPL-2170 Hydrodynamic Stability of a Boiling Channel, by A. B. Jones; 2 October 1961. 4.4-28 KAPL-2208 Hydrodynamic Stability of a Boiling Channel Part 2, by A. B. Jones; 20 April 1962.
4.4-29 KAPL-2290 Hydrodynamic Stability of a Boiling Channel Part 3, by A. B. Jones and D. G. Dight; 28 June 1963.
4.4-30 KAPL-3070 Hydrodynamic Stability of a Boiling Channel Part 4, by A. B. Jones; 18 August 1964.
4.4-31 KAPL-3072 Reactivity Stability of a Boiling Reactor Part 1, by A. B. Jones and W. M. Yarbrough; 14 September 1964.
4.4-32 KAPL-3093 Reactivity Stability of a Boiling Reactor Part 2, by A. B. Jones, 1 March 1965.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-21 4.4-33 Deleted
4.4-34 McBeth, R.V., R. Trenberth, and R. W. Wood, "An Investigation Into the Effects of Crud Deposits on Surface Temperature, Dry-Out, and Pressure Drop, with Forced Convection Boiling of Water at 69 Bar in an Annular Test Section," AEEW-R-705, 1971.
4.4-35 Green, S.J., B. W. LeTourneau, A.C. Peterson, "Thermal and Hydraulic Effects of Crud Deposited on Electrically Heated Rod Bundles," WAPD-TM-918, Sept. 1970.
4.4-36 H.T. Kim and H.S. Smith, "Core Flow Distribution in a General Electric Boiling Water Reactor as Measured in Quad Cities Unit 1," NEDO-10722A, August, 1976.
4.4-37 Licensing Topical Report, "Stability and Dynamic Performance of the General Electric Boiling Water Reactor," January, 1977 (NEDO-21506).
4.4-38 Deleted
4.4-39 Deleted 4.4-40 Deleted
4.4-41 Deleted
4.4-42 ANF-524 (P)(A), Revision 2 and Supplements 1 and 2, "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," November 1990.
4.4-43 "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies", XN-NF-79-59(P)(A), November 1983.
4.4-44 MICROBURN-B2 Based Impact of Failed / Bypassed LPRMs and TIPs, Extended LPRM Calibration interval on Single Loop Operation on Measured Radial Bundle Power Uncertainty, "EMF-2493(P), Rev. 0, December 2000.
4.4-45 "Thermal-Hydraulic Characteristics of the ATRIUM-10 Fuel Design for Susquehanna", EMF-95-066(P), June 1995.
4.4-46 "Single Phase Hydraulic Performance of Exxon Nuclear BWR 9x9 Fuel Assembly", XN-NF-683(P), February 1983.
4.4-47 "Generic Mechanical Design Criteria for BWR Fuel Designs", ANF-89-98(P)(A) Revision 1, and Revision 1 Supplement 1, May 1995.
4.4-48 EMF-CC-074(P)(A), Volume 4, Revision 0, "BWR Stability Analysis - Assessment of STAIF with input from MOCROBURN-B2,"Siemens Power Corporation, August 2000.
4.4-49 "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description", XN-NF-80-19(P)(A) Volume 3 Revision 2, January 1987.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-22 4.4-50 "Impact of Failed/Bypassed LPRMs and TIPs and Extended LPRM Calibration Interval on Radial Bundle Power Uncertainty", EMF-1903 Revision 2, October 1996.
4.4-51 "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4 / MICROBURN-B2," emf-2158(P)(A), Rev. 0, October 1999.
4.4-52 "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, August 1996
4.4-53 "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960-A, June 1991
4.4-54 "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960-A Supplement 1, March 1992
4.4-55 "ABB Option III Oscillation Power Range Monitor (OPRM)," CENPD-400-P-A, Revision 1, May 1995 4.4-56 NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip Function", October 1995.
4.4-57 NEDC-32410P-A Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip Function", November 1997.
4.4-58 EMF-2209 (P) (A), Revision 2, "SPCB Critical Power Correlation, "Framatome ANP, September 2003.
4.4-59 NRC Letter from R. V. Guzman (NRC) to B. T. McKinney (PPL), January 30, 2008,
Subject:
Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Regarding 13-Percent Extended Power Uprate (TAC Nos. MD3309 and MD 3310) [Accessi on ML 080020182]
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-4, Rev. 54 AutoCAD Figure 4_4_4.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-10, Rev. 54 AutoCAD Figure 4_4_10.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-6A, Rev. 54 AutoCAD Figure 4_4_6A.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-6B, Rev. 54 AutoCAD Figure 4_4_6B.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-6C, Rev. 54 AutoCAD Figure 4_4_6C.doc
THIS FIGURE HAS BEEN REPLACED BY DWG.
M1-C12-8, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 4.6-6 replaced by dwg. M1-C12-8, Sh. 1 FIGURE 4.6-6, Rev. 49 AutoCAD Figure 4_6_6.doc
THIS FIGURE HAS BEEN REPLACED BY DWG.
M-146, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 4.6-5A replaced by dwg. M-146, Sh. 1 FIGURE 4.6-5A, Rev. 50 AutoCAD Figure 4_6_5A.doc THIS FIGURE HAS BEEN REPLACED BY DWG.
M-147, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 4.6-5B replaced by dwg. M-147, Sh. 1 FIGURE 4.6-5B, Rev. 55 AutoCAD Figure 4_6_5B.doc SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-1 4.1
SUMMARY
DESCRIPTION Susquehanna Units 1 and 2 are General Electric BWR/4 Boiling Water Reactors. Each reactor contains 764 fuel assemblies and 185 control rods arranged in an upright cylindrical configuration. Light water acts as both moderator and coolant. The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator and dryer assemblies, and jet pumps. Also included in the reactor assembly are the control rods, control rod drive housings, and the control rod drives. Figure 3.9-3, shows the arrangement of reactor assembly components. Important design and performance characteristics are discussed in Sections 4.2, 4.3 and 4.4. Loading conditions for reactor assembly components are specified in Section 3.9.
4.1.1 REACTOR
VESSEL The reactor vessel design and description are covered in Section 5.3.
4.1.2 REACTOR
INTERNAL COMPONENTS The major reactor internal components are the core (fuel, channels, control rods, and instrumentation), the core support structure (including the shroud, top guide and core plate), the shroud head and steam separator assembly, the steam dryer assembly, the feedwater spargers, the core spray spargers, and the jet pumps. Except for the Zircaloy in the reactor core, these reactor internals are stainless steel or other corrosion resistant alloys. All major internal components of the vessel can be removed except the jet pump diffusers, the jet pump risers, the shroud, the core spray lines, spargers, and the feedwater sparger. The removal of the steam dryers, shroud head and steam separators, fuel assemblies, in-core instrumentation, control rods, orificed fuel supports, and control rod guide tubes, can be accomplished on a routine basis.
4.1.2.1 Reactor Core 4.1.2.1.1 General The design of the boiling water reactor core, including fuel, is based on the proper combination of many design variables and operating experience. These factors contribute to the achievement of high reliability. A number of important features of the boiling water reactor core design are summarized in the following paragraphs:
(1)The BWR core mechanical design is based on conservative application of stress limits, operating experience, and experimental test results. The moderate pressu re level characteristics of a direct cycle reactor (approximately 1050 psia) result in moderate cladding te mperatures and stress levels.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-2 (2)The low coolant saturation temperature, high heat transfer coefficients, and near-neutral water chemistry of the BWR are significant, adva ntageous factors in minimizing Zirca loy temperature and associated temperature-dependent corrosion and hydride buildup.The relatively uniform fuel cladding temperatures throughout the core minimize migration of the hydrides to cold cladding zones and reduce thermal stresses.
(3)The basic thermal and mechanical criteria applied in the design have been proven by irradiation of statistically significant quantities of fuel. The design heat transfer rates an d linear heat generation rates are similar to values proven in fuel assembly irradiation.
(4)The design power distribution used in sizing the core represents a worst expected state of operation.
(5)The AREVA critical power methodology for boiling water reactors (References 4.1-1 2, and 4.1-29) is applied to assure that more than 99.9% of the fuel rods are expected to avoid boiling transition for the most severe abnormal operational transient described in Chapter 15. The possibility of boiling transition o ccurring during normal reactor operation is insignificant.
(6)Because of the large negative moderator density coefficient of reactivity during normal power operation, the BWR has a number of inherent advantages. These are the uses of coolant flow for power maneuvering, the inherent self-flattening of the radial power distribution, the ease of control, the spatial xenon stability, and the abilit y to override xenon, in order to follow load.Boiling water reactors do not have instability problems due to xenon. This has been demonstrated by special tests which have been conducted on operating BWRs in an attempt to force the reactor into xenon instability, and by calculations. No xenon instabilities have ever been observed in the test results. All of these indicators have proven that xenon transients are highly damped in a BWR due to the large negative power coefficient of reactivity (Reference 4.1-1). Important features of the reactor core arrangement are as follows:
(1)The bottom-entry cruciform control rods consist of B 4C in stainless steel tubes (i.e., B 4 C rods) only or a combination of B 4C rods and solid hafnium rods surrounde d by stainless steel. Control Rods are further described in subsections 4.1.3.2 and subs equent sections.(2)The fixed in-core ion chambers provide continuo us power range neutron flux monitoring.
A probe tube in each in-core assembly provides for a traversing ion cham ber for calibration and axial detail. Source and intermediate range monitors are located in-core and are axially retractable. The in-core location of the startup and source r ange instruments provides coverage of the large reactor core and provides an acceptable signal-to-noise ratio and neutron-to-gamma ratio. All in-core instrument leads enter fr om the bottom and the instruments are in service during refueling. In-core instr umentation is further discussed in Subsection 7.7.
1.6.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-3 (3) As shown by experience, the operator, utilizing the in-core flux monitor system, can maintain the desired power distribution within a large core by proper control rod scheduling.
(4) The channels (Zircaloy-2 or Zircaloy-4) provide a fixed flow path for the boiling coolant, serve as a guiding surface for the control rods, and protect the fuel during handling operations.
(5) The core is designed to be subcritical at any time in its operating history with any one control rod fully withdrawn.
(6) The selected control rod pitch provides the ability to finely control the power distribution in the assemblies contained in the reactor core. The pitch also allows ample clearance below the pressure vessel between control rod drive mechanisms for ease of maintenance and removal.
4.1.2.1.2 Core Configuration The reactor core is arranged as an upright circular cylinder containing a large number of fuel cells and is located within the reactor vessel. The coolant flows upward through the core. The Susquehanna SES Units utilize a conventional scatter loading with the lowest reactivity bundles placed in the peripheral region of the core. At periodic refueling intervals, each unit will enter an outage. During this time, fuel assemblies are identified to be discharged, new fuel is loaded and fuel assemblies in the reactor core may be "shuffled" to new locations. Therefore, the core loading patterns are both unit and cycle specific. The core configurations for each unit are discussed in Section 4.3.
4.1.2.1.3 Fuel Assembly Description The fuel assembly is composed of fuel and water rods (or interior water channels), structural components and a fuel channel. The mechanical design of the assembly is described in Section 4.2. The nuclear design of the assembly is described in Section 4.3. Thermal hydraulic design of the assembly is described in Section 4.4.
4.1.2.1.3.1 Fuel Rod A fuel rod consists of UO 2 pellets and a Zircaloy-2 cladding tube. A fuel rod is made by stacking pellets into a Zircaloy-2 cladding tube which is evacuated and back-filled with helium, and sealed by welding Zircaloy end plugs in each end of the tube.
The BWR fuel rod is designed as a pressure vessel. The ASME Boiler and Pressure Vessel Code,Section III, is used as a guide in the mechanical design and stress analysis of the fuel
rod.
The rod is designed to withstand the applied loads, both external and internal. The fuel pellet is sized to provide sufficient volume within the fuel tube to accommodate differential expansion between fuel and clad. Overall fuel rod design is conservative in its accommodation of the mechanisms affecting fuel in a BWR environment. Fuel rod design bases are discussed in more detail in Section 4.2.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-4 4.1.2.1.3.2 Fuel Bundle The fuel bundle has two important design features:
(1)The bundle design places minimum external forces on a fuel rod; each fuel rod is free to expand in the axial direction.
(2)The unique structural design permits the removal and replacem ent, if required, of individual fu el rods.Fuel bundles are designed to meet all the criteria for core performance and to provide ease of handling. Selected fuel rods in each assembly may differ from the others in initial uranium enrichment, burnable poison content, and fuel rod length. The variation in enrichment and burnable poison distribution produces more uniform power production across the fuel assembly, and thus allows a significant reduction in the amount of heat transfer surface required to satisfy the design thermal limitations. The inclusion of part length fuel rods in the assembly improves the two phase pressure drop, enhances the inherent stability of the bundle, and improves the required shutdown margin of the core design. Section 4.2 provides a more detailed description of the mechanical design aspects of the fuel bundles in use at Susquehanna. 4.1.2.1.4 Assembly Support and Control Rod Location All peripheral fuel assemblies are supported by the core plate. Otherwise, individual fuel assemblies in the core rest on fuel support pieces mounted on top of the control rod guide tubes. Each guide tube, with its fuel support piece, bears the weight of four assemblies and is supported by a control rod drive penetration nozzle in the bottom head of the reactor vessel. The core plate provides lateral support and guidance at the top of each control rod guide tube. The top guide, mounted inside the shroud, provides lateral support and guidance for each fuel assembly. The reactivity of the core is controlled by cruciform control rods which occupy alternate spaces between fuel assemblies. The position of each control rod is controlled by independent mechanical hydraulic drive systems. These systems insert and withdraw the control rod from the bottom of the core and can accurately position its associated control rod during normal operation and yet exert approximately ten times the force of gravity to insert the control rod during the scram mode of operation. Bottom entry allows optimum power shaping in the core, ease of refueling, and convenient drive maintenance.
4.1.2.2 Shroud The shroud is a cylindrical, stainless steel structure which surrounds the core and provides a barrier to separate the upward flow through the core from the downward flow in the annulus, and also provides a floodable volume in the unlikely event of an accident which tends to drain the reactor pressure vessel. A flange at the top of the shroud mates with a flange on the shroud head and steam separators. The upper cylindrical wall of the shroud and the shroud head form the core discharge plenum. The jet pump diffusers penetrate the shroud support below the core SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-5 elevation to introduce the coolant to the bottom head volume. The shroud support is designed to support and locate the jet pumps, core support structure, peripheral fuel assemblies and to separate the inlet and outlet flows of the recirculation loops. Mounted inside the upper shroud cylinder in the space between the top of the core and the upper shroud flange are the core spray spargers with spray nozzles for injection of cooling water. The core spray spargers and nozzles do not interfere with the installation or removal of fuel from the core.
4.1.2.3 Shroud Head and Steam Separators The shroud head consists of a flange and dome onto which is welded an array of standpipes, with a steam separator located at the top of each standpipe. The shroud head mounts on the flange at the top of the cylinder and forms the cover of the core discharge plenum region. The joint between the shroud head and shroud flange does not require a gasket or other replacement sealing technique. The fixed axial flow-type steam separators have no moving parts and are made of stainless steel. In each separator, the steam-water mixture rising from the standpipe impinges on vanes which give the mixture a spin to establish a vortex wherein the centrifugal forces separate the steam from the water. Steam leaves the separator at the top and passes into the wet steam plenum below the dryer. The separated water exits from the lower end of the separator and enters the pool that surrounds the standpipes to enter the downcomer annulus. For ease of removal, the shroud head is bolted to the shroud top flange by long shroud head bolts that extend above the separators for easy access during refueling. The shroud head is guided into position on the shroud via guide rods on the inside of the vessel and locating pins located on the shroud head. The objective of the shroud head bolt design is to provide direct access to the bolts during reactor refueling operations with underwater tool manipulation during the removal and installation of the assemblies. 4.1.2.4 Steam Dryer Assembly The steam dryer assembly is mounted in the reactor vessel above the shroud head and forms the top and sides of the wet steam plenum. Vertical guide rods on the inside of the vessel provide alignment for the dryer assembly during installation. The dryer assembly is supported by pads extending from the vessel wall and is prevented from lifting during postulated transients by brackets welded to the reactor vessel top head. Steam from the separators flows upward into the dryer assembly. Moisture is removed by the dryer vanes and flows first through a system of troughs and pipes to the pool surrounding the separators and then into the downcomer annulus between the core shroud and reactor vessel wall. The steam leaving the top of the dryer assembly flows into vessel steam outlet nozzles which are located alongside the steam dryer assembly. The schematics of a typical steam dryer panel are shown in Figures 4.1-2 and 4.1-3. 4.1.3 REACTIVITY CONTROL SYSTEMS 4.1.3.1 Operation SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-6 The control rods perform dual functions of power distribution shaping and reactivity control. Power distribution in the core is controlled during operation of the reactor by manipulation of selected patterns of rods. The rods, which enter from the bottom of the near-cylindrical reactor core, are positioned in such a manner to counter-balance steam voids in the top of the core and which results in significant power flattening.
These groups of control elements, used for power flattening, experience a somewhat higher duty cycle and neutron exposure than the other rods in the control system.
The reactivity control function requires that all rods be available for either reactor "scram" (prompt shutdown) or reactivity regulation. Because of this, the control elements are mechanically designed to withstand the dynamic forces resulting from a scram. They are connected to bottom-mounted, hydraulically actuated drive mechanisms which allow either axial positioning for reactivity regulation or rapid scram insertion.
The design of the rod-to-drive connection permits each blade to be attached or detached from its drive without disturbing the remainder of the control system. The bottom-mounted drives permit the entire control system to be left intact and operable for tests with the reactor vessel
open.
4.1.3.2 Description of Rods
For the original equipment and Duralife D160-C control rods the neutron absorber portion of the control rod is contained in the wings of the cruciform shaped control blades which are inserted in the bypass region between four fuel assemblies. The original equipment control blades contain boron-carbide (B 4C) powder filled stainless steel absorber tubes. Newer generation control blades contain a combination of B 4C filled tubes and solid hafnium rods. The boron-carbide absorber tubes are seal welded with end plugs on either end. Stainless steel balls are used to separate the tubes into individual compartments. The stainless steel balls are held in position by a slight crimp in the tube. The individual tubes act as pressure vessels to contain the helium gas released by the boron-neutron capture reaction. The tubes or rods are held in a cruciform array by a stainless steel sheath extending the full length of the tubes.
A top handle aligns the tubes and provides structural rigidity at the top of the control rod.
Rollers, housed in the handle, provide guidance for control rod insertion and withdrawal. A bottom casting is also used to provide structural rigidity and contains positioning rollers and a parachute-shaped velocity limiter. The handle and lower casting are welded into a single structure by means of a small cruciform post located in the center of the control rod.
