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Revision 12 4-i January, 2003 TABLE OF CONTENTS Section Title Page 4.0 REACTOR 4.0-1 4.1
SUMMARY
DESCRIPTION 4.1-1 4.1.1 REACTOR VESSEL 4.1-1 4.1.2 REACTOR INTERNAL COMPONENTS 4.1-1 4.1.2.1 Reactor Core 4.1-2 4.1.2.2 Shroud 4.1-7 4.1.2.3 Shroud Head and Steam Separators 4.1-7 4.1.2.4 Steam Dryer Assembly 4.1-7 4.1.3 REACTIVITY CONTROL SYSTEMS 4.1-8 4.1.3.1 Operation 4.1-8 4.1.3.2 Description of Control Rods 4.1-8 4.1.3.3 Supplementary Reactivity Control 4.1-8 4.1.4 ANALYSIS TECHNIQUES 4.1-9 4.1.4.1 Reactor Internal Components 4.1-9 4.1.4.2 Fuel Rod Thermal Analysis 4.1-12 4.1.4.3 Reactor Systems Dynamics 4.1-12 4.1.4.4 Nuclear Engineering Analysis 4.1-13 4.1.4.5 Neutron Fluence Calculations 4.1-13 4.1.4.6 Thermal Hydraulic Calculations 4.1-14 4.
1.5 REFERENCES
FOR SECTION 4.1 4.1-14 4.2 FUEL SYSTEM DESIGN 4.2-1 4.2.1 DESIGN BASES 4.2-1 4.
2.2 DESCRIPTION
AND DESIGN DRAWINGS 4.2-1 4.2.2.1 Reactivity Control Assembly 4.2-1 4.2.3 DESIGN EVALUATIONS 4.2-5 4.2.4 TESTING, INSPECTION AND SURVEILLANCE PLANS 4.2-6 4.2.5 OPERATING AND DEVELOPMENTAL EXPERIENCE 4.2-6 4.
2.6 REFERENCES
FOR SECTION 4.2 4.2-6
Revision 12 4-ii January, 2003 TABLE OF CONTENTS (Continued)
Section Title Page 4.3 NUCLEAR DESIGN 4.3-1 4.3.1 DESIGN BASES 4.3-1 4.3.1.1 (Deleted) 4.3-1 4.3.1.2 (Deleted) 4.3-1 4.
3.2 DESCRIPTION
4.3-1 4.3.2.1 Nuclear Design Description 4.3-2 4.3.2.2 Power Distribution 4.3-3 4.3.2.3 Reactivity Coefficients 4.3-4 4.3.2.4 Control Requirements 4.3-5 4.3.2.5 Control Rod Patterns and Reactivity Worths 4.3-6 4.3.2.6 Criticality of Reactor During Refueling 4.3-9 4.3.2.7 Stability 4.3-10 4.3.2.8 Vessel Irradiation 4.3-10 4.3.3 ANALYTICAL METHODS 4.3-12 4.3.4 CHANGES 4.3-12 4.
3.5 REFERENCES
FOR SECTION 4.3 4.3-13 4.4 THERMAL AND HYDRAULIC DESIGN 4.4-1 4.4.1 DESIGN BASIS 4.4-1 4.4.1.1 Safety Design Bases 4.4-1 4.4.1.2 (Deleted) 4.4-1 4.4.1.3 Requirements for Steady-State Conditions 4.4-1 4.4.1.4 Requirements for Transient Conditions 4.4-2 4.4.1.5 Summary of Design Bases 4.4-2 4.
4.2 DESCRIPTION
OF THERMAL-HYDRAULIC DESIGN OF THE REACTOR CORE 4.4-2 4.4.2.1 Summary Comparison 4.4-2 4.4.2.2 Critical Power Ratio 4.4-2 4.4.2.3 Linear Heat Generation Rate (LHGR) 4.4-2 4.4.2.4 Void Fraction Distribution 4.4-3 4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern 4.4-3 4.4.2.6 Core Pressure Drop and Hydraulic Loads 4.4-3 4.4.2.7 Correlation and Physical Data 4.4-3 4.4.2.8 Thermal Effects of Operational Transients 4.4-3
Revision 13 4-iii December, 2003 TABLE OF CONTENTS (Continued)
Section Title Page 4.4.2.9 Uncertainties in Estimates 4.4-3 4.4.2.10 Flux Tilt Considerations 4.4-3 4.
4.3 DESCRIPTION
OF THE THERMAL AND HYDRAULIC DESIGN OF THE REACTOR COOLANT SYSTEM 4.4-4 4.4.3.1 Plant Configuration Data 4.4-4 4.4.3.2 Operating Restrictions on Pumps 4.4-5 4.4.3.3 Power-Flow Operating Map 4.4-5 4.4.3.4 Temperature-Power Operating Map (PWR) 4.4-10 4.4.3.5 Load Following Characteristics 4.4-10 4.4.3.6 (Deleted) 4.4-10 4.4.4 EVALUATION 4.4-10 4.4.4.1 Critical Power 4.4-10 4.4.4.2 Core Hydraulics 4.4-11 4.4.4.3 Influence of Power Distributions 4.4-11 4.4.4.4 Core Thermal Response 4.4-11 4.4.4.5 Analytical Methods 4.4-11 4.4.4.6 Thermal-Hydraulic Stability Analysis 4.4-11 4.4.5 TESTING AND VERIFICATION 4.4-11 4.4.6 INSTRUMENTATION REQUIREMENTS 4.4-12 4.4.6.1 Loose Parts Monitoring 4.4-12 4.4.7 (Deleted) 4.5 REACTOR MATERIALS 4.5-1 4.5.1 CONTROL ROD DRIVE SYSTEM STRUCTURAL MATERIALS 4.5-1 4.5.1.1 Material Specifications 4.5-1 4.5.1.2 Austenitic Stainless Steel Components 4.5-4 4.5.1.3 Other Materials 4.5-6 4.5.1.4 Cleaning and Cleanliness Control 4.5-6 4.5.2 REACTOR INTERNAL MATERIALS 4.5-8 4.5.2.1 Material Specifications 4.5-8 4.5.2.2 Controls on Welding 4.5-11 4.5.2.3 Nondestructive Examination of Wrought Seamless Tubular Products 4.5-11
Revision 12 4-iv January, 2003 TABLE OF CONTENTS (Continued)
Section Title Page 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel - Regulatory Guide Conformance 4.5-11 4.5.2.5 Other Materials 4.5-13 4.5.3 CONTROL ROD DRIVE HOUSING SUPPORTS 4.5-14 4.
5.4 REFERENCES
FOR SECTION 4.5 4.5-15 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 4.6-1 4.6.1 INFORMATION FOR CRDS 4.6-1 4.6.1.1 Control Rod Drive System Design 4.6-1 4.6.1.2 Control Rod Drive Housing Supports 4.6-23 4.6.2 EVALUATIONS OF THE CRDS 4.6-25 4.6.2.1 Failure Mode and Effects Analysis 4.6-25 4.6.2.2 Protection from Common Mode Failures 4.6-25 4.6.2.3 Safety Evaluation 4.6-26 4.6.3 TESTING AND VERIFICATION OF THE CRDS 4.6-43 4.6.3.1 Control Rod Drives 4.6-43 4.6.3.2 Control Rod Drive Housing Supports 4.6-50 4.6.4 INFORMATION FOR COMBINED PERFORMANCE OF REACTIVITY CONTROL SYSTEMS 4.6-50 4.6.4.1 Vulnerability to Common Mode Failures 4.6-50 4.6.4.2 Accidents Taking Credit for Multiple Reactivity Systems 4.6-50 4.6.5 EVALUATION OF COMBINED PERFORMANCE 4.6-51 4.
6.6 REFERENCES
FOR SECTION 4.6 4.6-61
Revision 13 4-v December, 2003 LIST OF TABLES Table Title Page 4.3-1 (Deleted) 4.3-14 4.3-2 Neutron Calculation and Dosimetry Results Used to Evaluate Vessel Irradiation 4.3-15 4.4-1 through 4.4-4 (Deleted) 4.4-13 4.4-5 Reactor Coolant System Geometric Data 4.4-14 4.4-6 Lengths of Safety Injection Lines 4.4-15
Revision 12 4.0-1 January, 2003 4.0 REACTOR This chapter was prepared using the latest approved version of the licensing topical report General Electric Standard Application for Reactor Fuel (GESTAR) NEDE-24011-P-A including the United States Supplement, NEDE-24011-P-A-US. Applicable sections of this report are referenced as noted in <Section 4.1>, <Section 4.2>, <Section 4.3>, and
<Section 4.4>. Reference is made to standardized information contained in the topical report, consistent with the NRC overall standardization philosophy.
