ML16357A131
Text
LGS UFSAR CHAPTER 4 - REACTOR TABLE OF CONTENTS
4.0 INTRODUCTION
4.1
SUMMARY
DESCRIPTION 4.1.1 Reactor Vessel 4.1.2 Reactor Internal Components 4.1.2.1 Reactor Core 4.1.2.1.1 General 4.1.2.1.2 Core Configuration 4.1.2.1.3 Fuel Assembly Description 4.1.2.1.4 Fuel Assembly Support and Control Rod Location 4.1.2.2 Shroud 4.1.2.3 Shroud Head and Steam Separator Assembly 4.1.2.4 Steam Dryer Assembly 4.1.3 Reactivity Control Systems 4.1.3.1 Operation 4.1.3.2 Description of Control Rods 4.1.3.3 Supplementary Reactivity Control 4.1.4 Analysis Techniques 4.1.4.1 Reactor Internal Components 4.1.4.1.1 MASS (Mechanical Analysis of Space Structure) 4.1.4.1.2 SNAP (MULTISHELL) 4.1.4.1.3 GASP 4.1.4.1.4 NOHEAT 4.1.4.1.5 FINITE 4.1.4.1.6 DYSEA 4.1.4.1.7 SHELL 5 4.1.4.1.8 HEATER 4.1.4.1.9 FAP-71 (Fatigue Analysis Program) 4.1.4.1.10 CREEP/PLAST 4.1.4.1.11 ANSYS 4.1.4.2 Fuel Rod Thermal Analysis 4.1.4.3 Reactor Systems Dynamics 4.1.4.4 Nuclear Engineering Analysis 4.1.4.5 Neutron Fluence Calculations 4.1.4.6 Thermal-Hydraulic Calculations 4.1.5 References 4.2 FUEL SYSTEM DESIGN 4.2.1 Design Bases CHAPTER 04 4-i REV. 16, SEPTEMBER 2012
LGS UFSAR TABLE OF CONTENTS (cont'd) 4.2.2 Description and Design Drawings 4.2.2.1 Reactivity Control Assembly (Control Rods) 4.2.2.2 Reactivity Control Assembly Evaluation 4.2.3 Design Evaluation 4.2.3.1 Exceptions to GESTAR II 4.2.4 Testing, Inspection, and Surveillance Plans 4.2.5 References 4.3 NUCLEAR DESIGN 4.3.1 Design Bases 4.3.2 Description 4.3.2.1 Nuclear Design Description 4.3.2.2 Power Distribution 4.3.2.3 Reactivity Coefficient 4.3.2.4 Control Requirements 4.3.2.4.1 Shutdown Reactivity 4.3.2.4.1 Reactivity Variations 4.3.2.5 Control Rod Patterns and Reactivity Worths 4.3.2.5.1 Scram Reactivity 4.3.2.6 Criticality of Reactor During Refueling 4.3.2.7 Stability 4.3.2.7.1 Xenon Transients 4.3.2.7.2 Thermal-Hydraulic Stability 4.3.2.8 Vessel Irradiations 4.3.3 Analytical Methods 4.3.4 Changes 4.3.5 References 4.4 THERMAL AND HYDRAULIC DESIGN 4.4.1 Design Basis 4.4.2 Description of Thermal-Hydraulic Design of the Reactor Core 4.4.2.1 Summary Comparison 4.4.2.2 Critical Power Ratio 4.4.2.3 Linear Heat Generation Rate 4.4.2.4 Void Fraction Distribution 4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern 4.4.2.6 Core Pressure Drop and Hydraulic Loads 4.4.2.7 Correlation and Physical Data 4.4.2.8 Thermal Effects of Operational Transients CHAPTER 04 4-ii REV. 16, SEPTEMBER 2012
LGS UFSAR TABLE OF CONTENTS (cont'd) 4.4.2.9 Uncertainties in Estimates 4.4.2.10 Flux Tilt Considerations 4.4.3 Description of the Thermal and Hydraulic Design of the Reactor Coolant System 4.4.3.1 Plant Configuration Data 4.4.3.1.1 Reactor Coolant System Configuration 4.4.3.1.2 Reactor Coolant System Thermal-Hydraulic Data 4.4.3.1.3 Reactor Coolant System Geometric Data 4.4.3.2 Operating Restrictions on Pumps 4.4.3.3 Power-Flow Operating Map 4.4.3.3.1 Limits for Normal Operation 4.4.3.3.2 Regions of the Power-Flow Map 4.4.3.4 Temperature-Power Operating Map (PWR) 4.4.3.5 Load-Following Characteristics 4.4.3.6 Thermal and Hydraulic Characteristics Summary Table 4.4.4 Evaluation 4.4.5 Testing and Verification 4.4.6 Instrumentation Requirements 4.4.6.1 Deleted 4.5 REACTOR MATERIALS 4.5.1 Control Rod System Structural Materials 4.5.1.1 Material Specifications 4.5.1.1.1 Material List 4.5.1.1.2 Special Materials 4.5.1.2 Austenitic Stainless Steel Components: Processes, Inspections, and Tests 4.5.1.3 Other Materials 4.5.1.4 Cleaning and Cleanliness Control 4.5.1.4.1 Protection of Materials During Fabrication, Shipping, and Storage 4.5.2 Reactor Internal Materials 4.5.2.1 Material Specifications 4.5.2.2 Controls on Welding 4.5.2.3 Nondestructive Examination of Wrought Seamless Tubular Products 4.5.2.4 Regulatory Guide Conformance for Fabrication and Processing of Austenitic Stainless Steel 4.5.2.5 Other Materials 4.5.3 Control Rod Drive Housing Supports 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 4.6.1 Control Rod Drive System 4.6.1.1 Design CHAPTER 04 4-iii REV. 16, SEPTEMBER 2012
LGS UFSAR TABLE OF CONTENTS (cont'd) 4.6.1.1.1 Safety Design Bases 4.6.1.1.2 Power Generation Design Basis 4.6.1.2 Description 4.6.1.2.1 Control Rod Drive Mechanisms 4.6.1.2.2 Drive Components 4.6.1.2.3 Materials of Construction 4.6.1.2.4 Control Rod Drive Hydraulic System 4.6.1.2.5 Control Rod Drive System Operation 4.6.1.2.6 Instrumentation 4.6.1.3 Control Rod Drive Housing Supports 4.6.1.3.1 Safety Objective 4.6.1.3.2 Safety Design Bases 4.6.1.3.3 Description 4.6.2 Evaluations of the CRD System 4.6.2.1 Failure Mode and Effects Analysis 4.6.2.2 Protection from Common Mode Failures 4.6.2.3 Safety Evaluation 4.6.2.3.1 Control Rods 4.6.2.3.2 Control Rod Drives 4.6.2.3.3 Control Rod Drive Housing Supports 4.6.2.3.4 Loss of Scram Discharge Valve Air Pressure 4.6.3 Testing and Verification of the Control Rod Drives 4.6.3.1 Control Rod Drives 4.6.3.1.1 Testing and Inspection 4.6.3.2 Control Rod Drive Housing Supports 4.6.3.2.1 Testing and Inspection 4.6.4 Information for Combined Performance of Reactivity Systems 4.6.4.1 Vulnerability to Common Mode Failures 4.6.4.2 Accidents Taking Credit for Multiple Reactivity Systems 4.6.5 Evaluation of Combined Performance 4.6.6 References CHAPTER 04 4-iv REV. 16, SEPTEMBER 2012
LGS UFSAR CHAPTER 4 - REACTOR LIST OF TABLES TABLE TITLE 4.3-1 Calculated Neutron Fluxes (Used to Evaluate Vessel Irradiation) 4.3-2 Calculated Neutron Flux at Core Equivalent Boundary 4.3-3 Calculated Initial Core Effective Multiplication and Control System Worth - No Voids, 20oC 4.4-1 Thermal and Hydraulic Design Characteristics of the Reactor Core 4.4-2 Reactor Coolant System Geometric Data 4.4-3 Axial Void Fraction Distribution 4.4-4 Axial Flow Quality Distribution 4.4.5 Axial Power Distribution Used to Generate Void and Quality Distributions 4.4-6 Deleted 4.4-7 Safety Injection Line Lengths 4.4-8 Stability Analysis Results (Cycle-1 Most Limiting Conditions) 4.6-1 Identification of Scram Discharge System Components CHAPTER 04 4-v REV. 16, SEPTEMBER 2012
LGS UFSAR CHAPTER 4 - REACTOR LIST OF FIGURES FIGURE TITLE 4.3-1 Model for One-Dimensional Transport Analysis of Vessel Fluence 4.3-2 Radial Power Distributions Used in the Vessel Fluence Calculation 4.3-3 Reference Loading Pattern 4.4-1 Power-Flow Operating Map 4.4-2 Total Core Stability 4.4-3 10 psi Pressure Regulator Setpoint Step at 51.5% Rated Power (Natural Circulation) 4.4-4 10 Cent Rod Reactivity Step at 51.5% Rated Power (Natural Circulation) 4.4-5 6 Inch Water Level Setpoint Step at 51.5% Rated Power (Natural Circulation) 4.4-6 6 Inch Water Level Setpoint Step at 68% Rated Power and 51.5% Rated Flow 4.4-7 Deleted 4.6-1 Control Rod to Control Rod Drive Coupling 4.6-2 Control Rod Drive Unit 4.6-3 Control Rod Drive Schematic BWR/4&5 4.6-4 Control Rod Drive 4.6-5 Deleted 4.6-6 Deleted 4.6-7 Deleted 4.6-8 Control Rod Drive Hydraulic Control Unit 4.6-9 Control Rod Drive Housing Support CHAPTER 04 4-vi REV. 16, SEPTEMBER 2012
4.0 INTRODUCTION
This chapter was prepared using the latest approved revision of the topical report, "General Electric Standard Application for Reactor Fuel" (GESTAR-II) including the "United States Supplement," NEDE-24011-P-A and NEDE-24011-P-A-US. Applicable sections of this report are referenced as noted in Sections 4.1 through 4.4.
CHAPTER 04 4.0-1 REV. 13, SEPTEMBER 2006
SUMMARY
DESCRIPTION The reactor assembly consists of the reactor vessel, its internal components of the core, steam separator and dryer assemblies, and jet pumps. Also included in this assembly are the control rods, CRD housings, and the CRDs. Figure 3.9-4 shows the arrangement of reactor assembly components. A summary of the important design and performance characteristics is given in Section 1.3. Loading conditions for reactor assembly components are discussed in Section 3.9.
4.1.1 REACTOR VESSEL The reactor vessel design and description is discussed in Section 5.3.
4.1.2 REACTOR INTERNAL COMPONENTS The major reactor internal components are the core (fuel, channels, control blades, and incore instrumentation), core support structure (including the shroud, top guide, and core plate), shroud head and steam separator assembly, steam dryer assembly, feedwater spargers, core spray spargers, and jet pumps. Except for the Zircaloy in the reactor core, these reactor internals are made of stainless steel or other corrosion-resistant alloys. Of the preceding components, the fuel assemblies (including fuel rods and channel), control blades, incore instrumentation, shroud head and steam separator assembly, steam dryers and jet pump subassemblies are removable when the reactor vessel is open.
4.1.2.1 Reactor Core 4.1.2.1.1 General The design of the BWR core (including fuel) is based on the proper combination of many design variables and operating experience. This contributes to the achievement of high reliability.
Important features of the reactor core arrangement are as follows:
- a. The bottom-entry control blades are cruciform-shaped and consist of stainless steel clad neutron absorbing material as described in Section 4.2. The neutron-absorbing material is surrounded by a stainless steel sheath in the conventional designs. In the newer designs, (e.g., GE Marathon Series) the absorber tube and sheath arrangement is replaced with an array of welded square tubes containing the neutron absorbing material.
- b. Fixed incore fission chambers provide continuous power range neutron flux monitoring. A guide tube in each incore assembly provides for a traversing ion chamber for calibration and axial detail. Source and intermediate range detectors are located incore and are axially retractable. The incore location of the startup and source range instruments provides coverage of the large reactor core and provides an acceptable signal-to-noise ratio and neutron-to-gamma ratio. All incore instrument leads enter from the bottom. Source and intermediate range instruments are in service during refueling. Incore instrumentation is further discussed in Section 7.6 and 7.7 .
CHAPTER 04 4.1-1 REV. 15, SEPTEMBER 2010
- c. Experience has shown that the operator, utilizing the incore flux monitoring system, can maintain the desired power distribution within a large core by proper control rod positioning.
- d. The reusable channels 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.
- e. Mechanical reactivity control permits criticality checks during refueling and provides maximum plant safety. The core is designed to be subcritical at any time in its operating history, with any one control rod fully withdrawn.
- f. The selected control rod pitch represents a practical value of individual control rod reactivity worth, and allows adequate clearance below the pressure vessel, between CRDMs, 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 core arrangement (plan view) and the lattice configuration are given in Section 4.3.2.
4.1.2.1.3 Fuel Assembly Description Descriptions of the fuel assembly and the fuel rods are given in Section 4.2.
4.1.2.1.4 Fuel Assembly Support and Control Rod Location Twenty four peripheral fuel assemblies and their individual peripheral fuel support pieces 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 CRD 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-shaped control rods (Section 4.2.2) and their associated mechanical-hydraulic drive system. The control rods occupy alternate spaces between fuel assemblies. Each independent CRD enters the core from the bottom, accurately positions its associated control rod during normal operation, and yet exerts 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 CRD maintenance.
4.1.2.2 Shroud Information on the shroud is contained in Section 3.9.5.
4.1.2.3 Shroud Head and Steam Separator Assembly CHAPTER 04 4.1-2 REV. 15, SEPTEMBER 2010
LGS UFSAR Information on the shroud head and steam separators is contained in Section 3.9.5.
4.1.2.4 Steam Dryer Assembly Information on the steam dryer assembly is contained in Section 3.9.5.
4.1.3 REACTIVITY CONTROL SYSTEMS 4.1.3.1 Operation 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 to counterbalance steam voids in the top of the core and effect 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 designed for both reactor scram and 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 Control Rods A description of the control rods is given in Section 4.2.2.1.
4.1.3.3 Supplementary Reactivity Control The initial and reload core control requirements are met by use of the combined effects of the movable control rods, supplementary burnable poison, and the variation of reactor coolant flow. A description of the supplementary burnable poison is given in Section 4.2.2.
4.1.4 ANALYSIS TECHNIQUES 4.1.4.1 Reactor Internal Components Computer codes used for the analysis of the internal components are:
- a. MASS
- b. SNAP (MULTISHELL)
- c. GASP
- d. NOHEAT CHAPTER 04 4.1-3 REV. 15, SEPTEMBER 2010
- e. FINITE
- f. DYSEA
- g. SHELL 5
- h. HEATER
- i. FAP 71
- j. CREEP PLAST
- k. ANSYS Detailed descriptions of these programs are given below:
4.1.4.1.1 MASS (Mechanical Analysis of Space Structure) 4.1.4.1.1.1 Program Description This program, proprietary to GE, 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 can analyze curved beam, plate, and shell elements. It can accommodate mechanical and thermal loads in a static analysis, and can predict natural frequencies and mode shapes in a dynamic analysis.
4.1.4.1.1.2 Program Version and Computer GE's Nuclear Energy Division used revision "0" of MASS. The program operated on a Honeywell 6000 computer.
4.1.4.1.1.3 History of Use Since its development in the early 1960s, 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 by GE's Nuclear Energy Division began shortly after its development.
4.1.4.1.1.4 Extent of Application Besides the Jet Engine and Nuclear Energy Divisions, GE's Missile and Space Division, the Appliance Division, and the Turbine Division have also applied the program to a wide range of engineering problems. The Nuclear Energy Division uses it mainly for piping and reactor internals analyses.
4.1.4.1.2 SNAP (MULTISHELL) 4.1.4.1.2.1 Program Description CHAPTER 04 4.1-4 REV. 15, SEPTEMBER 2010
LGS UFSAR The SNAP program, which is also called MULTISHELL, is the GE 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 The version used for the Limerick design, was maintained by GE's Jet Engine Division at Evandale, Ohio, and was used on the Honeywell 6000 computer by the Nuclear Energy Division.
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 by 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. It has been one of the main shell analysis programs in GE's Nuclear Energy Division, because of its efficiency, economy, and reliability.
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 load inputs can be made 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, operated on the Honeywell 6000 computer.
4.1.4.1.3.3 History of Use The program was developed in 1965 (Reference 4.1-3). The present version used by GE's Nuclear Energy Division has been in operation since 1967.
4.1.4.1.3.4 Extent of Application CHAPTER 04 4.1-5 REV. 15, SEPTEMBER 2010
LGS UFSAR The application of GASP by the Nuclear Energy Division is mainly for elastic analysis of axisymmetric and plane structures, such as the jet pump riser brace, under thermal and pressure loads. The GE version has been extensively tested and used by GE engineers.
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 a 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 the temperature of all the nodal points for the time instants required by the user. The program can accommodate multitransient temperature input.
4.1.4.1.4.2 Program Version and Computer The version of the program used for the Limerick design was an improvement of the program originally developed by I. Farhoomand and Professor E.L. Wilson (Reference 4.1-4). The program operated on the Honeywell 6000 computer.
4.1.4.1.4.3 History of Use The program was developed in 1971 and installed in GE's 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 discussed above. Such compatibility simplifies the connection of the two analyses and minimizes 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 plane stress, plane strain, and 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 orthotropic. The development of the FINITE program is based on the GASP program discussed above.
4.1.4.1.5.2 Program Version and Computer CHAPTER 04 4.1-6 REV. 15, SEPTEMBER 2010
LGS UFSAR The present version of the program used for the Limerick design at GE's Nuclear Energy Division was obtained from the developer, J.E. McConnelee of GE's Gas Turbine Department, in 1969 (Reference 4.1-5). The Nuclear Energy Division version operated 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 GE's Gas Turbine and Jet Engine Departments for turbine component analysis.
4.1.4.1.5.4 Extent of Usage The program is used at GE's Nuclear Energy Division 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/internals and building system. It calculates the dynamic response of linear structural systems by either temporal modal superposition or response spectrum method. The fluid-structure interaction effect in the RPV is taken into account by way of hydrodynamic mass.
DYSEA is based on program SAPIV with added capability to handle the hydrodynamic mass effect. Structural stiffness and mass matrices are formulated similarly to SAPIV. Solution is obtained in time domain by calculating the dynamic response mode by mode. Time integration is performed by using Newmark's Beta method. Response spectrum solution is also available as an option.
4.1.4.1.6.2 Program Version and Computer The DYSEA version used for the Limerick design on the Honeywell 6000 computer 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 could accommodate three-dimensional dynamic problems with beam, trusses, and springs. Both acceleration time histories and response spectra could be used as input.
4.1.4.1.6.3 History of Use The DYSEA program was developed in 1976. Adopted as a standard production program in 1977, it has been used extensively in all dynamic and seismic analysis of the RPV/internals and building system.
4.1.4.1.6.4 Extent of Application CHAPTER 04 4.1-7 REV. 15, SEPTEMBER 2010
LGS UFSAR 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 inplane 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 inplane rotation can be significant.