Replacement Marathon control rods may use a modified handle assembly that eliminates pins and rollers present in the earlier design.
Marathon control blade wings are made up of an array of square tubes welded together. The tube arrays are welded to center tie rods to form the cruciform blade shape. The square tubes are loaded with either B 4 C or Hafnium. The B 4C is contained in separate capsules to prevent migration within the tubes. The square tubes are sealed at each end to prevent the neutron poisons from washing out into the coolant. The blade handle and velocity limiter are equivalent to previous control blade designs, (Reference 4.1-24). The Marathon Ultra - HD Control Rod design was introduced in U2C18, (Reference 4.1-30).
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-7 Westinghouse CR 99 control rods, introduced in U1C20, are designed similar to the original equipment and newer style Marathon control rod, (Reference 4.1-31). Like the above GE control rods, the Westinghouse CR 99 control rods have a cruciform blade shape, a handle, a B 4C loaded absorber zone and a velocity limiter. Different from the aforementioned GE control rods, the Westinghouse CR 99 control rod has horizontal absorber holes drilled into solid stainless steel wings and uses guide pads (buttons) or no guide pads, rather than upper pins and rollers, to guide control rod motion. Reference 4.1-31 provides additional discussion on the
design of the CR 99 control rod. The control rods can be positioned at 6-in. steps and have a nominal withdrawal and insertion speed of 3 in/sec. The velocity limiter, an engineered safety feature (ESF), is a device which is an integral part of the control rod and protects against the low probability of a rod drop accident. It is designed to limit the free fall velocity and reactivity insertion rate of a control rod so that minimum fuel damage would occur. It is a one-way device, in that control rod scram time is not significantly affected. Control rods are cooled by the core leakage (bypass) flow. The core leakage flow is made up of recirculation flow that leaks through the several leakage flow paths, which are:
(1)The area between fuel channel and fuel assembly nosepiece
- (2)The area between fuel assembly nosepiece and fuel support piece; (3)The area between fuel support piece and core plate; (4)The area between core plate and sh roud; and (5)The bypass flow holes in the fuel assembly nosepiece.
Further details of the control blade design are provided in Section 4.2. 4.1.4 ANALYSIS TECHNIQUES 4.1.4.1 Reactor Internal Components The following computer codes were used for initial design of the reactor internal components. Code descriptions are provided for historical purposes only. Computer codes used for the analysis of the internal components are listed as follows: (1)MASS(2)SNAP (MULTISHELL)(3)GASP (4)NOHEAT (5)FINITE (6)DYSEA (7)SHELL 5(8)HEATER(9)FAP-71(10)CREEP-PL ASTDetailed descriptions of these programs are given in the following sections:
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-8 4.1.4.1.1 MASS (Mechanical Analysis of Space Structure) 4.1.4.1.1.1 Program Description The program, proprietary of the General Electric Company, is an outgrowth of the PAPA (Plate
and Panel Analysis) program originally developed by L. Beitch in the early 1960s. The program is based on the principle of the finite element method. Governing matrix equations are formed in terms of joint displacements using a "stiffness-influence-coefficient" concept originally proposed by L. Beitch (Reference 4.1-2). The program offers curved beam, plate, and shell elements. It can handle mechanical and thermal loads in a static analysis and predict natural frequencies and mode shapes in a dynamic analysis.
4.1.4.1.1.2 Program Version and Computer The GE Nuclear Energy Division is using a past revision of MASS. This revision is identified as revision "0" in the computer production library. The program operates on the Honeywell 6000
computer.
4.1.4.1.1.3 History of Use Since its development in the early 60s, the program has been successfully applied to a wide variety of jet-engine structural problems, many of which involve extremely complex geometries.
The use of the program in the Nuclear Energy Division also started shortly after its development.
4.1.4.1.1.4 Extent of Application
Besides the Jet Engine and Nuclear Energy Divisions, the Missile and Space Division, the Appliance Division, and the Turbine Division of General Electric have also applied the program to a wide range of engineering problems. The Nuclear Energy Division (NED) uses it mainly for piping and reactor internals analyses.
4.1.4.1.2 SNAP (MULTISHELL)
4.1.4.1.2.1 Program Description
The SNAP Program, which is also called MULTISHELL, is the General Electric Code which determines the loads, deformations, and stresses of axisymmetric shells of revolution (cylinders, cones, discs, toroids, and rings) for axisymmetric thermal boundary and surface load conditions. Thin shell theory is inherent in the solution of E. Peissner's differential equations for each shell's influence coefficients. Surface loading capability includes pressure, average temperature, and linear-through-wall gradients; the latter two may be linearly varied over the shell meridian. The theoretical limitations of this program are the same as those of classical theory.
4.1.4.1.2.2 Program Version and Computer
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-9 The current version maintained by the General Electric Jet Engine Division at Evandale, Ohio is being used on the Honeywell 6000 computer in GE/NED. 4.1.4.1.2.3 History of Use The initial version of the Shell Analysis Program was completed by the Jet Engine Division in 1961. Since then, a considerable amount of modification and addition has been made to accommodate its broadening area of application. Its application in the Nuclear Energy Division has a history longer than ten years. 4.1.4.1.2.4 Extent of Application The program has been used to analyze jet engine, space vehicle and nuclear reactor components. Because of its efficiency and economy, in addition to reliability, it has been one of the main shell analysis programs in the Nuclear Energy Division of General Electric.
4.1.4.1.3 GASP 4.1.4.1.3.1 Program Description GASP is a finite element program for the stress analysis of axisymmetric or plane two-dimensional geometries. The element representations can be either quadrilateral or triangular. Axisymmetric or plane structural loads can be input at nodal points. Displacements, temperatures, pressure loads, and axial inertia can be accommodated. Effective plastic stress and strain distributions can be calculated using a bilinear stress-strain relationship by means of an iterative convergence procedure. 4.1.4.1.3.2 Program Version and Computer The GE version, originally obtained from the developer, Professor E. L. Wilson, operates on the Honeywell 6000 computer. 4.1.4.1.3.3 History of Use The program was developed by E. L. Wilson in 1965 (Reference 4.1-3). The present version in GE/NED has been in operation since 1967. 4.1.4.1.3.4 Extent of Application The application of GASP in GE/NED is mainly for elastic analysis of axisymmetric and plane structures under thermal and pressure loads. The GE version has been extensively tested and used by engineers in GE.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-10 4.1.4.1.4 NOHEAT 4.1.4.1.4.1 Program Description The NOHEAT program is a two-dimensional and axisymmetric transient nonlinear temperature analysis program. An unconditionally stable numerical integration scheme is combined with an iteration procedure to compute temperature distribution within the body subjected to arbitrary time- and temperature-dependent boundary conditions. This program utilizes the finite element method. Included in the analysis are the three basic forms of heat transfer, conduction, radiation, and convection, as well as internal heat generation. In addition, cooling pipe boundary conditions are also treated. The output includes temperature of all the nodal points for the time instants by the user. The program can handle multitransient temperature input. 4.1.4.1.4.2 Program Version and Computer The current version of the program is an improvement of the program originally developed by I.
Farhoomand and Professor E. L. Wilson of University of California at Berkeley (Reference 4.1-4). The program operates on the Honeywell 6000 computer. 4.1.4.1.4.3 History of Use The program was developed in 1971 and installed in General Electric Honeywell computer by one of its original developers, I. Farhoomand, in 1972. A number of heat transfer problems related to the reactor pedestal have been satisfactorily solved using the program. 4.1.4.1.4.4 Extent of Application The program using finite element formulation is compatible with the finite element stress-analysis computer program GASP. Such compatibility simplified the connection of the two analyses and minimizes human error. 4.1.4.1.5 FINITE 4.1.4.1.5.1 Program Description FINITE is a general-purpose finite element computer program for elastic stress analysis of two-dimensional structural problems including (1) plane stress, (2) plane strain, and (3) axisymmetric structures. It has provision for thermal, mechanical and body force loads. The materials of the structure may be homogeneous or nonhomogeneous and isotropic or orthotopic. The development of the FINITE program is based on the GASP program. (See
Subsection 4.1.4.1.3.) 4.1.4.1.5.2 Program Version and Computer SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-11 The present version of the program at GE/NED was obtained from the developer J. E. McConnelee of GE/Gas Turbine Department in 1969 (Reference 4.1-5). The NED version is used on the Honeywell 6000 computer. 4.1.4.1.5.3 History of Use Since its completion in 1969, the program has been widely used in the Gas Turbine and the Jet Engine Departments of the General Electric Company for the analysis of turbine components. 4.1.4.1.5.4 Extent of Usage The program is used at GE/NED in the analysis of axisymmetric or nearly axisymmetric BWR internals.
4.1.4.1.6 DYSEA 4.1.4.1.6.1 Program Description The DYSEA (Dynamic and Seismic Analysis) program is a GE proprietary program developed specifically for seismic and dynamic analysis of RPV and internals/building system. It calculates the dynamic response of linear structural system by either temporal modal superposition or response spectrum method. Fluid-structure interaction effect in the RPV is taken into account by way of hydrodynamic mass. Program DYSEA was based on program SAPIV with added capability to handle the hydrodynamic mass effect. Structural stiffness and mass matrices are formulated similar to SAPIV. Solution is obtained in time domain by calculating the dynamic response mode by mode. Time integration is performed by using Newmark's b-method. Response spectrum solution is also available as an option. 4.1.4.1.6.2 Program Version and Computer The DYSEA version now operating on the Honeywell 6000 computer of GE, Nuclear Energy Systems Division, was developed at GE by modifying the SAPIV program. Capability was added to handle the hydrodynamic mass effect due to fluid-structure interaction in the reactor. It can handle 3-Dimensional dynamic problem with beam, trusses, and springs. Both acceleration time histories and response spectra may be used as input. 4.1.4.1.6.3 History of Use The DYSEA program wa s developed in the summer of 1976. It has been adopted as a standard production program since 1977 and it has been used extensively in all dynamic and seismic analysis of the RPV and internals/ building system.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-12 4.1.4.1.6.4 Extent of Application The current version of DYSEA has been used in all dynamic and seismic analysis since its development. Results from test problems were found to be in close agreement with those obtained from either verified programs or analytic solutions.
4.1.4.1.7 SHELL 5 4.1.4.1.7.1 Program Description SHELL 5 is a finite shell element program used to analyze smoothly curved thin shell structures with any distribution of elastic material properties, boundary constraints, and mechanical thermal and displacement loading conditions. The basic element is triangular whose membrane displacement fields are linear polynomial functions, and whose bending displacement field is a cubic polynomial function (Reference 4.1-6). Five degrees of freedom (three displacements and two bending rotations) are obtained at each nodal point. Output displacements and stresses are in a local (tangent) surface coordinate system. Due to the approximation of element membrane displacements by linear functions, the in-plane rotation about the surface normal is neglected. Therefore, the only rotations considered are due to bending of the shell cross section and application of the method is not recommended for shell intersection (or discontinuous surface) problems where in-plane rotation can be significant. 4.1.4.1.7.2 Program Version and Computer A copy of the source deck of SHELL 5 is maintained in GE/NED by Y. R. Rashid, one of the originators of the program. SHELL 5 operates on the UNIVAC 1108 computer. 4.1.4.1.7.3 History of Use SHELL 5 is a program developed by Gulf General Atomic Incorporated (Reference 4.1-7) in 1969. The program has been in production status at Gulf General Atomic, General Electric, and at other major computer operating systems since 1970. 4.1.4.1.7.4 Extent of Application SHELL 5 has been used at General Electric to analyze reactor shroud support and torus. Satisfactory results were obtained.
4.1.4.1.8 HEATER 4.1.4.1.8.1 Program Description HEATER is a computer program used in the hydraulic design of feedwater sparger s and their associated delivery header and piping. The program utilizes test data obtained by GE using full SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-13 scale mockups of feedwater spargers combined with a series of models which represent the complex mixing processes obtained in the upper plenum, downcomer, and lower plenum. Mass and energy balances throughout the nuclear steam supply system are modeled in detail (Reference 4.1-8). 4.1.4.1.8.2 Program Version and Computer This program was developed at GE/NED in FORTRAN IV for the Honeywell 6000 computer. 4.1.4.1.8.3 History of Use The program was developed by various individuals in GE/NED beginning in 1970. The present version of the program has been in operation since January 1972. 4.1.4.1.8.4 Extent of Application The program is used in the hydraulic design of the feedwater spargers for each BWR plant, in the evaluation of design modifications, and the evaluation of unusual operational conditions.
4.1.4.1.9 FAP-71 (Fatigue Analysis Program) 4.1.4.1.9.1 Program Description The FAP-71 computer code, or Fatigue Analysis Program, is a stress analysis tool used to aid in performing ASME-III Nuclear Vessel Code structural design calculations. Specifically, FAP-71 is used in determining the primary plus secondary stress range and number of allowable fatigue cycles at points of interest. For structural locations at which the 3Sm (P+Q) ASME Code limit is exceeded, the program can perform either (or both) of two elastic-plastic fatigue life evaluations: 1)the method reported in ASME Pa per 68-PVP-3, 2) the present method documented in Paragraph NB-3228.3 of the 1971 Edition of the ASME Section III Nuclear Vessel Code. The program can accommodate up to 25 transient stress states of as many as 20 struct ural locations.
4.1.4.1.9.2 Program Version and Computer The present version of FAP-71 was completed by L. Young of GE/NED in 1971 (Reference 4.1-9). The program currently is on the NED Honeywell 6000 computer. 4.1.4.1.9.3 History of Use Since its completion in 1971, the program has been applied to several design analyses of GE BWR vessels.
4.1.4.1.9.4 Extent of Use SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-14 The program is used in conjunction with several shell analysis programs in determining the fatigue life of BWR mechanical components subject to thermal transients. 4.1.4.1.10 CREEP/PLAST 4.1.4.1.10.1 Program Description A finite element program is used for the analysis of two-dimensional (plane and axisymmetric) problems under conditions of creep and plasticity. The creep formulation is based on the memory theory of creep in which the constitutive relations are cast in the form of hereditary integrals. The material creep properties are built into the program and they represent annealed 304 stainless steel. Any other creep properties can be included if required. The plasticity treatment is based on kinematic hardening and von Mises yield criterion. The hardening modulus can be constant or a function of strain. 4.1.4.1.10.2 Program Version and Computer The program can be used for elastic-plastic analysis with or without the presence of creep. It can also be used for creep analysis without the presence of instantaneous plasticity. A detailed description of theory is given in Reference 4.1-11. The program is operative on Univac-1108. 4.1.4.1.10.3 History of Use This program was developed by Y. R. Rashid (Ref. 4.1-11) in 1971. It underwent extensive program testing before it was put on production status.
4.1.4.1.10.4 Extent of Application The program is used at GE/NED in the channel cross section mechanical analysis.
4.1.4.2 Fuel Rod Thermal Analysis Fuel Rod Thermal Design Analyses are performed utilizing the classical relationships for heat transfer in cylindrical coordinate geometry with internal heat generation. Steady state fuel rod thermal-mechanical analyses are performed to assure that fuel rod thermal-mechanical limits (e.g., steady state cladding strain and stress, hydrogen absorption, and, corrosion, etc.) are not exceeded. Abnormal operational transients are also evaluated to assure that the damage limit of 1.0% cladding plastic strain is not violated. Fuel rod analyses were performed with RAMPEX and approved versions of RODEX2, RODEX2A, and COLAPX codes. The fuel rod performance characteristics modeled by the RODEX2 and RODEX2A codes are:
-Gas release-Radial thermal conduction and gap conductance
-Free rod volume and gas pressure calculation s
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-15 - Pellet clad interaction (PCI) - Fuel swelling, densification, cracking, and crack healing
- Cladding creep deformation and irradiation induced growth RODEX2 determines the initial conditions for fuel rod power ramping analysis, performed using RAMPEX.
RODEX2A (Reference 4.1-28) determines the steady state strain, internal pressure, fuel cladding temperature, corrosion, hydrogen absorption, fuel temperature, and the fuel rod internal pressure for creep collapse analysis. This computer code is used to determine gap conductance for transient analysis.
Creep collapse analysis is performed using the COLAPX code.
Section 4.2 presents the fuel rod mechanical design and associated methodology.
4.1.4.3 Reactor Systems Dynamics The analysis techniques and computer codes used in reactor systems dynamics are described in section 4 of Reference 4.1-10. Subsection 4.4.4.6 also provides a complete stability analysis for the reactor coolant system.
Channel and core stability analyses were performed by Framatome ANP, Inc. on a fuel design and cycle specific basis using the STAIF code. A description of the methods employed by the STAIF code is provided in Reference 4.1-20. Using the RAMONA 5 code (References 4.1-26 and 27), FANP also performs transient stability analyses in support of the generation of OPRM setpoints.
4.1.4.4 Nuclear Engineering Analysis A brief summary of principal computer codes used in reactor core design and analysis is
provided below.
4.1.4.4.1 CASMO-4
The CASMO-4 computer code (Reference 4.1-25) was developed by STUDSVIK of America to perform steady state modeling of fuel bundles. CASMO-4 uses deterministic transport methods. At the pin cell level it exclusively uses a collision probability method to collapse the energy nuclear data into multi-group data. At the lattice level, it uses a method of characteristics for the neutron equation solution. CASMO-4, as opposed to CASMO-3G, does not need to do pin cell homogenization to perform a 2-D lattice wide transport calculation. The code is used to model each unique fuel lattice in the reactor to calculate few-group cross sections, bundle reactivities and relative fuel rod powers within a fuel bundle. The effects of conditions such as void, control rod presence, moderator temperature, fuel temperature, soluble boron, etc., are included in the model.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-16 4.1.4.4.2 MICROBURN-B2 The MICROBURN-B2 code (Reference 4.1-25) solves a two-group neutron diffusion equation based on an interface current method. It calculates the burnup chain equation for heavy nuclides and burnable poison nuclides, determines the three-dimensional core nodal power distribution, bundle flow, and void distributions. It also determines pin power distributions and thermal margins to technical specification limits. MICROBURN-B2 is used with CASMO-4. 4.1.4.5 Neutron Fluence Calculations Vessel neutron fluence calculations were performed to determine the azimuthal and axial variation of fluence at the vessel inside surface and at 1/4 T depth. The azimuthal and axial results were synthesized to obtain the fluence profile at the vessel inside surface and 1/4 T depth. The calculations also evaluate vessel fluence at power uprate conditions. Sections 4.3 and 5.3 provide additional detail regarding reactor pressure vessel irradiation. 4.1.4.6 Thermal Hydraulic Calculations XCOBRA (References 4.1-21, 4.1-22, and 4.1-23) calculates the steady state thermal hydraulic performance of a BWR. The code determines the flow and local fluid conditions at various axial positions in the core and represents the core as a collection of discrete parallel channels. The only interaction allowed between channels is the equalization of pressure in the inlet and outlet plenums. This is achieved by allowing the core flow to distribute among the various flow channels until the pressure drop in each channel is equalized. Pressure drop in each channel is determined through the application of two-phase pressure drop correlations and various data which hydraulically characterize the fluid channel. At a given axial position in the core, XCOBRA calculates a core-wide distribution of flow, enthalpy, density, quality, void fraction, and mass velocity.