<Appendix 15B>, Reload Safety Analysis provides a summary description of the fuel designs, corresponding nuclear and thermal-hydraulic characteristics, stability considerations, etc. for the current cycle reload core.
Revision 12 4.1-1 January, 2003 4.1
SUMMARY
DESCRIPTION The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator and dryer assemblies, and jet pumps. Also included in the reactor assembly are the control rods, control rod drive housing and the control rod drives.
<Figure 3.9-19>, Reactor Vessel Cutaway, shows the arrangement of reactor assembly components. A summary of the important design and performance characteristics is given in <Section 1.3.1.1>. Loading conditions for reactor assembly components are specified in
<Section 3.9>.
4.1.1 REACTOR VESSEL The reactor vessel design and description are covered in <Section 5.3>.
4.1.2 REACTOR INTERNAL COMPONENTS The major reactor internal components are the core (fuel, channels, control blades, and instrumentation), the core support structure (including the shroud, top guide and core plate), the shroud head and steam separator assembly, the steam dryer assembly, the feedwater spargers, the core spray spargers, and the jet pumps. Except for the Zircaloy in the reactor core, these reactor internals are stainless steel or other corrosion resistant alloys. All major internal components of the vessel can be removed except the jet pump diffusers, the jet pump risers, the shroud, and the core spray lines.
The steam dryers, shroud head and steam separators, fuel assemblies, in-core assemblies, control rods, orificed fuel supports, feedwater spargers, core spray spargers, and control rod guide tubes, can be removed.
Revision 12 4.1-2 January, 2003 4.1.2.1 Reactor Core 4.1.2.1.1 General The design of the boiling water reactor core, including fuel, is based on the proper combination of many design variables and operating experience. These factors contribute to the achievement of high reliability.
A number of important features of the boiling water reactor core design are summarized in the following paragraphs:
- a.
The BWR core mechanical design is based on conservative application of stress limits, operating experience and experimental test results. The moderate pressure level characteristic of a direct cycle reactor (approximately 1,000 psia) results in moderate cladding temperatures and stress levels.
- b.
The low coolant saturation temperature, high heat transfer coefficients and neutral water chemistry of the BWR are significant, advantageous factors in minimizing Zircaloy temperature and associated temperature-dependent corrosion and hydride buildup.
The relatively uniform fuel cladding temperatures throughout the core minimize migration of the hydrides to cold cladding zones and reduce thermal stresses.
- c.
The basic thermal and mechanical criteria applied in the design have been proven by irradiation of statistically significant quantities of fuel. The design heat transfer rates and linear heat generation rates are similar to values proven in fuel assembly irradiation.
Revision 12 4.1-3 January, 2003
- d.
The design power distribution used in sizing the core represents a worst expected state of operation.
- e.
The General Electric thermal analysis basis, GETAB, is applied to assure that more than 99.9% of the fuel rods in the core are expected to avoid boiling transition for the most severe moderate frequency per <Regulatory Guide 1.70>, (Revision 3) transient described in <Chapter 15>. The possibility of boiling transition occurring during normal reactor operation is insignificant.
- f.
Because of the large negative moderator density coefficient of reactivity, the BWR has a number of inherent advantages. These are the uses of coolant flow for load following, the inherent self-flattening of the radial power distribution, the ease of control, the spatial xenon stability, and the ability to override xenon, in order to follow load.
Boiling water reactors do not have instability problems due to xenon.
This has been demonstrated by special tests which have been conducted on operating BWRs in an attempt to force the reactor into xenon instability, and by calculations. No xenon instabilities have ever been observed in the test results. All of these indicators have proven that xenon transients are highly damped in a BWR due to the large negative power coefficient of reactivity (Reference 1).
Important features of the reactor core arrangement are as follows:
- a.
The bottom-entry cruciform control rod designs consist of several absorber tubes filled with neutron absorbing material such as B4C and/or hafnium.
Revision 17 4.1-4 October, 2011 Control rods typical of the Original Equipment control rod design have been irradiated for more than eight years in the Dresden-1 reactor and have accumulated thousands of hours of service without significant failure in operating BWRs.
The lead Marathon control rod was loaded in Oyster Creek in November 1988. Inspection after one cycle indicated the integrity of the overall assembly as well as that of the absorber tubes and welds was maintained.
- b.
The fixed in-core fission chambers provide continuous power range neutron flux monitoring. A guide tube in each in-core assembly provides for a traversing ion chamber for calibration and axial detail. Source and intermediate range detectors are located in-core and are axially retractable. The in-core location of the source and intermediate range instruments provides coverage of the large reactor core and provides an acceptable signal-to-noise ratio and neutron-to-gamma ratio. All in-core instrument leads enter from the bottom and the instruments are in service during refueling. In-core instrumentation is discussed in <Section 7.6>.
- c.
As shown by experience obtained at Dresden-1 and other plants, the operator, utilizing the in-core flux monitor system, can maintain the desired power distribution within a large core by proper control rod scheduling.
- d.
The Zircaloy 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 of the assembly.
- e.
The 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.
Revision 12 4.1-5 January, 2003
- 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 control rod drive mechanisms for ease of maintenance and removal.
4.1.2.1.2 Core Configuration The reactor core is arranged as an upright circular cylinder containing a large number of fuel cells and is located within the reactor vessel.
The coolant flows upward through the core. The core arrangement (plan view) and the lattice configuration are described in <Section 4.3>.
4.1.2.1.3 Fuel Assembly Description The boiling water reactor core is composed of essentially two components--fuel assemblies and control rods. The fuel assembly
<Section 4.2> and control rod mechanical configurations <Figure 4.2-1>,
<Figure 4.2-2>, and <Figure 4.2-3>, are basically the same as used in Dresden-1 and in all subsequent General Electric boiling water reactors.
A description of the fuel assembly including fuel rods, water rods, other fuel assembly components, and channels are given in <Section 4.2>
which references Section 2.1 of GESTAR (Reference 5). A discussion of the fuel designs utilized for the current cycle is contained in
<Appendix 15B>, Reload Safety Analysis. A general description of the fuel rods and bundle is given below.
4.1.2.1.3.1 Fuel Rod A fuel rod consists of UO2 pellets and a Zircaloy cladding tube.
Barrier fuel bundles consist of fuel rods with a thin, high purity zirconium liner, i.e., barrier, mechanically bonded to the cladding tube. A fuel rod is made by stacking pellets into the Zircaloy cladding tube which is evacuated, back-filled with helium and sealed by welding
Revision 12 4.1-6 January, 2003 Zircaloy end plugs in each end of the tube. The rod is designed to withstand applied loads, both external and internal. The fuel pellet is sized to provide sufficient clearance within the fuel tube to accommodate axial and radial differential expansion between fuel and clad. Overall fuel rod design is conservative in its accommodation of the mechanisms affecting fuel in a BWR environment. Fuel rod design bases are discussed in more detail in <Section 4.2.1>.
4.1.2.1.3.2 Fuel Bundle Each fuel bundle contains fuel rods and water rods which are spaced and supported in a square (nxn) array by spacers and a lower and upper tie plate. Fuel bundle design descriptions are contained in GESTAR (Reference 5). The fuel bundle has two important design features:
- a.
The bundle design places minimum external forces on a fuel rod; each fuel rod is free to expand in the axial direction.
- b.
The unique structural design permits the removal and replacement, if required, of individual fuel rods.
The fuel assemblies, of which the core is comprised, are designed to meet all the criteria for core performance and to provide ease of handling. Selected fuel rods in each assembly differ from the others in uranium enrichment. This arrangement produces more uniform power production across the fuel assembly, and thus allows a significant reduction in the amount of heat transfer surface required to satisfy the design thermal limitations.