4.1.4.1.7.2 Program Version and Computer A copy of the source deck of SHELL 5 is retained in GE's Nuclear Energy Division. SHELL 5 was used 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, GE, and at other major computer operating systems since 1970.
4.1.4.1.7.4 Extent of Application SHELL 5 has been used at GE 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 used in the hydraulic design of feedwater spargers, and their associated delivery header and piping. The program utilizes test data obtained by GE using full-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 CHAPTER 04 4.1-8 REV. 15, SEPTEMBER 2010
LGS UFSAR This program was developed at GE's Nuclear Energy Division in FORTRAN IV for the Honeywell 6000 computer.
4.1.4.1.8.3 History of Use The program was developed in the Nuclear Energy Division beginning in 1970. The version used for the Limerick design was 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 in 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 Section III structural design calculations. Specifically, FAP-71 is used in determining the primary plus secondary stress range, and the number of allowable fatigue cycles at points of interest. For structural locations at which the 3 Sm (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 Paper 68-PVP-3, 2) the method of paragraph NB-3228.3 of the ASME Section III, 1971 edition. The program can accommodate up to 25 transient stress states of as many as 20 structural locations.
4.1.4.1.9.2 Program Version and Computer The version of FAP-71 used for the Limerick design was completed by L. Young of GE's Nuclear Energy Division in 1971 (Reference 4.1-9). The program was used on the 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 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 This finite-element program is used for the analysis of two-dimensional (plane and axisymmetric) problems under the 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 CHAPTER 04 4.1-9 REV. 15, SEPTEMBER 2010
LGS UFSAR 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 be a function of strain.
4.1.4.1.10.2 Program Version and Computer The program was developed for elastic-plastic analysis with or without the presence of creep. It was also designed for creep analysis without the presence of instantaneous plasticity. A detailed description of theory is given in Reference 4.1-11. The program was used on the UNIVAC 1108 computer .
4.1.4.1.10.3 History of Use This program was developed by Y.R. Rashid (Reference 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's Nuclear Energy Division in the channel cross-section mechanical analysis.
4.1.4.1.11 ANSYS 4.1.4.1.11.1 Program Description ANSYS is a general purpose finite-element computer program designed to solve a variety of problems in engineering analysis. The ANSYS program features the following capabilities:
- a. Structural analysis including the following analyses:
- 1. Static Elastic
- 2. Plastic and Creep
- 3. Dynamic
- 4. Seismic and Dynamic Plastic
- 5. Large Deflection
- 6. Stability
- b. One-dimensional fluid flow analyses
- c. Transient heat transfer analysis including conduction, convection, and radiation with direct input to thermal-stress analyses CHAPTER 04 4.1-10 REV. 15, SEPTEMBER 2010
- d. An extensive finite-element library, including gaps, friction interfaces, springs, cables (tension only), direct interfaces (compression only), curved elbows, etc.
Many of the elements contain complete plastic, creep, and swelling capabilities.
- e. Geometry plotting available for all elements in the ANSYS library, including isometric and perspective views of three-dimensional structures
- f. Restart Capability - The ANSYS program has restart capability for several types of analyses. An option is also available for saving the stiffness matrix once it is calculated for the structure, and for using it for other loading conditions.
4.1.4.1.11.2 Program Version & Computer The program used in the Limerick design was maintained current by Swanson Analysis Systems, Inc. of Pittsburgh, Pennsylvania, and was supplied to GE for use on the Honeywell 6000.
4.1.4.1.11.3 History of Use The ANSYS program has been used for productive analyses since 1970. Users now include nuclear, pressure vessel, piping, mining, structures, bridge, chemical, and automotive industries, as well as many consulting firms.
4.1.4.1.11.4 Extent of Application ANSYS is used extensively by GE's Nuclear Energy Division for elastic and elastic-plastic analysis of the RPV, core support structures, and reactor internals.
4.1.4.2 Fuel Rod Thermal Analysis Fuel rod thermal design analyses are referenced in sections A.4.2.1 and A.4.2.3 of GESTAR II (Reference 4.1-1).
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. Section 4.4.4.6 of Reference 4.1-10 also provides a complete stability analysis for the RCS.
4.1.4.4 Nuclear Engineering Analysis The analysis techniques are described and referenced in section 3 of GESTAR II (Reference 4.1-1).
4.1.4.5 Neutron Fluence Calculations The original vessel neutron fluence calculations were carried out using a one-dimensional discrete ordinates (Sn) transport code with general anisotropic scattering.
This code is a modification of a widely used discrete ordinates code which solves a wide variety of radiation transport problems. The program solves both fixed source and multiplication problems.
CHAPTER 04 4.1-11 REV. 15, SEPTEMBER 2010
LGS UFSAR Slab, cylinder, and spherical geometry are allowed with various boundary conditions. The fluence calculations were done as a fixed source problem in cylindrical geometry. The fixed source was developed from neutron fission distribution data prepared as a part of the core physics data.
Anisotropic scattering was considered for all regions. The cross-sections were prepared with a l/E flux weighted and (P) matrices for anisotropic scattering, but did not include resonance self-shielding factors. Fast neutron fluxes at locations other than the core midplane were calculated using a two-dimensional discrete ordinate code. The two-dimension code is an extension of the one-dimensional code.
This calculational procedure was also used on PBAPS Units 2 and 3. These units are similar in design, and are of the same core size and power rating as the LGS units. The neutron flux at these units has been measured using a threshold foil technique on the wire samples contained in the vessel surveillance samples. Flux greater than 1 MeV was determined to be 1.3x109 2 8 2 n/cm -sec for Unit 2, and 9.8x10 n/cm -sec for Unit 3. This is equivalent to a fluence over 40 years of 1.3x10 nvt and 9.8x10 nvt. The calculated value using these methods is 1.4x1018 nvt.
18 17 Following the original fluence calculations, the NRC issued Regulatory Guide (RG) 1.190, which provides state of the art calculation and measurement procedures that are acceptable to the NRC for determining Reactor Pressure Vessel (RPV) neutron fluence. LGS RPV fluence has been evaluated using a method in accordance with the recommendations of RG 1.190. Future evaluations of RPV fluence will be completed using a method in accordance with the recommendations of RG 1.190 (as noted in Reference 5.3-14).
4.1.4.6 Thermal-Hydraulic Calculations A description of the thermal-hydraulic models is given in section 4 of GESTAR II (Reference 4.1-1).
4.
1.5 REFERENCES
4.1-1 "General Electric Standard Application for Reactor Fuel," including the United States Supplement, NEDE-24011-P-A and NEDE-24011-P-A-US, (latest approved revision).
4.1-2 L. Beitch, "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 Analysis 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, "Nonlinear Heat Transfer Analysis of Axisymmetric Solids", SESM Report SESM71 6, University of California at Berkeley, Berkeley, California, (1971).
4.1-5 J.E. McConnelee, "Finite-Users Manual", GE TIS Report DF 69SL20 6, (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).
CHAPTER 04 4.1-12 REV. 15, SEPTEMBER 2010
LGS UFSAR 4.1-7 "A Computer Program For the Structural Analysis of Arbitrary Three-Dimensional Thin-Shells", Report No. GA 9952, Gulf General Atomic, (January 1970).
4.1-8 A.B. Burgess, "User Guide and Engineering Description of HEATER Computer Program", NEDE-20371-02, (March 1974).
4.1-9 L.J. Young, "FAP-71 (Fatigue Analysis Program) Computer Code", GE's Nuclear Energy Division Design Analysis Unit R. A. Report No. 49, (January 1972).
4.1-10 L.A. Carmichael and G.J. Scatena, "Stability and Dynamic Performance of the General Electric Boiling Water Reactor", APED 5652, (April 1969).
4.1-11 Y.R. Rashid, "Users Manual for CRPLSO1 Computer Program", NEDO-23538, (December 1976).
4.1-12 DELETED.
4.1-13 Methodology and Uncertainties for Safety Limit MCPR Evaluations, GE Nuclear Energy document No. NEDC-32601P-A, (latest approved revision).
CHAPTER 04 4.1-13 REV. 15, SEPTEMBER 2010
LGS UFSAR 4.2 FUEL SYSTEM DESIGN The format of this section corresponds to SRP Section 4.2 in NUREG-0800. Most of the information is presented by reference to GESTAR II (Reference 4.1-1).
4.2.1 DESIGN BASES References to design bases are given in subsection A.4.2.1 of GESTAR II (Reference 4.1-1).
4.
2.2 DESCRIPTION
AND DESIGN DRAWINGS References to the fuel system description and design drawings are given in subsections A.1.2.2.3.1 and A.4.2.2 of GESTAR II (Reference 4.1-1). Details of General Electric fuel assembly designs are provided in Reference 4.2-5. For the advanced Nuclear Fuels Corp. 9X9-9, the ABB Atom Inc. SVEA-96, and the General Electric GE11 Qualification Fuel Bundles first used in Limerick 2, Cycle 2, the fuel design description can be found in Reference 4.2-6.
4.2.2.1 Reactivity Control Assembly (Control Rods)
The Unit 1 and Unit 2 cores contain a mix of control rod designs including the standard boron carbide and the advanced long-life hafnium rods. The long-life control rods are inserted in subsequent cycles as the standard rods are depleted. A description of the standard rods is given in subsection 2.2.4 and is shown in figures 2.6a, 2.6b, and 2.7 of NEDE-20944-P-1 (Reference 4.2-1). The advanced long-life control rods are described in subsection 2.1 and shown in figures 2.2 and 2.3 of NEDE-22290-A, Supplement 2 (Reference 4.2-3) and subsection 2.0 and Figures 2-1 and 2-2 of NEDE-31758P-A (Reference 4.2-7).
4.2.2.2 Reactivity Control Assembly Evaluation The standard boron carbide control rod and the advanced long-life hafnium control rod differ primarily in the type of neutron- absorbing materials. Other minor design differences are described in NEDE-22290-A, Supplement 2 (Reference 4.2-3). The evaluation of the standard control rod is given in subsection 2.3.3 of NEDE-20944-P-1 (Reference 4.2-1). The advanced long-life control rod assemblies are directly interchangeable with the standard assemblies and are compatible with existing hardware.
4.2.3 DESIGN EVALUATION Compliance with the design bases is discussed in subsection A.4.2.3 of GESTAR II (Reference 4.1-1), with the exceptions as provided below.
4.2.3.1 Exceptions to GESTAR II PARAGRAPH 4.2.3.2.9 - MECHANICAL FRACTURING EVALUATION Mechanical adequacy of the fuel assembly is demonstrated by the mechanical design analyses described in section 2.0 of Reference 4.2-4. Corrosion, oxidation, and hydriding effects are specifically addressed in subsection 2.2.1.4 of Reference 4.2-4.
CHAPTER 04 4.2-1 REV. 16, SEPTEMBER 2012
LGS UFSAR All mechanical breaking under normal operation and abnormal operational transients is bounded by the analysis for LOCA plus SSE given in Section 3.9.1.4.10 and Table 3.9-6(x).
PARAGRAPH 4.2.3.3.5 - STRUCTURAL DEFORMATION EVALUATION Results of the LGS specific SSE plus LOCA analysis are documented in Section 3.9.1.4.10.
4.2.4 TESTING, INSPECTION, AND SURVEILLANCE PLANS Descriptions of fuel assembly testing, inspection, and surveillance are referenced in subsection A.4.2.4 of GESTAR II (Reference 4.1-1), with the exception that the following postirradiation fuel surveillance program is intended to conform with the objectives of Regulatory Guide 1.70.
Routine visual inspection of representative (usually discharged) fuel will be performed in accordance with the GE Postirradiation Fuel Surveillance Program (Reference 4.2-2).
4.
2.5 REFERENCES
4.2-1 "BWR/4 and BWR/5 Fuel Design," NEDE-20944-P-1 (Proprietary) and NEDO-20944-1, (October 1976 and Amendment 1, January 1977).
4.2-2 Letter from L.S. Rubenstein (NRC) to R.L. Gridley (GE) "Acceptance of GE Proposed Fuel Surveillance Program," (June 27, 1984).
4.2-3 "Safety Evaluation of the General Electric Hybrid 1 Control Rod Assembly",
NEDE-22290-A (Proprietary), (September 1983), and NEDE-22290-A, Supplement 2, (August 1985).
4.2-4 NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel", (latest approved version).
4.2-5 "GE Fuel Bundle Designs", NEDE-31152P, (latest revision).
4.2-6 G. J. Beck to USNRC, "Limerick Generating Station Unit 2 Qualification Fuel Bundles in Operating Cycle 2", March 11, 1991.
4.2-7 "GE Marathon Control Rod Assembly," NEDE-31758P-A, (October 1991).
CHAPTER 04 4.2-2 REV. 16, SEPTEMBER 2012
LGS UFSAR 4.3 NUCLEAR DESIGN Most of the information of Section 4.3 is provided in the licensing topical report, GESTAR II (Reference 4.1-1). The subsection numbers in Section 4.3 directly correspond to the subsection numbers of Appendix A of GESTAR II. Any additions or differences are given below for each applicable subsection.
4.3.1 DESIGN BASES 4.
3.2 DESCRIPTION
4.3.2.1 Nuclear Design Description The nuclear design description in GESTAR II is referenced in subsection A.4.3.2.1 of GESTAR II (Reference 4.1-1). The reference loading pattern for the initial core for Limerick 1 and 2 is shown in Figure 4.3-3. The reference loading pattern for subsequent cycle of Limerick 1 and Limerick 2 can be found in the unit and cycle specific Supplemental Reload Licensing Report. The actual as loaded core may be different than the reference loading pattern. The as loaded core can be found in the unit and cycle specific Core Design Report.
4.3.2.2 Power Distribution Power distribution is referenced in subsection A.4.3.2.2 of GESTAR II (Reference 4.1-1).
4.3.2.3 Reactivity Coefficient Directly corresponds to information in Appendix A of GESTAR II (Reference 4.1-1).
4.3.2.4 Control Requirements Directly corresponds to information in Appendix A of GESTAR II (Reference 4.1-1).
4.3.2.4.1 Shutdown Reactivity Information on shutdown reactivity is referenced in subsection A.4.3.2.4.1 of GESTAR II (Reference 4.1-1), except for the cold shutdown margin for the reference initial core loading pattern, which is given in Table 4.3-3.
4.3.2.4.2 Reactivity Variations Information on reactivity variations is referenced in subsection A.4.3.2.4.2 of GESTAR II (Reference 4.1-1). The combined effects of the individual constituents of reactivity are accounted for in each Keff in Table 4.3-3.
4.3.2.5 Control Rod Patterns and Reactivity Worths Control rod patterns and reactivity worths are discussed in section 3.2.5 of NEDE-20944-P-1 (Reference 4.2-1). Typical control rod patterns and the associated power distributions are presented in appendix A of NEDE-20944-P-1 (Reference 4.2-1). These control rod patterns are CHAPTER 04 4.3-1 REV. 16, SEPTEMBER 2012
LGS UFSAR calculated with the BWR Core Simulator. Qualification for this model is discussed and referenced in section 3.1 of GESTAR II (Reference 4.1-1) 4.3.2.5.1 Scram Reactivity Scram reactivity is calculated as described in section S.2 of GESTAR II (Reference 4.1-1) and is discussed in section 3.2.5.3 of NEDE-20944-P-1 (Reference 4.2-1).
4.3.2.6 Criticality of Reactor During Refueling Directly corresponds to information in Appendix A of GESTAR II (Reference 4.1-1).
4.3.2.7 Stability General Design Criterion 12 of 10CFR50 Appendix A states that the reactor core and associated coolant control and protection systems shall be designed to assure power oscillations that can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed. That requirement is met as described in GESTAR II (Reference 4.1-1) and this section.
4.3.2.7.1 Xenon Transients As documented in GESTAR II (Reference 4.1-1), Xenon transients do not cause instability in BWR reactors.
4.3.2.7.2 Thermal-Hydraulic Stability LGS has implemented the BWROG Long Term Stability Solution Option III. An OPRM Upscale Function is incorporated into each APRM channel to reliably detect power oscillations in the operating ranges where thermal-hydraulic instability has been determined to be credible. Upon detection of power oscillations, the OPRM Upscale Function generates a trip signal to RPS, which results in an automatic scram to suppress the oscillation before the MCPR Safety Limit is reached. The OPRM Upscale Function is described in References 7.6-1 through 7.6-4 and in Reference 4.3-1.
Stability-based MCPR Operating Limits are calculated for each operating cycle. These calculated values validate the selected OPRM setpoints for a given core configuration. Thus, the core design combined with the OPRM Upscale Function and selected system setpoints for detection and suppression of thermal-hydraulic power oscillation conform to the requirements of General Design Criterion 12 of 10CFR50, Appendix A.
4.3.2.8 Vessel Irradiations The neutron fluxes at the vessel were originally calculated using the one-dimensional discrete ordinates transport code described in Section 4.1.4.5. The discrete ordinates code was used in a distributed source mode with cylindrical geometry. The geometry described six regions from the center of the core to a point beyond the vessel. The core region was modeled as a single homogenized cylindrical region. The coolant water region between the fuel channel and the shroud was described as containing saturated water at 550oF and 1050 psi. The material CHAPTER 04 4.3-2 REV. 16, SEPTEMBER 2012
LGS UFSAR compositions for the stainless steel in the shroud and the carbon steel in the vessel contain the mixtures by weight, as specified in the material specifications for ASME SA240, 304L, and ASME SA533 Grade B. In the region between the shroud and the vessel (for conservatism), the presence of the jet pumps was ignored. A simple diagram showing the regions, dimensions, and weight fractions is shown in Figure 4.3-1.
The distributed source used for this analysis was obtained from the gross radial power description. The distributed source at any point in the core is the product of the power from the power description and the neutron yield from fission. By using the neutron energy spectrum, the distributed source is obtained for position and energy. The integral over position and energy is normalized to the total number of neutrons in the core region. The core region is defined as a 1 cm thick disc with no transverse leakage. The power in this core region is set equal to the maximum power in the axial direction. The radial power distribution is shown in Figure 4.3-2 The neutron fluence is determined from the calculated flux by assuming that the plant is operated 90% of the time at 90% power level for 40 years, or equivalent to 1x109 full power seconds. The calculated fluxes and fluence are shown in Table 4.3-1. The calculated neutron flux leaving the cylindrical core is shown in Table 4.3-2 Following the original fluence calculations, the NRC issued Regulatory Guide (RG) 1.190, which provides state of the art calculation and measurement procedures that are acceptable to the NRC for determining Reactor Pressure Vessel (RPV) neutron fluence. LGS RPV fluence has been evaluated using a method in accordance with the recommendations of RG 1.190. Future evaluations of RPV fluence will be completed using a method in accordance with the recommendations of RG 1.190 (as noted in Reference 5.3-14).
4.3.3 ANALYTICAL METHODS Analytical methods are referenced in section A.4.3.3 of GESTAR II (Reference 4.1-1).
4.3.4 CHANGES Changes are referenced in section A.4.3.4 of GESTAR II (Reference 4.1-1).
4.