4.
1.5 REFERENCES
4.1-1 Crowther, R. L. "Xenon Considerations in Design of Boiling Water Reactors," APED-5640, June 1968. 4.1-2 Beitch, L., "Shell Structures Solved Numerically by Using a Network of Partial Panels," AIAA Journal, Volume 5, No. 3, March 1967.
4.1-3 E.L. Wilson, "A Digital Computer Program For the Finite Element Analys is of Solids With Non Linear Material Properties," Aerojet General Technical Memo No. 23, Aerojet General, July 1965.
4.1-4 I.Farhoomand and E. L. Wilson, "No n-Linear Heat Transfer Analysis of Axisymmetric Solids," SESM Report SESM71-6, University of California at Berkeley, Berkeley, California, 1971.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-17 4.1-5 J. E. McConnelee, "Finite-Users Manual," General Electric TIS Report DF 69SL206, March 1969.
4.1-6 R. W. Clough and C. P. Johnson, "A Finite Element Approximation For the Analysis of Thin Shells," International Journal Solid Structures, Vol. 4, 1968.
4.1-7 "A Computer Program For the Structural Analysis of Arbitrary Three-Dimensional Thin Shells," Report No. GA-9952, Gulf General Atomic.
4.1-8 Burgess, A. B., "User Guide and Engineering Description of HEATER Computer Program," March 1974.
4.1-9 Young, L. J., "FAP-71 (Fatigue Analysis Program) Computer Code," GE/NED Design Analysis Unit R. A. Report No. 49, January 1972.
4.1-10 Carmichael, L.A. and Scatena, G. J., "Stability and Dynamic Performance of the General Electric Boiling Water Reactor," APED-5652.
4.1-11 Y. R. Rashid, "Theory Report for Creep-Plast Computer Program," GEAP-10546, AEC Research and Development Report, January 1972.
4.1-12 "Advanced Nuclear Fuels Corporation Critical Power Methodology," ANF-524(P)(A), Revision 2, and Supplement 1, Revision 2, November 1990.
4.1-13 Deleted 4.1-14 Deleted
4.1-15 Deleted
4.1-16 Deleted
4.1-17 Deleted
4.1-18 Deleted
4.1-19 Deleted
4.1-20 "STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain", EMF-CC-074(P)(A) Volumes 1 and 2, July 1994, and Volume 4, August
2000, Siemens Power Corporation, Richland WA 99352.
4.1-21 "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description", XN-NF-80-19(P)(A) Volume 3, Revision 2, Siemens Power Corporation, January 1987.
4.1-22 "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies", XN-NF-79-59(P)(A), November 1979.
4.1-23 "XCOBRA Users Manual", EMF-CC-43, Rev. 3, December 1995.
SSES-FSAR NIMS Rev. 65 FSAR Rev. 68 4.1-18 4.1-24 NEDE-31758P-A, "GE Marathon Control Rod Assembly" GE Nuclear Energy, October 1991. 4.1-25 EMF-2158 (P) (A) "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, " October 1999. 4.1-26 NEDO-32465-A "Reactor Stability Detect and Suppress Solution Licensing Basis Methodology for Reload Applications". 4.1-27 BAW-10255(P), "Cycle Specific Divom Methodology Using the ROMONA5-FA Code," Framatome ANP, Inc. September 2004. 4.1-28 XN-NF-85-74(P)(A). "RODEX2A(BWR) Fuel Thermal-Mechanical Evaluation Model", Exxon-Nuclear Company, Inc. February 1998. 4.1-29 EMF-2209 (P) (A), "SPCB SPCB Critical power Correlation," Framatome ANP, September 2003. 4.1-30 NEDE-33284 Supplement 1 P-A, Revision 1 March 2012, Licensing Topical Report, "Marathon - Ultra Control Rod Assembly" 4.1-31 WCAP-16182-P-A, Revision 1 October 2009, "Westinghouse BWR Control Rod CR 99 Licensing Report - Update to Mechanical Design Limits" THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.1-1, Rev. 54 AutoCAD Figure 4_1_1.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.1-4, Rev. 54 AutoCAD Figure 4_1_4.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.1-5, Rev. 54 AutoCAD Figure 4_1_5.doc THIS FIGURE HAS BEEN RENUMBERED TO 4.2-14 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure renumbered from 4.1-6 to 4.2-14 FIGURE 4.1-6, Rev. 54 AutoCAD Figure 4_1_6.doc THIS FIGURE HAS BEEN RENUMBERED TO 4.2-20 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure renumbered from 4.1-7 to 4.2-20 FIGURE 4.1-7, Rev. 54 AutoCAD Figure 4_1_7.doc THIS FIGURE HAS BEEN RENUMBERED TO 4.2-21 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure renumbered from 4.1-8 to 4.2-21 FIGURE 4.1-8, Rev. 54 AutoCAD Figure 4_1_8.doc Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-1 4.2 FUEL SYSTEM DESIGN The fuel system includes the fuel assembly (channeled fuel bundle) and the portion of the control
rod assembly which extends above the coupling mechanism on the control rod drive. The following
sections discuss the thermal/mechanical design bases, design descriptions, and design evaluations
for the fuel system components. Nuclear design is described in Section 4.3 and thermal hydraulic
design is described in Section 4.4.
4.2.1 Design
Bases
4.2.1.1 Fuel Assembly
The core designs described in Section 4.3 contain one fuel assembly design. The FANP
ATRIUM TM-10 is the primary fuel type loaded into the core. Occasionally, Lead Use Assemblies (LUAs) are also used in the core to provide operating experience with alternative fuel designs.
When used, LUAs are loaded in non-limiting locations in the core.
FANP ATRIUM TM-10 Fuel The mechanical design for the ATRIUM-10 TM assembly is based on compliance with generic mechanical design criteria established by FANP and approved by the NRC in Reference 4.2.6-10.
In accordance with the requirements of the approved mechanical design criteria, the ATRIUM TM-10 mechanical analyses were performed to provide the following assurances.
- 1) The fuel assembly shall not fail as a result of normal operation and AOO's.
- 2) Damage to fuel assemblies shall never prevent control rod insertion when required.
- 3) The number of fuel rod failures is not underestimated for postulated accidents.
- 4) Fuel coolability shall always be maintained.
Mechanical design analyses have been performed to evaluate the cladding stress and strain limits, fretting wear, oxidation, hydriding and crud buildup, fuel rod bowing, differential fuel rod growth, internal hydriding, cladding collapse, and cladding and fuel pellet overheat. The RODEX2, RODEX2A, RAMPEX, and COLAPX codes were used in the mechanical design analyses.
Mechanical analyses have also been performed to evaluate the ATRIUM TM-10 fuel design for Seismic/LOCA loads and for normal shipping and handling. In addition, FANP ATRIUM TM-10 fuel has been evaluated for power uprate conditions.
Results of these analyses and evaluations are discussed in Section 4.2.3.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-2 Lead Use Assemblies (LUA)
Occasionally, LUAs are loaded into the reactor core. The core loading of the LUAs will be such that
the assemblies are not loaded into thermally limiting locations. Mechanical design analyses are
performed for the LUA to evaluate fuel design parameters similar to that provided for the reload fuel.
4.2.1.2 Original Equipment Control Rod Assembly
The design bases for the original equipment control rod assembly are presented in Reference
4.2.6-3. The End-of-Life evaluation for the original equipment control rod assembly was modified in
accordance with PPL's response to IE Bulletin 79-26, Rev. 1. PPL has committed to replacing these
control rod assemblies prior to exceeding a limit of 34% B 10 depletion averaged over the upper one-fourth of the control rod assembly.
4.2.1.3 GE Duralife 160C Control Rod Assembly
The Duralife 160C control rod has been evaluated to assure it has adequate structural margin
under loading due to handling, and normal, emergency, and faulted operating modes. The loads
evaluated include those due to normal operating transients (scram and jogging), pressure
differentials, thermal gradients, seismic deflection, irradiation growth, and all other lateral and
vertical loads expected for each condition. The Duralife 160C control rod assembly design bases
have been reviewed and approved by the NRC (References 4.2.6-4 and 4.2.6-5).
4.2.1.4 GE Marathon Control Rod Assembly
The Marathon control rod has been evaluated to assure it has adequate structural margin under
loading due to handling, and normal, emergency, and faulted operating modes. The loads
evaluated include those due to normal operating transients (scram and jogging), pressure
differentials, thermal gradients, seismic deflection, irradiation growth, and all other lateral and
vertical loads expected for each condition. The Marathon control rod assembly design bases have
been reviewed and approved by the NRC (References 4.2.6-12 and 4.2.6-16).
4.2.1.5 Westinghouse CR 99 Control Rod Assembly
The Westinghouse CR 99 control rod has been evaluated to assure it has adequate structural
margin under loading due to handling, and normal, emergency, and faulted operating modes. The
loads evaluated include those due to normal operating transients (scram and jogging), pressure
differentials, thermal gradients, seismic deflection, irradiation growth, and all other lateral and
vertical loads expected for each condition. The Westinghouse CR 99 control rod assembly design
bases have been reviewed and approved by the NRC (Reference 4.2.6-17).
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-3
4.2.2 General
Design Description A summary of fuel design characteristics is provided in Table 4.2-14 for fuel designs loaded in the
core at SSES.
4.2.2.1 Core Cell
A core cell consists of a control rod assembly and the four fuel assemblies which immediately
surround the control rod. Figure 4.2-15 provides nominal dimensions for a core cell loaded with
FANP ATRIUM TM-10 fuel and a Duralife 160C control rod. Figures 4.2-15A and 4.2.15B provide nominal dimensions for a core cell loaded with FANP ATRIUM TM-10 fuel and a Marathon C+ control rod and a Marathon Ultra - HD control rod, respectively. Figure 4.2-15C provides the nominal
dimensions for a core cell loaded with FANP ATRIUM TM-10 fuel and a Westinghouse CR 99 control rod. These figures illustrate the general layout of a core cell while providing nominal dimensions for
fuel and control rods. A core cell may contain multiple fuel types, regardless of control rod type
utilized.
Each core cell is associated with a four-lobed fuel support piece. Around the outer edge of the
core, certain fuel assemblies are not immediately adjacent to a control rod and are supported by
individual peripheral fuel support pieces.
4.2.2.2 Fuel Bundle
FANP ATRIUM TM-10 Fuel An FANP ATRIUM TM-10 fuel bundle contains 83 full length and 8 part length fuel rods. The 8 part length fuel rods are provided to decrease the two phase pressure loss in the top of the bundle
thereby providing fuel bundle design that is more stable. A central water channel, which displaces
a 3x3 array of fuel rods near the center of the bundle, provides additional moderation within the
bundle thereby enhancing fuel utilization. The water channel is also used to fix the spacer locations
within the fuel bundle and serves as the main structural member connecting the upper and lower tie
plates. A total of 8 spacers are used to maintain fuel rod spacing. Reference 4.2.6-11 provides
detailed discussion of the various components of the ATRIUM TM-10 fuel bundle. Nominal dimensions for the ATRIUM TM-10 fuel bundle are provided in Figure 4.2-15. A schematic of an ATRIUM TM-10 fuel bundle is shown in Figure 4.2-17.
4.2.2.3 Fuel Assembly
A fuel assembly is a fuel bundle including the surrounding fuel channel. The fuel assemblies are
arranged in the reactor core to approximate a right circular cylinder inside the core shroud. Each
fuel assembly is supported by a fuel support piece and the top guide.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-4 The fuel channel enclosing the FANP fuel bundles is fabricated from Zircaloy-4 or Zircaloy-2 and performs these functions: the channel separates flow inside the bundle from the bypass flow
between channels; the channel guides the control rod and provides a bearing surface for it; the
channel provides rigidity for the fuel bundle. A channel fastener attaches the fuel channel to the
fuel bundle using a threaded post on the upper tie plate. Once the channel fastener has been
installed and the fuel assembly has been positioned in the core, the spring on the channel fastener
helps to hold the assembly against the core grid. A schematic of a fuel channel is shown in Figure
4.2-18. The fuel channel for U2C12 has a slightly different design and is shown in Figure 4.2-18-1.
The fuel channel for U1C14 reload fuel and subsequent reloads for Units 1 and 2 has been
changed to a 100 mill wall thickness and is shown in Figure 4.2-18-2, (Reference 4.2.6-9).
Co-resident fuel assemblies use the 80-mil fuel channel design. The difference between the 80-mil
and 100-mil channel is the thickness of the channel wall. Because the inside of the channel is the
same for the two designs, the top 20.1" of the 100-mil channel outer sides that face the upper guide
is reduced to 80-mil. This wall thickness will then allow for the bundle to be in the same lateral
position as the 80-mil channel. To maintain compatibility with an adjacent 80-mil channel fastener
in a control cell, there are also slight modifications to the 100-mil channel fastener to accommodate
the 100-mil channel. Starting with U1C15 the fuel channel mechanical design is based on
Reference 4.2.6-8.
The Advanced Fuel Channel (AFC) has been introduced on fuel for Units 1 and 2 beginning with
the reload for U1C20 and is shown in Figure 4.2-18-3. Co-resident fuel assemblies use the
standard 100-mil fuel channel design. The AFC dimensions are the same as the standard 100-mil
fuel channels with the difference that the AFC has thinned side-walls in the active core region. The
AFC use similar channel fasteners as the standard 100-mil fuel channels. Details regarding the
AFC are available in Reference 4.2.6-8.
U1C14 reload fuel uses the Framatome-ANP FUELGUARD lower tie plate design. The
FUELGUARD lower tie plate has the same outer envelope dimensions as the small hole lower tie
plate, including the same seal/finger springs. The basic difference between the two types of lower
tie plates is the grid. The FUELGUARD grid is constructed from wavy plates with support bars for
the fuel pins to sit on. The small hole design has machined holes to form a grid for the coolant to
flow through and the fuel pins to sit on. Framatome ANP performed flow tests to demonstrate that
the FUELGUARD lower tie plate is hydraulically compatibility with the small hole lower tie plate.
Both lower tie plates basically have the same loss coefficient. Subsequent reloads for Units 1 and 2
will use the Framatome-ANP FUELGUARD lower tie plate design.
A design change of the upper locking mechanism was incorporated into the ATRIUM-10 fuel
assemblies commencing with the Unit 2 Cycle 16 core loading. This design change is referred to
as the "Harmonized Advanced Load Chain" or HALC and replaces the previous load chain design.
The HALC was made to improve manufacturing reliability and improve the ability to remove the
upper tie plate of the fuel assembly. The design change does not affect the thermal, hydraulic or
mechanical performance of the fuel assembly. Subsequent reloads for Units 1 and 2 will use the
HALC design.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-5 Proper assembly orientation in the core is verified by visual inspection and is assured by verification procedures during core loading. Five visual indications of proper fuel assembly orientation exist.
These indications are:
- 1) The channel fastener assemblies are located at one corner of each fuel assembly adjacent to the center of the control rod.
- 2) The orientation boss on the fuel assembly handle points toward the adjacent control rod.
- 3) The channel spacing buttons are adjacent to the control rod passage area.
- 4) The assembly identification numbers which are located on the fuel assembly handles are all readable from the direction of the center of the cell.
- 5) There is cell-to-cell replication.
Proper assembly orientation in the core is shown in Figure 4.2-19.
4.2.2.4 Reactivity Control Assembly
4.2.2.4.1 Original Equipment Control Rod Assembly
The control rod consists of a sheathed cruciform array of commercial grade stainless steel tubes
filled with B 4 C powder. The main structural member of a control rod is made of Type-304 stainless steel and consists of a top handle, a bottom casting with a velocity limiter and control rod drive
coupling, a vertical cruciform center post, and four U-shaped absorber tube sheaths. Control rods
are cooled by core bypass flow.
Stellite rollers with Haynes Alloy 25 pins located at the top and bottom of the control rod help guide
the control rod as it is inserted and withdrawn from the core.
The control rod velocity limiter consists of cast austenitic stainless steel (Grade CF-8) and is an
integral part of the control rod bottom casting. The velocity limiter protects against high reactivity
insertion rate by limiting the control rod velocity in the event of a control rod drop accident.
Reference 4.2.6-3 provides the general design characteristics of the original equipment control rod
assembly. A diagram of this control rod is provided in Figure 4.2-20.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-6 4.2.2.4.2 Duralife 160C Control Rod Assembly The main differences between the Duralife 160C control rods and the original equipment control
rods are:
the Duralife 160C control rods utilize three solid hafnium rods at each edge of the cruciform to replace the three B 4 C rods that are most susceptible to cracking and to increase control rod life; the Duralife 160C control rods utilize improved B 4 C tube material (i.e. high purity type 304 stainless steel vs. commercial purity stainless steel) to eliminate cracking in the remaining
B 4 C rods during the lifetime of the control rod; the Duralife 160C control rods use GE's crevice-free structure design, which includes additional B 4 C tubes in place of the stiffeners, an increased sheath thickness, a full length weld to attach the handle and velocity limiter, and additional coolant holes at the top and
bottom of the sheath; the Duralife 160C control rods utilize low cobalt-bearing pin and roller materials in place of stellite which was previously used (PH13-8 Mo for pins, Inconel X750 for rollers);
the Duralife 160C control rod handles are longer by approximately 3.1 inches. The extended handle provides lateral support against the top grid and facilitates fuel moves
within the reactor vessel during refueling outages; the Duralife 160C control rods are roughly 10% to 15% heavier (depending on the velocity limiter design) as a result of the design changes described above; and the Duralife 160C control rod velocity limiter material is cast austenitic stainless steel grade CF-3.
References 4.2.6-4 and 4.2.6-5 provide additional discussion on the design of the various Duralife
control rod assemblies, including the features of the Duralife 160C inserted in the SSES Units. A
cross section of the Duralife 160C control rod is shown in Figure 4.2-15, and a diagram of this
control rod is provided in Figure 4.2-21.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-7 4.2.2.4.3 GE Marathon Control Rod Assembly The main difference between the Marathon control rod and the Duralife design control rods are:
the absorber tube and sheath arrangement of the Duralife designs is replaced with an array of square tubes resulting in reduced weight and increased absorber volume; the full length center tie rod is replaced with a segmented tie rod which also reduces weight.