4.1.2.1.4 Assembly Support and Control Rod Location A few peripheral fuel assemblies are supported by fuel support pieces mounted on the core plate. Otherwise, individual fuel assemblies in the
Revision 12 4.1-7 January, 2003 core rest on fuel support pieces mounted on top of the control rod guide tubes. Each guide tube, with its fuel support piece, bears the weight of four assemblies and is supported by a control rod drive penetration nozzle in the bottom head of the reactor vessel. The core plate provides lateral support and guidance at the top of each control rod guide tube.
The top guide, mounted on top of the shroud, provides lateral support and guidance for the top of each fuel assembly. The reactivity of the core is controlled by cruciform control rods and their associated mechanical hydraulic drive system. The control rods occupy alternate spaces between fuel assemblies. Each independent drive enters the core from the bottom, and can accurately position its associated control rod during normal operation and yet exert approximately ten times the force of gravity to insert the control rod during the scram mode of operation.
4.1.2.2 Shroud The information on the shroud is contained in <Section 3.9.5.1>.
4.1.2.3 Shroud Head and Steam Separators The information on the shroud head and steam separators is contained in
<Section 3.9.5.1>.
4.1.2.4 Steam Dryer Assembly The information on the steam dryer assembly is contained in
<Section 3.9.5.1>.
Revision 12 4.1-8 January, 2003 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 in such a manner to counter-balance steam voids in the top of the core and effect significant power flattening.
The reactivity control function requires that all rods be available for either reactor scram (prompt shutdown) or reactivity regulation.
Because of this, the control elements are mechanically designed to withstand the dynamic forces resulting from a scram. They are connected to bottom-mounted, hydraulically actuated drive mechanisms which allow either axial positioning for reactivity regulation or rapid scram insertion. The design of the rod-to-drive connection permits each blade to be attached or detached from its drive without disturbing the remainder of the control system. The bottom-mounted drives permit the entire control system to be left intact and operable for tests with the reactor vessel open.
4.1.3.2 Description of Control Rods The description for control rod assembly designs applicable to Perry are 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 variation of reactor coolant flow. The supplementary
Revision 12 4.1-9 January, 2003 burnable poison is gadolinia (Gd2O3) mixed with UO2 in selected fuel rods in some fuel bundles.
4.1.4 ANALYSIS TECHNIQUES 4.1.4.1 Reactor Internal Components Computer codes used for the analysis of the internal components are listed as follows:
- a.
MASS
- b.
DYSEA
- c.
FAP-71
- d.
ANSYS Detailed descriptions of these programs are given in the sections that follow.
4.1.4.1.1 MASS (Mechanical Analysis of Space Structure) 4.1.4.1.1.1 Program Description The program, proprietary of the General Electric (GE) Company, is an outgrowth of the PAPA (Plate and Panel Analysis) program originally developed by L. Beitch in the early 1960s. The program is based on the principle of the finite element method. Governing matrix equations are formed in terms of joint displacements using a stiffness-influence-coefficient concept originally proposed by L. Beitch (Reference 2).
The program offers curved beam, plate and shell elements. It can handle mechanical and thermal loads in a static analysis and predict natural frequencies and mode shapes in a dynamic analysis.
Revision 12 4.1-10 January, 2003 4.1.4.1.1.2 (Deleted) 4.1.4.1.1.3 (Deleted) 4.1.4.1.1.4 (Deleted) 4.1.4.1.2 DYSEA 4.1.4.1.2.1 Program Description The DYSEA (Dynamic and Seismic Analysis) program is a GE proprietary program developed specifically for seismic and dynamic analysis of RPV and internals/building system. It calculates the dynamic response of linear structural systems by either temporal modal superposition or response spectrum method. Fluid-structure interaction effect in the RPV is taken into account by way of hydrodynamic mass.
Program DYSEA was based on program Structural Analysis Program (SAP) IV with added capability to handle the hydrodynamic mass effect.
Structural stiffness and mass matrices are formulated similar to SAP IV.
Solution is obtained in time domain by calculating the dynamic response mode by mode. Time integration is performed by using Newmarks B-method. Response spectrum solution is also available as an option.
4.1.4.1.2.2 (Deleted) 4.1.4.1.2.3 (Deleted) 4.1.4.1.2.4 (Deleted)
Revision 12 4.1-11 January, 2003 4.1.4.1.3 FAP-71 (Fatigue Analysis Program) 4.1.4.1.3.1 Program Description The FAP-71 computer code, or Fatigue Analysis Program (Reference 3), is a stress analysis tool used to aid in performing ASME-III Nuclear Vessel Code structural design calculations. Specifically, FAP-71 is used in determining the primary plus secondary stress range and number of allowable fatigue cycles at points of interest. For structural locations at which the 3Sm (PQ) 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, or (2) the present method documented in Paragraph NB-3228.3 of the 1971 Edition of the ASME Section III Nuclear Vessel Code. The program can accommodate up to 25 transient stress states of as many as 20 structural locations.
4.1.4.1.3.2 (Deleted) 4.1.4.1.3.3 (Deleted) 4.1.4.1.3.4 (Deleted) 4.1.4.1.4 ANSYS 4.1.4.1.4.1 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 static, elastic, plastic and creep, dynamic, seismic and dynamic plastic, and large deflection and stability analysis.
Revision 12 4.1-12 January, 2003
- b.
One-dimensional fluid flow analyses.
- c.
Transient heat transfer analysis including conduction, convection and radiation with direct input to thermal-stress analyses.
- 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.
Plotting - Geometry plotting is 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 analyses types. An option is also available for saving the stiffness matrix once it is calculated for the structure, and using it for other loading conditions.
4.1.4.1.4.2 (Deleted) 4.1.4.1.4.3 (Deleted) 4.1.4.1.4.4 (Deleted) 4.1.4.2 Fuel Rod Thermal Analysis Fuel rod thermal analyses are described in Section 2 of GESTAR II (Reference 5).
4.1.4.3 Reactor Systems Dynamics The analysis techniques and computer codes used in reactor systems dynamics are described in Section 4 (Reference 4).
Revision 13 4.1-13 December, 2003 4.1.4.4 Nuclear Engineering Analysis The analysis techniques are described in Subsection A.4.3.3 of GESTAR II (Reference 5). The codes used in the analysis are:
Computer Code Function Lattice Physics Calculates average few-group cross sections, bundle Model reactivities and relative fuel rod powers within the fuel bundle.
BWR Reactor Calculates three-dimensional nodal power Simulator distributions, exposures and thermal hydraulic characteristics as burnup progresses.
4.1.4.5 Neutron Fluence Calculations Neutron flux at the reactor vessel ID was calculated using the transport codes and assumptions described below.
4.1.4.5.1 Unit 1 Neutron Fluence Calculations Unit 1 neutron vessel fluence calculations were performed with DORT, which is the two-dimensional module of the TORT (Reference 6) three-dimensional, discrete ordinates, Sn transport code system. This code will solve a wide variety of radiation transport problems including both fixed source and multiplication problems. Slab, cylinder, and spherical geometries are allowed with various boundary conditions.
The fluence calculations incorporate, as an initial starting point, a neutron source distribution prepared from core power distribution data.
Anisotropic scattering was considered for all regions. The cross sections were prepared with 1/E flux weighted, PL matrices for anisotropic scattering. A two-dimensional transport calculation in (R,) coordinates was performed to obtain fast neutron fluxes at core midplane. Fast neutron fluxes at locations other than the core mid-plane were calculated using a second two-dimensional calculation in (R,Z) coordinates.
Revision 13 4.1-14 December, 2003 The fast neutron flux calculations are used to establish the lead factor, which is the ratio of flux between the surveillance capsule locations and the location of peak vessel inside surface flux. Use of the lead factor is discussed in <Section 4.3.2.8.1>.
4.1.4.5.2 (Deleted) 4.1.4.6 Thermal Hydraulic Calculations The digital computer program uses a parallel flow path model to perform the steady-state BWR reactor core thermal-hydraulic analysis. Program input includes the core geometry, operating power, pressure, coolant flow rate and inlet enthalpy, and power distribution within the core.
Output from the program includes core pressure drop, coolant flow distribution, critical power ratio, and axial variations of quality, density and enthalpy for each channel type. A description of the thermal-hydraulic models is given in Section 4 of GESTAR (Reference 5).
4.
1.5 REFERENCES
FOR SECTION 4.1
- 1.
Crowther, R. L. Xenon Considerations in Design of Boiling Water Reactors, APED-5640, June 1968.
- 2.