3.5 REFERENCES
4.3-1 LEAM-MUR-0042, Rev. 0, Thermal Hydraulic Stability, dated December, 2009 CHAPTER 04 4.3-3 REV. 16, SEPTEMBER 2012
LGS UFSAR Table 4.3-1 CALCULATED NEUTRON FLUXES (USED TO EVALUATE VESSEL IRRADIATION)(1)
FLUX AT THE AVERAGE FLUX FLUX AT THE INSIDE SURFACE NEUTRON ENERGY IN THE CORE CORE BOUNDARY VESSEL Mev (n/cm2-sec) (n/cm2-sec) (n/cm2-sec)
>3.0 1.5x1013 5.4x1012 3.3x108 1.0 - 3.0 3.3x1013 1.2x1013 2.8x108 0.1 - 1.0 5.3x1013 1.7x1013 4.9x108 Maximum Fluence(2) >1.0 MeV = 1.4x1018 n/cm2
_________________
(1)
The calculated flux is a maximum in the axial direction, but is an average over the azimuthal angle.
(2)
The maximum fluence is calculated using the flux and a capacity factor of 80% or 1x109 full power seconds. The fluence includes an azimuthal peaking factor and a factor to cover analytical uncertainties.
CHAPTER 04 4.3-4 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.3-2 CALCULATED NEUTRON FLUX AT CORE EQUIVALENT BOUNDARY LOWER ENERGY FLUX GROUP BOUND (eV) (n/cm2-sec) 1 10.0x106 4.6x1010 2 6.065x106 6.1x1011 3 3.679x106 2.1x1012 4 2.231x106 4.2x1012 6 12 5 1.353x10 4.4x10 6 8.208x105 3.9x1012 7 4.979x105 4.0x1012 8 3.020x105 2.8x1012 5 12 9 1.832x10 2.3x10 10 1.111x105 1.8x1012 11 6.732x104 1.4x1012 4 12 12 4.087x10 1.1x10 13 2.478x104 1.0x1012 14 1.503x104 1.0x1012 15 9.119x103 9.6x1011 16 5.531x103 9.4x1011 3 11 17 3.355x10 9.4x10 18 2.034x103 9.1x1011 3 12 19 1.010x10 1.3x10 20 2.492x102 2.5x1012 21 5.560x101 2.6x1012 1 12 22 1.240x10 2.5x10 23 0.625 4.0x1012 13 24 0.0 2.5x10 CHAPTER 04 4.3-5 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.3-3 CALCULATED INITIAL CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20oC Beginning of Cycle-1, K-effective Unit 1 Unit 2 Uncontrolled 1.1047 1.1051 Fully Controlled 0.9230 0.9330 Strongest Control Rod Out (26-55) 0.9826 0.9799 R, Maximum Increase in Cold Core Reactivity with Exposure Cycle-1, k 0.0 0.0 CHAPTER 04 4.3-6 REV. 13, SEPTEMBER 2006
LGS UFSAR 4.4 THERMAL AND HYDRAULIC DESIGN Most of the information of Section 4.4 is provided in the licensing topic report GESTAR II (Reference 4.1-1). The subsection numbers in Section 4.4 directly correspond to the subsection numbers of Appendix A of GESTAR II. The differences are discussed below.
4.4.1 DESIGN BASIS The thermal and hydraulic design bases are referenced in GESTAR II (Reference 4.1-1).
The design steady-state initial operating cycle limit for the MCPR and the LHGR are given in Table 4.4-1. The MCPR and LHGR operating limits for subsequent cycles can be found in the unit and cycle specific Core Operating Limits Report.
4.4.2 Description Of Thermal-Hydraulic Design Of The Reactor Core A description of the thermal-hydraulic design of the reactor core is referenced in GESTAR II (Reference 4.1-1). An evaluation of plant performance from a thermal and hydraulic standpoint is provided in Section 4.4.3.
4.4.2.1 Summary Comparison A summary comparison of the thermal and hydraulic design parameters of the reactor with reactors of similar design is provided in Table 4.4-1.
4.4.2.2 Critical Power Ratio Critical power ratio is referenced in GESTAR II (Reference 4.1-1).
4.4.2.3 Linear Heat Generation Rate LHGR models are referenced in GESTAR II (Reference 4.1-1).
4.4.2.4 Void Fraction Distribution The core average and maximum exit void fractions in the core at rated condition are given in Table 4.4-1. The axial distribution of core void fractions for the average radial channel and the maximum radial channel (end of node value) for the core are given in Table 4.4-3. The core average and maximum exit values are also provided. Similar distributions for steam quality are provided in Table 4.4-4. The core average axial power distribution used to produce these tables is given in Table 4.4-5.
4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern The core coolant flow distribution and orificing pattern are referenced in GESTAR II (Reference 4.1-1).
4.4.2.6 Core Pressure Drop and Hydraulic Loads Core pressure drop and hydraulic loads are referenced in GESTAR II (Reference 4.1-1).
CHAPTER 04 4.4-1 REV. 17, SEPTEMBER 2014
LGS UFSAR 4.4.2.7 Correlation and Physical Data GE has obtained substantial amounts of physical data in support of the pressure drop and thermal-hydraulic loads. This information is given in appendix US-B of GESTAR II (Reference 4.1-1) where responses are provided to NRC questions on section 4 of GESTAR II.
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.
4.4.2.9 Uncertainties in Estimates Ultrasonic flow meters have been installed to reduce the uncertainty in the feedwater flow measurements to 0.35%.
The uncertainties in thermal-hydraulic parameters are referenced in GESTAR II (Reference 4.1-1) and NEDC-32601P-A (Reference 4.1-13). The licensing basis power uncertainty for U.S. BWRs is 2.0% as documented in GESTAR II. This power uncertainty is based, in part, on an assumed 1.80% uncertainty in feedwater flow measurement. Feedwater flow uncertainty input to the cycle-specific Safety Limit MCPR calculation is 1.80%.
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 Technical Specifications.
4.
4.3 DESCRIPTION
OF THE THERMAL AND HYDRAULIC DESIGN OF THE REACTOR COOLANT SYSTEM The thermal and hydraulic design of the RCS is described in this section.
4.4.3.1 Plant Configuration Data 4.4.3.1.1 Reactor Coolant System Configuration The RCS 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 5.4-1.
4.4.3.1.2 Reactor Coolant System Thermal-Hydraulic Data The steady-state distribution of temperature, pressure, and flow rate for each flow path in the RCS is shown in Figure 5.1-1.
4.4.3.1.3 Reactor Coolant System Geometric Data CHAPTER 04 4.4-2 REV. 17, SEPTEMBER 2014
LGS UFSAR The volumes of regions and components within the reactor vessel are shown in Figure 5.1-2.
Table 4.4-2 provides the flow path length, height, liquid level, minimum elevations, and minimum flow areas for each major flow path volume within the reactor vessel and recirculation loops of the RCS.
Table 4.4-7 provides the lengths and sizes of all safety injection lines to the RCS.
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 Section 5.4.1. Section 4.4.3.3 gives the operating 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 BWR must operate with certain restrictions because of pump NPSH, overall plant control characteristics, core thermal power limits, etc. The general power-flow map characteristics for the power range of operation is shown in Figure 4.4-1. The LGS specific power flow map is discussed in Chapter 15. The nuclear system equipment, nuclear instrumentation, and the RPS, 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:
- a. Natural circulation line A: The operating state of the reactor moves along this line for the normal control rod withdrawal sequence in the absence of recirculation pump operation.
- b. 20% Recirculation pump constant speed line B: Startup operations of the plant are normally carried out with the recirculation pumps operating at approximately 20%
speed. The operating state for the reactor follows this line for the normal control rod withdrawal sequence.
- c. Rated flow control line: The rated flow control line (100% rod line) passes through 100% power at 100% 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 constant xenon concentration.
- d. 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 28% speed when the feedwater flow drops below a preset value.
4.4.3.3.1.1 Performance Characteristics Other performance characteristics shown on the power-flow operating map are as follows:
CHAPTER 04 4.4-3 REV. 17, SEPTEMBER 2014
- a. Recirculation pump constant speed Line C: This line shows the change in flow associated with power changes while maintaining constant recirculation pump speed.
- b. Constant rod line: These lines show the change in power associated with flow changes while maintaining constant control rod position, e.g., 50% rod line, etc.
4.4.3.3.2 Regions of the Power-Flow Map For normal operating conditions, the nuclear system equipment, nuclear instrumentation, and the RPS, 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:
- a. Region I: This is the transition region between natural circulation operation and 20% pump speed operation. Steady-state conditions cannot exist in this area because the Adjustable Speed Drive (ASD) has been programmed to limit operator manipulation of pump speed to a minimum of 28%. However, the ASD speed output may drop below 28% momentarily upon a power cell bypass or speed controller swap-over to backup controller but will maintain speed above 20% at all times. Normal startup is along the 20% pump speed boundary of this region.
- b. Region II: 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.
- c. Region III: 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 that trip the recirculation pumps to 28% speed whenever feedwater flow is less than a preset value (typically 20% of rated).
4.4.3.4 Temperature-Power Operating Map (PWR)
This section is not applicable to LGS.
4.4.3.5 Load-Following Characteristics The following simple description of BWR 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 following the sequence shown as points 1 to 7 in Figure 4.4-1. The first part of the sequence (1 to 3) is achieved with control rod withdrawal and manual, individual recirculation pump control. Individual pump startup procedures are provided that achieve 28% of full pump speed in each loop. Power, steam flow, and feedwater flow are increased as control rods are manually withdrawn until the feedwater flow has reached approximately 20%. An interlock prevents low power/high recirculation flow combinations that create recirculation pump and jet pump NPSH problems.
Reactor power increases as the operating state moves from point 2 to 3 due to the inherent flow control characteristics of the BWR. Once the feedwater interlock is cleared, the operator can manually increase recirculation flow in each loop until the operating state reaches point 3, the lower limit of the flow control range. At point 3, the operator can switch to simultaneous recirculation pump ontrol. Thermal output can then be increased by either control rod withdrawal CHAPTER 04 4.4-4 REV. 17, SEPTEMBER 2014
LGS UFSAR or recirculation flow increase. For example, the operator can reach 50% power in the ways indicated by points 4 or 5. With a slight rod withdrawal and an increase of recirculation flow to rated flow, point 4 can be achieved. If, however, it is desired to maintain lowest recirculation flow, 50% power can be reached by withdrawing control rods until point 5 is reached. The operating map, Figure 4.4-1, shows the expected flow control range.
The curve labeled "rated flow control line" represents a typical steady-state power-flow characteristic for a fixed rod pattern. It is slightly affected by xenon, core leakage flow assumptions, and reactor vessel pressure variations; however, for this example, these effects have been neglected.
Normal power range operation is along or below the "rated flow control line." If load-following response is desired in either direction, plant operation near 90% power provides the most capability. If maximum load pickup capability is desired, the nuclear system can be operated near point 6, with fast load response available all the way up to point 7, rated power.
The large negative operating reactivity and power coefficients, which are inherent in the BWR, provide important advantages as follows:
- a. Good load-following with well-damped behavior and little undershoot or overshoot in the heat transfer response
- b. Load-following with recirculation flow control
- c. Strong damping of spatial power disturbances The reactor power level can be controlled over approximately 45% of the power level on the rated rod line 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 that 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 moderator will cause a decrease in the reactor power level to a level 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, 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 in tables of Section 5.4 for other portions of the RCS.
CHAPTER 04 4.4-5 REV. 17, SEPTEMBER 2014
LGS UFSAR 4.4.4 EVALUATION Refer to GESTAR II (Reference 4.1-1). The results of the cycle-1 stability analysis are given in Table 4.4-8 and Figures 4.4-2 through 4.4-6.
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 (Initial Test Program).
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 reactor vessel sensors are discussed in Sections 7.6 and 7.7.
4.4.6.1 Deleted CHAPTER 04 4.4-6 REV. 17, SEPTEMBER 2014
LGS UFSAR Table 4.4-1 THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE LGS(2) SSES BWR/4 BWR/4 BWR/4 GENERAL OPERATING CONDITIONS (251-764) (218-560) (251-764)
Reference design thermal 3293 2436 3293 output, MWt Power level for engineered 3435 2558 3435 safety features, MWt Steam flow rate, at 420F 14.159 10.5 13.48(1) final feedwater temperature, millions lb/hr Core coolant flow rate, 100 77.0 100 millions lb/hr Feedwater flow rate, 14.127 10.4 13.44 millions lb/hr System pressure, nominal in 1020 1020 1020 steam dome, psia System pressure, nominal 1035 1035 1035 core design, psia Coolant saturation temperature 548 549 549 at core design pressure, F Average power density, kW/liter 48.7 49.2 48.7 Maximum linear heat generation 13.4 13.4 13.4 rate, kW/ft Average linear heat generation 5.3 5.4 5.3 rate, kW/ft Core total heat transfer area, 74,871 54,879 74,871 ft2 Maximum heat flux, Btu/hr-ft2 361,600 361,600 361,600 Average heat flux, Btu/hr-ft2 144,100 145,060 143,700 Design operating MCPR 1.22 1.22 1.25 CHAPTER 04 4.4-7 REV. 16, SEPTEMBER 2012
LGS UFSAR Table 4.4-1 (Cont'd)
(2)
LGS SSES BWR/4 BWR/4 BWR/4 GENERAL OPERATING CONDITIONS (251-764) (218-560) (251-764)
Core inlet enthalpy at 420F (1)
FFWT, Btu/lb 526.1 526.9 521.8 Core inlet temperature, at 420F FFWT, F 532 532 528(1)
Core maximum exit voids within assemblies, % 77.1 76.0 76.6 Core average void fraction, active coolant 0.418 0.422 0.410 Maximum fuel temperature, F 3435 3435 3435 Active coolant flow area per assembly, in2 15.824 15.824 15.824 Core average inlet velocity, ft/sec 6.44 6.65 6.31 Maximum inlet velocity, ft/sec 7.78 7.1 7.5 Total core pressure drop, psi 21.95 23.89 21.52 Core support plate pressure drop, psi 17.52 19.46 17.08 Average orifice pressure drop Central region, psi 5.10 8.0 5.09 Peripheral region, psi 14.32 16.52 14.03 Maximum channel pressure loading, psi 11.63 12.86 11.24 Average-power assembly channel pressure loading (bottom), psi 10.39 11.47 9.96 Shroud support ring and lower shroud pressure loading, psi 23.57 26.46 23.13 Upper shroud pressure loading, psi 6.05 7.0 6.05
__________________
(1)
At 383F final feedwater temperature (2)
Based on original licensed operating conditions.
CHAPTER 04 4.4-8 REV. 16, SEPTEMBER 2012
LGS UFSAR Table 4.4-2 REACTOR COOLANT SYSTEM GEOMETRIC DATA HEIGHT ELEVATION FLOW AND OF BOTTOM MINIMUM PATH LIQUID OF EACH FLOW LENGTH LEVEL VOLUME(1) AREAS (in)
(in) (in) (ft2)
Lower plenum 216.5 216.5 -161.5 92.5 216.5 Core 152.0 163.0 163.0 55.0 (includes 163.0 bypass)
Upper plenum and 185.0 185.0 217.5 45.0 separators 185.0 Dome (above normal 299.5 299.5 402.5 352.0 water level) 0 Downcomer area 311.0 311.0 -30.0 118.0 311.0 Recirculation loops 97.0 77.5 in2 492.0 and jet pumps (one -472.5 (one 492.0 loop) loop)
_________________
(1)
The reference point is the recirculation nozzle outlet centerline.
CHAPTER 04 4.4-9 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.4-3 AXIAL VOID FRACTION DISTRIBUTION MAXIMUM CHANNEL CORE AVERAGE (END OF NODE (AVERAGE NODE NODE VALUE) VALUE)
(Bottom of core) 1 0. 0.
2 0.006 0.000 3 0.073 0.007 4 0.186 0.041 5 0.293 0.105 6 0.380 0.183 7 0.453 0.261 8 0.510 0.331 9 0.556 0.389 10 0.593 0.435 11 0.622 0.473 12 0.647 0.504 13 0.667 0.529 14 0.683 0.550 15 0.693 0.567 16 0.711 0.583 17 0.722 0.597 18 0.734 0.610 19 0.744 0.622 20 0.753 0.633 21 0.761 0.643 22 0.767 0.651 23 0.770 0.656 (Top of core) 24 0.771 0.659
_________________
Core average value = .418 Maximum exit value = .772 Active fuel length = 150 inches
_________________________________________________________________
CHAPTER 04 4.4-10 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.4-4 AXIAL FLOW QUALITY DISTRIBUTION MAXIMUM CHANNEL CORE AVERAGE (END OF NODE (AVERAGE NODE NODE VALUE) VALUE)
(Bottom of core) 1 0. 0.
2 0. 0.
3 0.003 0.000 4 0.012 0.001 5 0.027 0.005 6 0.045 0.011 7 0.065 0.019 8 0.086 0.030 9 0.107 0.042 10 0.127 0.054 11 0.146 0.066 12 0.164 0.076 13 0.180 0.096 14 0.195 0.095 15 0.209 0.104 16 0.223 0.112 17 0.236 0.120 18 0.249 0.127 19 0.263 0.135 20 0.275 0.143 21 0.286 0.150 22 0.294 0.155 23 0.300 0.159 (Top of core) 24 0.301 0.161
__________________
Core average value = 0.077 Maximum exit value = 0.301 Active fuel length = 150 inches CHAPTER 04 4.4-11 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.4-5 AXIAL POWER DISTRIBUTION USED TO GENERATE VOID AND QUALITY DISTRIBUTIONS NODE AXIAL POWER FACTOR (Bottom of core) 1 0.38 2 0.69 3 0.93 4 1.10 5 1.21 6 1.30 7 1.47 8 1.51 9 1.49 10 1.44 11 1.36 12 1.28 13 1.16 14 1.06 15 1.01 16 0.97 17 0.94 18 0.97 19 0.96 20 0.91 21 0.77 22 0.59 23 0.38 (Top of core) 24 0.12 CHAPTER 04 4.4-12 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.4-6 The information on this page has been deleted.
This page is intentionally left blank.
CHAPTER 04 4.4-13 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.4-7 SAFETY INJECTION LINE LENGTHS PRELIMINARY DESIGN DATA - NOT AS-BUILT CONDITIONS LINE O.D. LINE INJECTION LINE (in) LENGTH High Pressure Coolant Injection(2)
Pump discharge to valve 14/12 337'-8" 108(1)
Inside containment to RPV 12 42'-3" Low Pressure Coolant Injection(3)
Loop A Pump A discharge to valve 18 331'-9" F017A(1)
Inside containment to RPV 12 49' Loop B Pump discharge to valve 18 318'-4" F017B(1)
Inside containment to RPV 12 49'-1" Loop C Pump discharge to valve 18 270'-8" F017C(1)
Inside containment to RPV 12 49'-7" Loop D Pump discharge to valve 18 256'-3" F017D(1)
Inside containment to RPV 12 64'-5" CHAPTER 04 4.4-14 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.4-7 (Cont'd)
LINE O.D. LINE INJECTION LINE (in) LENGTH Core Spray(4)
Loop A Pump A discharge to common 12 57'-5" discharge Pump B discharge to common 12 23'-4" discharge Common discharge to valve 14 199'-8" F005(1)
Inside containment to RPV 12 43'-3" Loop B Pump B discharge to common 12 59'-11" discharge Pump D discharge to common 12 23'-2" discharge Common discharge to valve 14 150'-8" 108(1)
Inside containment to RPV 12 43'-3"
__________________
(1)
Valve located as close as possible to outside of containment wall.