References 4.2.6-12 and 4.2.6-16 provide additional discussion on the designs of the Marathon
control rod assemblies. A cross section of the Marathon C+ control rod is shown in Figure 4.2-15A.
A cross section of the Marathon Ultra HD Control Rod is shown in Figure 4.2-15B. A diagram of the
control rod is provided in Figure 4.2-22.
Replacement Marathon control rods may use a modified handle assembly that eliminates pins and
rollers present in the earlier design. Figure 4.2-22 is applicable to the rollerless design; the only
difference is that the small circles depicting the rollers would be removed from the diagram for the
rollerless design. The overall shape and dimensions of the upper handle remains unchanged.
4.2.2.4.4 Westinghouse CR 99 Control Rod Assembly
The Westinghouse BWR CR 99 control rod design is comparable to that of the GE design. Both
the Westinghouse and GE control blade designs have a coupling socket, velocity limiter, coupling
release handle, 143-inch active absorber zone, B 4 C as their main absorber material and a handle.
The overall length of the Westinghouse CR 99 control rod is the same as the OE GE control rod
design at 173-inches. The main difference in the design between the Westinghouse and GE
control rod is that the Westinghouse CR 99 control blade has horizontal absorber holes drilled in
solid stainless steel wings. Reference 4.2.6-17 provides additional discussion on the design of the
CR 99 control rod. A diagram of the CR 99 control rod is provided in Figure 4.2-23.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-8
4.2.3 Design
Evaluations 4.2.3.1 Fuel Design Evaluations
FANP ATRIUM TM-10 Fuel For the FANP ATRIUM TM-10 fuel, the design is such that adequate margins to fuel mechanical design limits (e.g., centerline melting temperature, transient strain, etc.) are assured for all
anticipated operational occurrences (AOOs) throughout the life of the fuel as demonstrated by the
fuel mechanical design analyses (References 4.2.6-10 and 4.2.6-11), provided that the steady state
fuel rod power remains within the power history assumed in the analyses. The design steady state
power history for the FANP ATRIUM TM-10 fuel is shown in Reference 4.2.6-11 and is incorporated into the Unit/Cycle specific Core Operating Limits Report (COLR) as an operating limit. The
operating limit may be in terms of planar or pellet exposure. ARTS has been implemented for Unit
1 and Unit 2. The COLR has a flow dependent LHGR multiplier and a power dependent multiplier, which are used to adjust the LHGR limit at off-rated conditions to assure that design limits are not
exceeded. The mechanical analyses support a maximum assembly average exposure of 49,400
MWD/MTU for fresh ATRIUM TM-10 fuel loaded in Cycles 12 and 13 on Unit 1 and Cycles 10 and 11 on Unit 2. Commencing with Unit 1 Cycle 14 and Unit 2 Cycle 12, the mechanical analyses for
fresh ATRIUM TM-10 assemblies support a maximum assembly average exposure of 54,000 MWd/MTU.
For U2C14, ARTS has been implemented. With ARTS the need for the FRTP/MFLPD adjustment
factor has been eliminated from the U2C14 Technical Specifications and the COLR. For U2C14, the COLR has a flow dependent LHGR multiplier and a power dependent multiplier, which are used
to adjust the LHGR limit at off-rated conditions.
FANP has evaluated the performance of ATRIUM TM-10 fuel assemblies under Susquehanna Seismic LOCA conditions. For this evaluation, maximum loads and/or stresses were calculated for
the fuel components under an acceleration load equivalent to a maximum dynamic load which
bounds the allowable bending moment in BWR/4 80 mil and 100 mil fuel channels, (Reference
4.2.6-9). The large margin that resulted from these analyses shows that the ATRIUM TM-10 fuel assembly with an 80 mil or 100 mil fuel channel demonstrates adequate structural integrity in the
Susquehanna Units under Seismic LOCA conditions. With regard to assembly liftoff, the net force
for the ATRIUM TM-10 fuel assembly was found to be downwards. Starting with U1C15 the fuel channel mechanical design is based on Reference 4.2.6-8.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-9 4.2.3.2 Results of Control Rod Assembly Design Evaluations 4.2.3.2.1 Original Equipment Control Rod Assembly
The original equipment control rod assembly design evaluations are discussed in Reference
4.2.6-3. Subsequent to the completion of the above referenced evaluations, a new failure
mechanism was identified for the original equipment control rod assembly. IE Bulletin 79-26 Rev-1
discusses this failure mechanism and recommends a reduction in the end-of-life criteria for the
original equipment control rod assembly. PPL committed to replacing the original equipment control
rod assemblies in accordance with this IE bulletin.
4.2.3.2.2 Duralife 160C Control Rod Assembly
The Duralife 160C control rod stresses, strains, and cumulative fatigue have been evaluated and
result in an acceptable margin to safety. The control rod insertion capability has been evaluated
and found to be acceptable during all modes of plant operation within the limits of plant analyses.
The Duralife 160C control rod coupling mechanism is equivalent to the original equipment coupling
mechanism, and is therefore fully compatible with the existing control rod drives in the plant. In
addition, the materials used in the Duralife 160C are compatible with the reactor environment. The
Duralife 160C control rods are roughly 10% to 15% heavier than the original equipment control rod
assembly, depending on the velocity limiter design utilized. The impact of the increased weight of
the control rods on the seismic and hydrodynamic load evaluation of the reactor vessel and
internals has been evaluated and found to be negligible.
With the exception of the crevice-free structure and the extended handle, the Duralife 160C control
rod is equivalent to the NRC approved Hybrid I Control Rod Assembly (Reference 4.2.6-4). The
mechanical aspects of the crevice-free structure were approved by the NRC for all control rod
designs in Reference 4.2.6-5. A neutronics evaluation of the crevice-free structure for the Duralife
160C design was performed by GE using the same NRC approved nuclear interchangeability
evaluation methodology as described in Reference 4.2.6-4. These calculations were performed for
the original equipment control rods and the Duralife 160C control rods assuming an infinite array of
FANP 9x9-2 fuel. The Duralife 160C control rod has a slightly higher worth than the original
equipment design, but the increase in worth is within the criterion for nuclear interchangeability.
The increase in rod worth has been taken into account in the appropriate reload analyses.
In Reference 4.2.6-4, the NRC approved the Hybrid I (Duralife 160C) control rod which weighs less
than the D lattice control rod. The basis of the Control Rod Drop Accident analysis continues to be
conservative with respect to control rod drop speed since the Duralife 160C control rod (including the extended handle, crevice free structure, and heavier velocity limiter) weighs less than the D
lattice control rod, and the heavier D lattice control rod drop speed is used in the
analysis. In addition, GE performed scram time analyses and determined that the Duralife 160C
control rod scram times are not significantly different than the original equipment control rod scram
times.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-10 Also, the scram speeds are monitored in the plant to assure compliance with safety analysis assumptions and technical specification limits.
IE Bulletin 79-26, Rev. 1 was issued to address B 4 C rod cracking and subsequent loss of boron in GE original equipment control rods. The Duralife 160C control rod design contains solid hafnium
absorber rods in locations where B 4 C tubes have historically failed. The remaining B 4 C rods are manufactured with an improved tubing material (high purity stainless steel vs. commercial purity
stainless steel), thus, boron loss due to cracking is not expected to occur.
Due to the control rod design, IE Bulletin 79-26, Rev. 1 does not apply to Duralife 160C control
rods. However, PPL plans to continue tracking the depletion of each control rod and discharge any
control rod prior to a ten percent loss in reactivity worth.
4.2.3.2.3 GE Marathon Control Rod Assembly
The form, fit and function of the Marathon control rod design are equivalent to the original
equipment control rods used at Susquehanna. Reference 4.2.6-12 documents NRC acceptance of
the GE Marathon control rod mechanical design.
The control rod stresses, strains, and cumulative fatigue were evaluated by GE Nuclear and result
in acceptable margins to safety. The control rod insertion capability was evaluated and found to be
acceptable during all modes of plant operation within the limits of plant analyses. In addition, the
coupling mechanism is fully compatible with the existing control rod drives in the plant. The
materials used in the Marathon control rods were also evaluated and are compatible with the
reactor environment. The Marathon control rods are approximately the same weight as the original
equipment control rods and, therefore, there is no impact on the seismic and hydrodynamic load
evaluation for the reactor vessel and internals. With lighter weight than the D160 control rods and
envelope dimensions less than or the same as the original equipment, the Marathon design is
compatible with existing NSSS hardware and there is no change in scram performance or drop
time.
Neutronics evaluations of the Marathon control rods by GE Nuclear using the methodologies
described in Reference 4.2.6-12 indicate the C lattice Marathon design for Susquehanna slightly
exceeds the +5% beginning-of-life reactivity worth constraint relative to the original equipment all
B 4 C design. Therefore, the effect of the increased reactivity worth on plant analyses had to be considered. The increased reactivity worth was found to not adversely impact normal operation, and is considered in the analysis of abnormal operational occurrences, infrequent events, or
accidents. The Marathon Ultra - HD control rod design satisfies the +5% beginning-of-life reactivity
worth constraint, (Reference 4.2.6-16).
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-11 The Marathon control rods used improved materials and contain significant design improvements to eliminate cracking and the associated loss of Boron experienced by the original equipment. GE
defines the end of life as a 10% reduction in cold reactivity worth in any 1/4 axial segment relative to
the initial undepleted state of the original equipment control rods. PPL will track the depletion of the
Marathon control rods and discharge any control rod prior to reaching the defined end of life, or
provide technical justification for its continued use.
4.2.3.2.4 Westinghouse CR 99 Control Rod Assembly
The form, fit and function of the Westinghouse CR 99 control rod design are equivalent to the
OE control rods used at Susquehanna. Reference 4.2.6-17 documents NRC acceptance of the
CR 99 control rod mechanical design.
The CR 99 control rod stresses, strains, and cumulative fatigue were evaluated by Westinghouse
and result in acceptable margins to safety. The CR 99 insertion capability was evaluated and found
to be acceptable during all modes of plant operation within the limits of plant analyses. The CR 99 coupling mechanism is fully compatible with the existing control rod drives in the plant. The
materials used in the CR 99 control rods were also evaluated and are compatible with the reactor
internals and the reactor environment. The CR 99 control rods are similar in nominal weight of the
OE control rods and, therefore, there is no impact on the seismic and hydrodynamic load evaluation
for the reactor vessel and internals. Scram speeds and settling times in the reactor are not
adversely affected by the CR 99 control rods. The CR 99 velocity limiter design is identical to the
design of the OE control rods and meets the assumption for the control rod drop accident.
The total worth of the CR 99 control rod is within +/-5% of the OE control rod. There is no
negative impact to shutdown margin and minimal impact on LPRM detector indications. The
nuclear end of life criteria is maintained as 10% reactivity worth decrease relative to the OE
control rod (Reference 4.2.6-17).
The CR 99 use of an improved high density absorber material, which is less sensitive to both
powder densification and absorber swelling due to neutron absorption reactions, minimizes the
possibility of absorber swelling causing contact with the surrounding stainless steel and
contributing stress. The CR 99 use of AISI 316L stainless steel, with its better resistance to fast
neutron irradiation assisted stress corrosion cracking (IASCC), also reduces the potential for
control blade cracking. SSES will track the depletion of the CR 99 control rod and discharge
any control rods prior to reaching the defined end of life, or provide technical justification for its
continued use.
4.2.4 Testing
and Inspection
4.2.4.1 Fuel Hardware and Assembly
Framatome - ANP, Inc (FANP) has developed Quality Control Standards for manufacturing, testing, and inspection of FANP components and fuel bundles. Details regarding FANP
manufacturing, testing, and inspection processes are available in Reference 4.2.6-6.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-12 On-site inspection of all new fuel bundles, fuel channels, and control rods is performed prior to installation into a Sus quehanna Unit. These inspections are controlled by plant procedures. The procedures were developed bas ed on guidelines provided by the fuel, channel, and control rod suppliers for rec eipt inspection and include acceptance criteria which are verified for each fuel bundle, fuel channel, or control rod.
The fuel channel management practices in place at Susquehanna are consistent with the recommendations contained in GE SIL 320, 'Recommendations for Mitigation of the Effects of Fuel Channel Bowing'. In addition, PPL will only use fuel channels for only one fuel bundle lifetime and will not reuse them. The fuel c hannel management practices are continuously reviewed against plant operation and industry practices.
4.2.4.2 Enrichment, Burnable Poison, and Absorber Rod Concentrations FANP has established adequate measures, in accordance with Reference 4.2.6-6, to assure that nuclear materials of v arying enrichment and form are pos itively identified and physically segregated as required to assure no inadvertent intermixing of enrichment forms. These measures include, as appropriate, identification of storage and processing containers, gamma sc an verific ation of powder, nuclear rod assay, analy tical ex aminations, in-process inspections, cleanouts of processing equipment between enrichments, administrative controls on the handling of materials, and audits of processing and product.
FANP fuel pellets are manufactured in accordance with approved procedures and are controlled by Product Design Specifications which define the allowable concentration tolerances and confidence levels required to verify enrichment and burnable poison concentrations.
General Electric (GE) supplied the original equipment control rods for both Sus quehanna Units and is the supplier of the Duralife-160C and Marathon replacement control rods. The absorber materials, boron carbide and hafnium, are c ertified by GE to meet GE Material Specific ations. The isotopic B 10 content and boron content is verified for eac h powder lot received by General Electric.
All boron c arbide absorber rod assemblies are s ubjected to a leakage test to insure abs orber rod integrity. GE performs analysis of hafnium absorber rod lots to insure chemical composition is in c onformance with GE Material Specifications.
Wes tinghouse supplied the CR 99 control rods for use at both Susquehanna Units. The boron c arbide absorber material is certified by Westinghouse to meet Westinghouse Material Specifications. The isotopic B 10 content and boron content is verified for each powder lot receiv ed by Westinghous
- e. Eac h CR 99 control rod blade is leak tested with helium per the Westinghouse Materials Specifications.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-13 4.2.4.3 Surveillance, Inspection, and Testing PPL has a fuel reliability program that includes fuel performance monitoring and fuel failure
response. On-line fuel performance monitoring is conducted to determine whether there is a fuel
failure and may include evaluation of the general location of the failed assembly, the number of fuel
assemblies suspected, when the failure occurred, and the approximate exposure of the failed
assembly. Determination of this information prior to refueling allows preparation for changes in the
following cycle's core design. In addition, control rod sequence exchanges and full power control
rod patterns can be developed to minimize the offgas release from the failed rod(s) and stress on
the suspect assembly during power maneuvering.
On-line fuel performance monitoring is performed at the Susquehanna station by periodic
evaluation of pretreatment offgas activity and/or reactor coolant samples. Verification of failed fuel
is made by periodic evaluation of the pretreatment xenon and krypton offgas activity and reactor
water cleanup system iodine and cesium activity. The general location of the failed fuel assembly
is, typically, identified by control rod motion testing and monitoring of the pretreatment offgas
activity. Identification of the exact assembly may be performed by sipping or ultrasonic testing (UT)
of the suspect assemblies.
Post-irradiation fuel failure evaluations are, typically, performed to determine the exact fuel rod
location within the assembly and the root cause of a fuel rod failure. The exact location of the failed
rod may be determined by UT or eddy current testing. Root cause evaluations may include review
of manufacturing and inspection records, visual examination of the failure location, and destructive
examination of the failed fuel rod.
Post-irradiation inspection programs have been developed by FANP to evaluate fuel design
performance. Reference 4.2.6-11 discusses the FANP inspection and surveillance program for
irradiated ATRIUM TM-10 fuel.
4.2.5 Operating
and Developmental Experience
FANP ATRIUM TM-10 fuel has been utilized at SSES beginning with Unit 1 Cycle 11 and Unit 2 Cycle 9.
PPL continually tracks the performance of all fuel in the Susquehanna Units in an effort to identify
indications of potential fuel rod failures.
Prior to the implementation of a mechanical fuel design into either Susquehanna Unit, that
introduces features not currently in other operating plants, a plan will be developed to evaluate the
performance of this fuel design in the Susquehanna Units. This plan may include pre and post irradiation fuel assembly characterization, visual inspection, power maneuvering evaluations, fuel
clad corrosion evaluations, and UT inspections.
PPL occasionally participates in Lead Use Assembly programs. These programs allow the
company to evaluate and gain operating experience with new fuel designs.
Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-14
4.2.6 References
4.2.6-1 Deleted 4.2.6-2 Deleted 4.2.6-3 "BWR/4 and BWR/5 Fuel Design," NEDE-20944(P), General Electric Company, October 1976, and Letter from Olan D. Parr (NRC) to Dr. G. G. Sherwood (GE), "Review of General Electric Topical Report NEDE-20944-P, BWR/4 and BWR/5
Fuel Design (NEDO-20944 Non-Proprietary Version)", September 30, 1977. 4.2.6-4 "Safety Evaluation of the General Electric Hybrid I Control Rod Assembly for the BWR 4/5 C Lattice, "NEDE-22290-A, General Electric Company, September 1983, and Supplement 1, General Electric Company, July 1985. 4.2.6-5 "Safety Evaluation of the General Electric Duralife 230 Control Rod Assembly, "
NEDE-22290-A, Supplement 3, General Electric Company, May 1988. 4.2.6-6 "Nuclear Fuel Business Group Quality Management Manual, "NFQM, Rev. 0, Framatome-ANP,U.S. Version, June 2002.
4.2.6-7 Deleted 4.2.6-8 "Mechanical Design for BWR Fuel Channels", EMF-93-177 (P) (A) Rev. 1, August 2005. 4.2.6-9 "Mechanical Design for BWR Fuel Channels," EMF-93-177 (P)(A) and Supplement 1, Siemens Power Corporation, August 1995. 4.2.6-10 "Generic Mechanical Design Criteria for BWR Fuel Designs," ANF-89-98(P)(A), Rev.
1 and Rev. 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995. 4.2.6-11 "Mechanical Design Evaluation for Siemens Power Corporation ATRIUM TM BWR Reload Fuel, " EMF-95-52(P), Rev. 2, Siemens Power Corporation -
Nuclear Division, December 1998. 4.2.6-12 "GE Marathon Control Rod Assembly, " NEDE-31758P-A, GE Nuclear Energy, October 1991.
4.2.6-13 Deleted
4.2.6-14 Deleted
4.2.6-15 Deleted Text Rev. 66 SSES-FSAR FSAR Rev. 68 4.2-15 4.2.6-16 NEDE - 33284 Supplement 1 P-A, Revision 1 March 2012, Licensing Topical Report, "Marathon Ultra Control Rod Assembly". 4.2.6-17 WCAP-16182-P-A, Revision 1 October 2009, "Westinghouse BWR Control Rod CR 99 Licensing Report - Update to Mechanical Design Limits".