Beitch, L., Shell Structures Solved Numerically by Using a Network of Partial Panels, AIAA Journal, Volume 5, No. 3, March 1967.
- 3.
Young, L. J., FAP-71 (Fatigue Analysis Program) Computer Code, GE/NED Design Analysis Unit R. A. Report No. 49, January 1972.
- 4.
Carmichael, L. A. and Scatena, G. J., Stability and Dynamic Performance of the General Electric Boiling Water Reactor, APED-5652.
Revision 13 4.1-15 December, 2003
- 5.
General Electric Company 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).
- 6.
CCC-543, TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport Version 2.8.14 Radiation Shielding Information Center, Oak Ridge National Laboratory.
Revision 19 4.2-1 October, 2015 4.2 FUEL SYSTEM DESIGN The format of this section corresponds to Standard Review Plan Section 4.2 in <NUREG-0800>. Most of the information is presented by reference to the licensing topical report GESTAR (Reference 1). The subsection numbers in <Section 4.2> generally correspond to the subsection numbers of Appendix A of GESTAR. Any additional information or differences are given for the applicable subsection.
<Appendix 15B>, Reload Safety Analysis provides summary information on the fuel system design.
4.2.1 DESIGN BASES Information on fuel system design bases is provided in (Reference 1)
(Subsection A.4.2.1).
4.
2.2 DESCRIPTION
AND DESIGN DRAWINGS Information on fuel system description and design drawings is provided in (Reference 12) and (Reference 13) for the GE14 fuel product line and in (Reference 14) for the GNF2 fuel product line, except for the reactivity control assembly description, which is described below.
4.2.2.1 Reactivity Control Assembly The main structural member of a control rod is made of stainless steel and consists of a top handle, a bottom casting with a velocity limiter and a control rod drive coupling, and four wings attached to a vertical cruciform center post. The top handle, bottom casting and center post are welded into a single skeletal structure. The top handle provides structural rigidity at the top of the control rod. The bottom casting also provides for structural rigidity and contains positioning rollers
Revision 16 4.2-2 October, 2009 and a parachute-shaped velocity limiter. Rollers, housed in the top handle and bottom casting of the control rod, provide guidance for control rod insertion and withdrawal. Marathon Control Rods supplied after 2007 do not contain rollers in the top hands.
The control rods are separated uniformly throughout the core on a 12-inch pitch maximum. Each control rod is surrounded by four fuel assemblies. The control rods can be positioned at 6-inch steps and have a nominal withdrawal and insertion speed of 3 in./sec.
Control rods are cooled by the core leakage (bypass) flow. The core leakage flow is made up of recirculation flow that leaks through the several leakage flow paths, the most important of which are:
- a.
The area between the fuel channel and the fuel assembly lower tie plate;
- b.
Holes in the lower tie plate;
- c.
The area between the fuel assembly lower tie plate and the fuel support piece;
- d.
The area between the fuel support piece and the control rod guide tube;
- e.
The area between the control rod guide tube and the core support plate; and
- f.
The area between the core support plate and the shroud.
4.2.2.1.1 Original Equipment Control Rods The original equipment control rod consists of a sheathed cruciform array of 72 Type-304 stainless steel absorber tubes (18 tubes in each wing of the cruciform) filled with vibration compacted boron carbide
Revision 16 4.2-3 October, 2009 powder shown in <Figure 4.2-1>. The boron carbide (B4C) powder in the absorber tubes is compacted to about 70 percent of its theoretical density. The B4C contains a minimum of 76.5 percent by weight natural boron. The Boron-10 (B-10) minimum content of the boron is 18 percent by weight. The top handle aligns the absorber tubes which are seal welded with end plugs on either end. Each absorber tube is 0.22 inch outside diameter and has a 0.027 inch wall thickness. The B4C is longitudinally separated into individual compartments by stainless steel balls at approximately 17-inch intervals. The balls are held in position by a slight crimp in the tube. The individual tubes provide containment of the helium gas released by the boron-neutron capture reaction. Should the B4C compact further in service, the steel balls will distribute the resulting voids over the length of the absorber.
The absorber tubes are held in a cruciform array by a stainless steel U-shaped sheath extending the full length of the tubes. The sheaths are perforated to allow the coolant to circulate freely about the absorber tubes. Operating experience has shown that control rods constructed as described above are not susceptible to dimensional distortions.
4.2.2.1.2 Marathon Control Rods The General Electric Marathon control rod consists of a cruciform array of externally-square absorber tubes that are welded full length to each other to form a straight line array called a wing, shown in
<Figure 4.2-4>. Each wing is comprised of 14 absorber tubes with each tube acting as an individual chamber to hold the helium released from the boron carbide (B4C). The four wings are welded to tie rod segments to form the cruciform-shaped member of the control rod. The square absorber tubes are circular inside and are loaded with either B4C capsules or hafnium metal rods. The B4C powder is compacted to about 70 percent of its theoretical density into thin-walled, stainless steel capsules with stainless steel end caps to prevent B4C migration and to allow helium release from the capsules into the absorber tube. The
Revision 19 4.2-4 October, 2015 capsules are smaller than the absorber tube inside diameter, allowing B4C to swell before it makes contact with the absorber tube to prevent excessive absorber tube strains. The B4C contains a minimum of 76.5 percent by weight natural boron. The Boron-10 (B-10) minimum content of the boron is 18 percent by weight. The B4C capsules are either 3.00 or 11.41 inch minimum length.
Two hafnium rods each 71.40 inch minimum length are located in the three edge absorber tubes. Hafnium does not emit gases during its depletion.
However, hafnium has demonstrated swelling due to hydriding when clad with stainless steel resulting in high strains and cracking of the control rods. The hafnium metal is sized smaller than the absorber tube inside diameter to accommodate the swelling and to prevent excessive absorber tube strains.
The Marathon design uses an enhanced grade of high purity Type-304 stainless steel referred to as RAD RESIST 304S which provides a high resistance to irradiation-assisted corrosion cracking. Niobium and Tantalum are added to provide greater protection against stress corrosion cracking. Material hardening characteristics are the same as the Type-304 stainless steel used in previously approved designs.
The mechanical design for the GE Marathon control rods are described in (Reference 8) and accepted by the NRC for licensing applications in the Safety Evaluation Report in (Reference 8).
The GE Marathon Ultra HD control rods are considered to be direct replacements for the GE Marathon control rods, with respect to fit, form and function, from both a mechanical and nuclear perspective. The minor differences between the GE Marathon control rods and the GE Marathon Ultra HD control rods are shown in Table 2-1 of (Reference 10). These differences are justified as acceptable and as having a negligible impact on the function of the control rods. The mechanical design for
Revision 19 4.2-4a October, 2015 the GE Marathon Ultra HD control rods are described in (Reference 10) and accepted by the NRC for licensing applications in the Safety Evaluation Report in (Reference 10).
An additional GE-provided equivalency evaluation comparing the GE Marathon Ultra HD control rods to the GE Marathon control rods is contained in (Reference 11).
4.2.2.1.3 Velocity Limiter The control rod velocity limiter <Figure 4.2-3> is an integral part of the bottom assembly of each control rod. This engineered safeguard protects against high reactivity insertion rate by limiting the control rod free fall velocity in the event of a control rod drop accident. It
Revision 12 4.2-5 January, 2003 is a one-way device in that the control rod scram velocity (control rod scram time) is not significantly affected but the control rod dropout velocity is reduced to a permissible limit.
The velocity limiter is in the form of two nearly mated, conical elements that act as a large clearance piston inside the control rod guide tube. The lower conical element is separated from the upper conical element by four radial spacers 90 degrees apart and is at a 15 degree angle relative to the upper conical element, with the peripheral separation less than the central separation.
The hydraulic drag forces on a control rod are proportional to approximately the square of the rod velocity and are negligible at normal rod withdrawal or rod insertion speeds. However, during the scram stroke the rod reaches high velocity, and the drag forces must be overcome by the drive mechanism.
To limit control rod velocity during dropout, but not during scram, the velocity limiter is provided with a streamlined profile in the scram (upward) direction.
Thus, when the control rod is scrammed, water flows over the smooth surface of the upper conical element into the annulus between the guide tube and the limiter. In the dropout direction, however, water is trapped by the lower conical element and discharged through the annulus between the two conical sections. Because this water is jetted in a partially reversed direction into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod assembly to less than 3.11 ft/sec.