(2)
See drawings M-55 and M-56.
(3)
See drawing M-51.
(4)
See drawing M-52.
CHAPTER 04 4.4-15 REV. 13, SEPTEMBER 2006
LGS UFSAR Table 4.4-8 STABILITY ANALYSIS RESULTS (Cycle-1 Most Limiting Conditions)
Rod Line Analyzed Natural circulation 51.5% rated core flow Rod pattern 105.0% rated thermal power Decay Ratio Total system stability, x2/xo = See Figure 4.4-2 Reactor core stability, x2/xo = 0.78 (0.32 Hz resonant frequency)
Channel hydrodynamic = 0.59 (0.36 Hz performance, x2/xo resonant frequency)
CHAPTER 04 4.4-16 REV. 13, SEPTEMBER 2006
LGS UFSAR 4.5 REACTOR MATERIALS 4.5.1 CONTROL ROD SYSTEM STRUCTURAL MATERIALS 4.5.1.1 Material Specifications 4.5.1.1.1 Material List The following material listing applies to the CRDM supplied for this application. The position indicator and minor nonstructural items are omitted.
- a. Cylinder, tube, and flange assembly:
Flange ASME SA182 Grade F304 Plugs ASME SA182 Grade F304 Cylinder ASTM A269 Grade TP 304 Outer Tube ASTM A269 Grade TP 304 Tube ASTM A269 Grade TP 304 Spacer ASTM A269 Grade TP 304 or ASTM A511 Grade MT 304
- b. Piston tube assembly:
Piston Tube ASTM A269 Grade TP 304 Stud ASTM A276 Type 304 Head ASME SA182 Grade F304 Ind. Tube ASME SA312 Type 316 Cap ASME SA182 Grade F304
- c. Drive line assembly:
Coupling Spud Inconel X-750 Index Tube ASTM A269 Grade TP 304 Piston Head ARMCO 17-4 PH Coupling ASME SA312 Grade TP 304 or ASTM A511 Grade MT 304 Magnet Housing ASME SA312 Grade TP 304 or ASTM A511 Grade MT 304
- d. Collet assembly:
CHAPTER 04 4.5-1 REV. 15, SEPTEMBER 2010
LGS UFSAR Collet piston ASTM A269 Grade TP 304 or ASME SA312 Grade TP 304 Finger Inconel X-750 Retainer ASTM A269 Grade TP 304 or ASTM A511 Grade MT 304 Guide cap ASTM A269 Grade TP 304 Collet piston Haynes 25 Alloy Rings
- e. Miscellaneous parts:
Stop piston ASTM A276 Type 304 Connector ASTM A276 Type 304 O-Ring spacer ASME SA240 Type 304 Nut ASME SA193 Grade B8 Barrel ASTM A269 Grade TP 304 or ASME SA312 Grade TP 304 or ASME SA240 Type 304 O-Ring seals Ethylene Propylene Collet spring Inconel X-750 Ring flange ASME SA182 Grade F304 Ball check valve Haynes, Deloro, or Equivalent Stellite Cobalt-Base Alloy The materials listed under an ASTM specification number are all in the annealed condition (with the exception of the outer tube in the cylinder, tube, and flange assembly), and their properties are readily available. The outer tube is approximately 1/8 hard and has a tensile strength of 90,000/
125,000 psi, yield strength of 50,000/85,000 psi, and minimum elongation of 25%.
The coupling spud, collet fingers, and collet spring are fabricated from Inconel X-750 in the annealed or equalized condition and are aged 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> at 1300 F to produce a tensile strength of 165,000 psi minimum, yield of 105,000 psi minimum, and elongation of 20% minimum. The piston head is ARMCO 17-4 PH in condition H-1100 (aged 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 1100 F), with a tensile of 140,000 psi minimum, yield of 115,000 psi minimum, and elongation of 15% minimum.
These are widely used materials whose properties are well-known. The parts are readily accessible for inspection and replacement if necessary.
CHAPTER 04 4.5-2 REV. 15, SEPTEMBER 2010
LGS UFSAR All materials have been successfully used for the past 10 to 15 years in similar drive mechanisms.
4.5.1.1.2 Special Materials No cold-worked austenitic stainless steels with a yield strength greater than 90,000 psi are employed in the CRD system. Hardenable martensitic stainless steels are not used. ARMCO 17-4 PH (precipitation-hardened stainless steel) is used for the piston head. This material is aged to the H-1100 condition to produce resistance to stress-corrosion cracking in the BWR environments. ARMCO 17-4 PH (H-1100) has been successfully used for the past 10 to 15 years in BWR drive mechanisms.
4.5.1.2 Austenitic Stainless Steel Components: Processes, Inspections, and Tests All austenitic stainless steel used in the CRD system is solution annealed material with one exception, the outer tube in the cylinder, tube, and flange assembly (Section 4.5.1.1). Proper solution annealing is verified by testing in accordance with ASTM A262, "Recommended Practices for Detecting Susceptibility to Intergranular Attack in Stainless Steels".
Two special processes are employed that subject selected components to temperatures in the sensitization range.
- a. The cylinder (cylinder, tube, and flange assembly) and the retainer (collet assembly) are hard-surfaced with Colmonoy-6.
- b. The following components are nitrided to provide a wear-resistant surface:
- 1. Tube (cylinder, tube, and flange assembly)
- 2. Piston tube (piston tube assembly)
- 3. Index tube (drive line assembly)
- 4. Collet piston and guide cap (collet assembly)
Colmonoy hard-surfacing is applied by the flame spray process. Parts are preheated to 550 F-800 F and then sprayed with Colmonoy. The sprayed coating is fused at about 2000 F using an oxyacetylene torch followed by air cooling.
Nitriding is accomplished using a proprietary process called New Malcomizing. Components are exposed to a temperature of about 1080 F for about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> during the nitriding cycle.
Colmonoy hard-surfaced components have performed successfully for the past 10 to 15 years in drive mechanisms. Nitrided components have been used in BWR CRDs since 1967. It is normal practice to remove some CRDs at each refueling outage. At this time, both the Colmonoy hard-surfaced parts and nitrided surfaces are accessible for visual examination. In addition, dye penetrant examinations have been performed on nitrided surfaces of the longest service drives.
This inspection program is adequate to detect any incipient defects before they can become serious enough to cause operating problems.
CHAPTER 04 4.5-3 REV. 15, SEPTEMBER 2010
LGS UFSAR Pickling is prohibited on stainless steel components that have been exposed to temperatures in the sensitization range during nitriding or hard-surfacing operations. Pickling of welded stainless steel is also prohibited unless the welded material has been solution annealed subsequent to all welding operations.
Welding is performed in accordance with ASME Section IX. Heat input for stainless steel welds is restricted to a maximum of 50,000 Joules/in and interpass temperature to 350oF. Heating above 800 F (except for welding) is prohibited unless the welds are subsequently solution annealed.
These controls are employed to avoid severe sensitization and comply with the intent of Regulatory Guide 1.44.
Control of Delta Ferrite Content All type 308 and 308L weld metal is purchased to a specification that requires a minimum of 5%
delta ferrite. This amount of ferrite is adequate to prevent any microfissuring (hot-cracking) in austenitic stainless steel welds.
An extensive test program performed by GE, with the concurrence of the regulatory staff, has demonstrated that controlling weld filler metal ferrite at 5% minimum produces production welds that meet the requirements of Regulatory Guide 1.31, "Control of Stainless Steel Welding." A total of approximately 400 production welds in five BWR plants was measured, and all welds met the requirements of BTP MTEB 5-1, "Interim Regulatory Position on Regulatory Guide 1.31".
4.5.1.3 Other Materials These are discussed in Section 4.5.1.1.2.
4.5.1.4 Cleaning and Cleanliness Control 4.5.1.4.1 Protection of Materials During Fabrication, Shipping, and Storage All the CRD parts listed above (Section 4.5.1.1) are fabricated under a process specification that limits contaminants in cutting, grinding, and tapping coolants and lubricants. It also restricts all other processing materials (marking inks, tape, etc) to those that are completely removable by the applied cleaning process. All contaminants are then required to be removed by the appropriate cleaning process before any of the following:
o
- a. Any processing that increases part temperature above 200 F
- b. Assembly that results in a decrease of accessibility for cleaning
- c. Release of parts for shipment The specification for packaging and shipping the CRD provides the following:
The drive is rinsed in hot deionized water and dried in preparation for shipment. The ends of the drive are then covered with a vapor-tight barrier with desiccant. Packaging is designed to protect the drive and prevent damage to the vapor barrier. The planned storage period considered in the design of the container and packaging is two years. This packaging has been qualified and in use for a number of years. Periodic audits have indicated satisfactory protection.
CHAPTER 04 4.5-4 REV. 15, SEPTEMBER 2010
LGS UFSAR The degree of surface cleanliness obtained by these procedures meets the requirements of Regulatory Guide 1.37 (Rev 0).
Site or warehouse storage specifications require inside heated storage comparable to level B of ANSI N45.2.2. After the second year, a yearly inspection 2% of the units themselves, or a semi-annual examination of the humidity indicators, where provided, associated with 10% of the units is required to verify that the units are dry and in satisfactory condition. Position indicator probes are not subject to this inspection.
4.5.2 REACTOR INTERNAL MATERIALS 4.5.2.1 Material Specifications Materials used for the core support structure are as follows:
- a. Shroud support: SB168
- b. Shroud, core plate, and top guide: ASME SA240, SA193, SA194, SA182, SA479, SA312; ASTM A276, A249, A213, (Types 304L or 304)
- c. Peripheral fuel supports: ASTM A269 Grade TP 304, ASTM A312 Grade TP 304, ASTM A511, S1 and S2 Grade MT 304
- f. Control rod guide tube: ASTM A358 Class 2; or ASTM A376 Grade TP 304; or ASTM A312 Grade TP 304; ASTM A351 Grade CF8; ASTM A358 Class 2; or ASTM A213 TP 304; or ASTM A249 TP 304
- g. Orificed fuel support: ASME A351 Grade CF8 Materials employed in other reactor internal structures are as follows:
- a. Steam separator and steam dryer:
All materials are Type 304 stainless steel:
Plate, Sheet, ASTM A240 Type 304 and Strip Forgings ASTM A182 Grade F304 ASTM A182 Grade FXM-19 (Slip-fit joint clamps)
Bars ASTM A276 Type 304 Pipe ASTM A312 Grade TP 304 CHAPTER 04 4.5-5 REV. 15, SEPTEMBER 2010
LGS UFSAR Tube ASTM A269 Grade TP 304 Castings ASTM A351 Grade CF8
- b. Jet pump assemblies:
The components in the jet pump assemblies are a riser, inlet-mixer, diffuser, adaptor, and riser brace. Materials used for these components are to the following specifications:
Castings ASTM A351 Grade CF8 Bars ASTM A276 Type 304 Bolts ASTM A193 Grade B8 or B8M Nut ASTM A194 Grade B8 Sheet and Plate ASTM A240 Type 304 and 304L Tubing ASTM A269 Grade TP 304 Pipe ASTM A358 Type 304 and ASTM A312 Grade TP 304 Weld Coupling ASTM A403 Grade WP304 Forgings ASTM A182 Grade F304 ASTM A182 Grade FXM-19 (slip-fit joint clamps)
Inconel Forgings ASTM B166 Materials in the jet pump assemblies that are not Type 304 stainless steel are:
- 1. The inlet-mixer adaptor casting, the wedge casting, bracket casting adjusting screw, and the diffuser collar casting are Type 304 hard-surfaced with Stellite-6 for slip-fit joints. The auxiliary spring wedge assembly and the improved inlet-mixer wedge assembly are made of Iconel X-750, ASTM B-637. The wedge is given a high temperature anneal followed by an aging heat treatment.
- 2. The adaptor is a bimetallic component made by welding a Type 304 forged ring to a forged Inconel-600 ring, made to Specification ASTM B166.
- 3. The inlet-mixer contains a pin, insert, and beam made of Inconel X-750 to Specification ASTM A637 Grade 688. The beam is given a high temperature anneal after forging.
- 4. The jet pump beam bolt is SS316L conforming to ASME SA479.
CHAPTER 04 4.5-6 REV. 15, SEPTEMBER 2010
- 5. The inlet-mixer to diffuser slip-fit joint clamp, if installed, contains small parts made of precipitation-hardened, heat treated Nickel Alloy to Specification ASTM B-637 (UNS N07750 Type 3).
- c. Core spray spargers and core spray lines:
Materials used for these components are as follows: ASME SA240, SA182, SA312, SA351, SA479, SA193, SA194, or ASTM A276, A249, A213, A269, A403, Types 304, 304L, 316, 316L, CF3 or CF8.
- d. Feedwater Sparger:
Materials used in this component are: 304, 304L, and 316L.
All core support structures are fabricated from ASME and ASTM specified materials and are designed using ASME Section III as a guide. The other reactor internals are noncoded and are fabricated from ASTM or ASME specification materials. Material requirements in the ASTM specifications are identical to requirements in corresponding ASME material specifications.
4.5.2.2 Controls on Welding The requirements of the ASME Section IX are followed in fabrication of core support structures and other internals.
4.5.2.3 Nondestructive Examination of Wrought Seamless Tubular Products Core support structures and other internals for LGS were noncoded because they were fabricated before the issuance of subsection NG of ASME Section III. The only coded tubular parts are the CRD housings and incore housings, which are welded to the reactor vessel. These coded tubular parts were examined by radiographic and/or ultrasonic methods in accordance with the requirements of paragraph NB 2550 of the applicable ASME Section III issue that was in effect at that time.
Other noncoded tubular products were supplied in accordance with the applicable ASTM/ASME material specifications. These specifications require a hydrostatic test on each length of tubing.
4.5.2.4 Regulatory Guide Conformance for Fabrication and Processing of Austenitic Stainless Steel
- a. Regulatory Guide 1.31, "Control of Stainless Steel Welding" Cold-worked stainless steels are not used in the reactor internals. All austenitic stainless steel weld filler materials were supplied with a minimum of 5% delta ferrite. This amount of ferrite is considered adequate to prevent microfissuring in austenitic stainless steel welds.
An extensive test program performed by GE, with the concurrence of the regulatory staff, has demonstrated that controlling the weld filler metal ferrite at 5%
minimum produces production welds that meet the requirements of Regulatory Guide 1.31. A total of approximately 400 production welds in five BWR plants was CHAPTER 04 4.5-7 REV. 15, SEPTEMBER 2010
LGS UFSAR measured, and all welds met the requirements of BTP MTEB 5-1, "Interim Regulatory Position on Regulatory Guide 1.31".
- b. Regulatory Guide 1.37, "Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants" Exposure to contaminants is avoided by carefully controlling all cleaning and processing materials that contact stainless steel during manufacture and construction. Any inadvertent surface contamination is removed to avoid potential detrimental effects.
Special care is exercised to ensure removal of surface contaminants before any heating operation. Water quality for rinsing, flushing, and testing is controlled and monitored. The water quality meets or exceeds the requirements of ANSI N45.2.1.
The degree of cleanliness obtained by these procedures meets the requirements of Regulatory Guide 1.37.
- c. Regulatory Guide 1.44, "Control of the Use of Sensitized Steel" All wrought austenitic stainless steel was purchased in the solution heat treated condition. Heating above 800 F was prohibited (except for welding) unless the stainless steel was subsequently solution annealed. For 304 steel with carbon content in excess of 0.035% carbon, purchase specifications restricted the maximum weld heat input to 110,000 Joules/in and the weld interpass temperature to 350 F maximum. Welding was performed in accordance with ASME Section IX. These controls were employed to avoid severe sensitization and comply with the intent of Regulatory Guide 1.44.
- d. Regulatory Guide 1.71, "Welder Qualification for Areas of Limited Accessibility" Welder performance qualification tests are made in accordance with the requirements of ASME Section IX. Performance qualifications require testing of welders under simulated access conditions when the access to nonvolumetrically examined production welds is less than 12 inches in any direction and allow welding from one access only. Requalification is required when significantly more restricted access conditions occur or when any essential welding variable is changed.
4.5.2.5 Other Materials Hardenable martensitic stainless steels and precipitation hardening stainless steels are not used in the reactor internals.
Materials, other than Type 300 stainless steel, employed in vessel internals are: SB166, 167, and 168 Nickel-Chrome-Iron (Inconel 600); SA637 Grade 688 Inconel X-750.
Inconel 600 tubing, plate, and sheet are used in the annealed condition. Bar may be in the annealed or cold-drawn condition.
CHAPTER 04 4.5-8 REV. 15, SEPTEMBER 2010
LGS UFSAR Inconel X-750 components are fabricated in the annealed or equalized condition and aged 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> at 1300 F.
Stellite 6 hard-surfacing is applied to austenitic stainless steel castings using the gas tungsten arc welding or plasma arc surfacing processes.
All materials have been successfully used for the past 10 to 15 years in BWR applications.
4.5.3 CONTROL ROD DRIVE HOUSING SUPPORTS All CRD housing support subassemblies are fabricated of ASTM A36 structural steel, except for the following items:
ITEM MATERIAL
Grid ASTM A441 Disc springs Schnorr, Type BS-125-71-8 Hex bolts and nuts STM A307 6"x4"x3/8" tubes STM A500 Grade B For further CRD housing support information, see Section 4.6.1.2.
CHAPTER 04 4.5-9 REV. 15, SEPTEMBER 2010
LGS UFSAR 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS The reactivity control systems consist of control rods and the CRD system, supplementary reactivity control for the initial core (Section 4.1.3), and the SLCS (Section 9.3).
4.6.1 CONTROL ROD DRIVE SYSTEM 4.6.1.1 Design 4.6.1.1.1 Safety Design Bases The CRD mechanical system meets the following safety design bases:
- a. The design provides for a sufficiently rapid control rod insertion so that no fuel damage results from any abnormal operating transient.
- b. The design includes positioning devices, each of which individually supports and positions a control rod.
- c. Each positioning device:
- 1. Prevents its control rod from initiating withdrawal as a result of a single malfunction,
- 2. Is individually operated so that a failure in one positioning device does not affect the operation of any other positioning device, and
- 3. Is individually energized when rapid control rod insertion (scram) is signaled, so that failure of power sources external to the positioning device does not prevent other positioning devices' control rods from being inserted.
4.6.1.1.2 Power Generation Design Basis The CRD design provides for positioning the control rods to control power generation in the core.
4.6.1.2 Description The CRD system controls gross changes in core reactivity by incrementally positioning neutron-absorbing control rods within the reactor core, in response to manual control signals. It is also designed to quickly shut down the reactor (scram) in emergency situations by rapidly inserting withdrawn control rods into the core in response to a manual or automatic signal. The CRD system consists of locking piston CRDMs, and the CRD hydraulic system (including power supply and regulation, HCUs, interconnecting piping, instrumentation, and electrical controls).