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-1, Rev. 54 AutoCAD Figure 4_2_1.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-2, Rev. 54 AutoCAD Figure 4_2_2.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-3, Rev. 54 AutoCAD Figure 4_2_3.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-4, Rev. 54 AutoCAD Figure 4_2_4.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-5, Rev. 54 AutoCAD Figure 4_2_5.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6, Rev. 54 AutoCAD Figure 4_2_6.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-7, Rev. 54 AutoCAD Figure 4_2_7.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8, Rev. 54 AutoCAD Figure 4_2_8.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-9, Rev. 54 AutoCAD Figure 4_2_9.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-10, Rev. 54 AutoCAD Figure 4_2_10.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-11, Rev. 54 AutoCAD Figure 4_2_11.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-12, Rev. 54 AutoCAD Figure 4_2_12.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-13, Rev. 54 AutoCAD Figure 4_2_13.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-14, Rev. 55 AutoCAD Figure 4_2_14.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-16, Rev. 55 AutoCAD Figure 4_2_16.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6a, Rev. 48 AutoCAD Figure 4_2_6a.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6b, Rev. 48 AutoCAD Figure 4_2_6b.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-6c, Rev. 48 AutoCAD Figure 4_2_6c.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8a, Rev. 54 AutoCAD Figure 4_2_8a.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8b, Rev. 54 AutoCAD Figure 4_2_8b.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.2-8c, Rev. 54 AutoCAD Figure 4_2_8c.doc
For Information Only SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-1 4.3 NUCLEAR DESIGN The nuclear design of the initial cores for Susquehanna is described in References 4.3-1, 4.3-2, and 4.3-3. This section incorporates much of the general nuclear design information in Reference 4.3-1 and presents detailed design information for reload cores. 4.3.1 Design Bases Nuclear design bases fall into two categories: safety design bases and core performance design bases. Safety design bases are required by the General Design Criteria to ensure safe operation of the core. Core performance design bases are required to meet power production objectives. 4.3.1.1 Safety Design Bases Safety design bases protect the nuclear fuel from damage which would result in a release of radioactivity, representing an undue risk to the health and safety of the public. Safety design bases are listed below.
1)The core shall be capable of being rendered subcritical at any time or core condition wi th the highest worth control rod fully withdrawn.
2)The void coefficient shall be negative over the entire operating range.
3)Technical specification limits on Linear Heat Generation Rate (LHGR), Minimum CriticalPower Ratio (MCPR), and the Average Planar Linear Heat Generation Rate (APLHGR), shall not be exceeded during steady state operation.
4)The nuclear characteristics of the design shall not exhibit any tendency toward divergen t operation.5)Reload fuel lattice enrichments shall be such that the nuclear design bases are met for the new fuel storage racks (section 9.1.1.1.1.2) and spent fuel storage (section 9.1.2.1.1.2).4.3.1.2 Plant Performance Design Bases 1)The core design shall have adequate excess reactivity to reach the desired cycle length
.2)The core design shall be capable of operating without exceeding technical specification lim its.3)The core and fuel design and the reactivity control system shall allow continuous, stableregulation of reactivity
.4)The core and fuel design shall have adequate reactivity feedback to facilitate normaloperation.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-2 4.3.2 Description A general description of BWR nuclear characteristics is provided in Reference 4.3-1. A summary of reactor core characteristics for Susquehanna is listed in Table 4.3-1. 4.3.2.1 Nuclear Design Description The nuclear design of Susquehanna is both unit and cycle specific. A detailed description of the initial core nuclear design is available in Reference 4.3-1. Susquehanna Steam Electric Station Units 1 and 2 operate at power conditions in Table 4.3-1 with increased core flow. Fuel bundle and core reload designs have been developed using NRC approved methods. 4.3.2.1.1 Core Composition The core contains 764 fuel assemblies arranged in a conventional scatter loaded pattern. Typically, the lowest reactivity fuel assemblies are placed in the peripheral region of the core. SSES uses FANP ATRIUM TM-10 fuel as the primary reload fuel mechanical design. In addition, a limited number of Lead Use Assemblies (LUAs) may be loaded to evaluate new fuel designs. Detailed core compositions are presented in Tables 4.3-2 and 4.3-3 for Units 1 and 2, respectively. The core loading patterns for both units are shown in Figures 4.3-1 and 4.3-2. 4.3.2.1.2 Fuel Bundle Nuclear Design Reference 4.3-1 describes the first core bundle designs and related fuel nuclear properties. Reload fuel bundle design descriptions are presented below. The burnup dependent behavior of certain nuclear properties is primarily a function of enrichment and does not change significantly with bundle mechanical design. These characteristics include Uranium depletion and Plutonium buildup, fission fraction, delayed neutron fraction, and neutron lifetime. Figures 4.3-3 through 4.3-7 show the typical response of these characteristics with burnup for an enriched reload fuel bundle lattice. Several FANP ATRIUM TM-10 bundle designs are in use at SSES. Each design may have a unique axial enrichment distribution, radial enrichment distribution, or burnable absorber loading. Figure series 4.3-8 shows the nominal axial zoning for the various fuel bundles used in the reload cores. Figure series 4.3-9 shows the nominal radial enrichment distributions for the various lattice types used in the fuel bundles. Table 4.3-6 lists the fuel types currently used and the associated 4.3-8 series figure numbers. Table 4.3-6 also lists the lattice numbers used by each fuel type. Table 4.3-7 lists the lattice currently used and the associated 4.3-9 series figure numbers. 4.3.2.2 Power Distributions This section presents typical power distributions for SSES reload cores. Typical local, core radial, and core axial power distributions for the initial core are described in Reference 4.3-1. The core is designed such that the resulting power distributions meet the thermal limits identified in the plant Technical Specifications. The primary criteria for thermal limits are the Maximum Linear Heat Generation Rate (MLHGR) and the Minimum Critical Power Ratio (MCPR). In addition, a Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit is applied to the plant.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-3 Each of these parameters is a function of the core 3-D power distribution and the local rod-to-rod power distribution. Design calculations are performed to ensure that the core meets thermal limits and to demonstrate that the power distributions comply with the cycle design envelope.
The local peaking factor is defined as the ratio of the power density in the highest power rod in a lattice to the average power density in the lattice. Local effects on Critical Power Ratio are characterized by F-effective. Both the local peaking factor and F-effective have associated target values which typically satisfy the design envelope. Gross power peaking in the core is defined as the ratio of the maximum power density in any axial segment of any bundle in the core to the average core power density. Design allowances are included in the design stage to ensure that thermal limits are met. During plant operation, the power distributions are measured by the in-core instrumentation system and thermal margins are calculated by the core monitoring system.
4.3.2.2.1 Local Power Distribution
The local rod-to-rod power distribution and associated F-effective distribution are a direct function of the lattice enrichment distribution. Near the outside of the lattice where thermal flux peaks due to interbundle water gaps, low enrichment fuel rods are utilized to reduce power peaking. Closer to the center of the bundle, higher enrichment rods are used to increase power peaking and flatten the bundle power distribution. In addition, the water rods (or water channels) in the center of the lattice increase thermal flux and cause more power to be produced in the center of the lattice. The combination of enrichment and water channels results in a relatively flat power distribution.
To control bundle reactivity, Gd 2 O 3 is utilized as a burnable absorber. Power is suppressed in gadolinia bearing fuel rods early in bundle life. As gadolinia is depleted, power in these rods initially increases, then decreases as fuel is depleted. Local power distributions are calculated using licensed methodology described in Section 4.3.3.
Figure 4.3-11-1 shows bundle reactivity (k) as a function of void fraction and burnup for an FANP ATRIUM TM-10 fuel assembly dominant lattice. At low exposure, reactivity is higher for lower void fractions. As exposure increases the curves cross, largely due to the effect of void history and the increase in plutonium buildup.
Figures 4.3-11-2 to 4.3-11-4 show typical unrodded local power distributions for an FANP ATRIUM TM-10 fuel assembly dominant lattice as a function of burnup with a constant void fraction. Figures 4.3-11-2, 4.3-11-5, and 4.3-11-6 show typical unrodded local power distributions for a fresh ATRIUM TM-10 fuel dominant lattice as a function of void fraction at BOC. Figure 4.3-11-7 shows the typical response of the unrodded maximum local peaking factor as a function of void fraction and burnup.
4.3.2.2.2 Radial Power Distribution The integrated bundle power, commonly referred to as the radial power, is the primary factor for determining MCPR. At rated conditions the MCPR is directly proportional to the radial power. The radial power distribution is a function of the control rod pattern in the core, the fuel bundle type and loading pattern, and void distribution. Radial power is calculated using the licensed methodology described in Section 4.3.3.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-4 4.3.2.2.3 Axial Power Distribution Axial power distributions in a BWR are a function of control rod position, steam voids, axial gadolinia distribution, and the exposure distribution. Voids tend to skew power toward the bottom of the core; bottom entry control rods reduce the power in the bottom of the core; and the axial gadolinia distribution assists in flattening the power in the bottom of the core. Since the void distribution is primarily determined by the power shape, the two means available for axial power shape optimization are the control rods and gadolinia. Typically, the core axial power shape is bottom peaked at BOC and becomes top peaked at EOC. Axial power shapes are calculated using the licensed methodology described in Section 4.3.3. 4.3.2.2.4 Power Distribution Measurements Power distribution measurement methodology and measurement uncertainties are described in References and 4.3-10 and 4.3.13. 4.3.2.2.5 Power Distribution Accuracy The accuracy of calculated power distributions is discussed in References 4.3-10 and 4.3-13. 4.3.2.3 Reactivity Coefficients Reactivity coefficients are differential changes in reactivity produced by differential changes in core conditions. These coefficients are useful in calculating the response of the core to varying plant conditions. The initial condition of the core and the postulated initiating event determine which of the coefficients are significant in evaluating core response. The dynamic behavior of BWRs over all operating states can be characterized by three reactivity coefficients. These coefficients are the Doppler coefficient, the moderator temperature coefficient, and the void coefficient. The Power coefficient is also associated with a BWR; however, this coefficient is the combination of the Doppler and void coefficients in the operating range. Reactivity coefficients are calculated using the licensed methods described in Section 4.3.3. 4.3.2.3.1 Void Coefficient The most important reactivity coefficient in a BWR is the void coefficient. The void coefficient must be large enough to prevent power oscillation due to spatial xenon changes, but it must be small enough that pressurization transients do not limit plant operation. The void coefficient inherently flattens the radial power distribution during normal operation and provides enhanced reactor control through the void feedback mechanism. The overall void coefficient is always negative over the complete operating range since the BWR design is typically undermoderated. Void formation changes reactivity by reducing the amount of water available for neutron moderation, thus increasing neutron leakage. Typical values for the void coefficient are listed in Table 4.3-4.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-5 4.3.2.3.2 Moderator Temperature Coefficient The moderator temperature coefficient (MTC) is the least important of the reactivity coefficients since its effect is limited to a very small portion of the reactor operating range. Once the reactor reaches the power production range, boiling begins and the MTC remains essentially constant.
Like the void coefficient, the moderator coefficient is associated with the amount of neutron moderation in the water. The MTC is negative during power operation; however, under cold
conditions beginning soon after BOC, the MTC may become slightly positive. The range of values of MTCs in reload lattices does not include any that are significant from a safety point of view. Typical values for the MTC are listed in Table 4.3-4. The small magnitude of this coefficient, relative to that associated with steam voids, combined with the long time-constant associated with heat transfer from fuel to coolant, makes the reactivity contribution of a change in moderator temperature insignificant during rapid transients. 4.3.2.3.3 Doppler Temperature Coefficient The Doppler Temperature coefficient (DTC) is the change in reactivity due to a change in fuel temperature. This change in reactivity occurs due to the broadening of the fuel resonance absorption cross sections as temperature increases. The DTC is primarily a measure of the Doppler broadening of U238 and Pu240 resonance absorption peaks. An increase in fuel temperature increases the effective resonance absorption cross section of the fuel and produces a corresponding reduction in reactivity. The Doppler coefficient changes as a function of core life representing the combined effects of fuel temperature reduction with burnup and the buildup of Pu240. Typical values for the Doppler coefficient are listed in Table 4.3-4. 4.3.2.3.4 Power Coefficient The power coefficient is determined from the composite of all the significant individual sources of reactivity change associated with a differential change in reactor power. This coefficient assumes constant xenon. Typical values for the power coefficient may be obtained from Reference 4.3-1 for the initial cores. 4.3.2.4 Control Requirements The core and fuel design in conjunction with the reactivity control system provide a stable system for BWRs. The control rod system is designed to provide adequate control of the maximum excess reactivity anticipated during the equilibrium fuel cycle operation. Since fuel reactivity is a maximum and control rod worth is a minimum at ambient temperature, the shutdown capability is evaluated assuming a cold, xenon free core. The safety design basis requires that the core, in its maximum reactivity condition, shall be subcritical with all control rods inserted except with the highest worth rod completely withdrawn. This limit allows control rod testing at any time in core life and assures that the reactor can be made subcritical by control rods alone. The typical behavior of hot excess reactivity as a function of cycle exposure for SSES Units 1 and 2 is shown in Figure 4.3-12.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-6 4.3.2.4.1 Shutdown Reactivity Core Shutdown Margin calculations are performed to assess whether the basic criterion for reactivity control is met by the reload design. This criterion requires that the core, under cold, no xenon conditions, must be subcritical with the highest worth control rod fully withdrawn and all other rods fully inserted. SSES Technical Requirements Manual requires a shutdown margin of at least 0.38% k/k. The shutdown margin requirement is based on the uncertainties associated with the statistical variance of cold criticality calculations at a given exposure, plus a manufacturing uncertainty. The manufacturing uncertainty results from the fact that the calculated highest worth control rod may not be the highest worth rod in reality due to the stackup of manufacturing tolerances in a control cell.
Core Shutdown margin is very dependent on bundle and core designs and is a function of core exposure. Gadolinia loading, enrichment loading, and core loading all significantly affect core and local cell reactivity as a function of exposure. As a result, shutdown margin must be evaluated throughout the expected cycle operation to assure adequate margin to Technical Specification requirements. For design purposes, an additional uncertainty is added to the Technical Specification value to account for prediction uncertainties.
Shutdown margin is calculated as a function of cycle exposure in the following manner:
%100*)E (bias)E (k)E (bias)E (k 1)E (SDM eff eff where; SDM(E) = core shutdown margin (%k/k) at cycle exposure E, K eff(E) = core k-effective at cycle exposure (E) with all rods in except the strongest worth rod (68F with no xenon),
Bias(E) = core k-effective bias for cold core simulation model at cycle exposure (E). The bias equals the target cold core simulation model critical k-effective minus 1.0.
The Cycle R value is determined from the evaluation of shutdown margin as a function of cycle exposure. The R value is used to determine shutdown margin testing requirements, and it is defined as the difference between the calculated beginning of cycle shutdown margin minus the calculated minimum shutdown margin during the cycle, where shutdown margin is a positive number. The value of R must be either positive or zero and must be determined for each fuel loading cycle.
Typical behavior of shutdown margin as a function of cycle exposure for SSES Units 1 and 2 is shown in Figure 4.3-13.
A description of the methods used to calculate shutdown margin is provided in Section 4.3.3.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-7 4.3.2.4.2 Reactivity Variations Reference 4.3-1 provides a general discussion of reactivity variations in a BWR/4. The reference provides tables showing typical k-effective values for various power levels, control fractions, and Xenon concentrations. From this data, the general reactivity effect of changing a single core variable can be determined. 4.3.2.5 Control Rod Patterns and Reactivity Worths 4.3.2.5.1 RWM Range Below the low power setpoint, control rod patterns follow prescribed withdrawal and insertion sequences restricted by the Rod Worth Minimizer (RWM). The sequences are established to assure that the maximum insequence control rod or rod segment reactivity worth would not be sufficient to result in a deposited fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal or insertion. Further discussion of the RWM and control rod sequence limitations is provided in Section 15.4.9 (Control Rod Drop Accident). 4.3.2.5.2 Operating Range In the power range, above the low power setpoint, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak fuel enthalpy of 280 cal/gm. Therefore, restrictions on control rod patterns are not required to minimize control rod worths. During power operation the control rod patterns are selected based on the measured core power distributions. For reload design purposes, optimized control rod patterns are selected for the cycle depletion. The series of design control rod patterns form the Cycle Step Out. Control rod sequence identification (A2, B2, A1, B1) is defined in Reference 4.3-1. 4.3.2.5.3 SCRAM Reactivity The reactor protection system (RPS) is capable of shutting down the reactor by initiating a SCRAM. The RPS and the control rod drive (CRD) system act quickly enough to prevent the initiating event from driving the fuel beyond transient limits. During a SCRAM from operating conditions, the control rod worth, reactor power, delayed neutron fraction, and void distributions must be properly accounted for as a function of time. The methodology used to account for these variables and determine SCRAM reactivity is described in Section 4.3.3. 4.3.2.6 Criticality of Reactor During Refueling Criticality of fuel assemblies in the core during refueling is avoided by assuring that the Technical Specification shutdown margin requirement is met. For core shuffles, a shutdo wn margin design criterion is defined to account for prediction uncertainties. This criterion helps determine the acceptability of a fuel move for meeting the Technical Specification limit. A description of the methods used to calculate shutdown margin is provided in Section 4.3.3.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-8 4.3.2.7 Stability Boiling Water Reactors do not have instability problems due to Xenon. Xenon transients are highly damped in a BWR due to the large negative power coefficient. References 4.3-1 and 4.3-3 provide additional discussion of Xenon instability. Thermal hydraulic stability is discussed in detail in Section 4.4. 4.3.2.8 Vessel Irradiations The RAMA Fluence Methodology (Reference 4.3-14) is used to evaluate the Reactor Pressure Vessel (RPV) fluence for both units. This methodology has been reviewed and approved by the NRC for RPV fluence evaluations (Reference 4.3-15) and is consistent with applicable regulatory guidance (Reference 4.3-16). Detailed descriptions of the calculations for each unit are provided in References 4.3-7 and 4.3-8. The fast fluence evaluations are based on the RAMA Code Methodology. The Methodology includes a transport code, model builder codes, a fluence calculator code, an uncertainty methodology, and a nuclear data library. The transport code, fluence calculator, and nuclear data library are the primary software components for calculating the neutron flux and fluence. The transport code uses a deterministic, three-dimensional, multigroup nuclear particle transport theory to perform the neutron flux calculations. The transport code couples the nuclear transport method with a general geometry modeling capability to provide a flexible and accurate tool for calculating fluxes in light water reactors. The fluence calculator uses reactor operating history information with isotopic production and decay data to estimate activation and fluence in the reactor components over the operating life of the reactor. The nuclear data
library contains nuclear cross-section data and response functions that are needed in the flux, fluence, and reaction rate calculations. The cross sections and response functions are based on the BUGLE-96 nuclear data library. Fluxes are calculated at the inner vessel surface, at 1/4 T and 3/4 T depths. The RAMA methodology calculates RPV fluence and uncertainty at all locations in the RPV in the active core region in accordance with applicable regulatory guidance. The results from the vessel fluence coupon analyses are solely used to support the methodology uncertainty analysis. The RAMA methodology directly calculates the fluence at all RPV locations in the active core region. Therefore, lead factors, which were historically used to extrapolate the measured fluence at the coupon locations to the RPV 1/4 T depths are no longer used or calculated. Previous fluence calculations were performed using the DORT computer code, which is described in Section 4.1. The RAMA Fluence Methodology will continue to be used to calculate the fluence for both units and is described in BWRVIP-114 (Reference 4.3-14). The analytical model for (R, ) geometry is shown in Figure 4.3-14. The model consists of an inner and outer core region, the shroud, water regions inside and outside the shroud, jet pump components, the vessel wall, inner and outer Cavity, Mirror Insulation and the Biological Shield.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-9 Neutron fluence was determined based on actual and expected operating history for each unit. This included the effects of several power uprates that have occurred during the operating history. Final end of life RPV fluence is calculated for both units at 32 EFPY at the RPV [both inner diameter (ID) and 1/4 T (1/4 of the distance from the inside diameter to the outside diameter)] based on actual and expected operating history. Details on the power history assumed in the fluence analysis are provided in footnotes to the data in Table 4.3-5. Table 4.3-5 lists the 32 EFPY maximum fast fluence results and also provides historical results from the original analyses for comparison. 4.3.3 Analytical Methods Reload design for SSES Units 1 and 2 is performed using NRC approved methodology. The approved methods used for nuclear design are fully described in Reference 4.3-13. A summary description of several nuclear design codes is provided in Section 4.1. Reference 4.3-1 describes the methods used for initial core nuclear design. 4.3.4 Changes Reference 4.3-1 lists several changes made to the initial reactor nuclear design. Reload core nuclear designs incorporate the following significant changes. Unit 2, Cycle 9 and Unit 1, Cycle 11 were the first cores to utilize the FANP ATRIUM TM-10 fuel design at SSES. ATRIUM TM-10 has a 10x10 lattice which is significantly different from the 9x9 lattice utilized in previous cycles. Nuclear characteristics of ATRIUM TM-10 fuel are discussed in Section 4.3. Mechanical design of ATRIUM TM-10 fuel is discussed in Section 4.2. The CASMO-3G lattice physics code was first used to support the U1C10 reload design. Unit 2, Cycle 9 was designed for a 24 month cycle. This cycle length represents a change from the 18 month cycle used for previous core designs. The effects of a 24 month cycle on the U2C9 reload were evaluated in Reference 4.3-11. Unit 1, Cycle 11 was designed for a 24 month cycle. This cycle length represents a change from the 18 month cycle used for previous core designs. The effects of a 24 month cycle on the U1C11 reload were evaluated in Reference 4.3-12. The CASMO-4/MICROBURN-B2 code system was first used to support the U1C14 reload design. A summary description of CASMO-4 and MICROBURN-B2 is provided in Section 4.1. Unit 1 Cycle 14 was the first cycle to utilize 100 mil fuel channels and the Framatome-ANP FUELGUARD Lower Tie Plate design. The 100 mil fuel channel and FUELGUARD Lower Tie Plate are described in Section 4.2. Unit 1 Cycle 20 was the first cycle to utilize the Advanced Fuel Channel (AFC). The AFC is described in Section 4.2.
SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-10 4.3.5 References 4.3-1 "BWR/4 and BWR/5 Fuel Design", NEDE-20944(P), General Electric Company, October 1976. 4.3-2 "BWR/4 and BWR/5 Fuel Design", Amendment 1 NEDE-20944-1(P), General Electric Company, January 1977. 4.3-3 Letter from Olan. D. Parr (NRC) to Dr. G. G. Sherwood (GE), "Review of General Electric Topical Report NEDE-20944-P, BWR/4 and BWR/5 Fuel Design (NEDO-20944 Non-Proprietary Version)", September 30, 1977.
4.3-4D e leted 4.3-5D e leted 4.3-6D e leted 4.3-7 "Susquehanna Unit 1 Reactor Pressure Vessel Fluence Evaluation, "PPL-FLU-002-R-002, Rev. 1, TransWare Enterprises, Inc., October 2005. 4.3-8 "Susquehanna Unit 2 Reactor Pressure Vessel Fluence Evaluation,"PPL-FLU-002-R-001, Rev. 0, TransWare Enterprises, Inc., May 2005.4.3-9 "Power Uprate Engineering Report for Susquehanna Steam Electric Station Units 1 and 2", NEDC-32161P, GE Nuclear Energy, December 1993. 4.3-10 "Advanced Nuclear Fuels Methodology For Boiling Water Reactors", XN-NF-80-19 (P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990. 4.3-11 "Susquehanna SES Unit 2 Cycle 9 Reload Summary Report", PL-NF-97-003, Rev. 1, PP&L, September 1997. 4.3-12 "Susquehanna SES Unit 1 Cycle 11 Reload Summary Report", PL-NF-98-002, Rev. 1, PP&L, Inc., July 1998. 4.3-13 EMF-2158 (P) (A), Siemens Power Corporation Methodology For Boiling Water Reactors Evaluation and Validation of Casmo-4/Microburn-B2", October 1999. 4.3-14 "BWR vessel and Internals Project RAMA Fluence Methodology Manual," BWRVIP-114, May 2003.
4.3-15 Safety Evaluation of proprietary EPRI Reports, "BWR Vessel And Internals Project, RAMA Fluence Methodology Manual (BWRVIP-114)," "RAMA Fluence Methodology Benchmark Manual - Evaluation of Regulatory Guide 1.190 Benchmark Problems (BWRVIP-115)," "RAMA Fluence Methodology - Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5 (BWRVIP-117)," and "RAMA Fluence Methodology Procedures Manual (BWRVIP-121)," and "Hope Creek Flux Wire Dosimeter Activation Evaluation for SSES-FSAR Text Rev. 62 FSAR Rev. 68 4.3-11 Cycle 1 (TWE-PSE-001-R-001)" (TAC No. MB9765), BWRVIP 2005-208B, William H. Bateman, NRC to Bill Eaton, BWRVIP Chairman, May 13, 2005.
4.3-16 "Calculation and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Guide 1.190, March 2001.
SSES-FSAR Table Rev. 56 FSAR Rev. 65 Page 1 of 1 TABLE 4.3-1 REACTOR CORE CHARACTERISTICS Reactor Type/Configuration BWR-4/2 Loop Jet Pump Recirculation System, C-Lattice Rated Thermal Power, Unit 1 3952 MWt Rated Thermal Power, Unit 2 3952 Mwt Number of Fuel Assemblies 764 Number of Control Rods 185 Number of Traversing In-core Probe Locations 43 Active Core Height, ft 12.45 Control Rod Pitch, inches 12.0 Fuel Assembly Pitch, inches
6.0 AutoCAD
Figure Fsar 4_3_2.dwg FIGURE 4.3-2, Rev 63 UNIT 2 CORE LOADING MAP SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-8, Rev. 54 AutoCAD Figure 4_3_8.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9, Rev. 54 AutoCAD Figure 4_3_9.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10, Rev. 54 AutoCAD Figure 4_3_10.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-11, Rev. 54 AutoCAD Figure 4_3_11.doc
SUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 AND 2FINAL SAFETY ANALYSIS REPORTFIGURE 4.3-14, Rev. 55VESSEL FLUENCE (R.0) MODEL FORAZIMUTHAL FLUX DISTRIBUTIONAuto Cad: Figure Fsar 4_3_14.dwg THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-15, Rev. 55 AutoCAD Figure 4_3_15.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-16, Rev. 55 AutoCAD Figure 4_3_16.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-17, Rev. 54 AutoCAD Figure 4_3_17.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-18, Rev. 54 AutoCAD Figure 4_3_18.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-19, Rev. 54 AutoCAD Figure 4_3_19.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-20, Rev. 54 AutoCAD Figure 4_3_20.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-21, Rev. 54 AutoCAD Figure 4_3_21.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-22, Rev. 54 AutoCAD Figure 4_3_22.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-23, Rev. 54 AutoCAD Figure 4_3_23.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-24, Rev. 54 AutoCAD Figure 4_3_24.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-25, Rev. 54 AutoCAD Figure 4_3_25.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-26, Rev. 54 AutoCAD Figure 4_3_26.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-27, Rev. 54 AutoCAD Figure 4_3_27.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-28, Rev. 54 AutoCAD Figure 4_3_28.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-29, Rev. 54 AutoCAD Figure 4_3_29.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-30, Rev. 54 AutoCAD Figure 4_3_30.doc
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THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-5, Rev. 57 AutoCAD Figure 4_3_9_5.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-6, Rev. 55 AutoCAD Figure 4_3_9_6.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-9-7, Rev. 55 AutoCAD Figure 4_3_9_7.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-1, Rev. 55 AutoCAD Figure 4_3_10_1.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-2, Rev. 55 AutoCAD Figure 4_3_10_2.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-5, Rev. 55 AutoCAD Figure 4_3_10_5.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-6, Rev. 55 AutoCAD Figure 4_3_10_6.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.3-10-7, Rev. 55 AutoCAD Figure 4_3_10_7.doc
AutoCAD: Figure Fsar 4_3_8_32.dwgFIGURE 4.3-8-32, Rev 0ATRIUM -10 FUEL AXIAL ENRICHMENT(NOMINAL)PPL ASSEMBLY TYPE 79SUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 & 2FINAL SAFETY ANALYSIS REPORT TM AutoCAD: Figure Fsar 4_3_8_33.dwgFIGURE 4.3-8-33, Rev 0ATRIUM -10 FUEL AXIAL ENRICHMENT(NOMINAL)PPL ASSEMBLY TYPE 80SUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 & 2FINAL SAFETY ANALYSIS REPORT TM
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SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-1 4.4 THERMAL AND HYDRAULIC DESIGN This section addresses the original plant thermal hydraulic design (number of assemblies, core power and flow, etc), the compatibility of co-resident fuel designs and the relative stability of reload
cores. 4.4.1 DESIGN BASES 4.4.1.1 Safety Design Bases Thermal-hydraulic design of the core shall establish:
(1)Actuation limits for the devices of the nuclear safety systems such that no fuel damag e occurs as a result of moderate frequency transient events. For example, the Minimu m Critical Power Ratio (MCPR) operating limit is specified such that at least 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during the mostsevere moderate (Per Regulatory Guide 1.70 Revision 2) frequency transient even ts.(2)The thermal-hydraulic safety limits for use in evaluating the safety margin relating theconsequences of fuel barrier failure to public safety
.(3)That the nuclear system exhibits no inherent tendency toward divergent or limit cyc le oscillations which would compromise the integrity of the fuel or nuclear system proces s barri er.4.4.1.2 Power Generation Design Bases The thermal-hydraulic design of the core shall provide the following operational characteristics:
(1)The ability to achieve rated core power output throughout the design life of the fuel witho ut sustaining premature fuel failure.
(2)Flexibility to adjust core output over the range of plant load and load maneuveringrequirements in a stable, predictable manner without sustaining fuel dama ge.4.4.1.3 Requirements for Steady-State Conditions Steady-State Limits For purposes of maintaining adequate thermal margin during normal steady-state operation, the minimum critical power ratio must not be less than the required MCPR operating limit, and the maximum linear heat generation rate (LHGR) must be maintained below the LHGR limit. This does not specify the operating power nor does it specify peaking factors. These parameters are determined subject to a number of constraints including the thermal limits given previously. The core and fuel design basis for steady-state operation, i.e., MCPR and LHGR limits, have been defined to provide margin between the steady-state operating conditions and any fuel damage SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-2 condition to accommodate uncertainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time in life.
Steady-state limits also exist on the maximum average planar linear heat generation rate (MAPLHGR). The MAPLHGR limits protect against violation of the ECCS acceptance criteria during a Loss of Coolant Accident and are derived from the LOCA analyses described in Section 6.3.
4.4.1.4 Requirements for Transient Conditions Transient Limits
The transient thermal limits are established such that no fuel damage is expected to occur during the most severe moderate frequency transient event. Fuel damage is defined as perforation of the cladding that permits release of fission products. Mechanisms that cause fuel damage in reactor transients are:
(1) Severe overheating of fuel cladding caused by inadequate cooling, and (2) Fracture of the fuel cladding caused by relative expansion of the uranium dioxide pellet inside the fuel cladding.
For design purposes, the transient limit requirement relating to cladding overheating is met if at least 99.9 percent of the fuel rods in the core do not experience boiling transition during any moderate frequency transient event. No fuel damage would be expected to occur even if a fuel rod actually experienced a boiling transition.
A value of 1 percent plastic strain of Zircaloy cladding is conservatively defined as the limit below which fuel damage from overstraining the fuel cladding is not expected to occur. The linear heat generation rate required to cause this amount of cladding strain depends on the fuel type and burnup. The linear heat generation rates are discussed on a fuel type specific basis in Section 4.2.3.
4.4.1.5 Summary of Design Bases
In summary, the steady-state operating limits have been established to assure that the design basis is satisfied for the most severe moderate frequency transient event. There is no steady-state design overpower basis. An overpower which occurs during an incident of a moderate frequency transient event must meet the plant transient MCPR limit and 1% plastic strain limit. Demonstration that the transient limits are not exceeded is sufficient to conclude that the design basis is satisfied.
The MCPR, MAPLHGR, and LHGR limits are sufficiently general so that no other limits need to be stated. For example, cladding surface temperatures will always be maintained within 10 to 15 oF of the coolant temperature as long as the boiling process is in the nucleate regime. The cladding and fuel bundle integrity criterion is assured as long as MCPR, MAPLHGR, and LHGR limits are met. There are no additional design criteria on coolant void fraction, core coolant flow-velocities, or flow distribution, nor are they needed. The coolant flow velocities and void fraction become constraints upon the mechanical and physics design of reactor components and are partially constrained by stability and control requirements.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-3 4.
4.2 DESCRIPTION
OF THERMAL-HYDRAULIC DESIGN OF THE REACTOR CORE 4.4.2.1 Summary Comparison
An evaluation of plant performance from a thermal and hydraulic standpoint is provided in Subsection 4.4.3.
A tabulation of thermal and hydraulic parameters of the core is given in Table 4.4-1.
4.4.2.2 Critical Power Ratio
There are three different types of boiling heat transfer to water in a forced convection system:
nucleate boiling, transition boiling, and film boiling. Nucleate boiling, at lower heat transfer rates, is an extremely efficient mode of heat transfer, allowing large quantities of heat to be transferred with a very small temperature rise at the heated wall. As heat transfer rate is increased the boiling heat transfer surface alternates between film and nucleate boiling, leading to fluctuations in heated wall temperatures. The point of departure from the nucleate boiling region into the transition boiling region is called the boiling transition. Transition boiling begins at the critical power and is characterized by fluctuations in cladding surface temperature. Film boiling occurs at the highest heat transfer rates; it begins as transition boiling comes to an end. Film boiling heat transfer is characterized by stable wall temperatures which are higher than those experienced during nucleate boiling.
4.4.2.2.1 Boiling Correlations The occurrence of boiling transition is a function of the fluid enthalpy, mass flow rate, pressure, flow geometry and assembly power distribution. Framatome ANP, Inc. (FANP) has conducted extensive experimental investigations of these parameters. These parametric studies encompass the entire design range of these variables. The SPCB critical power correlation, Reference 4.4-58, is used for ATRIUMŽ-10 fuel. This correlation is based on accurate test data of full-scale prototypic simulations of reactor fuel assemblies operating under conditions typical of those in actual reactor designs. The correlation is a "best fit" to the data and is used together with a statistical analysis to assure adequate reactor thermal margins (Reference 4.4-42).
The figure of merit used for reactor design and operation is the Critical Power Ratio (CPR). This is defined as the ratio of the bundle power at which boiling transition occurs to the bundle power at the reactor condition of interest (i.e., the ratio of critical bundle power to operating bundle power).
In this definition, the critical power is determined at the same mass flux, inlet temperature, and pressure which exist at the specified reactor condition.
4.4.2.3 Thermal Operating Limits
The limiting constraints in the design of the reactor core are stated in terms of the MCPR, MAPLHGR, and LHGR limits. The design philosophy used to assure that these limits are met involves the selection of one or more power distributions which are more limiting than expected operating conditions and subsequent verification that under these more stringent conditions, the design limits are met. Therefore, the "design power distributions" represent extreme conditions of power. Use of these power distributions in the analyses is a fair and stringent test of the operability SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-4 of the reactor as designed to comply with the foregoing limits. Expected operating conditions are less severe than those represented by the design power distributions which give the MCPR, MAPLHGR and LHGR limits.
However, it must be established that operation with a less severe power distribution is not a necessary condition for the safety of the reactor. Because there are an infinite number of operating reactor states which can exist (with variations in rod patterns, time in cycle, power level, distribution, flow etc.) which are within the design constraints, it is not possible to determine them all. However, constant monitoring of operating conditions using the available plant measurements can ensure compliance with design objectives.
4.4.2.3.1 Design Power Distribution Thermal design of the reactor--including the selection of the core size and effective heat transfer area, the design steam quality, the total recirculation flow, the inlet subcooling, and the specification of internal flow distribution -- was performed by the NSSS vendor and is based on the concept and application of a design power distribution. The design power distribution was an appropriately conservative representation of the most limiting thermal operating state at rated conditions and included design allowances for the combined effects (on the fuel rod, and the fuel assembly heat flux and temperature) of the gross and local steady-state power density distributions and adjustments of the control rods.
4.4.2.3.2 Design Linear Heat Generation Rates
The maximum and core average linear heat generation rates are shown in Table 4.4-1. The maximum linear heat generation rate at any location is the average linear heat generation rate at a given axial location multiplied by the total peaking factor of that location.
Fuel type specific LHGR limits and MAPLHGR limits are provided in the Core Operating Limits Report for each unit (see FSAR section 16.3, Technical Requirements Manuals).
4.4.2.4 Void Fraction Distribution Typical core average and maximum exit void fractions in the core at rated condition are given in Table 4.4-2. The axial distribution of core void fractions for the average radial channel and the maximum radial channel (end of node value) are also given in Table 4.4-2. Similar distributions for steam quality are provided in Table 4.4-3. The core average axial power distributions used to produce these tables are given in Table 4.4-2a.