4.2.3 DESIGN EVALUATIONS Information on the fuel system evaluation for compliance with the design bases is provided in (Reference 1) (Subsection A.4.2.3).
Revision 19 4.2-6 October, 2015 4.2.4 TESTING, INSPECTION AND SURVEILLANCE PLANS Information on testing, inspection and surveillance is provided in (Reference 1) (Subsection A.4.2.4). Fuel assembly surveillance plans are further described in (Reference 2), (Reference 3), (Reference 4),
and (Reference 5).
4.2.5 OPERATING AND DEVELOPMENTAL EXPERIENCE For a discussion of fuel experience, see (Reference 6) and (Reference 7).
4.
2.6 REFERENCES
FOR SECTION 4.2
- 1.
Global Nuclear Fuel 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).
- 2.
J. S. Charnley (GE) to C. H. Berlinger (NRC), Post-Irradiation Fuel Surveillance Program, November 23, 1983.
- 3.
L. S. Rubenstein (NRC) to R. L. Gridley (GE), Post-Irradiation Fuel Surveillance, January 18, 1984.
- 4.
J. S. Charnley (GE) to L. S. Rubenstein (NRC), Fuel Surveillance Program, February 29, 1984.
- 5.
J. S. Charnley (GE) to L. S. Rubenstein (NRC), Additional Details Regarding Fuel Surveillance Program, May 25, 1984.
- 6.
Experience with BWR Fuel through January 1981, NEDE-24343, May 1981.
Revision 19 4.2-7 October, 2015
- 7.
J. S. Charnley (GE) to L. S. Rubenstein (NRC), 1985 Fuel Experience Report, August 13, 1986.
- 8.
GE Marathon Control Rod Assembly, NEDE-31758P-A, October 1991.
- 9.
(Deleted)
- 10. GE Marathon-Ultra Control Rod Assembly, NEDE-33284 Supplement 1P-A Revision 1, March 2012.
- 11. GEH Nuclear Energy Part Equivalency Evaluation Number 112-0495, Revision 0, May 29, 2012.
- 12. General Electric Company, Global Nuclear Fuels Fuel Bundle Designs, NEDC-31152P, (latest approved revision).
- 13. Global Nuclear Fuel, GE14 Compliance with Amendment 22 of NEDE-24011-P-A (GESTAR II), NEDC-32868P, Revision 5, May 2013.
- 14. Global Nuclear Fuel, GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II), NEDC-33270P, Revision 5, May 2013.
Revision 12 4.3-1 January, 2003 4.3 NUCLEAR DESIGN Most of the information in <Section 4.3> is provided in the licensing topical report GESTAR (Reference 1). The subsection numbers in
<Section 4.3> generally correspond to the subsection numbers of Appendix A of GESTAR. Additional information or differences are given for each applicable subsection below.
<Appendix 15B>, Reload Safety Analysis provides information on the current cycle nuclear design.
4.3.1 DESIGN BASES Information on nuclear design bases is provided in (Reference 1)
(Subsection A.4.3).
4.3.1.1 (Deleted) 4.3.1.1.1 Reactivity Bases Information on reactivity bases is provided in (Reference 1)
(Subsection A.4.3.1.1).
4.3.1.1.2 Overpower Bases Information on overpower bases is provided in (Reference 1)
(Subsection A.4.3.1.2).
4.3.1.2 (Deleted) 4.
3.2 DESCRIPTION
Information on the nuclear design description is provided in (Reference 1) (Subsection A.4.3.2).
Revision 17 4.3-2 October, 2011 4.3.2.1 Nuclear Design Description The nuclear design description is provided in (Reference 1)
(Subsection A.4.3.2.1) with the exception of the core loading pattern.
The loading pattern for the current reload cycle is given in
<Appendix 15B>, Reload Safety Analysis. The fuel bundle description is provided in (Reference 1) (Subsection A.4.2.2) and the applicable bundle types for the current reload cycle are given in <Appendix 15B>, Reload Safety Analysis.
4.3.2.1.1 (Deleted)
Revision 17 4.3-3 October, 2011 4.3.2.1.2 (Deleted) 4.3.2.2 Power Distribution Information on power distribution is provided in (Reference 1)
(Subsection A.4.3.2.2).
4.3.2.2.1 (Deleted) 4.3.2.2.2 (Deleted) 4.3.2.2.3 (Deleted)
Revision 12 4.3-4 January, 2003 4.3.2.2.4 Power Distribution Calculations Information on power distribution calculations is provided by the Perry nuclear fuel vendor in the cycle management report for each reload cycle, and discussed further in <Section 4.3.2.5>.
A full range of calculated power distributions along with the resultant exposure shapes and corresponding control rod patterns are also shown in Appendix 4A of (Reference 7), for a typical BWR/6.
4.3.2.2.5 Power Distribution Measurements Information on power distribution measurements is provided in (Reference 1) (Subsection A.4.3.2.2.2).
4.3.2.2.6 Power Distribution Accuracy Information on power distribution accuracy is provided in (Reference 1)
(Subsection A.4.3.2.2.3).
4.3.2.2.7 Power Distribution Anomalies Information on power distribution anomalies is provided in (Reference 1)
(Subsection A.4.3.2.2.4).
4.3.2.3 Reactivity Coefficients Information on reactivity coefficients including void, moderator, temperature, doppler and power coefficients is provided in (Reference 1)
(Subsection A.4.3.2.3).
Revision 12 4.3-5 January, 2003 4.3.2.4 Control Requirements Information on control requirements is provided in (Reference 1)
(Subsection A.4.3.2.4). Further information is provided below.
The control rod system is designed to provide adequate control of the maximum excess reactivity anticipated during the equilibrium fuel cycle operation.
Thus, the basis for design of the burnable poison loading is that it shall compensate for the reactivity difference between the control rod system capability and the core fuel. Because fuel reactivity is at a maximum and control at a minimum at ambient temperature, the shutdown capability is evaluated assuming a cold, xenon free core.
The safety design basis requires that the core, in its maximum reactivity condition, be subcritical with the control rod of the highest worth fully withdrawn and all others fully inserted. This limit allows control rod testing at any time in core life and assures that the reactor can be made subcritical by control rods alone.
4.3.2.4.1 Shutdown Reactivity Information on shutdown reactivity is provided in (Reference 1)
(Subsection A.4.3.2.4.1). See <Appendix 15B>, Reload Safety Analysis for the cold shutdown margin for the current cycle reference loading pattern.
4.3.2.4.2 Reactivity Variations Information on reactivity variations is provided in (Reference 1)
(Subsection A.4.3.2.4.2). The combined effects of the individual
Revision 17 4.3-6 October, 2011 constituents of reactivity for the current reload cycle are accounted for in each Keff in the Reload Safety Analysis <Appendix 15B>.
The excess reactivity designed into the core is controlled by a control rod system supplemented by gadolinia-urania fuel rods. The average fuel enrichment for the core load is chosen to provide excess reactivity in the fuel assemblies sufficient to overcome the neutron losses caused by core neutron leakage, moderator heating and boiling, fuel temperature rise, equilibrium xenon and samarium poisoning, plus an allowance for fuel depletion.
4.3.2.5 Control Rod Patterns and Reactivity Worths Information on control rod patterns and reactivity worths is provided to the Perry staff by the Perry nuclear fuel vendor in the cycle management report and the beginning of cycle cold startup report for each reload cycle.
Revision 17 4.3-7 October, 2011 4.3.2.5.1 Rod Control and Information System Control rod patterns and associated control rod reactivity worths are regulated by the Rod Control and Information System (RCIS). This system utilizes redundant inputs to provide rod pattern control over the complete range of reactor operations. The control rod worths are limited to such an extent that the Rod Drop Accident (RDA) and the Power Range Rod Withdrawal Error (RWE) become unimportant. The RCIS provides for stable control of core reactivity in both the single rod or rod gang mode of operation. The Rod Pattern Controller (RPC) mode of RCIS provides protection from an RDA from startup to about 19 percent of rated power. The Rod Withdrawal Limiter (RWL) provides protection from the RWE for all conditions above the low power setpoint (LPSP). Each of these modes is described in the following sections.