4.6.1.2.1 Control Rod Drive Mechanisms CHAPTER 04 4.6-1 REV. 18, SEPTEMBER 2016
LGS UFSAR The CRDM (drive) used for positioning the control rod in the reactor core is a double-acting, mechanically latched, hydraulic cylinder using demineralized water as its operating fluid (Figures 4.6-1, 4.6-2, 4.6-3, and 4.6-4). The individual drives are mounted on the bottom head of the RPV.
The drives do not interfere with refueling and are operative even when the head is removed from the reactor vessel.
The drives are also readily accessible for inspection and servicing. The bottom location makes maximum utilization of the water in the reactor as a neutron shield, and gives the least possible neutron exposure to the drive components. The use of water from the condensate treatment system, and/or the CST as the operating fluid, eliminates the need for a special hydraulic fluid.
The drives utilize simple piston seals whose leakage provides cooling for the CRDMs. This leakage does not contaminate the reactor water.
The drives are capable of inserting or withdrawing a control rod at a slow, controlled rate, in addition to providing rapid insertion when required. A mechanism, on the drive, locks the control rod at 6 inch increments of stroke over the length of the core.
A coupling spud at the top end of the drive index tube (piston rod) engages and locks into a mating socket at the base of the control rod. The weight of the control rod is sufficient to engage and lock this coupling. Once locked, the drive and rod form an integral unit that must be manually unlocked by specific procedures before the components can be separated.
The drive holds its control rod in distinct latch positions until the hydraulic system actuates movement to a new position. Withdrawal of each rod is limited by the seating of the rod in its guide tube. Withdrawal of the drive beyond this position to the overtravel limit can be accomplished only if the rod and drive are uncoupled. Withdrawal to the overtravel limit is annunciated by an alarm.
Individual rod position indicators, grouped in one control panel display, correspond to relative rod locations in the core. A separate, smaller display, located just below the large display presents the positions of the control rod selected for movement and of the other rods in the affected rod group.
For display purposes, the control rods are considered in groups of four adjacent rods centered around a common core volume. Each group is monitored by four LPRM strings (Section 7.6.1.4).
Rod groups at the periphery of the core may have less than four rods. The small display shows the positions, in digital form, of the rods in the group to which the selected rod belongs.
Backlighting indicates which of the four rods is the one selected for movement.
4.6.1.2.2 Drive Components Figure 4.6-2 illustrates the operating principle of a CRD. Figures 4.6-3 and 4.6-4 illustrate the drive in more detail. The main components of the drive and their functions are described below.
4.6.1.2.2.1 Drive Piston The drive piston is mounted at the lower end of the index tube. The function of the index tube is similar to that of a piston rod in a conventional hydraulic cylinder. The drive piston and index tube make up the main moving assembly in the drive. The drive piston operates between positive end stops, with a hydraulic cushion provided at the upper end only. The piston has both inside and outside seal-rings, and operates in an annular space between an inner cylinder (fixed piston tube)
CHAPTER 04 4.6-2 REV. 18, SEPTEMBER 2016
LGS UFSAR and an outer cylinder (drive cylinder). Because the type of inner seal used is effective in only one direction, the lower sets of seal-rings are mounted with one set sealing in each direction.
A pair of nonmetallic bushings prevent metal-to-metal contact between the piston assembly and the inner cylinder surface. The outer piston rings are segmented, step-cut seals with expander springs holding the segments against the cylinder wall. A pair of split bushings on the outside of the piston prevents piston contact with the cylinder wall. The effective piston area for downtravel, or withdrawal, is approximately 1.2 in2 versus approximately 4.1 in2 for uptravel, or insertion. This difference in driving area tends to balance the control rod weight and ensures a higher force for insertion than for withdrawal.
4.6.1.2.2.2 Index Tube The index tube is a long hollow shaft made of nitrided stainless steel. Circumferential locking grooves, spaced every 6 inches along the outer surface, transmit the weight of the control rod to the collet assembly.
4.6.1.2.2.3 Collet Assembly The collet assembly serves as the index tube locking mechanism. It is located in the upper part of the drive unit. This assembly prevents the index tube from accidentally moving downward. The assembly consists of the collet fingers, a return spring, a guide cap, a collet housing (part of the cylinder, tube, and flange), and the collet piston.
Locking is accomplished by fingers mounted on the collet piston at the top of the drive cylinder. In the locked or latched position, the fingers engage a locking groove in the index tube.
The collet piston is normally held in the latched position by a force of approximately 150 pounds supplied by a spring. Metal piston rings are used to seal the collet piston from reactor vessel pressure. The collet assembly does not unlatch until the collet fingers are unloaded by a short, automatically sequenced drive-in signal. A pressure, approximately 180 psi above reactor vessel pressure, must then be applied to the collet piston to overcome spring force, slide the collet up against the conical surface in the guide cap, and spread the fingers out so they do not engage a locking groove.
A guide cap is fixed in the upper end of the drive assembly. This member provides the unlocking cam surface for the collet fingers and serves as the upper bushing for the index tube.
If reactor water is used during a scram to supplement accumulator pressure, it is drawn through a filter on the guide cap.
4.6.1.2.2.4 Piston Tube The piston tube is an inner cylinder, or column, extending upward inside the drive piston and index tube. The piston tube is fixed to the bottom flange of the drive and remains stationary. Water is brought to the upper side of the drive piston through this tube. A series of orifices at the top of the tube provides progressive water shutoff to cushion the drive piston at the end of its scram stroke.
4.6.1.2.2.5 Stop Piston CHAPTER 04 4.6-3 REV. 18, SEPTEMBER 2016
LGS UFSAR A stationary piston, called the stop piston, is mounted on the upper end of the piston tube. This piston provides the seal between reactor vessel pressure and the space above the drive piston. It also functions as a positive end stop at the upper limit of control rod travel. A stack of spring washers just below the stop piston helps absorb the final mechanical shock at the end of control rod travel. The piston rings are similar to the drive piston outer rings. A bleed-off passage to the center of the piston tube is located between the two pairs of rings. This arrangement allows seal leakage from the reactor vessel (during a scram) to be bled directly to the discharge line. The lower pair of seals is used only during the cushioning of the drive piston at the upper end of the stroke.
The center tube of the drive mechanism forms a well to contain the position indicator probe. The probe is an aluminum extrusion attached to a cast aluminum housing. Mounted on the extrusion are hermetically sealed, magnetically operated reed switches. The entire probe assembly is protected by a thin-walled stainless steel tube. The switches are actuated by a ring magnet located at the bottom of the drive piston.
The drive piston, piston tube, and indicator tube are all of nonmagnetic stainless steel, allowing the individual switches to be operated by the magnet as the piston passes. One switch is located at each position corresponding to an index tube groove, thus allowing indication at each latching point. An additional switch is located at each midpoint between latching points to indicate the intermediate positions during drive motion. Thus, indication is provided for each 3 inches of travel.
Duplicate switches are provided for the full-in and full-out positions. One additional switch (an overtravel switch), is located at a position below the normal full-out position. Because the limit of downtravel is normally provided by the control rod itself as it reaches the back-seat position, the drive can pass this position and actuate the overtravel switch only if it is uncoupled from its control rod. A convenient means is thus provided to verify that the drive and control rod are coupled after installation of a drive and at any time during plant operation.
4.6.1.2.2.6 Flange and Cylinder Assembly A flange and cylinder assembly is made up of a heavy flange welded to the drive cylinder. A sealing surface on the upper face of this flange forms the seal to the drive housing flange. The seals contain reactor pressure and the two hydraulic control pressures. Teflon-coated, stainless steel rings are used for these seals. The drive flange contains the integral ball, or two-way check (ball-shuttle) valve. This valve directs either the reactor vessel pressure or the driving pressure, whichever is higher, to the underside of the drive piston. Reactor vessel pressure is admitted to this valve from the annular space between the drive and drive housing through passages in the flange.
Water used to operate the collet piston passes between the inner and outer tubes of the drive cylinder. The inside of the cylinder tube is honed to provide the surface required for the drive piston seals.
Both the cylinder tube and outer tube are welded to the drive flange. The upper ends of these tubes have a sliding fit to allow for differential expansion.
The upper end of the index tube is threaded to receive a coupling spud. The coupling (Figure 4.6-1) accommodates a small amount of angular misalignment between the drive and the control rod. Six spring fingers allow the coupling spud to enter the mating socket on the control rod. A lock plug then enters the spud and prevents uncoupling.
CHAPTER 04 4.6-4 REV. 18, SEPTEMBER 2016
LGS UFSAR 4.6.1.2.2.7 Lock Plug Two means of uncoupling are provided. With the reactor vessel head removed, the lock plug can be raised against the spring force of approximately 50 pounds by a rod extending up through the center of the control rod to an unlocking handle located above the control rod velocity limiter. The control rod, with the lock plug raised, is then lifted from the drive.
If it is desired to uncouple a drive without removing the reactor vessel head for access, the lock plug can also be pushed up from below. In this case, the piston tube assembly is pushed up against the uncoupling rod assembly, which raises the lock plug and allows the coupling spud to disengage the socket as the drive piston and index tube are driven down.
The control rod is heavy enough to force the spud fingers to enter the socket and push the lock plug up, allowing the spud to enter the socket completely, and the lock plug to snap back into place. Therefore, the drive is coupled to the control rod using only the weight of the control rod.
4.6.1.2.3 Materials of Construction Factors that determine the choice of construction materials are discussed in the following sections.
4.6.1.2.3.1 Index Tube The index tube must withstand the locking and unlocking action of the collet fingers. A compatible bearing combination must be provided that is able to withstand moderate misalignment forces.
The reactor environment limits the choice of materials suitable for corrosion resistance. The column and tensile loads can be satisfied by an annealed, 300-series stainless steel. The wear and bearing requirements are provided by Malcomizing the completed tube. To obtain suitable corrosion resistance, a carefully controlled process of surface preparation is employed.
4.6.1.2.3.2 Coupling Spud The coupling spud is made of Inconel X-750 that is aged for maximum physical strength and for the required corrosion resistance. Because misalignment tends to cause chafing in the semispherical contact area, this component is protected by a thin chromium plating (electrolyzed).
This plating also prevents galling of the threads attaching the coupling spud to the index tube.
4.6.1.2.3.3 Collet Fingers Inconel X-750 is used for the collet fingers, which must function as leaf springs when cammed open to the unlocked position. Colmonoy-6 hard-facing provides a long-wearing surface, adequate for design life, for the area contacting the index tube and unlocking the cam surface of the guide cap.
4.6.1.2.3.4 Seals and Bushings Graphitar-3030 is selected for seals and bushings on the drive piston and stop piston. The material is inert and has a low friction coefficient when water-lubricated. Because some loss of Graphitar-3030 strength is experienced at higher temperatures, the drive is supplied with cooling water to hold temperatures below 250 F. The Graphitar-3030 is relatively soft, which is CHAPTER 04 4.6-5 REV. 18, SEPTEMBER 2016
LGS UFSAR advantageous when an occasional particle of foreign matter reaches a seal. The resulting scratches in the seal reduce sealing efficiency until worn smooth, but the drive design can tolerate considerable water leakage past the seals into the reactor vessel.
4.6.1.2.3.5 Summary All drive components exposed to reactor vessel water are made of austenitic stainless steel except the following:
- a. Seals and bushings on the drive piston and stop piston are Graphitar-3030.
- b. All springs and members requiring spring action (collet fingers, coupling spud, and spring washers) are made of Inconel X-750.
- c. The ball check valve is a Haynes Stellite cobalt-base alloy.
- d. Elastomeric O-ring seals are ethylene propylene.
- e. Metal piston rings are Haynes-25 alloy.
- f. Certain wear surfaces are hard-faced with Colmonoy-6.
- g. Nitriding by a proprietary new Malcomizing process and chromium plating are used in certain areas where resistance to abrasion is necessary.
- h. The drive piston head is made of ARMCO 17-4PH.
Pressure-containing portions of the drives are designed and fabricated in accordance with requirements of ASME Section III.
4.6.1.2.4 Control Rod Drive Hydraulic System The CRD hydraulic system supplies and controls the pressure and flow to and from the drives through HCUs. The water discharged from the drives during a scram flows through the HCUs to the scram discharge volume. The water discharged from a drive during a normal control rod positioning operation flows through the HCU, the exhaust header, and is returned to the reactor vessel via the HCUs of nonmoving drives. There are as many HCUs as the number of CRDs. The design criterion for the LGS CRD hydraulic system requires that no single failure of a component, structure, or service function shall prevent a reactor shutdown. Each CRD acts somewhat independently, and reactor shutdown can be achieved assuming failure of a single rod to insert.
In addition, the CRD scram hydraulic system is designed to be fail-safe (i.e., to scram on loss of electrical or pneumatic power.)
4.6.1.2.4.1 Hydraulic Requirements The CRD hydraulic system piping and instrumentation diagram is shown in drawings M-46 and M-47, and the process diagrams are shown in drawings C11-1020-G-004, C11-1020-G-005, C11-1020-G-006, and C11-1020-G-007. The hydraulic requirements, identified by the function they perform, are as follows:
CHAPTER 04 4.6-6 REV. 18, SEPTEMBER 2016
- a. An accumulator hydraulic charging pressure of approximately 1400-1500 psig is required. Flow to the accumulators is required only during scram reset or system startup.
- b. Drive pressure of 260 psi (minimum) above reactor vessel pressure is required. A flow rate of approximately 4 gpm to insert a control rod and 2 gpm to withdraw a control rod is required.
- c. Cooling water to the drives is required at approximately 15 psi above reactor vessel pressure and at a flow rate of approximately 0.2-0.34 gpm per drive unit.
- d. The scram discharge volume is sized to receive, and contain, all the water discharged by the drives during a scram; a minimum volume of 3.34 gallons per drive is required (excluding the instrument volume).
4.6.1.2.4.2 System Description The CRD hydraulic system provides the required functions with the pumps, filters, strainers, valves, instrumentation, and piping shown in drawings M-46 and M-47, listed in Table 4.6-1, and described in the following paragraphs.
Duplicate components are included, where necessary, to ensure continuous system operation if an inservice component requires maintenance. The scram system design for LGS provides numerous indications and alarms in the control room to inform the operator of the condition of vital components, including the following:
- 1. SDV water level
- 2. Low scram accumulator pressure alarm
- 3. Scram accumulator H2O leakage alarm
- 5. SDV vent and drain valve position indication.
4.6.1.2.4.2.1 Supply Pump One supply pump pressurizes the system with water from the condensate treatment system and/or CST. One spare pump is provided for standby. A discharge check valve prevents backflow through the nonoperating pump. During normal operation, a portion of the pump discharge flow is diverted through a minimum flow bypass line to the CST. This flow is controlled by an orifice and is sufficient to prevent pump damage if the pump discharge is inadvertently closed. An alternative alignment allows pump operation with the minimum flow bypass line closed for the purpose of CRD hydraulic system maintenance, testing or trouble shooting.
Condensate water is processed by two filters. The pump suction filter is a disposable element-type with a 25 micron absolute rating. A differential pressure indicator and control room alarm monitor the filter element as it collects foreign material. A 250 micron strainer, downstream of the suction filter at each pump suction protects the pumps when the suction filter is being serviced.
The drive water filter downstream of the pump is a cleanable-element-type with a maximum 50 micron absolute rating. One spare drive water filter is provided for standby service. A 250 micron CHAPTER 04 4.6-7 REV. 18, SEPTEMBER 2016
LGS UFSAR strainer in each drive water filter/discharge line guards the hydraulic system if there is a filter element failure.
4.6.1.2.4.2.2 Accumulator Charging Pressure Accumulator charging pressure is established by precharging the nitrogen accumulator to a precisely controlled pressure at known temperature. During a scram, the scram inlet (and outlet) valves open and permit the stored energy in the accumulators to discharge into the drives. The resulting pressure decrease in the charging water header allows the CRD supply pump to "run-out" (i.e., flow rate to increase substantially) into the CRDs via the charging water header. The flow element, upstream of the accumulator charging water header, senses high flow and provides a signal to the manual auto-flow control station, which in turn closes the system flow control valve.
This action maintains increased flow through the charging water header, while avoiding prolonged pump operation at run-out conditions.
Pressure in the charging water header is monitored in the control room with a pressure indicator and a low pressure alarm.
During normal operation, the flow control valve maintains a constant system flow rate. This flow is used for drive flow and drive cooling.
4.6.1.2.4.2.3 Drive Water Pressure Drive water pressure required in the drive water header is maintained by the drive/cooling pressure control valve, which is manually adjusted from the control room. A flow rate of approximately 6 gpm (the sum of the flow rate required to insert and withdraw a control rod) normally passes from the drive water header through two solenoid-operated stabilizing valves (arranged in parallel) into the cooling water header. The flow through one stabilizing valve equals the drive insert flow; that through the other stabilizing valve equals the drive withdrawal flow.
When operating a drive, the required flow is diverted to that drive by closing the appropriate stabilizing valve, and at the same time opening the drive and exhaust directional control solenoid valves. Thus, flow through the drive pressure control valve is always constant.
Flow indicators in the drive water header and in the line downstream from the stabilizing valves allow the flow rate through the stabilizing valves to be adjusted when necessary. Differential pressure between the reactor vessel and the drive header is indicated in the control room.
On Unit 2, a permanent CRD friction test facility is connected to the drive water header and to a reactor vessel pressure instrument line. The device measures the friction of the control rods during rod movement. This facility eliminates the need for repetitive instrumentation setups while testing the CRDs during startup testing, refueling outages, and power plant operation. The friction test facility will be isolated from the CRD hydraulic system except when it is being used for friction testing or diagnostics.
4.6.1.2.4.2.4 Cooling Water Header The cooling water header is located downstream from the drive/cooling pressure valve. The drive/cooling pressure control valve is manually adjusted from the control room to produce the required drive water/cooling water pressure balance.
The flow through the flow control valve is virtually constant. Therefore, once adjusted, the drive/cooling pressure control valve maintains the required drive pressure and cooling water CHAPTER 04 4.6-8 REV. 18, SEPTEMBER 2016
LGS UFSAR pressure, independent of reactor vessel pressure. Changes in setting of the pressure control valves are required only to adjust for changes in the cooling requirements of the drives, as the drive seal characteristics change with time. A flow indicator in the control room monitors cooling water flow. A differential pressure indicator in the control room indicates the difference between reactor vessel pressure and drive cooling water pressure. Although the drives can function without cooling water, seal life is shortened by long-term exposure to reactor temperatures. The temperature of each drive is indicated and recorded, and excessive temperatures are annunciated in the control room.
4.6.1.2.4.2.5 Exhaust Water Header The exhaust water header connects to each HCU and provides a low pressure plenum and discharge path for the fluid expelled from the drives during control rod insert and withdraw operations. The fluid injected into the exhaust water header during rod movements is discharged back up to the RPV via reverse flow through the insert exhaust directional solenoid valves of adjoining HCUs. The pressure in the exhaust water header is, therefore, maintained at essentially reactor pressure. To ensure that the pressure in the exhaust water header is maintained near reactor pressure during the period of vessel pressurization, redundant pressure equalizing valves connect the exhaust water header to the cooling water header.