4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern
Correct distribution of core coolant flow among the fuel assemblies is accomplished by the use of an accurately calibrated fixed orifice at the inlet of each fuel assembly. The orifices are located in the fuel support piece. They serve to control the flow distribution and, hence, the coolant conditions within prescribed bounds throughout the design range of core operation.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-5 The sizing and design of the orifices ensure stable flow in each fuel assembly during all phases of operation at normal operating conditions.
The core is divided into two orificed flow zones. The outer zone is a narrow, reduced-power region around the periphery of the core. The inner zone consists of the core center region. No other control of flow and steam distribution, other than that incidentally supplied by adjusting the power distribution with the control rods, is used or needed. The orifices can be changed during refueling, if necessary.
Design core flow distribution calculations were performed by the NSSS vendor using a design power distribution which consists of a hot and average powered assembly in each of the two orifice zones. The design bundle power and resulting relative flow distribution are given in Table 4.4-4.
The flow distribution to the fuel assemblies is calculated on the assumption that the pressure drop across all fuel assemblies is the same. This assumption has been confirmed by measuring the flow distribution in a modern boiling water reactor as reported in References 4.4-2 and 4.4-36.
There is reasonable assurance, therefore, that the calculated flow distribution throughout the core is in close agreement with the actual flow distribution of an operating reactor.
The use of the design power distribution discussed previously ensures the orificing chosen covers the range of normal operation. The expected shifts in power production during core life are less severe and are bounded by the design power distribution.
4.4.2.6 Core Pressure Drop and Hydraulic Loads The pressure drop across various core components under the steady state design conditions is included in Table 4.4-1. Analyses for the most limiting conditions, the recirculation line break and the steam line break are reported in Chapter 15.
The components of bundle pressure drop considered are friction, local elevation and acceleration.
Reference 4.4-43 presents the methodology and constitutive relationships used by FANP for the calculation of pressure drop in BWR fuel assemblies. These are implemented in the XCOBRA computer code which is used to perform steady state thermal-hydraulic analyses, Reference 4.4-49. The thermal hydraulic loads on the fuel rods during steady-state operation, transient, and accident conditions are negligible, primarily because of the channel confinement, thereby resulting in small cross flow between rods (i.e., essentially constant pressure at any given elevation in the fuel bundle). The loads (i.e. horizontal) across the control blades are minimal or negligible primarily due to the flat interchannel velocity profile as given in Reference 4.4-13.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-6 4.4.2.6.1 Friction Pressure Drop Friction pressure drop is calculated using the model relation 2 TPFchH 2 f fL 2g w = P A D 2 where P f = friction pressure drop, psi w = mass flow rate, g = acceleration of gravity, = water density, D H = channel hydraulic diameter, A ch = channel flow area, L = length, f = friction factor, and 2TPF = two-phase friction multiplier This basic model is similar to that used throughout the nuclear power industry. The formulation for the two-phase multiplier used by FANP is the correlation determined by Jones, Reference 4.4-43, which represents a mass velocity correction to the Martinelli-Nelson correlation, Reference 4.4-3.
Significant amounts of friction pressure drop data in multirod geometries representative of modern BWR plant fuel bundles have been taken and both the friction factor and two-phase multipliers have been correlated on a best-fit basis using the above pressure drop formulation.
4.4.2.6.2 Local Pressure Drop
The local pressure drop is defined as the irreversible pressure loss associated with an area change such as the orifice, lower tie plates, and spacers of a fuel assembly.
The general local pressure drop model is similar to the friction pressure drop and is:
L 2 2 2 TPL 2 P = w 2g K A where P L = local pressure drop, psi, K = local pressure drop loss coefficient, A 2 = reference area for local loss coefficient, and 2 TPL = two-phase local multiplier, and w and g are defined the same as for friction. This basic model is similar to that used throughout the nuclear power industry. The two-phase multiplier used by FANP is given by the ratio of the saturated water and two-phase mixture densities. Tests are performed in both single and two-phase flow to arrive at best-fit design values for spacer and upper tie plate pressure drop. The range of test variables is specified to include the range of interest to boiling water reactors.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-7 New data are taken whenever there is a significant design change to ensure the most applicable methods are in use at all times. For ATRIUM-10 fuel, a simple multiplier based on local quality is also used to calculate the spacer pressure drop for the two-phase conditions, Reference 4.4-45.
4.4.2.6.3 Elevation Pressure Drop
The elevation pressure drop is based on the well-known relation
E P gL=
=+fg ()1 where PE = elevation pressure drop, psi L = incremental length
= average water density
= average void fraction over the length L f, g = saturated water and vapor density, respectively. g = acceleration of gravity
4.4.2.6.4 Acceleration Pressure Drop A reversible pressure change occurs when an area change is encountered, and an irreversible loss occurs when the fluid is accelerated through the boiling process. The basic formulation for the reversible pressure change resulting from a flow area change is given by:
P W gA A A ACC==();1 2 2 2 2 2 2 1 where PACC = acceleration pressure drop, A 2 = final flow area, A 1 = initial flow area, and other terms are as previously defined. The basic formulation for the acceleration pressure change due to density change is:
P W gAACC ch M out M in=2 2 11 where
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-8
()1 1 1 2 2 1Mg x x=+() M = momentum density, x = steam quality, 1 = saturated liquid density and other terms are as previously defined. The total acceleration pressure drop in boiling water reactors is on the order of a few percent of the total pressure drop.
4.4.2.7 Correlation and Physical Data
Substantial amounts of physical data support the pressure drop and thermal hydraulic loads discussed in Subsection 4.4.2.6. Correlations have been developed to fit these data to the formulations discussed.
4.4.2.7.1 Pressure Drop Correlations
Pressure drop data in multirod geometries representative of modern BWR plant fuel bundles has been correlated to the friction factor and two-phase multipliers on a best fit basis using the pressure drop formulations reported in Subsections 4.4.2.6.1 and 4.4.2.6.2 FANP's pressure drop methodology is described in Reference 4.4-43.
New data are taken whenever there is a significant design change. Applicability of the pressure drop correlations is confirmed by full scale prototype flow tests. Pressure drop tests for the FANP ATRIUM-10 fuel designs is reported in Reference 4.4-45. The range of tests variables is specified to include the range of interest to boiling water reactors.
4.4.2.7.2 Void Fraction Correlation
The void fraction is determined by a Zuber-Findlay model with constitutive relations as supplied by Ohkawa and Lahey, Reference 4.4-43.
4.4.2.7.3 Heat Transfer Correlation
The Jens-Lottes (Reference 4.4-5) wall superheat equation is used in fuel design to determine the cladding-to-coolant heat transfer coefficients for nucleate boiling.
4.4.2.8 Thermal Effects of Operational Transients
The evaluation of the core's capability to withstand the thermal effects resulting from anticipated operational transients is covered in Chapter 15.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-9 4.4.2.9 Uncertainties in Estimates Uncertainties in thermal-hydraulic parameters are considered in the statistical analysis which is performed to establish the fuel cladding integrity safety MCPR limit such that at least 99.9% of the fuel rods in the core are expected not to experience boiling transition during any moderate frequency transient event. The statistical model and analytical procedure are described in Reference 4.4-42.
The MCPR safety limit is determined by a statistical convolution of the uncertainties associated with the calculation of thermal margin. Some uncertainties are fuel related and others are characteristics of the reactor system. Examples of fuel related uncertainties are those introduced by the critical power correlation, the calculation of core wide power peaking and the calculation of core wide flow distribution. Examples of uncertainties which are characteristics of the reactor system are the measurement uncertainties associated with reactor pressure, total core flow, feedwater flow and feedwater temperature. The uncertainties which are considered are shown in Table 4.4-6.
4.4.2.10 Flux Tilt Considerations
The inherent design characteristics of the BWR are particularly well suited to handle perturbations due to flux tilt. The stabilizing nature of the moderator void coefficient effectively damps oscillations in the power distribution. In addition to this damping, the incore instrumentation system and the associated on-line computer provide the operator with prompt and reliable power distribution information. Thus, the operator can readily use control rods or other means to effectively limit the undesirable effects of flux tilting. Because of these features and capabilities, it is not necessary to allocate a specific peaking factor margin to account for flux tilt. If for some reason, the power distribution could not be maintained within normal limits using control rods, then the operating power limits would have to be reduced as prescribed in the Plant Technical Specifications. The power distributions will be maintained such that the operating limits given in the Core Operating Limits Report will not be exceeded.
4.4.2.11 Crud Deposition
In general, the CPR is not affected as crud accumulates on fuel rods (References 4.4-34 and 4.4-35). Therefore, no modifications to the critical power correlation are made to account for crud deposition. The effect of crud deposition on pressure drop and flow is to increase the pressure drop and decrease the flow. An increase in crud deposition for high exposure assemblies would tend to reduce the flow in these assemblies and increase the flow in low exposure, CPR limiting assemblies. No credit is taken for the increase in CPR margin due to crud deposition.
The effects of crud deposition are included in thermal and rod internal pressure calculations, Reference 4.4-47.
4.
4.3 DESCRIPTION
OF THE THERMAL AND HYDRAULIC DESIGN OF THE REACTOR COOLANT SYSTEM The thermal and hydraulic design of the reactor coolant system is described in this subsection.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-10 4.4.3.1 Plant Configuration Data 4.4.3.1.1 Reactor Coolant System Configuration The reactor coolant system is described in Section 5.4 and shown in isometric perspective in Figure 5.4-1. The piping sizes, fittings and valves are listed in Table 5.1-1.
4.4.3.1.2 Reactor Coolant System Thermal Hydraulic Data Table 5.1-1 provides design temperatures, pressures and flow rates for the reactor coolant system and its components.
4.4.3.1.3 Reactor Coolant System Geometric Data
Coolant volumes of regions and components within the reactor vessel are shown in Figure 5.1-2.
Table 4.4-8 provides the flow path length, height and liquid level, minimum elevations, and minimum flow areas for each major flow path volume within the reactor vessel and recirculation loops of the reactor coolant systems.
Table 4.4-9 provides the lengths and sizes of all safety injection lines to the reactor coolant system.
4.4.3.2 Operating Restrictions on Pumps
Expected recirculation pump performance curves are shown in Figure 5.4-3. These curves are valid for all conditions with a normal operating range varying from approximately 20% to 115% of rated pump flow.
The pump characteristics including considerations of NPSH requirements are the same for the conditions of two pump and one pump operation as described in Subsection 5.4.1. Subsection 4.4.3.3 gives the limits imposed on the recirculation pumps by cavitation, pump loads, bearing design, flow starvation, and pump speed.
4.4.3.3 Power-Flow Operating Map
4.4.3.3.1 Limits for Normal Operation
A boiling water reactor must operate with certain restrictions because of pump net positive suction head (NPSH), overall plant control characteristics, core thermal power limits, etc. A representation of a simplified power-flow map for the power range of operation is shown in Figure 4.4-5. The actual power-flow maps for Units 1 and 2 are found in the respective COLR, FSAR Section 16.3.
The nuclear system equipment, nuclear instrumentation, and the reactor protection system, in conjunction with operating procedures, maintain operations within the area of this map for normal operating conditions. The boundaries on this map are as follows:
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-11 Natural Circulation Line: The operating state of the reactor moves along this line for the normal control rod withdrawal sequence in the absence of recirculation pump operation.
30 Percent Recirculation Pump Constant Speed Line: Startup operations of the plant are normally carried out with the recirculation pumps operating at approximately 30 percent speed. The operating state for the reactor follows this line for the normal control rod withdrawal sequence.
Rated Flow Control Line: The rated flow control line (100% rod line) passes through 100 percent power at 108 Mlb/hr flow. The operating state for the reactor follows this line for recirculation flow changes with a fixed control rod pattern. The line is based on full power constant xenon concentration.
Cavitation Protection Line: This line (minimum power line) results from the recirculation pump and jet pump NPSH requirements. The recirculation pumps are automatically switched to 30 percent speed when the feedwater flow drops below a preset value.
Note that an actual power-flow map will contain stability related regions. The actual Unit 1 and Unit 2 power-flow maps are included in their respective COLR, FSAR Section16.3.
4.4.3.3.1.1 Performance Characteristics Other Power Flow Operating Map performance characteristics are:
Recirculation Pump Constant Speed Line: This line shows the change in flow associated with power changes while maintaining constant recirculation pump speed.
Constant Rod Lines: These lines show the change in power associated with flow changes while maintaining constant control rod position (e.g. 80% rod line).
4.4.3.3.2 Regions of the Power Flow Map For normal operating conditions, the nuclear system equipment, nuclear instrumentation, and the reactor protection system, in conjunction with operating procedures, maintain operation outside the exclusion areas of the power flow map. Main regions of the map are discussed below to clarify operational capabilities.
Region A - This is the transition region between natural circulation operation and 30% pump speed operation. Operation at less than 30% pump speed with two recirculation loops results in flow instabilities (causing flow induced vibrations), therefore the recirculation pumps are not continually operated below 30% pump speed. Normal startup is along the 30% pump speed boundary of this region.
Region B - This region represents the normal operating zone of the map where power changes can be made either by control rod movement or by core flow changes by changing recirculation pump drive speed.
Region C - This is the low power area of the map where cavitation can be expected in the recirculation pumps and in the jet pumps. Operation within this region is precluded by system interlocks which runback the recirculation pumps to 30% speed whenever feedwater flow is less than a preset value (typically 20% of rated).
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-12 4.4.3.4 Temperature-Power Operating Map (PWR)
Not Applicable.
4.4.3.5 Load Following Characteristics
The following simple description of boiling water reactor operation with recirculation flow control summarizes the principal modes of normal power range operation. Assuming the plant to be initially hot with the reactor critical, full power operation can be approached by initially moving along the two pump 30% speed line until power is at least above the minimum power line (cavitation interlock) of Region C (see Figure 4.4-5). Note, other low power restrictions may apply as a result of cycle specific transient analyses. This initial sequence may be achieved with control rod withdrawal and manual, individual recirculation pump control. Individual pump startup procedures are provided which achieve 30 percent of full pump speed in each loop. Power, steam flow, and feedwater flow are increased as control rods are manually withdrawn. An interlock prevents low power-high recirculation flow combinations which create recirculation pump and jet pump NPSH problems.
Reactor power increases as the operating state moves to the right on Figure 4.4-5 as the operator manually increases recirculation flow in each loop. Eventually, the operator can switch to simultaneous recirculation pump control. Thermal output can then be increased by either control rod withdrawal or recirculation flow increase. Both combinations are required to achieve full power.
The operating map is shown in Figure 4.4-5 with the designated flow control range expected.
The curve labeled "100% Xe Rod Line" (i.e., the "Rated Flow Control Line") represents a typical steady state power flow characteristic for a fixed rod pattern. It is affected by xenon, core leakage flow assumptions, and reactor vessel pressure variations.
Normal power range operation is along the "Rated Flow Control Line", below the APRM Rod Block Trip Setpoint, and below 100% rated power.
The large negative operating reactivity and power coefficients, which are inherent in the boiling water reactor, provide important advantages as follows:
(1) Good load following with well damped behavior and little undershoot or overshoot in the heat transfer response.
(2) Load following with recirculation flow control.
(3) Strong damping of spatial power disturbances.
The reactor power level can be controlled by flow control over approximately 35 percent of the power level on the rated rod line. Load following is accomplished by varying the recirculation flow to the reactor. This method of power level control takes advantage of the reactor negative void coefficient. To increase reactor power, it is necessary to increase the recirculation flow rate which sweeps some of the voids from the moderator, causing an increase in core reactivity. As the reactor power increases, more steam is formed and the reactor stabilizes at a new power level with the transient excess reactivity balanced by the new void formation. No control rods are moved to accomplish this power level change. Conversely, when a power reduction is required, it is necessary only to reduce the recirculation flow rate. When this is done, more voids in the SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-13 moderator automatically decrease the reactor power level to that commensurate with the new recirculation flow rate. Again, no control rods are moved to accomplish the power reduction.
Varying the recirculation flow rate (flow control) is more advantageous, relative to load-following, than using control rod positioning. Flow variations perturb the reactor uniformly in the horizontal planes and ensure a flatter power distribution and reduced transient allowances. As flow is varied, the power and void distributions remain approximately constant at the steady state end points for a wide range of flow variations. After adjusting the power distribution by positioning the control rods at a reduced power and flow, and taking into account any effects due to Xe variations, the operator can then bring the reactor to rated conditions by increasing flow, with the assurance that the power distribution will remain approximately constant. Section 7.7 describes how recirculation flow is
varied.
4.4.3.6 Thermal and Hydraulic Characteristics Summary Table The thermal hydraulic characteristics are provided in Table 4.4-1 for the core and tables of Sections 5.1 and 5.4 for other portions of the reactor coolant system.
4.4.4 EVALUATION
The design basis employed for the thermal and hydraulic characteristics incorporated in the core design, in conjunction with the plant equipment characteristics, nuclear instrumentation, and the reactor protection system, is to require that no fuel damage occur during normal operation or during abnormal operation transients. Demonstration that the applicable thermal-hydraulic limits are not exceeded is given by analyses.
4.4.4.1 Critical Power
The SPCB critical power correlation is utilized in thermal-hydraulic evaluations. This correlation is discussed in more detail in Subsection 4.4.2.2.1.
4.4.4.2 Core Hydraulics
Core hydraulic models and correlations are discussed in Subsections 4.4.2.6, 4.4.2.7, and 4.4.4.5.
4.4.4.3 Influence of Power Distributions
The influence of power distributions on the thermal-hydraulic design is discussed in Reference 4.4-1, Appendix V for the initial core. The influence of power distribution is included in the cycle specific licensing calculations.
4.4.4.4 Core Thermal Response The thermal response of the core for accidents and expected transient conditions is discussed in Chapter 15.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-14 4.4.4.5 Analytical Methods The analytical methods, thermodynamic data, and hydrodynamic data used in determining the thermal and hydraulic characteristics of the core are similar to those used throughout the nuclear power industry.
Core thermal-hydraulic analyses are performed with the aid of a digital computer program. This program models the reactor core through a hydraulic description of orifices, lower tie plates, fuel rods, fuel rod spacers, upper tie plates, fuel channel, and the core bypass flow paths.
4.4.4.5.1 Reactor Model
The reactor model includes a hydraulic representation of the orifice, lower tie plate, fuel rods, water rods or inner water channel, spacers, upper tie plate and the fuel channel.
The code can handle a number of fuel channel types and bypass flow paths. Usually there is one fuel assembly representing each of the "hot" fuel types. The average types then make up the balance of the core.
The computer program iterates on flow through each flow path (fuel assemblies and bypass paths) until the total differential pressure (plenum to plenum) across each path is equal, and the sum of the flows through each path equals the total core flow.