Revision 16 4.3-8 October, 2009 4.3.2.5.2 Rod Pattern Controller (RPC) Mode The RPC mode restricts control rod patterns to prescribed withdrawal sequences from the all-rods-inserted startup condition to about 19 percent of rated power. This mode minimizes control rod worths to the extent that they are not an important concern in the operation of a BWR. The consequences of an RDA or an RWE in this range are significantly less severe than that required to violate fuel safety limits. This system is described in detail in (Reference 4). Exception to (Reference 4) may be taken for Alternate control rod scram time testing provided that the exception does not result in exceeding the bounding analysis criteria used in (Reference 4). The supporting documents are provided in (Reference 8). Above 19 percent of rated power, control rod worths are very small due to the formation of voids in the moderator. Therefore, restrictions on control rod patterns are not required to minimize control rod worths.
The RPC Mode restrictions are also applied during a reactor shutdown, except that the RPC Mode restrictions may be bypassed for a reactor shutdown using the Improved BPWS Control Rod Insertion Process (Reference 12) and (Reference 13) provided:
Withdrawn Control Rods have a confirmed coupling check.
Control Rods, which do not have a confirmed coupling check, are fully inserted before bypassing the RPC Mode restrictions.
A coupling check is considered to be confirmed if no Single Operator Error can result in an incorrect coupling check, i.e., the coupling confirmation is performed once with two operators involved who both verify the rod is coupled, or the coupling confirmation is performed on two separate occasions. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core. Details on this coupling confirmation requirement are provided in <Section 4> and <Section 5> of (Reference 13).
Revision 16 4.3-8a October, 2009 Once the above conditions are met, the RPC Mode restrictions may be bypassed. Operable control rod insertions may continue by continuously inserting the control rods to position 00.
If control rods without a confirmed coupling check can not be inserted before reducing power below the low power setpoint, then control rod insertions must be performed using the RPC Mode restrictions. Once all rods without a confirmed coupling check are inserted then the RPC Mode restrictions may be bypassed. Operable control rod insertions may continue by continuously inserting the control rods to position 00.
Normally, following bypassing of the RPC Mode restrictions, control rods are continuously inserted from their current position to the full in position in approximately the reverse order of the RPC Mode restrictions. During a shutdown, it may be necessary to bank a group of rods or to insert other control rods to control Thermal Limits. Other restrictions for unique situations such as for shutdowns with one stuck rod are provided in (Reference 13) and in plant procedures.
Once the RPC Mode restrictions have been bypassed and control rod insertions have begun using the Improved BPWS Control Rod Insertion Process, control rod withdrawals are not permitted unless compliance with the RPC requirements are re-established.
4.3.2.5.3 Rod Withdrawal Limiter (RWL) Mode Above the low power set point the RCIS relies on the RWL mode to provide regulation of control rod withdrawals in order to prevent the occurrence of a rod withdrawal error. This mode limits the withdrawal of a single control rod or a gang of control rods to a predetermined increment depending on the power level. The system senses the location of the rod or rods and automatically blocks withdrawal when the preset increment is reached. The preset limit is determined by generic analyses such that the MCPR and LHGR are less than the limiting transient. At rated
Revision 16 4.3-8b October, 2009 conditions (above the high power set point) the rod will block at a 12-inch withdrawal. Between the low power and high power set points, the increment is allowed to increase, to 24-inches. Below the low power set point the RWL mode does not apply.
Revision 17 4.3-9 October, 2011 4.3.2.5.4 Control Rod Operation The control rods can be operated either individually or in a gang composed of up to four rods. The purpose of the ganged rods is to reduce the time required for plant startup or recovery from a scram.
The RCIS provides regulation of control rod operation regardless of whether rods are being moved in single or ganged mode. The assignment of control rods to RCIS groups is shown in <Figure 4.3-4>,
<Figure 4.3-5>, <Figure 4.3-6>, and <Figure 4.3-7>, for the A and B patterns respectively. Also shown in these figures is the division of the groups into gangs of 1 to 4 rods which can be moved simultaneously.
4.3.2.5.5 Scram Reactivity The Reactor Protection System (RPS) responds to some abnormal operational transients by initiating a scram. The RPS and the Control Rod Drive (CRD) System act quickly enough to prevent the initiating disturbance from driving the fuel beyond transient limits. Additional information on scram reactivity is provided in (Reference 1)
(Subsection S.5.1.5.2).
4.3.2.6 Criticality of Reactor During Refueling Information on criticality of the reactor during refueling is provided in (Reference 1) (Subsection A.4.3.2.6).
The maximum allowable value of k-effective is less than 1.000 at any time.
Revision 12 4.3-10 January, 2003 4.3.2.7 Stability 4.3.2.7.1 Xenon Transients Information on xenon transients is provided in (Reference 1)
(Subsection A.4.3.2.7.1).
4.3.2.7.2 Thermal-Hydraulic Stability Information on thermal-hydraulic stability is provided in (Reference 1)
(Subsection A.4.3.2.7.2) and is also covered in <Section 4.4.4.6>.
Thermal-hydraulic stability for the current reload cycle core is discussed in <Appendix 15B>, Reload Safety Analysis.
4.3.2.8 Vessel Irradiation Neutron fluence at the reactor vessel is calculated as described below.
4.3.2.8.1 Unit 1 Vessel Irradiation The lead factor for the RPV inside wall was determined by using a combination of two separate two-dimensional neutron transport computer analyses. The first of these established the azimuthal and radial variation of flux in the vessel at the fuel midplane elevation. The second analysis determined the relative variation of flux with elevation. The azimuthal and axial distribution results were combined to provide a simulation of the three-dimensional distribution of flux.
The ratio of fluxes, or lead factor, between the surveillance capsule location and the peak flux locations was obtained from this distribution.
The DORT computer program <Section 4.1.4.5.1>, which utilizes the discrete ordinates method to solve the Boltzmann transport equation in two dimensions, was used to calculate the spatial flux distribution
Revision 13 4.3-11 December, 2003 produced by a fixed source of neutrons in the core region. The azimuthal distribution was obtained with a model specified in (R,)
geometry. A schematic illustration of the (R,) vessel model is shown in <Figure 4.3-9> for 1/4-core geometry. The actual calculation utilized a 1/4-core model with reflective boundary conditions at 0 and
- 90. The model incorporates inner and outer core regions, the shroud, water regions inside and outside the shroud, jet pump components, and the vessel wall. A spatial mesh consisting of 194 radial intervals and 181 azimuthal intervals was used. The core region material compositions and neutron source densities were representative of values at the core midplane elevation (75 inches above the bottom of active fuel), which is near the elevation of the wires. The distributed source, which is assumed to be separate in space and energy, was obtained from the core power distribution and fission neutron spectra.
The integral over position and energy is normalized to the total fission neutron source in the region. Neutron cross-sections were specified for a 26 energy group set, with angular dependence of the scattering cross-sections approximated by a third-order Legendre polynomial expansion. The output of this calculation provided the distribution of flux as a function of azimuth and radius at reactor midplane. The azimuth of the peak flux and its magnitude relative to the flux at the 3 azimuth, which is the azimuth of the flux wires, were determined from this distribution.
The calculation of the axial flux distribution was performed in (R,Z) geometry, using a simplified cylindrical representation of the core configuration and realistic simulations of the axial variations of power density and coolant mass density. The core description was based on conditions near the azimuth angle of 21.8 where the edge of the core is closest to the vessel wall. The elevation of the peak flux was determined, as well as its magnitude relative to the flux at the surveillance capsule elevation.
Revision 13 4.3-12 December, 2003 The two-dimensional transport calculations indicate that flux maxima occur at azimuthal locations which are displaced by 25.5 from the RPV quadrant references (0, 90, etc.), at an elevation about 101 inches above the bottom of the active fuel. Calculated fluxes were obtained for the capsule position at the 3 azimuth and at the peak flux location on the vessel inside surface by combining the (R,) and (R,Z) flux distributions. The lead factor, as determined from the ratio of the calculated fluxes at these locations, is 0.52.
Dosimetry located on the inside surface of the vessel was removed after the first fuel cycle and tested to determine the flux at that location.
The lead factor relating the dosimeter location to the peak location was used to calculate the peak vessel inside surface flux. Assuming an 80%
capacity factor, or 32 effective full power years (EFPY) in 40 years of operation, the fluence for this operating period was estimated (Reference 9). Results are shown in
. Dosimetry measurements were repeated after 5.5 EFPY after removing the surveillance capsule at the 3 azimuth (Reference 10) and (Reference 11). Results are provided in3.5 REFERENCES
FOR SECTION 4.3
- 1.