4.6.1.2.4.2.6 Scram Discharge Volume The SDV consists of header piping which connects to each HCU and drains into an instrument volume. The header piping is sized to receive and contain all the water discharged by the drives during a scram, independent of the instrument volume. A minimum SDV of 3.34 gallons per drive is specified through the system design specifications. This minimum SDV is based on conservative assumptions such that scram performance is not adversely affected.
The LGS CRD scram discharge header is connected to the instrument volume by piping of the same diameter as the SDV and is sloped towards the instrument volume, which monitors any water accumulation in the header. As the instrument volume starts to fill, the instrumentation triggers an alarm and initiates a scram when the instrument volume is full, but before the scram header begins to fill. This ensures that the scram header is able to accept the scram discharge water. The instrumentation for the instrument volume is located on the vertical instrument volume, and not on connecting piping.
There are two sets of headers, each with its own directly connected SDIV attached to the low point of the header piping. The large diameter pipe of the instrument volume thus serves as a vertical extension of the SDV that monitors water accumulation in the header piping. There is no reduction in the pipe size of the header piping going from the HCUs to the scram discharge instrument volume. This volume in turn is directly connected to the SDV without restriction.
The LGS design provides for redundant and diverse level sensing instrumentation. Two float switches and differential pressure (level) transmitters with level indicating switches provide high water level input to RPS from each SDIV. (There are two SDIVs which are hydraulically coupled via the common drain line). Each device provides an input to one channel. The devices are arranged so that no single failure prevents a reactor scram caused by high water level. These level instruments are connected to the instrument volume and trigger alarms in the control room on SDV High Level Trip.
In addition to the redundant and diverse devices which input to the RPS, there are two level switches for the instrument volume to provide operator warning. At the lowest level, a level switch CHAPTER 04 4.6-9 REV. 18, SEPTEMBER 2016
LGS UFSAR actuates to indicate that the volume is not completely empty during postscram draining, or to indicate that the volume starts to fill through leakage accumulation at other times during reactor operation. At the second level, a level switch produces a rod withdrawal block to prevent further withdrawal of any control rod when leakage accumulates to half the capacity of the instrument volume.
The instrumentation on the LGS instrument volume is designed so that no more than half of the scram trip level instruments shall be of the same manufacture, design, and operating principle to minimize the potential for common mode failure.
The SDV level instruments are located on the large diameter SDIV piping to minimize hydrodynamic forces associated with water flow following a scram.
During normal plant operation, the SDV is empty and vented to the atmosphere through its open vent and drain valves. When a scram occurs, upon a signal from the RPS, these vent and drain valves are closed to conserve reactor water. The position of the SDV vent and drain valves is continuously monitored. Lights in the control room indicate the position of these valves.
Redundant scram discharge system vent and drain valves are provided on LGS to ensure that no single failure can result in an uncontrolled loss of reactor coolant. Redundant solenoid-operated pilot valves control the vent and drain valves. The vent and drain system is therefore sufficiently redundant to avoid a failure to isolate the SDV single active failures. In addition, the solenoid-operated pilot valves are fail-safe (i.e., SDV isolates) on loss of electric or pneumatic power. The SDV drain line discharges to the equipment drain collection tank, and the vent line discharges to dirty radwaste. Vent lines are protected by vacuum breakers. The redundant isolation valves on the vent and drain lines are normally open during power operation and would not prevent draining of those lines in the nonscrammed condition. The position of these valves is indicated in the control room. Neither the vent and drain functions would be adversely affected by other system interfaces. Because of the redundant vent and drain valves, repair operations on any one valve will not prevent isolation of the vent and drain lines when scramming.
The operation of the scram discharge system is either automatic or remote manual, thus minimizing the time in which an operator would be in close contact with system components. The instrument volumes and level instruments are installed in corner locations to take advantage of shielding provided by concrete structures and to permit the installation of about two feet of temporary shielding for personnel protection. The scram discharge volume drain valves are located one floor below the instrument volumes which minimize the need for temporary shielding.
The SDV and associated vent and drain piping are classified as important to safety and required to meet the ASME Section III, Class 2 and seismic Category I requirements. Design parameters such as temperature, pressure, and frequency for limiting modes of operation provide a design basis for supply of equipment as well as the interfacing piping design and analysis.
During a scram, the SDV partly fills with water discharged from above the drive pistons. After scram is completed, the CRD seal leakage from the reactor continues to flow into the SDV until the discharge volume pressure equals the reactor vessel pressure. A check valve in each HCU prevents reverse flow from the scram discharge header volume to the drive. When the initial scram signal is cleared from the RPS, the SDV signal is overridden with a key-lock override switch, and the SDV is drained and returned to atmospheric pressure.
CHAPTER 04 4.6-10 REV. 18, SEPTEMBER 2016
LGS UFSAR Remote manual switches in the pilot valve solenoid circuits allow the discharge volume vent and drain valves to be tested without disturbing the RPS. Closing the SDV valves, allows the outlet scram valve seats to be leak tested by timing the accumulation of leakage into the SDV.
System operating conditions that are required for scram are continuously monitored. Four liquid level switches and two level transmitters, connected to each instrument volume, monitor the volume for abnormal water level. They are set at three different levels. They provide redundant and diverse inputs to the RPS scram function, control room annunciation function, and control rod withdrawal block function (drawing M-47). At the lowest level, a level switch actuates to indicate that the volume is not completely empty during postscram draining, or to indicate that the volume starts to fill through leakage accumulation at other times during reactor operation. At the second level, a level switch produces a rod withdrawal block to prevent further withdrawal of any control rod when leakage accumulates to half the capacity of the instrument volume. The remaining two switches and two transmitters are interconnected with the trip channels of the RPS, and initiate a reactor scram on high water level in the SDIV. Electronic switches are actuated by level transmitters to provide diversity of signals to the RPS.
The SDIV scram level instrumentation arrangement and trip logic allows instrument adjustment or surveillance without bypassing the scram function or directly causing a scram. Each level instrument can be individually isolated without bypassing the scram function. For the repair or replacement of scram level instrumentation the "one-out-of-two-twice" logic could be temporarily modified to "one-out-of-two". Thus, the single failure criterion would not be violated and the scram function is available when needed. Provisions are included in the system design for functional testing and calibration of SDIV level instrumentation.
The SDV piping system leak rates, loading conditions and material properties are bounded by the limiting value for these parameters identified in the BWROG letter of May 10, 1984 (BWROG-8420) and accepted by the NRC in Generic Letter 86-01 (January 3, 1986). Leak tests and inspections are performed per the criterion for Class 2 piping contained in ASME Section XI.
Also, the CRD Scram Discharge System is in the Leakage Reduction Program described in UFSAR Section 6.2.8 The LGS EOP for secondary containment control, which is based on the BWROG EPGs, includes an entry condition for secondary containment sump water level being above the normal operating level. In the event of a break in the scram discharge piping system, the normal plant procedures call for a reset of the scram. If, however, the affected system cannot be isolated or the isolation proves ineffective in mitigating temperature or radiation increases, rapid depressurization of the reactor is required per the EOP if the condition causes a widespread problem in the reactor enclosure. Rapid depressurization of the reactor is accomplished by safety valve discharge into the suppression pool. This reduces the leakage into the reactor building. The EOP also specifies the operator actions for achieving isolation and reactor depressurization should they be needed to mitigate leak consequences.
4.6.1.2.4.3 Hydraulic Control Units Each HCU furnishes pressurized water, on signal, to a drive unit. The drive then positions its control rod as required. Operation of the electrical system that supplies scram and normal control rod positioning signals to the HCU is described in Section 7.7.1.2.
The basic components in each HCU are manual, pneumatic, and electrical valves; an accumulator; related piping; electrical connections; filters; and instrumentation (drawings M-46, M-CHAPTER 04 4.6-11 REV. 18, SEPTEMBER 2016
LGS UFSAR 47, and Figure 4.6-8). The components and their functions are described in the following paragraphs.
4.6.1.2.4.3.1 Insert Drive Valve The insert drive valve is solenoid-operated and opens on an insert signal. The valve supplies drive water to the bottom side of the main drive piston.
4.6.1.2.4.3.2 Insert Exhaust Valve The insert exhaust solenoid valve also opens on an insert signal. The valve discharges water from above the drive piston to the exhaust water header.
4.6.1.2.4.3.3 Withdraw Drive Valve The withdraw drive valve is solenoid-operated and opens on a withdraw signal. The valve supplies drive water to the top of the drive piston.
4.6.1.2.4.3.4 Withdraw Exhaust Valve The solenoid-operated withdraw exhaust valve opens on a withdraw signal, and discharges water from below the main drive piston to the exhaust header. It also serves as the settle valve which opens, following any normal drive movement (insert or withdraw), to allow the control rod and its drive to settle back into the nearest latch position.
4.6.1.2.4.3.5 Speed Control Units The insert drive valve and withdraw exhaust valve have a speed control unit. The speed control unit regulates the control rod insertion and withdrawal rates during normal operation. The manually adjustable flow control unit is used to regulate the water flow to and from the volume beneath the main drive piston. A correctly adjusted unit does not require readjustment except to compensate for changes in drive seal leakage.
4.6.1.2.4.3.6 Scram Pilot Valve The scram pilot valve is operated from the RPS. The single scram pilot valve with two solenoids, controls both the scram inlet valve and the scram exhaust valve. The valve is solenoid-operated and normally energized. On loss of electrical signal to the solenoids, such as the loss of external ac power, the inlet port closes and the exhaust port opens. The scram pilot valve is designed so that the trip system signal must be removed from both solenoids before air pressure can be discharged from the scram valve operators. This prevents inadvertent scram of a single drive if one of the scram pilot valve solenoids fails.
4.6.1.2.4.3.7 Scram Inlet Valve The scram inlet valve opens to supply pressurized water to the bottom of the drive piston. This quick-opening globe valve is operated by an internal spring and system pressure. It is closed by air pressure applied to the top of its diaphragm operator. A position indicator switch on this valve and on the scram exhaust valve energizes a light in the control room as soon as the valve starts to open.
CHAPTER 04 4.6-12 REV. 18, SEPTEMBER 2016
LGS UFSAR 4.6.1.2.4.3.8 Scram Exhaust Valve The scram exhaust valve opens slightly before the scram inlet valve, exhausting water from above the drive piston. The exhaust valve opens faster than the inlet valve because of the higher air pressure spring setting in the valve operator.
4.6.1.2.4.3.9 Scram Accumulator The scram accumulator stores sufficient energy to fully insert a control rod at lower vessel pressures. At higher vessel pressures the accumulator pressure is assisted or supplanted by reactor vessel pressure. The accumulator is a hydraulic cylinder with a free-floating piston. The piston separates the water on top from the nitrogen below. A check valve in the accumulator charging line prevents loss of water pressure if supply pressure is lost.
During normal plant operation, the accumulator piston is seated at the bottom of its cylinder. Loss of nitrogen decreases the nitrogen pressure which actuates a pressure switch which sounds an alarm and actuates an indicating light in the control room.
To ensure that the accumulator is able to produce a scram, it is monitored for water leakage. A float-type level switch actuates an alarm and an indicating light in the control room if water leaks past the piston barrier and collects in the accumulator instrumentation block.
4.6.1.2.5 Control Rod Drive System Operation The CRD system performs rod insertion, rod withdrawal, and scram. These operational functions are described below.
4.6.1.2.5.1 Rod Insertion Rod insertion is initiated by a signal from the operator to the insert valve solenoids. This signal causes both insert valves to open. The insert drive valve applies reactor pressure plus approximately 90 psi to the bottom of the drive piston. The insert exhaust valve allows water from above the drive piston to discharge to the exhaust header.
As is illustrated in Figure 4.6-3, the locking mechanism is a ratchet-type device and does not interfere with rod insertion. The speed at which the drive moves is determined by the flow through the insert speed control valve, which is set for approximately 4 gpm for a shim speed (nonscram operation) of 3 in/sec. During normal insertion, the pressure on the downstream side of the speed control valve is90-100 psi above reactor vessel pressure. However, if the drive slows for any reason, the flow through and the pressure drop across the insert speed control valve decreases; the full differential pressure (260 psi) is then available to cause continued insertion. With 260 psi differential pressure acting on the drive piston, the piston exerts an upward force of 1040 pounds.
4.6.1.2.5.2 Rod Withdrawal Rod withdrawal is, by design, more involved than insertion. The collet fingers (latch) must be raised to reach the unlocked position (Figure 4.6-3). The notches in the index tube and the collet fingers are shaped so that the downward force on the index tube holds the collet fingers in place.
The index tube must be lifted before the collet fingers can be released. This is done by opening the drive insert valves (in the manner described in the preceding paragraph) for approximately one second. The withdraw valves are then opened, applying driving pressure above the drive CHAPTER 04 4.6-13 REV. 18, SEPTEMBER 2016
LGS UFSAR piston, and opening the area below the piston to the exhaust header. Pressure is simultaneously applied to the collet piston. As the piston raises, the collet fingers are cammed outward, away from the index tube, by the guide cap.
The pressure required to release the latch is set and maintained at a level high enough to overcome the force of the latch return spring, plus the force of reactor pressure opposing movement of the collet piston. When this occurs, the index tube is unlatched and free to move in the withdraw direction. Water displaced by the drive piston flows out through the withdraw speed control valve, which is set to give a rod shim speed of approxiately 3 in/sec. The entire valving sequence is automatically controlled, and is initiated by a single operation of the rod withdraw switch.
Rod withdrawal will not occur without permissive operator action. Following a deliberate operator withdrawal action, a rod drift could occur due to failure of its collet assembly to return to the locked position. The operator can interrupt this withdrawal with a scram or an insert signal. No single failure can of itself initiate a rod withdrawal.
4.6.1.2.5.3 Scram During a scram, the scram pilot valves and scram valves are operated as previously described.
With the scram valves open, accumulator pressure is admitted under the drive piston, and the area over the drive piston is vented to the SDV.
The large differential pressure (approximately 1500 psi initially, and always several hundred psi, depending on reactor vessel pressure) produces a large upward force on the index tube and control rod. This force gives the rod a high initial acceleration and provides a large margin of force to overcome friction. After the initial acceleration is achieved, the drive continues at a nearly constant velocity. This characteristic provides a high initial rod insertion rate. As the drive piston nears the top of its stroke, the piston seals close off the large passage (buffer orifices) in the stop piston tube, providing a hydraulic cushion at the end of travel.
Prior to a scram signal, the accumulator in the HCU has approximately 1450-1510 psig on the water-side and 1050-1150 psig on the nitrogen side. As the inlet scram valve opens, the full 2
water-side pressure is available at the CRD acting on a 4.1 in area. As CRD motion begins, this pressure drops to the gas-side pressure, less line losses between the accumulator and the CRD.
At low vessel pressures, the accumulator completely discharges with a resulting gas-side pressure of approximately 575 psi. The CRD accumulators are required to scram the control rods when the reactor pressure is low, and the accumulators retain sufficient stored energy to ensure the complete insertion of the control rods in the required time.
The ball check valve in the drive flange allows reactor pressure to supply the scram force whenever reactor pressure exceeds the supply pressure at the drive. This occurs, due to accumulator pressure decay and inlet line losses, during all scrams at higher vessel pressures.
When the reactor is close to, or at, full operating pressure, reactor pressure alone inserts the control rod in the required time, although the accumulator does provide additional margin at the beginning of the stroke.
The CRD system, with accumulators, provides the following scram performances at full power operation, in terms of average elapsed time after the opening of the RPS trip actuator (scram signal) for the drives to attain the scram strokes listed (See Figure 15.0-2):
CHAPTER 04 4.6-14 REV. 18, SEPTEMBER 2016
% of full stroke 5 20 50 90 Stroke, in 7.2 28.8 72.0 129.6 Average time, sec 0.375 0.90 2.0 3.5 4.6.1.2.5.4 Alternate Rod Insertion The ARI feature of the RRCS is designed to increase the reliability of the CRD system. ARI provides for rod insertion by actuating valves that are redundant and diverse from the RPS scram pilot valves.
The RRCS signal to insert control rods results in energizing the eight ARI dc solenoid valves shown in drawings M-46, M-47, C11-1020-G-004, C11-1020-G-005, C11-1020-G-006, and C11-1020-G-007. Two of these valves, in series with the backup scram valves, also have parallel functioning check valves to ensure venting of air from the air supply line in the event that an ARI valve fails. Four ARI valves provide for venting of the HCU scram valve pilot air headers to atmosphere. Depressurizing the headers actuates rod insertion. The remaining two ARI valves vent the scram air header that supplies the air pressure to keep the vent and drain lines open, thereby resulting in isolation of the SDV.
4.6.1.2.6 Instrumentation The instrumentation, for both the CRDMs and HCUs, is defined by that given for the RMCS. The objective of the RMCS is to provide the operator with the means to make changes in nuclear reactivity so that reactor power level and power distribution can be controlled. The system allows the operator to manipulate control rods.
The design bases and further discussion are discussed in Chapter 7. Section 7.6 lists the leak detection systems provided. Operator action is relied on to determine possible break locations based on these and plant system observations as discussed in GE report NEDO-24342 for CRD system breaks.
The licensee participated in the development of NEDO-22209, which was transmitted to the NRC by the BWROG on August 23, 1982. The results of NEDO-22209 show a probability of failure of the SDV too small to warrant exclusive monitoring devices.
4.6.1.3 Control Rod Drive Housing Supports 4.6.1.3.1 Safety Objective The CRD housing supports prevent any significant nuclear transient from occurring if a drive housing breaks or separates from the bottom of the reactor vessel.
4.6.1.3.2 Safety Design Bases The CRD housing supports meet the following safety design bases:
- a. Following a postulated CRD housing failure, control rod downward motion is limited, so that any resulting nuclear transient could not be sufficient to cause fuel damage.
CHAPTER 04 4.6-15 REV. 18, SEPTEMBER 2016
- b. The clearance between the CRD housings and the supports is sufficient to prevent vertical contact stresses caused by thermal expansion during plant operation.
4.6.1.3.3 Description The CRD housing supports are shown in Figure 4.6-9. Horizontal beams are installed immediately below the bottom head of the reactor vessel, between the rows of CRD housings.
The beams are welded to brackets which are welded to the steel form liner of the drive room in the reactor support pedestal.
Hanger rods, approximately 10 feet long and 13/4 inches in diameter, are supported from the beams on stacks of disc springs. These springs compress approximately 2 inches under the design load.
The support bars are bolted between the bottom ends of the hanger rods. The spring pivots at the top, and the beveled, loose-fitting ends on the support bars prevent substantial bending moment in the hanger rods if the support bars are overloaded.
Individual grids rest on the support bars between adjacent beams. Because a single-piece grid would be difficult to handle in the limited work space and because it is necessary that CRDs, position indicators, and incore instrumentation components be accessible for inspection and maintenance, each grid is designed for in-place assembly or disassembly. Each grid assembly is made from two grid plates, a clamp, and a bolt. The top part of the clamp guides the grid to its correct position, directly below the respective CRD housing that it would support in the postulated accident.