For the initial core, orificing was selected to optimize the core flow distribution between orifice regions as discussed in Subsection 4.4.2.5. The core design pressure is determined from the required turbine throttle pressure, the steam line pressure drop, steam dryer pressure drop, and the steam separator pressure drop. The core inlet enthalpy is determined from the reactor and turbine heat balances. The required core flow is then determined by applying the procedures of this section and specifications such that the applicable thermal limits are satisfied. The results of applying these methods and specifications are:
(1) Flow for each bundle type, (2) Flow for each bypass path, (3) Core pressure drop, (4) Fluid property axial distribution for each bundle type, and (5) CPR calculations for each bundle type.
For reload cores, the appropriate orificing, core flow and system pressure are used as model input.
The same type of calculations that were used for the initial core are performed to calculate the parameters stated in (1)-(5) above.
4.4.4.5.2 System Flow Balances
The basic assumption used by the code in performing the hydraulic analysis is that the flow entering the core will divide itself between the fuel bundles and the bypass flow paths such that each assembly and bypass flow path experience the same pressure drop. The bypass flow paths considered are described in Table 4.4-7 and shown in Figure 4.4-1. Due to the large flow area, the pressure drop in the bypass region above the core plate is essentially all elevation head. Thus, the SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-15 sum of the core plate differential pressure and the bypass region elevation head is equal to the core differential pressure.
The total core flow less the control rod cooling flow enters the lower plenum through the jet pumps. A fraction of this passes through the various bypass paths. The remainder passes through the orifices in the fuel support (experiencing a pressure loss) where more flow is lost through the fit-up between the fuel support and the lower tie plate and also through the lower tie plate holes into the bypass region. The majority of the flow continues through the lower tie plate (experiencing a pressure loss) where some flow is lost through the flow path defined by the fuel channel and lower tie plate, and restricted by the finger springs, into the bypass region.
Full-scale tests have been performed to establish the flow coefficients for the major flow paths (Reference 4.4-14). The results of these tests were used to support the initial core design. These tests simulate actual plant configurations which have several parallel flow paths and therefore the flow coefficients for the individual paths could not be separated. However, analytical models of the individual flow paths were developed as an independent check of the tests. The models were derived for actual BWR design dimensions and considered the effects of dimensional variations.
These models predicted the test results when the "as built" dimensions were applied. When using these models for hydraulic design calculations, nominal drawing dimensions were used. This is done to yield the most accurate prediction of the expected bypass flow. With the large number of components in a typical BWR core, deviations from the nominal dimensions will tend to statistically cancel, resulting in a total bypass flow best represented by that calculated using nominal dimensions.
The bypass and active channel path loss coefficients are based on test data or analytical models. Use of these coefficients produces an accurate prediction of flow through the various flow paths.
The balance of the flow enters the fuel bundle from the lower tie plate and passes through either the fuel rod channel spaces or into a non-fueled water rod or water channel, depending on fuel type. This water rod or water channel flow, remixes with the active coolant channel flow below the upper tie plate. The uncertainties associated with the calculation of total core flow and assembly flow are considered in the MCPR safety limit calculation, Subsection 4.4.2.9.
4.4.4.5.3 System Heat Balances Within the fuel assembly, heat balances on the active coolant are performed nodally. Fluid properties are expressed as the bundle average at the particular node of interest. In evaluating fluid properties, a constant pressure model is used. The core power is divided into two parts: an active coolant power and a bypass flow power. The bypass flow is heated by neutron-slowing down and gamma heating transferred to the bypass flow from structures and control elements which are themselves heated by gamma absorption and by the (n, a) reaction in the control material. The fraction of total reactor power deposited in the bypass region is very nearly 2%. A similar phenomenon occurs within the fuel bundle relative to the active coolant and the water rod or inner water channel flows. The net effect is that approximately 96% of the core power is conducted through the fuel cladding and appears as heat flux.
In design analyses the power is allocated to the individual fuel bundles using a relative power factor. The power distribution along the length of the fuel bundle is specified with axial power factors which distribute the bundle's power among the axial nodes. A nodal local peaking factor is used to establish the peak heat flux at each nodal location.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-16 The relative (radial) and axial power distributions, when used with the bundle flow, determine the axial coolant property distribution resulting in sufficient information to calculate the pressure drop components within each fuel type. Once the equal pressure drop criterion has been satisfied, the critical bundle power is determined by an iterative process for each fuel type.
4.4.4.6 Thermal-Hydraulic Stability Analysis
4.4.4.6.1 Introduction
There are many definitions of stability, but for feedback processes and control systems it can be defined as follows: A system is stable if, following a disturbance, the transient settles to a steady, noncyclic state.
A system may also be acceptably safe even if oscillatory, provided that any limit cycle of the oscillations is less than a prescribed magnitude. Instability then, is either a continual departure from a final steady-state value or a greater-than-prescribed limit cycle about the final steady-state
value.
The mechanism for instability can be explained in terms of frequency response. Consider a sinusoidal input to a feedback control system which, for the moment, has the feedback disconnected. If there were no time lags or delays between input and output, the output would be in phase with the input. Connecting the output so as to subtract from the input (negative feedback or 180 o out-of-phase connection) would result in stable closed loop operation. However, natural laws can cause phase shift between output and input and should the phase shift reach 180 degrees, the feedback signal would be reinforcing the input signal rather than subtracting from it. If the feedback signal were equal to or larger than the input signal (loop gain equal to one or greater),
the input signal could be disconnected and the system would contin ue to oscillate.
If the feedback signal were less than the input signal (loop gains less than one), the oscillations would die out.
The design of the BWR is based on the premise that power oscillations can be readily detected and suppressed.
4.4.4.6.2 Description Three types of stability considered in the design of boiling water reactors are: (1) reactor core (reactivity) stability, (2) channel hydrodynamic stability, and (3) total system stability. Reactivity feedback instability of the reactor core could drive the reactor into power oscillations. Hydrodynamic channel instability could impede heat transfer to the moderator and drive the reactor into power oscillations. The total system stability considers control system dynamics combined with basic process dynamics. A stable system is analytically demonstrated if no inherent limit cycle or divergent oscillation develops within the system as a result of calculated step disturbances of any critical variable, such as steam flow, pressure, neutron flux, and recirculation flow.
The criteria to be considered are stated in terms of two compatible parameters. First is the decay ratio x 2/x 0, designated as the ratio of the magnitude of the second overshoot to the first overshoot resulting from a step perturbation. A plot of the decay ratio is a graphic representation of the physical responsiveness of the system, which is readily evaluated in a time-domain analysis. Second is the damping coefficient n, the definition of which corresponds to the pole pair closest to the j, axis in the s-plane for the system closed loop transfer function. This parameter also applies SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-17 to the frequency-domain interpretation. The damping coefficient is related to the decay ratio as shown in Figure 4.4-2.
4.4.4.6.3 Stability Criteria
The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable design limits are not possible or can be reliably and readily detected and suppressed.
The assurance that the total plant is stable and, therefore, has significant safety margin shall be demonstrated analytically when the decay ratio, x 2/x 0, is less than 1.0 or, equivalently, when the damping coefficient, n, is greater than zero for each type of stability discussed. Special attention is given to differentiate between inherent system limit cycles and small, acceptable limit cycles that are always present, even in the most stable reactors. The latter are caused by physical nonlinearities (deadband, striction, etc.) in real control systems and are not representative of inherent hydrodynamic or reactivity instabilities in the reactor. The ultimate performance limit criteria for the three types of dynamic performance are summarized below in terms of decay ratio and damping coefficient:
Channel hydrodynamic stability x 2/x 0 < 1, n > 0 Reactor core (reactivity) stability x 2/x 0 < 1, n > 0 Total system stability x 2/x 0 < 1, n > 0 These criteria shall be satisfied for all attainable conditions of the reactor that may be encountered in the course of plant operation. For stability purposes the most severe core power and core flow conditions to which these criteria will be applied correspond to the highest attainable rodline intersection with natural circulation flow.
New FANP fuel designs are designed to exhibit channel decay ratio characteristics equivalent to existing FANP fuel designs. Evaluation of the effect of all fuel designs present in the core on the core stability is currently made on a cycle specific basis. In support of these evaluations, FANP uses the STAIF computer code for stability calculations, Reference 4.4-48. SSES has implemented Option 3 (osc illation power range monitor system) for the long term stability solution.
4.4.4.6.4 Mathematical Model
For the initial core, the mathematical model representing the core examines the linearized reactivity response of a reactor system with density-dependent reactivity feedback caused by boiling. The core model (References 4.4-27 through 4.4-32), shown in block diagram form in Figure 4.4-3, solves the dynamic equations that represent the reactor core in the frequency domain.
The plant model considers the entire reactor system, neutronics, heat transfer, hydraulics, and the basic processes, as well as associated control systems such as the flow controller, pressure regulator, feedwater controller, etc. Although, the control systems may be stable when analyzed individually, final control system settings must be made in conjunction with the operating reactor so that the entire system is stable. The plant model yields results that are essentially equivalent to SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-18 those achieved with the core model and allows the addition of the controllers, which have adjustable features permitting the attainment of the desired performance.
The plant model solves the dynamic equations that present the BWR system in the time domain. The variables, such as steam flow and pressure, are represented as a function of time. The extensiveness of this model (Reference 4.4-10, which describes the version of the code used for Susquehanna system stability calculations) is shown in block diagram form in Figure 4.4-3. Many of the blocks are extensive systems in themselves.
For reload cores, the continued applicability of the exclusion region that has been established to assure thermal-hydraulic stability is demonstrated or the exclusion region is redefined. Stability calculations, when required are performed using the STAIF computer code (Reference 4.4-48).
4.4.4.6.5 Analytical Confirmation References 4.4-37 and 4.4-48 provide a description of the analytical methods used by GE and FANP as well as model qualification through comparison with test data.
4.4.4.6.6 Analysis Results
Using actual design parameters, the responses of important nuclear system variables for the first core to step disturbances were calculated for three different power/flow conditions. Figures 4.4-7A, 4.4-7B, and 4.4-7C show the responses at 51.5% power and natural circulation. Figures 4.4-8A, 4.4-8B, and 4.4-8C show the responses at rated power/flow conditions. Figures 4.4-9A, 4.4-9B, and 4.4-9C show the responses at the lower end of the automatic power-flow control path. For all of these cases the responses met the stability criterion.
For reload cores, a confirmatory analysis is performed to demonstrate the continued applicability of the core stability regions identified in the COLR. The analysis is based on comparison of core stability performance to previously analyzed cycles. A stability code is used to calculate the variations in decay ratio from cycle to cycle for operating conditions at representative state points near the stability exclusion region.
4.4.5 TESTING
AND VERIFICATION The testing and verification techniques to be used to assure that the planned thermal and hydraulic design characteristics of the core have been provided, and will remain within required limits throughout core lifetime, are discussed in Chapter 14. A summary is as follows:
(1) Preoperational Testing
Tests are performed during the preoperational test program to confirm that construction is complete and that all process and safety equipment is operational. Baseline data are taken to assist in the evaluation of subsequent tests. Heat balance instrumentation, jet pump flow and core temperature instrumentation, is calibrated and set points verified.
(2) Initial Start-Up
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-19 Hot functional tests are conducted with the reactor between 5 and 10% power. Core performance is monitored continuously to assure that the reactor is operating within allowable limits (e.g., peaking factors, linear heat generation rate, etc.) and is evaluated periodically to verify the core expected and actual performance margins.
4.4.6 INSTRUMENTATION
REQUIREMENTS
The reactor vessel instrumentation monitors the key reactor vessel operating parameters during planned operations. This ensures sufficient control of the parameters. The following reactor vessel sensors are discussed in Subsection 7.7.1.1.
(1) Reactor Vessel Temperature (2) Reactor Vessel Water Level (3) Reactor Vessel Coolant Flow Rates and Differential Pressures (4) Reactor Vessel Internal Pressure (5) Neutron Monitoring System
4.
4.7 REFERENCES
4.4-1 General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Company, January 1977, (NEDO-10958A).
4.4-2 Core Flow Distribution in a Modern Boiling Water Reactor as Measured in Monticello, August 1976, (NEDO-10722A).
4.4-3 R.C. Martinelli and D. E. Nelson, "Prediction of Pressure Drops During Forced Convection Boiling of Water," ASME Trans., 70, pp 695-702, 1948.
4.4-4 Deleted
4.4-5 Jens, W. H., and Lottes, P.A., Analysis of Heat Transfer, Burnout, Pressure Drop, and Density Data for High Pressure Water, USAEC Report-4627, 1972.
4.4-6 Deleted
4.4-7 Deleted
4.4-8 Deleted
4.4-9 Deleted
4.4-10 Analytical Methods of Plant Transient Evaluations for General Electric Boiling Water Reactor, General Electric Company, BWR Systems Department, February 1973, (NEDO-10802).
4.4-11 Deleted
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-20 4.4-12 Deleted
4.4-13 Peach Bottom Atomic Power Station Units 2 and 3, Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibration, General Electric Co.,
NEDO-20994, September, 1975.
4.4-14 "Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration," NEDE-21156, Class III, January 1976.
4.4-15 Deleted
4.4-16 Deleted
4.4-17 Deleted 4.4-18 Deleted
4.4-19 Deleted
4.4-20 Deleted
4.4-21 Deleted
4.4-22 Deleted
4.4-23 Deleted
4.4-24 Deleted
4.4-25 Deleted 4.4-26 Deleted
4.4-27 KAPL-2170 Hydrodynamic Stability of a Boiling Channel, by A. B. Jones; 2 October 1961. 4.4-28 KAPL-2208 Hydrodynamic Stability of a Boiling Channel Part 2, by A. B. Jones; 20 April 1962.
4.4-29 KAPL-2290 Hydrodynamic Stability of a Boiling Channel Part 3, by A. B. Jones and D. G. Dight; 28 June 1963.
4.4-30 KAPL-3070 Hydrodynamic Stability of a Boiling Channel Part 4, by A. B. Jones; 18 August 1964.
4.4-31 KAPL-3072 Reactivity Stability of a Boiling Reactor Part 1, by A. B. Jones and W. M. Yarbrough; 14 September 1964.
4.4-32 KAPL-3093 Reactivity Stability of a Boiling Reactor Part 2, by A. B. Jones, 1 March 1965.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-21 4.4-33 Deleted
4.4-34 McBeth, R.V., R. Trenberth, and R. W. Wood, "An Investigation Into the Effects of Crud Deposits on Surface Temperature, Dry-Out, and Pressure Drop, with Forced Convection Boiling of Water at 69 Bar in an Annular Test Section," AEEW-R-705, 1971.
4.4-35 Green, S.J., B. W. LeTourneau, A.C. Peterson, "Thermal and Hydraulic Effects of Crud Deposited on Electrically Heated Rod Bundles," WAPD-TM-918, Sept. 1970.
4.4-36 H.T. Kim and H.S. Smith, "Core Flow Distribution in a General Electric Boiling Water Reactor as Measured in Quad Cities Unit 1," NEDO-10722A, August, 1976.
4.4-37 Licensing Topical Report, "Stability and Dynamic Performance of the General Electric Boiling Water Reactor," January, 1977 (NEDO-21506).
4.4-38 Deleted
4.4-39 Deleted 4.4-40 Deleted
4.4-41 Deleted
4.4-42 ANF-524 (P)(A), Revision 2 and Supplements 1 and 2, "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," November 1990.
4.4-43 "Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies", XN-NF-79-59(P)(A), November 1983.
4.4-44 MICROBURN-B2 Based Impact of Failed / Bypassed LPRMs and TIPs, Extended LPRM Calibration interval on Single Loop Operation on Measured Radial Bundle Power Uncertainty, "EMF-2493(P), Rev. 0, December 2000.
4.4-45 "Thermal-Hydraulic Characteristics of the ATRIUM-10 Fuel Design for Susquehanna", EMF-95-066(P), June 1995.
4.4-46 "Single Phase Hydraulic Performance of Exxon Nuclear BWR 9x9 Fuel Assembly", XN-NF-683(P), February 1983.
4.4-47 "Generic Mechanical Design Criteria for BWR Fuel Designs", ANF-89-98(P)(A) Revision 1, and Revision 1 Supplement 1, May 1995.
4.4-48 EMF-CC-074(P)(A), Volume 4, Revision 0, "BWR Stability Analysis - Assessment of STAIF with input from MOCROBURN-B2,"Siemens Power Corporation, August 2000.
4.4-49 "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description", XN-NF-80-19(P)(A) Volume 3 Revision 2, January 1987.
SSES-FSAR Text Rev. 61 FSAR Rev. 64 4.4-22 4.4-50 "Impact of Failed/Bypassed LPRMs and TIPs and Extended LPRM Calibration Interval on Radial Bundle Power Uncertainty", EMF-1903 Revision 2, October 1996.
4.4-51 "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4 / MICROBURN-B2," emf-2158(P)(A), Rev. 0, October 1999.
4.4-52 "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, August 1996
4.4-53 "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960-A, June 1991
4.4-54 "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960-A Supplement 1, March 1992
4.4-55 "ABB Option III Oscillation Power Range Monitor (OPRM)," CENPD-400-P-A, Revision 1, May 1995 4.4-56 NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip Function", October 1995.
4.4-57 NEDC-32410P-A Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip Function", November 1997.
4.4-58 EMF-2209 (P) (A), Revision 2, "SPCB Critical Power Correlation, "Framatome ANP, September 2003.
4.4-59 NRC Letter from R. V. Guzman (NRC) to B. T. McKinney (PPL), January 30, 2008,
Subject:
Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Regarding 13-Percent Extended Power Uprate (TAC Nos. MD3309 and MD 3310) [Accessi on ML 080020182]
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-4, Rev. 54 AutoCAD Figure 4_4_4.doc
THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-10, Rev. 54 AutoCAD Figure 4_4_10.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-6A, Rev. 54 AutoCAD Figure 4_4_6A.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-6B, Rev. 54 AutoCAD Figure 4_4_6B.doc THIS FIGURE HAS BEEN DELETED FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure Deleted FIGURE 4.4-6C, Rev. 54 AutoCAD Figure 4_4_6C.doc
THIS FIGURE HAS BEEN REPLACED BY DWG.
M1-C12-8, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 4.6-6 replaced by dwg. M1-C12-8, Sh. 1 FIGURE 4.6-6, Rev. 49 AutoCAD Figure 4_6_6.doc
THIS FIGURE HAS BEEN REPLACED BY DWG.
M-146, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 4.6-5A replaced by dwg. M-146, Sh. 1 FIGURE 4.6-5A, Rev. 50 AutoCAD Figure 4_6_5A.doc THIS FIGURE HAS BEEN REPLACED BY DWG.
M-147, Sh. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT Figure 4.6-5B replaced by dwg. M-147, Sh. 1 FIGURE 4.6-5B, Rev. 55 AutoCAD Figure 4_6_5B.doc