Global Nuclear Fuel 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).
- 2.
J. A. Woolley, 3D BWR Core Simulator, May 1976 (NEDO-20953).
- 3.
G. R. Parkos, BWR Simulator Methods Verification, January 1977.
- 4.
C. J. Paone, Banked Position Withdrawal Sequence, January 1977 (NEDO-21231).
- 5.
(Deleted)
- 6.
(Deleted)
- 7.
General Electric Standard Safety Analysis Report (GESSAR).
- 8.
T. C. Lee (GE) to K. Donovan/P. Gilles (CEI), Control Rod Scram Time Testing Procedure, TCL-88039, TCL-8905, TCL-8910, TCL-9022.
- 9.
T. A. Caine, Implementation of <Regulatory Guide 1.99>, Revision 2 for Perry Nuclear Power Plant Unit 1, November 1989 (SASR 89-76/DRF 137-0010).
- 10. L. J. Tilly, Perry Unit 1 RPV Surveillance Materials Testing and Analysis, November 1996 (GE-NE-B1301793-01, Revision 0).
- 11. M. OConnor, Pressure-Temperature Curves for FirstEnergy Corporation, Using the KIc Methodology Perry Unit 1, April 2002 (GE-NE-0000-0000-8763-01, Revision 0).
Revision 16 4.3-13a October, 2009
- 12. License Amendment 150, Perry Nuclear Power Plant, Unit No. 1 -
Issuance of Amendment RE: TSTF-476, Improved Banked Position Withdrawal Sequence Control Rod Insertion Process, Per The Consolidated Line Item Improvement Process (TAC No. MD8184),
August 28, 2008.
- 13. NEDO-33091-A, Revision 2, July 2004, Improved BPWS Control Rod Insertion Process.
Revision 12 4.3-14 January, 2003
4.2 DESCRIPTION
OF THERMAL-HYDRAULIC DESIGN OF THE REACTOR CORE 4.4.2.1 Summary Comparison An evaluation of plant performance from a thermal and hydraulic standpoint is discussed in <Section 4.4.3>.
4.4.2.2 Critical Power Ratio Information on the critical power ratio including boiling correlations is provided in (Reference 1) (Subsection A.4.4.2.2). The current boiling correlation used is provided in <Appendix 15B>, Reload Safety Analysis.
4.4.2.3 Linear Heat Generation Rate (LHGR)
Information on linear heat generation rate is provided in (Reference 1)
(Subsection A.4.4.2.3).
Revision 13 4.4-3 December, 2003 4.4.2.4 Void Fraction Distribution Void fraction distributions are calculated by the Perry nuclear fuel vendor for each reload cycle.
4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern Information on core coolant flow distribution and orificing pattern is provided in (Reference 1) (Subsection A.4.4.2.5).
4.4.2.6 Core Pressure Drop and Hydraulic Loads Information on the core pressure drop and hydraulic loads is provided in (Reference 1) (Subsection A.4.4.2.6).
4.4.2.7 Correlation and Physical Data Information on the correlation and physical data is provided in (Reference 1) (Subsection A.4.4.2.7).
4.4.2.8 Thermal Effects of Operational Transients Information on thermal effects of operational transients is provided in (Reference 1) (Subsection A.4.4.2.8).
4.4.2.9 Uncertainties in Estimates Information on uncertainties in estimates is provided in (Reference 1)
(Subsection A.4.4.2.9).
4.4.2.10 Flux Tilt Considerations Information on flux tilt considerations is provided in (Reference 1)
(Subsection A.4.4.2.10) and in <Section 4.3.2.2.7>.
Revision 13 4.4-4 December, 2003 4.
4.3 DESCRIPTION
OF THE THERMAL AND HYDRAULIC DESIGN OF THE REACTOR COOLANT SYSTEM The thermal and hydraulic design of the reactor coolant system is provided in (Reference 1) (Subsection A.4.4.3).
4.4.3.1 Plant Configuration Data 4.4.3.1.1 Reactor Coolant System Configuration The reactor coolant system is described in <Section 5.4> and shown in isometric perspective in <Figure 5.4-1>. The piping sizes, fittings and valves are listed in- a.
- b.
- c.
- d.
- e.
- a.
- b.
- c.
- d.
- a.
- b.
- c.
- d.
- e.
- b.
- a.
- 1.
- 2.
- b.
- a.
- b.
- c.
- a.
- b.
- c.
- d.
- e.
- f.
- g.
- h.
- a.
- b.
- a.
- b.
- c.
- d.
5.4 REFERENCES
FOR SECTION 4.5
- 1.
J. Balcken (GE) to D. R. Frantz (FEC), letter dated January 29, 1999.
Revision 12 4.6-1 January, 2003 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS The reactivity control systems consist of control rods and control rod drives, supplementary reactivity control for the initial core
<Section 4.3> and the standby liquid control system described in
<Section 9.3.5>.
4.6.1 INFORMATION FOR CRDS 4.6.1.1 Control Rod Drive System Design 4.6.1.1.1 Design Bases 4.6.1.1.1.1 General Design Bases 4.6.1.1.1.1.1 Safety Design Bases The control rod drive mechanical system shall meet the following safety design bases:
- a.
The design shall provide for a sufficiently rapid control rod insertion such that no fuel damage results from any abnormal operating transient.
- b.
The design shall include positioning devices, each of which individually supports and positions a control rod.
- c.
Each positioning device shall:
- 1.
Prevent its control rod from initiating withdrawal as a result of a single malfunction.
- 2.
Be individually operated so that a failure in one positioning device does not affect the operation of any other positioning device.
Revision 12 4.6-2 January, 2003
- 3.
Be 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.
4.6.1.1.1.1.2 Power Generation Design Bases The control rod drive system design shall provide for positioning the control rods to control power generation in the core.
4.6.1.1.2 Description The control rod drive system (CRD) 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 required 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 from the reactor protection trip system. The control rod drive system consists of locking piston control rod drive mechanisms and the CRD hydraulic system (including power supply and regulation, hydraulic control units, interconnecting piping, instrumentation, and electrical controls).
4.6.1.1.2.1 Control Rod Drive Mechanisms The CRD mechanism (drive) used for positioning the control rod in the reactor core is a double-acting, mechanically latched, hydraulic cylinder using water as its operating fluid <Figure 4.6-1>,
<Figure 4.6-2>, <Figure 4.6-3>, and <Figure 4.6-4>. The individual drives are mounted on the bottom head of the reactor pressure vessel.
The drives do not interfere with refueling and are operative even when the head is removed from the reactor vessel.
Revision 12 4.6-3 January, 2003 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. Using water from the condensate treatment system, and/or condensate storage tanks as the operating fluid eliminates the need for special hydraulic fluid. Drives are able to utilize simple piston seals whose leakage does not contaminate the reactor water but provides cooling for the drive mechanisms and their seals.
The drives are capable of inserting or withdrawing a control rod at a slow, controlled rate, as well as 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 beyond this position to the over-travel limit can be accomplished only if the rod and drive are uncoupled. Withdrawal to the over-travel limit is annunciated by an alarm.
The individual rod indicators, grouped in one control panel display, correspond to relative rod locations in the core.
Changes in local flux during control rod motion at power may be observed by monitoring the readings of the appropriate local power range
Revision 12 4.6-4 January, 2003 monitor (LPRM) string. To facilitate this when a control rod is selected, the output of an appropriate LPRM string is displayed along with the position of the selected control rod. Except for certain peripheral control rods, the LPRM string used is diagonally adjacent to the selected control rod.
4.6.1.1.2.2 Drive Components
<Figure 4.6-2> illustrates the operating principle of a drive.
<Figure 4.6-3> and <Figure 4.6-4> illustrate the drive in more detail.
The main components of the drive and their functions are described below.
4.6.1.1.2.2.1 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) 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 prevents 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 sq in. versus 4.1 sq in. for uptravel, or insertion.
Revision 12 4.6-5 January, 2003 This difference in driving area tends to balance the control rod weight and assures a higher force for insertion than for withdrawal.
4.6.1.1.2.2.2 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.
4.6.1.1.2.2.3 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 will 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.