When the support bars and grids are installed, a gap of approximately l inch at room temperature (approximately 70 F) is provided between the grid and the bottom contact surface of the CRD flange. During system heatup, this gap is reduced by a net downward expansion of the housings with respect to the supports. In the hot operating condition, the gap is approximately 3/4 inch.
In the postulated CRD housing failure, the CRD housing supports are loaded when the lower contact surface of the CRD flange contacts the grid. The resulting load is then carried by two grid plates, two support bars, four hanger rods, their disc springs, and two adjacent beams.
The AISC Manual of Steel Construction, "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" is used in designing the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses used are 90% of yield, and the shear stress used is 60% of yield.
These design stresses are 1.5 times the AISC allowable stresses (60% and 40% of yield, respectively).
For purposes of mechanical design, the postulated failure resulting in the highest forces is an instantaneous circumferential separation of the CRD housing from the reactor vessel, with the reactor at an operating pressure of 1086 psig (at the bottom of the vessel) acting on the area of the separated housing. The weight of the separated housing, CRD, and blade, plus the pressure of 1086 psig acting on the area of the separated housing, gives a force of approximately 32,000 pounds. This force is used to calculate the impact force, conservatively assuming that the housing travels through a 1 inch gap before it contacts the supports. The impact force (109,000 lb) is then treated as a static load in design. The CRD housing supports are designed as seismic Category I CHAPTER 04 4.6-16 REV. 18, SEPTEMBER 2016
LGS UFSAR equipment as discussed in Section 3.2. Loading conditions and examples of stress analysis results and limits are shown in Table 3.9-6. A safety evaluation is provided in Section 4.6.2.3.3.
4.6.2 EVALUATIONS OF THE CRD SYSTEM 4.6.2.1 Failure Mode and Effects Analysis This subject is discussed in Section 15.9.
4.6.2.2 Protection from Common Mode Failures This subject is discussed in Section 15.9.
4.6.2.3 Safety Evaluation Safety evaluation of the control rods, CRDs, and CRD housing supports is described below.
Further description of control rods is contained in Section 4.2.
4.6.2.3.1 Control Rods 4.6.2.3.1.1 Materials Adequacy Throughout Design Lifetime The adequacy of the materials throughout the design life is evaluated in the mechanical design of the control rods. The primary materials, boron carbide (B4C) powder, solid hafnium, and 304 austenitic stainless steel, have been found suitable in meeting the demands of the BWR environment.
4.6.2.3.1.2 Dimensional and Tolerance Analysis Layout studies are done to ensure that, given the worst combination of part tolerance ranges at assembly, no interference exists that will restrict the passage of control rods. In addition, preoperational verification is made on each control blade system to show that the acceptable levels of operational performance are met.
4.6.2.3.1.3 Thermal Analysis of the Tendency to Warp The various parts of the control rod assembly remain at approximately the same temperature during reactor operation, negating the problem of distortion or warpage. What little differential thermal growth that could exist is allowed for in the mechanical design. A minimum axial gap is maintained between absorber rod tubes and the control rod frame assembly for this purpose. Use of dissimilar metals (stainless steel and hafnium) is evaluated to ensure that any effects due to thermal expansion or irradiation growth are acceptable.
4.6.2.3.1.4 Forces for Expulsion An analysis has been performed which evaluates the maximum pressure forces which could tend to eject a control rod from the core. The results of this analysis are given in Section 4.6.2.3.2.2.2.
In summary, if the collet were to remain open, which is unlikely, calculations indicate that the steady-state control rod withdrawal velocity would be 2 ft/sec for a pressure-under line break, the limiting case for rod withdrawal.
CHAPTER 04 4.6-17 REV. 18, SEPTEMBER 2016
LGS UFSAR 4.6.2.3.1.5 Functional Failure of Critical Components The consequences of a functional failure of critical components have been evaluated and the results are covered in Section 4.6.2.3.2.2.
4.6.2.3.1.6 Precluding Excessive Rates of Reactivity Addition In order to preclude excessive rates of reactivity addition, analysis has been performed both on the velocity limiter device, and the effect of probable control rod failures (Section 4.6.2.3.2.2).
4.6.2.3.1.7 Effect of Fuel Rod Failure on Control Rod Channel Clearances The CRD mechanical design ensures a sufficiently rapid insertion of control rods to preclude the occurrence of fuel rod failures, which could hinder reactor shutdown by causing significant distortions in channel clearances.
4.6.2.3.1.8 Mechanical Damage Analysis has been performed for all areas of the control system, showing that system mechanical damage does not affect the capability to continuously provide reactivity control.
In addition to the analysis performed on the CRDs (Sections 4.6.2.3.2.2 and 4.6.2.3.2.3) and the control rod blade, the following discussion summarizes the analysis performed on the control rod guide tube.
The guide tube can be subjected to any or all of the following loads:
- a. Inward load due to pressure differential
- b. Lateral loads due to flow across the guide tube
- c. Dead Weight
- d. Seismic (Vertical and Horizontal)
- e. Vibration In all cases, the analysis considers both a recirculation line break and a steam line break. These events result in the largest hydraulic loadings on a control rod guide tube.
Two primary modes of failure are considered in the guide tube analysis; exceeding allowable stress, and excessive elastic deformation. It was found that the allowable stress limit is not exceeded, and that the elastic deformations of the guide tube are never great enough to cause the free movement of the control rod to be jeopardized.
4.6.2.3.1.9 Evaluation of Control Rod Velocity Limiter The control rod velocity limiter limits the free fall velocity of the control rod to a value that cannot result in nuclear system process barrier damage. This velocity is evaluated in the rod-drop accident analysis in Chapter 15.
CHAPTER 04 4.6-18 REV. 18, SEPTEMBER 2016
LGS UFSAR 4.6.2.3.2 Control Rod Drives 4.6.2.3.2.1 Evaluation of Scram Time The rod scram function of the CRD system provides the negative reactivity insertion required by safety design basis as stated in Section 4.6.1.1.1. The scram time shown in the description is adequate as shown by the transient analyses of Chapter 15.
4.6.2.3.2.2 Analysis of Malfunction Relating to Rod Withdrawal There are no known single malfunctions that cause the unplanned withdrawal of even a single control rod. However, if multiple malfunctions are postulated, studies show that an unplanned rod withdrawal can occur at withdrawal speeds that vary with the combination of malfunctions postulated. In all cases the subsequent withdrawal speeds are less than that assumed in the rod-drop accident analysis as discussed in Chapter 15. Therefore, the physical and radiological consequences of such rod withdrawals are less than those analyzed in the rod-drop accident.
4.6.2.3.2.2.1 Drive Housing Failure at Attachment Weld The bottom head of the reactor vessel has a penetration for each CRD location. A drive housing is raised into position inside each penetration and is fastened by welding. The drive is raised into the drive housing and bolted to a flange at the bottom of the housing. The housing material is seamless, Type 304 stainless steel pipe with a minimum tensile strength of 75,000 psi. The basic failure considered here is a complete circumferential crack through the housing wall at an elevation just below the J-weld.
Static loads on the housing wall include the weight of the drive and the control rod, the weight of the housing below the J-weld, and the reactor pressure acting on the 6 inch diameter cross-sectional area of the housing and the drive. Dynamic loading results from the reaction force during drive operation.
If the housing were to fail as described, the following sequence of events is foreseen:
- a. The housing would separate from the reactor vessel.
- b. The CRD and housing would be blown downward against the support structure, by reactor pressure acting on the cross-sectional area of the housing and the drive.
- c. The downward motion of the drive and associated parts would be determined by the gap between the bottom of the drive and the support structure, and by the deflection of the support structure under the load.
- 1. In the current design, maximum deflection is approximately 3 inches.
- 2. If the collet remains latched, no further control rod ejection would occur (Reference 4.6-1); the housing would not drop far enough to clear the reactor vessel penetration.
- d. Reactor water would leak at a rate of approximately 180 gpm, through the 0.03 inch diameter clearance between the housing OD and reactor vessel penetration ID.
CHAPTER 04 4.6-19 REV. 18, SEPTEMBER 2016
LGS UFSAR If the basic housing failure were to occur while the control rod is being withdrawn (this is a small fraction of the total drive operating time) and if the collet were to stay unlatched, the following sequence of events is foreseen:
- a. The housing would separate from the reactor vessel.
- b. The control rod, CRD, and housing would be blown downward against the CRD housing support.
- c. Calculations indicate that the steady-state rod withdrawal velocity would be 0.3 ft/sec.
- d. During withdrawal, pressure under the collet piston would be approximately 250 psi greater than the pressure over it; therefore, the collet would be held in the unlatched position until driving pressure was removed from the pressure-over port.
4.6.2.3.2.2.2 Rupture of Hydraulic Line(s) to Drive Housing Flange There are three types of possible rupture of hydraulic lines to the drive housing flange:
pressure-under (insert) line break; pressure-over (withdrawn) line break; and coincident breakage of both of these lines.
4.6.2.3.2.2.2.1 Pressure-Under (Insert) Line Break For the case of a pressure-under (insert) line break, a partial or complete circumferential opening is postulated at or near the point where the line enters the housing flange. Failure is more likely to occur after another basic failure, wherein the drive housing or housing flange separates from the reactor vessel. Failure of the housing, however, does not necessarily lead directly to failure of the hydraulic lines.
If the pressure-under (insert) line were to fail and if the collet were latched, no control rod withdrawal would occur. There would be no pressure differential across the collet piston and, therefore, no tendency to unlatch the collet. Consequently, the associated control rod could not be withdrawn, but if reactor pressure is greater than 600 psig, it will insert on a scram signal.
The ball check valve is designed to seal off a broken pressure-under line by using reactor pressure to shift the check ball to its upper seat. If the ball check valve were prevented from seating, reactor water would leak to the reactor enclosure or containment. Because of the broken line, cooling water could not be supplied to the drive involved. Loss of cooling water would cause no immediate damage to the drive. However, prolonged exposure of the drive to temperatures at or near reactor temperature could lead to deterioration of material in the seals. High temperature would be indicated to the operator by the thermocouple in the position indicator probe, by high cooling water flow, and by operation of the containment sump pump.
If the basic line failure were to occur while the control rod is being withdrawn, the hydraulic force would not be sufficient to hold the collet open, and spring force normally would cause the collet to latch and stop rod withdrawal. However, if the collet were to remain open, calculations indicate that the steady-state control rod withdrawal velocity would be 2 ft/sec.
4.6.2.3.2.2.2.2 Pressure-Over (Withdrawn) Line Break CHAPTER 04 4.6-20 REV. 18, SEPTEMBER 2016
LGS UFSAR The case of the pressure-over (withdrawn) line breakage considers the complete breakage of the line at or near the point where it enters the housing flange. If the line were to break, pressure over the drive piston would drop from reactor pressure to atmospheric pressure. Any significant reactor pressure (approximately 600 psig or greater) would act on the bottom of the drive piston and fully insert the drive. Insertion would occur regardless of the operational mode at the time of the failure. After full insertion, reactor water would leak past the stop piston seals. This leakage would exhaust to the atmosphere through the broken pressure-over line. The leakage rate at 1000 psi reactor pressure is estimated to be 1-3 gpm; however, with the Graphitar seals of the stop piston removed, the leakage rate could be as high as 10 gpm, based on experimental measurements. If the reactor were hot, drive temperature would increase. This situation would be indicated to the reactor operator by the drift alarm, by the fully inserted drive, by a high drive temperature (annunciated in the control room), and by operation of the drywell sump pump.
4.6.2.3.2.2.2.3 Simultaneous Breakage of the Pressure-Over (Withdrawn) Pressure-Under (Insert) Lines For the simultaneous breakage of the pressure-over (withdrawn) and pressure-under (insert) lines, pressures above and below the drive piston would drop to zero, and the ball check valve would close the broken pressure-under line. Reactor water would flow from the annulus outside the drive, through the vessel ports, and to the space below the drive piston. As in the case of pressure-over line breakage, the drive would then insert (at reactor pressure approximately 600 psi or greater) at a speed dependent on reactor pressure. Full insertion would occur regardless of the operational mode at the time of failure. Reactor water would leak past the drive seals and out the broken pressure-over line to the reactor enclosure or containment, as described above. Drive temperature would increase. Indication in the control room would include the drift alarm, the fully inserted drive, the high drive temperature annunciated in the control room, and the operation of the drywell sump pump.
4.6.2.3.2.2.3 All Drive Flange Bolts Fail in Tension Each CRD is bolted to a flange at the bottom of a drive housing. The flange is welded to the drive housing. Bolts are made of AISI-4140 steel or AISI-4340 steel, with a minimum tensile strength of 125,000 psi. Each bolt has an allowable load capacity of at least 15,200 pounds. Capacity of the 8 bolts is at least 121,600 pounds. As a result of the reactor design pressure of 1250 psig, the major load on all 8 bolts is 30,400 pounds.
If a progressive or simultaneous failure of all bolts occurs, the drive separates from the housing.
The control rod and the drive would be blown downward against the support structure. Impact velocity and support structure loading would be slightly less than that for drive housing failure, because reactor pressure would act on the drive's cross-sectional area only and the housing would remain attached to the reactor vessel. The drive would be isolated from the cooling water supply. Reactor water would flow downward past the velocity limiter piston, through the large drive filter, and into the annular space between the thermal sleeve and the drive.
For worst case leakage calculations, the large filter is assumed to be deformed or swept out of the way so it would offer no significant flow restriction. At a point near the top of the annulus, where pressure would have dropped to 350 psi, the water would flash to steam and cause choke flow conditions. Steam would flow down the annulus and out the space between the housing and CHAPTER 04 4.6-21 REV. 18, SEPTEMBER 2016
LGS UFSAR the drive flanges to the drywell. Steam formation would limit the leakage rate to approximately 840 gpm.
If the collet were latched, control rod ejection would be limited to the distance the drive can drop before coming to rest on the support structure. There would be no tendency for the collet to unlatch, because pressure below the collet piston would drop to zero. Pressure forces, in fact, exert 1435 pounds to hold the collet in the latched position.
If the bolts failed during control rod withdrawal, pressure below the collet piston would drop to zero. The collet, with 1650 pounds return force, would latch and stop rod withdrawal.
4.6.2.3.2.2.4 Weld Joining Flange-to-Housing Failure in Tension The failure considered is a crack in or near the weld that joins the flange to the housing. This crack extends through the wall and completely around the housing. The flange material is forged, Type 304 stainless steel, with a minimum tensile strength of 75,000 psi. The housing material is seamless, Type 304 stainless steel pipe, with a minimum tensile strength of 75,000 psi. The conventional, full penetration weld of Type 308 stainless steel has a minimum tensile strength approximately the same as that for the parent metal. The design pressure and temperature are 1250 psig and 575oF. Reactor pressure acting on the cross-sectional area of the drive; the weight of the control rod, drive, and flange; and the dynamic reaction force during drive operation result in a maximum tensile stress at the weld of approximately 6000 psi.
If the basic flange-to-housing joint failure occurred, the flange and the attached drive would be blown downward against the support structure. The support structure loading would be slightly less than that for drive housing failure, because reactor pressure would act only on the drive cross-sectional area. Lack of differential pressure across the collet piston would cause the collet to remain latched and limit control rod motion to approximately 3 inches. Downward drive movement would be small; therefore, most of the drive would remain inside the housing. The pressure-under and pressure-over lines are flexible enough to withstand the small displacement and remain attached to the flange. Reactor water would follow the same leakage path described above for the flange bolt failure, except that exit to the drywell would be through the gap between the lower end of the housing and the top of the flange. Water would flash to steam in the annulus surrounding the drive. The leakage rate would be approximately 840 gpm.
If the basic failure were to occur during control rod withdrawal (a small fraction of the total operating time) and if the collet were held unlatched, the flange would separate from the housing.
The drive and flange would be blown downward against the support structure. The calculated steady-state rod withdrawal velocity would be 0.13 ft/sec. Because pressure-under and pressure-over lines remain intact, driving water pressure would continue to the drive, and the normal exhaust line restriction would exist. The pressure below the velocity limiter piston would drop below normal as a result of leakage from the gap between the housing and the flange. This differential pressure across the velocity limiter piston would result in a net downward force of approximately 70 pounds. Leakage out of the housing would greatly reduce the pressure in the annulus surrounding the drive. Thus, the net downward force on the drive piston would be less than normal. The overall effect of these events would be to reduce rod withdrawal to approximately one-half of normal speed. With a 560-psi differential across the collet piston, the collet would remain unlatched; however, it should relatch as soon as the drive signal is removed.
4.6.2.3.2.2.5 Housing Wall Ruptures CHAPTER 04 4.6-22 REV. 18, SEPTEMBER 2016
LGS UFSAR This failure is a vertical split in the drive housing wall just below the bottom head of the reactor vessel. The flow area of the hole is considered equivalent to the annular area between the drive and the thermal sleeve. Thus, flow through this annular area, rather than flow through the hole in the housing, would govern leakage flow. The housing is made of Type 304 stainless steel seamless pipe, with a minimum tensile strength of 75,000 psi. The maximum hoop stress of 11,900 psi results primarily from the reactor design pressure (1250 psig) acting on the inside of the housing.
If such a rupture were to occur, reactor water would flash to steam, and leak through the hole in the housing to the drywell at approximately 1030 gpm. Choke flow conditions would exist, as described previously for the flange bolt failure. However, leakage flow would be greater because flow resistance would be less, that is, the leaking water and steam would not have to flow down the length of the housing to reach the drywell. A critical pressure of 350 psi causes the water to flash to steam.
There would be no pressure differential acting across the collet piston to unlatch the collet; but the drive would insert as a result of loss of pressure in the drive housing causing a pressure drop in the space above the drive piston.
If this failure occurred during control rod withdrawal, drive withdrawal would stop, but the collet would remain unlatched. The drive would be stopped by a reduction of the net downward force action on the drive line. The net force reduction would occur when the leakage flow of 1030 gpm reduces the pressure in the annulus outside the drive to approximately 540 psig, thereby reducing the pressure acting on top of the drive piston to the same value. A pressure differential of approximately 710 psi would exist across the collet piston and holds the collet unlatched as long as the operator holds the withdraw signal.
4.6.2.3.2.2.6 Flange Plug Blows Out To connect the vessel ports with the bottom of the ball check valve, a hole of 0.75 inch diameter is drilled in the drive flange of this hole is sealed with a plug of 0.812 inch diameter and 0.25 inch thickness. A full penetration, Type 308 stainless steel weld holds the plug in place. The postulated failure is a full circumferential crack in this weld and subsequent blowout of the plug.
If the weld were to fail, the plug were to blow out, and the collet remained latched, there would be no control rod motion. There would be no pressure differential acting across the collet piston to unlatch the collet. Reactor water would leak past the velocity limiter piston, down the annulus between the drive and the thermal sleeve, through the vessel ports and drilled passage, and out the open plug hole to the drywell at approximately 320 gpm. Leakage calculations assume only liquid flows from the flange. Actually, hot reactor water would flash to steam, and choke flow conditions would exist. Thus, the expected leakage rate would be lower than the calculated value.
Drive temperature would increase and initiate an alarm in the control room.
If this failure were to occur during control rod withdrawal and if the collet were to stay unlatched, calculations indicate that control rod withdrawal speed would be approximately 0.24 ft/sec.
Leakage from the open plug hole in the flange would cause reactor water to flow downward past the velocity limiter piston. A small differential pressure across the piston would result in an insignificant driving force of approximately 10 pounds, tending to increase withdraw velocity.
CHAPTER 04 4.6-23 REV. 18, SEPTEMBER 2016
LGS UFSAR A pressure differential of 295 psi across the collet piston would hold the collet unlatched as long as the driving signal was maintained.
Flow resistance of the exhaust path from the drive would be normal because the ball check valve would be seated at the lower end of its travel by pressure under the drive piston.
4.6.2.3.2.2.7 Ball Check Valve Plug Blows Out As a means of access for machining the ball check valve cavity, a 1.25 inch diameter hole has been drilled in the flange forging. This hole is sealed with a plug with a 1.31 inch diameter and 0.38 inch thickness. A full penetration weld, utilizing Type 308 stainless steel filler, holds the plug in place. The failure postulated is a circumferential crack in this weld leading to a blowout of the plug.
If the plug were to blow out while the drive was latched, there would be no control rod motion. No pressure differential would exist across the collet piston to unlatch the collet. As in the previous failure, reactor water would flow past the velocity limiter, down the annulus between the drive and thermal sleeve, through the vessel ports and drilled passage, through the ball check valve cage and out the open plug hole to the drywell. The leakage calculations indicate that the flow rate would be 350 gpm. This calculation assumes liquid flow, but flashing of the hot reactor water to steam would reduce this rate to a lower value. Drive temperature would rapidly increase and initiate an alarm in the control room.
If the plug failure were to occur during control rod withdrawal, (it would not be possible to unlatch the drive after such a failure) the collet would relatch at the first locking groove. If the collet were to stick, calculations indicate the control rod withdrawal speed would be 11.8 ft/sec. There would be a large retarding force exerted by the velocity limiter due to a 35 psi pressure differential across the velocity limiter piston.
4.6.2.3.2.2.8 Drive/Cooling Water Pressure Control Valve Closure (Reactor Pressure, 0 psig)
The pressure to move a drive is generated by the pressure drop of practically the full system flow through the drive/cooling water pressure control valve. This valve is either a MOV or a standby manual valve; either one is adjusted to a fixed opening. The normal pressure drop across this valve develops a pressure 260 psi in excess of reactor pressure.
If the flow through the drive/cooling water pressure control valve were to be stopped, as by a valve closure or flow blockage, the drive pressure would increase to the shutoff pressure of the supply pump. The occurrence of this condition during withdrawal of a drive at zero vessel pressure will result in a drive pressure increase from 260 psig to no more than 1750 psig. Calculations indicate that the drive accelerates from 3 in/sec to approximately 6.5 in/sec. A pressure differential of 1670 psi across the collet piston would hold the collet unlatched. Flow would be upward, past the velocity limiter piston, but retarding force would be negligible. Rod movement would stop as soon as the driving signal was removed.
4.6.2.3.2.2.9 Ball Check Valve Fails to Close Passage to Vessel Ports Should the ball check valve sealing the passage to the vessel ports be dislodged and prevented from reseating following the insert portion of a drive withdrawal sequence, water below the drive piston would return to the reactor through the vessel ports and the annulus between the drive and the housing rather than through the speed control valve. Because the flow resistance of this CHAPTER 04 4.6-24 REV. 18, SEPTEMBER 2016
LGS UFSAR return path would be lower than normal, the calculated withdrawal speed would be 2 ft/sec.
During withdrawal, differential pressure across the collet piston would be approximately 40 psi.
Therefore, the collet would tend to latch and would have to stick open before continuous withdrawal at 2 ft/sec could occur. Water would flow upward past the velocity limiter piston, generating a small retarding force of approximately 120 pounds.
4.6.2.3.2.2.10 Hydraulic Control Unit Valve Failures Various failures of the valves in the HCU can be postulated, but none could produce differential pressures approaching those described in the preceding paragraphs, and none alone could produce a high velocity withdrawal. Leakage through either one or both of the scram valves produces a pressure that tends to insert the control rod rather than to withdraw it. If the pressure in the SDV should exceed reactor pressure following a scram, a check valve in the line to the scram discharge header prevents this pressure from operating the drive mechanisms.
4.6.2.3.2.2.11 Collet Fingers Fail to Latch The failure is presumed to occur when the drive withdraw signal is removed. If the collet fails to latch, the drive continues to withdraw at a fraction of the normal speed. This assumption is made because there is no known means for the collet fingers to become unlocked without some initiating signal. Because the collet fingers will not cam open under a load, accidental application of a down signal does not unlock them. (The drive must be given a short insert signal to unload the fingers and cam them open before the collet can be driven to the unlock position.) If the drive withdrawal valve fails to close following a rod withdrawal, the collet would remain open and the drive would continue to move at a reduced speed.
4.6.2.3.2.2.12 Withdrawal Speed Control Valve Failure Normal withdrawal speed is determined by differential pressures in the drive and is set for a nominal value of 3 in/sec. Withdrawal speed is maintained by the pressure regulating system and is independent of reactor vessel pressure. Tests show that accidental opening of the speed control valve to the fully open position produces a velocity of approximately 6 in/sec.
The CRD system prevents unplanned rod withdrawal, and it has been shown above that only multiple failures in a drive unit and in its control unit could cause an unplanned rod withdrawal.
4.6.2.3.2.3 Scram Reliability High scram reliability is the result of a number of features of the CRD system. For example:
- a. Two reliable sources of scram energy are used to insert each control rod:
individual accumulators at low reactor pressure, and the reactor vessel pressure itself at power.
- b. Each drive mechanism has its own scram valves and one scram pilot valve with two pilot solenoids, so only one drive can be affected if a scram valve fails to open.
Both pilot solenoids must be de-energized to initiate a scram.
- c. The RPS and the HCUs are designed so that the scram signal and mode of operation override all others.
CHAPTER 04 4.6-25 REV. 18, SEPTEMBER 2016
- d. The collet assembly and index tube are designed so they do not restrain or prevent control rod insertion during scram.
- e. The SDV is monitored for accumulated water and the reactor scrams before the volume is reduced to a point that could interfere with a scram.
4.6.2.3.2.4 Control Rod Support and Operation As described above, each control rod is independently supported and controlled as required by safety design bases.
4.6.2.3.3 Control Rod Drive Housing Supports Downward travel of the CRD housing and its control rod following the postulated housing failure equals the sum of these distances: the compression of the disc springs under dynamic loading, and the initial gap between the grid and the bottom contact surface of the CRD flange. If the reactor were cold and pressurized, the downward motion of the control rod would be limited to the spring compression (approximately 2 inches) plus a gap of approximately 3/4" +/- 1/4". If the reactor were hot and pressurized, the gap would be approximately 1/2" +/- 1/4" and the spring compression would be slightly less than in the cold condition. In either case, the control rod movement following a housing failure is substantially limited below one drive "notch" movement (6 inches). Sudden withdrawal of any control rod, through a distance of one drive notch at any position in the core, does not produce a transient sufficient to damage any radioactive material barrier.
The CRD housing supports are in place during power operation and when the nuclear system is pressurized. If a control rod is ejected during shutdown, the reactor remains subcritical because it is designed to remain subcritical with any one control rod fully withdrawn at any time.
At plant operating temperature, a gap of approximately 1/2" +/- 1/4" exists between the CRD housing and the supports. At lower temperatures the gap is greater. Because the supports do not contact any of the CRD housing, except during the postulated accident condition, vertical contact stresses are prevented. Inspection and testing are discussed in Section 4.6.3.2.
4.6.2.3.4 Loss of Scram Discharge Valve Air Pressure Slow or partial loss of air pressure to the scram discharge valves will not result in rapid filling of both the SDV and the SDIV, with consequent loss of adequate discharge volume. This is assured by the physical arrangement of this piping. This SDIV has a cross-sectional area greater than that of the SDV and is connected directly to the SDV, without restriction. In addition, the SDV piping slopes continuously toward the SDIV. Because of this arrangement, both volumes will not fill concurrently but rather the SDIV will fill, causing a reactor scram on high instrument volume level, prior to any filling of the SDV.
Likewise, a loss of reactor coolant will not occur for this event because the scram signal described above will also cause the SDV vent and drain valves to go closed.
4.6.3 TESTING AND VERIFICATION OF THE CONTROL ROD DRIVES 4.6.3.1 Control Rod Drives CHAPTER 04 4.6-26 REV. 18, SEPTEMBER 2016
LGS UFSAR 4.6.3.1.1 Testing and Inspection 4.6.3.1.1.1 Development Tests The initial development drive (prototypes of the standard locking piston design) testing included more than 5000 scrams and approximately 100,000 latching cycles. One prototype was exposed to simulated operating conditions for 5000 hours0.0579 days <br />1.389 hours <br />0.00827 weeks <br />0.0019 months <br />. These tests demonstrated the following:
- a. The drive easily withstands the forces, pressures, and temperatures imposed.
- b. Wear, abrasion, and corrosion of the nitrided stainless steel parts are negligible.
Mechanical performance of the nitrided surface is superior to that of materials used in earlier operating reactors.
- c. The basic scram speed of the drive has a satisfactory margin above minimum plant requirements at any reactor vessel pressure.
- d. Usable seal lifetimes in excess of 1000 scram cycles can be expected.
4.6.3.1.1.2 Factory Quality Control Tests Quality control of welding, heat treatment, dimensional tolerances, material verification, and similar factors is maintained throughout the manufacturing process to ensure reliable performance of the mechanical reactivity control components. Some of the quality control tests performed on the control rods, CRDMs, and HCUs are listed below:
- a. CRDM tests:
- 2. Electrical components are checked for electrical continuity and resistance to ground.
- 3. Drive parts that cannot be visually inspected for dirt are flushed with filtered water at high velocity. No significant foreign material is permitted in effluent water.
- 4. Seals are tested for leakage to demonstrate correct seal operation.
- 5. Each drive is tested for shim motion, latching, and control rod position indication.
- 6. Each drive is subjected to cold scram tests at various reactor pressures to verify correct scram performance.
- b. HCU tests:
- 1. Hydraulic systems are hydrostatically tested in accordance with the applicable code.
CHAPTER 04 4.6-27 REV. 18, SEPTEMBER 2016
- 2. Electrical components and systems are tested for electrical continuity and resistance to ground.
- 3. Correct operation of the accumulator pressure and level switches is verified.
- 4. The unit's ability to perform its part of a scram is demonstrated.
- 5. Correct operation and adjustment of the insert and withdrawal valves is demonstrated.
- c. Control rod tests:
- 1. Absorber rods are examined by nondestructive methods to ensure correct B4C density and weld integrity.
- 2. The absorber rod process control rod is inspected to ensure correct assembly.
- 3. The envelope of each control rod is inspected to ensure correct assembly.
4.6.3.1.1.3 Operational Tests After installation, all control rods and drive mechanisms are tested through their full stroke for operability.
During normal operation, each time a control rod is withdrawn a notch, the operator can observe the incore monitor indications to verify that the control rod is following the drive mechanism. All control rods that are partially withdrawn from the core can be tested for rod following by inserting or withdrawing the rod one notch and returning it to its original position, while the operator observes the incore monitor indications.
To make a positive test of control rod-to-CRD coupling integrity, the operator can withdraw a control rod to the end of its travel and then attempt to withdraw the drive to the overtravel position.
Failure of the drive to overtravel demonstrates rod-to-drive coupling integrity.
Hydraulic supply subsystem pressures can be observed from instrumentation in the control room.
Scram accumulator pressures can be observed on the nitrogen pressure gauges.
4.6.3.1.1.4 Acceptance Tests Criteria for acceptance of the individual CRDMs and the associated control and protection systems are incorporated in specifications and test procedures covering three distinct phases:
preinstallation; after installation prior to startup; and during startup testing.
The preinstallation specification will define criteria and acceptable ranges of such characteristics as seal leakage, friction, and scram performance under fixed test conditions which must be met before the component can be shipped.
The after installation, prestartup tests (Chapter 14) include normal and scram motion, and are primarily intended to verify that piping, valves, electrical components, and instrumentation are CHAPTER 04 4.6-28 REV. 18, SEPTEMBER 2016
LGS UFSAR properly installed. The test specifications will include criteria and acceptable ranges for drive speed, timer settings, scram valve response times, and control pressures. These are tests intended more to document system condition rather than to test performance.
As fuel is placed in the reactor, the startup test procedure (Chapter 14) is followed. The tests in this procedure are intended to demonstrate that the initial operational characteristics meet the limits of the specifications over the range of primary coolant temperatures and pressures from ambient to operating. The detailed specifications and procedures follow the general pattern established for such specifications and procedures in BWRs presently under construction and in operation.
4.6.3.1.1.5 Surveillance Tests The surveillance requirements for the CRD system are described below:
- a. Sufficient control rods are withdrawn, following a refueling outage when core alterations are performed, to demonstrate with a margin of 0.25% k that the core can be made subcritical at any time in the subsequent fuel cycle with the most reactive operable control rod fully withdrawn and all other operable rods fully inserted.
- b. Each partially or fully withdrawn control rod is exercised one notch periodically in accordance with the Technical Specifications.
In the event that operation is continuing with three or more rods valved out of service as a result of rod or drive system problems, this test shall be performed at least once each day.
The periodic control rod exercise test serves as a periodic check against deterioration of the control rod system, and also verifies the ability of the CRD to scram. If a rod can be moved with drive pressure, it may be expected to scram, since higher pressure is applied during scram. The frequency of exercising the control rods, under the conditions of three or more control rods valved out of service as a result of rod or drive system problems, provides even further assurance of the reliability of the remaining control rods.
- c. The coupling integrity is verified for each withdrawn control rod as follows:
- 1. When the rod is first withdrawn, observe discernible response of the nuclear instrumentation.
- 2. When the rod is fully withdrawn the first time, observe that the drive does not go to the overtravel position.
Observation of a response from the nuclear instrumentation during an attempt to withdraw a control rod indicates indirectly that the rod and drive are coupled. The overtravel position feature provides a positive check on the coupling integrity, for only an uncoupled drive can reach the overtravel position.
- d. During operation, accumulator pressure and level at the normal operating value is verified.
CHAPTER 04 4.6-29 REV. 18, SEPTEMBER 2016
LGS UFSAR Experience with CRD systems of the same type indicates that weekly verification of accumulator pressure and level is sufficient to ensure operability of the accumulator portion of the CRD system.
A channel functional test of the accumulator water leak detectors and a channel calibration of the accumulator pressure detectors, including alarm setpoint > 955 psig on decreasing pressure is performed at least once per 24 months.
- e. At the time of each major refueling outage, each operable control rod is subjected to scram time tests from the fully withdrawn position.
Experience indicates that the scram times of the control rods do not significantly change over the time interval between refueling outages. A test of the scram times at each refueling outage is sufficient to identify any significant increase in the scram times.
- f. The LGS Technical Specifications require periodic testing of the vent and drain valves.
- g. The LGS Technical Specifications include periodic testing and acceptance criteria to demonstrate SDV operability The operability of the entire system as an integrated whole shall be demonstrated periodically and during each operating cycle.
4.6.3.1.1.6 Functional Tests The functional testing program of the CRDs consists of the 5 year maintenance life and the 1.5 times design life test programs as described in Section 3.9.4.4.
There are a number of failures that can be postulated on the CRD, but it would be very difficult to test all possible failures. A partial test program with postulated accident conditions and imposed single failures is available.
The following tests with imposed single failures have been performed to evaluate the performance of the CRDs under these conditions:
- a. Simulated ruptured scram line test
- b. Stuck ball check valve in CRD flange
- c. HCU drive down inlet flow control valve (V122) failure
- d. HCU drive down outlet flow control valve (V120) failure
- f. HCU drive up outlet control valve (V121) failure
- g. HCU drive up inlet control valve (V123) failure CHAPTER 04 4.6-30 REV. 18, SEPTEMBER 2016
- h. Cooling water check valve (V138) leakage
- i. CRD flange check valve leakage
- j. CRD stabilization circuit failure
- k. HCU filter restriction
- l. Air trapped in CRD hydraulic system
- m. CRD collet drop test
- n. Control rod qualification velocity limiter drop test Additional postulated CRD failures are discussed in Sections 4.6.2.3.2.2.1 through 4.6.2.3.2.2.12.
4.6.3.2 Control Rod Drive Housing Supports 4.6.3.2.1 Testing and Inspection CRD housing supports are removed for inspection and maintenance of the CRDs. The supports for one control rod can be removed during reactor shutdown, even when the reactor is pressurized, because all control rods are then inserted. When the support structure is reinstalled, it is inspected for correct assembly with particular attention to maintaining the correct gap between the CRD flange lower contact surface and the grid.
4.6.4 INFORMATION FOR COMBINED PERFORMANCE OF REACTIVITY SYSTEMS 4.6.4.1 Vulnerability to Common Mode Failures The two reactivity control systems, CRD and SLCS, do not share any instrumentation or components. Thus, a common mode failure of the reactivity systems would be limited to an accident event which could damage essential equipment in the two independent systems.
A seismic event or the postulated accident environments (Section 3.11) are not considered potential common mode failures since the essential (scram) portions of the CRD system are designed to seismic Category I standards and to operate as required under postulated accident environmental conditions. The SLCS system is also designed to seismic Category I standards.
No common mode power failure is considered possible. The scram function of the CRD system is "fail-safe" on a loss of power and is designed to override any other CRD function. The SLCS system has three independent power supplies to its essential redundant pumps and valves. The power supplies to the SLCS system are considered vital and as such are switched to the onsite standby diesels on a loss of normal power sources.
4.6.4.2 Accidents Taking Credit for Multiple Reactivity Systems There are no postulated accidents evaluated in Chapter 15 that take credit for two or more reactivity control systems preventing or mitigating each accident.
CHAPTER 04 4.6-31 REV. 18, SEPTEMBER 2016
LGS UFSAR 4.6.5 EVALUATION OF COMBINED PERFORMANCE As indicated in Section 4.6.4.2, credit is not taken for multiple reactivity control systems for any postulated accidents in Chapter 15.
4.
6.6 REFERENCES
4.6-1 J.E. Benecki, "Impact Testing on Collet Assembly for Control Rod Drive Mechanism 7RD B144A", GE, Atomic Power Equipment Department, APED 5555, (November 1967).
CHAPTER 04 4.6-32 REV. 18, SEPTEMBER 2016
LGS UFSAR Table 4.6-1 IDENTIFICATION OF SCRAM DISCHARGE SYSTEM COMPONENTS(1)
Component Component No.
SDV Vent Valve F010 SDV Vent Valve F180 SDV Drain Valve F181 SDV Drain Valve F011 Instrument Volume Level Switch LS-N013-F (Alarm Only)
Instrument Volume Level Switch LS-N013-E (Rod Block)
Instrument Volume Level Switches LS-N013-A (Input to RPS) LS-N013-B LT-N012-C LT-N012-D Level Ind Switch LIS-N601-A LIS-N601-B LIS-N601-C LIS-N601-D
____________________
(1)
See drawing M-47.
CHAPTER 04 4.6-33 REV. 13, SEPTEMBER 2006