Revision 12 4.6-6 January, 2003 If reactor water is used during a scram to supplement accumulator pressure, it is drawn through a filter on the guide cap.
4.6.1.1.2.2.4 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 buffer shaft, at the upper end of the piston tube, supports the stop piston and buffer components.
4.6.1.1.2.2.5 Stop Piston 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. Piston rings and bushings, similar to those used on the drive piston, are mounted on the upper portion of the stop piston. The lower portion of the stop piston forms a thin-walled cylinder containing the buffer piston, its metal seal ring and the buffer piston return spring. As the drive piston reaches the upper end of the scram stroke it strikes the buffer piston. A series of orifices in the buffer shaft provides a progressive water shutoff to cushion the buffer piston as it is driven to its limit of travel. The high pressures generated in the buffer are confined to the cylinder portion of the stop piston, and are not applied to the stop piston and drive piston seals.
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
Revision 12 4.6-7 January, 2003 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. Two switches are located at each position corresponding to an index tube groove, thus allowing redundant indication at each latching point. Two additional switches are 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. Redundant overtravel switches are 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 backseat position, the drive can pass this position and actuate the overtravel switches 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 or at any time during plant operation.
4.6.1.1.2.2.6 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.
Revision 12 4.6-8 January, 2003 Water used to operate the collet piston passes between the outer tube and the cylinder tube. 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 plug then enters the spud and prevents uncoupling.
4.6.1.1.2.2.7 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, can then be lifted from the drive.
If it is desired to uncouple a drive without removing the reactor pressure 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
Revision 12 4.6-9 January, 2003 completely and the plug to snap back into place. Therefore, the drive can be coupled to the control rod using only the weight of the control rod.
4.6.1.1.2.3 Materials of Construction Factors that determine the choice of construction materials are discussed in the following subsections.
4.6.1.1.2.3.1 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. Large tensile and column loads are applied during scram. The reactor environment limits the choice of materials suitable for corrosion resistance. To meet these varied requirements, the index tube is made from annealed, single phase, nitrogen strengthened, austenitic stainless steel. The wear and bearing requirements are provided by Malcomizing the complete tube. To obtain suitable corrosion resistance, a carefully controlled process of surface preparation is employed.
4.6.1.1.2.3.2 Coupling Spud The coupling spud is made of Inconel X-750 that is aged for maximum physical strength and the required corrosion resistance. Because misalignment tends to cause chafing in the semispherical contact area, the part is protected by a thin chromium plating (electrolyzed). This plating also prevents galling of the threads attaching the coupling spud to the index tube.
Revision 19 4.6-10 October, 2015 4.6.1.1.2.3.3 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, to the area contacting the index tube and unlocking cam surface of the guide cap.
4.6.1.1.2.3.4 Seals and Bushings Graphite carbon seal material 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 Graphite carbon seal material strength is experienced at higher temperatures, the drive is supplied with cooling water to hold temperatures below 250F. The Graphite carbon seal material is relatively soft, which is 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.
4.6.1.1.2.3.5 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 Graphite carbon seal material.
- b.
All springs and members requiring spring action (collet fingers, coupling spud and spring washers) are made of Inconel X-750.
Revision 19 4.6-10a October, 2015
- c.
The ball check valve is a Haynes Stellite cobalt-base alloy.
- d.
Elastomeric O-ring seals are ethylene propylene.
Revision 12 4.6-11 January, 2003
- 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, stop piston, buffer shaft, and buffer piston are made of Armco 17-4 PH.
- i.
Certain fasteners and locking devices are made of Inconel X-750 or 600.
Pressure-containing portions of the drives are designed and fabricated in accordance with requirements of Section III of the ASME Boiler and Pressure Vessel Code.
The CRD return line is capped to avoid potential nozzle cracking as required by <NUREG-0619>.
4.6.1.1.2.4 Control Rod Drive Hydraulic System The control rod drive hydraulic system <Figure 4.6-5> supplies and controls the pressure and flow to and from the drives through hydraulic control units (HCU). 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 non-moving drives. There is one HCU for each control rod drive.
Revision 13 4.6-12 December, 2003 4.6.1.1.2.4.1 Hydraulic Requirements The CRD hydraulic system design is shown in <Figure 4.6-5> and
<Figure 4.6-7>. The hydraulic requirements, identified by the function they perform, are as follows:
- a.
An accumulator hydraulic charging pressure of approximately 1,750 to 2,000 psig is required. Flow to the accumulators is required only during scram reset or system startup.
- b.
Drive pressure of approximately 260 psi above reactor vessel pressure measured at a point immediately above the core plate is required. A flow rate of approximately 4 gpm to insert each control rod and 2 gpm to withdraw each control rod is required.
- c.
Cooling water to the drives is required at greater than reactor vessel pressure and at a flow rate of approximately 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 while maintaining a pressure less than 65 psig; a minimum volume of 3.34 gallons per drive is required (excluding the instrument volume).
- e.
Charging water header supplies approximately.008 gpm to level control instruments, used to prevent buildup of non-condensible gases.
4.6.1.1.2.4.2 System Description The CRD hydraulic systems provide the required functions with the pumps, filters, valves, instrumentation, and piping shown in <Figure 4.6-5> and described in the following paragraphs.
Revision 12 4.6-13 January, 2003 4.6.1.1.2.4.2.1 Supply Pump One supply pump pressurizes the system with water from the condensate treatment system and/or condensate storage tanks. One spare pump is provided for standby. A discharge check valve prevents backflow through the non-operating pump. A portion of the pump discharge flow is diverted through a minimum flow bypass line to the condensate storage tank. This flow is controlled by an orifice and is sufficient to prevent immediate pump damage if the pump discharge is inadvertently closed.
Condensate water is processed by two sets of filters in the system. The pump suction filters are a disposable element type designed to remove particulate which could reduce the operating life of the drive water filters. A 250-micron strainer in the filter bypass line protects the pump when these filters are being serviced. The drive water filters, downstream of the pump, are cleanable element types with a 50-micron absolute rating. A differential pressure indicator and control room alarm monitor the operating filter element as it collects foreign materials.
4.6.1.1.2.4.2.2 Accumulator Charging Pressure Accumulator charging pressure is established by precharging the nitrogen accumulator and then opening the charging water isolation valve. During 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 control rod drives via the charging water header. The flow element upstream of the accumulator charging 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 diverts increased flow to the charging water header.
Revision 12 4.6-14 January, 2003 Pressure in the charging header is monitored in the control room with a pressure indicator and 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.
4.6.1.1.2.4.2.3 Drive Water Pressure Drive water pressure required in the drive header is maintained by the drive pressure control valve, which is manually adjusted from the control room. A flow rate of approximately 16 gpm (the sum of the flow rate required to insert 4 control rods) normally passes from the drive water pressure stage through eight solenoid operated stabilizing valves (arranged in parallel) into the cooling water header. The flow through two stabilizing valves equals the drive insert flow for one drive; that of one stabilizing valve equals the drive withdrawal flow for one drive.
When operating a drive(s), the required flow is diverted to the drives by closing the appropriate stabilizing valves, at the same time opening the drive directional control and exhaust 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 pressure stage is indicated in the control room.
4.6.1.1.2.4.2.4 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/cooling water pressure balance.
Revision 12 4.6-15 January, 2003 The flow through the flow control valve is virtually constant.
Therefore, once adjusted, the drive/cooling pressure control valve will maintain the correct drive pressure and cooling water 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.
4.6.1.1.2.4.2.5 Scram Discharge Volume (SDV)
The scram discharge volume 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. Each of the two sets of headers has its own directly-connected scram discharge instrument volume (SDIV) attached to the low point of the header piping.
The large diameter pipe of the instrument volume serves as a vertical extension of the SDV (though no credit is taken for it in determining SDV sizing requirements).
During normal plant operation the scram discharge volume is empty, and vented to atmosphere through its open vent and drain valve. When a scram occurs, upon a signal from the safety circuit these vent and drain valves are closed to conserve reactor water. Redundant vent and drain valves are provided to assure against loss of reactor coolant from the SDV following a scram. Lights in the control room indicate the position of these valves.
Revision 12 4.6-16 January, 2003 During a scram, the scram discharge volume partly fills with water discharged from above the drive pistons. After scram is completed, the control rod drive seal leakage from the reactor continues to flow into the scram discharge volume 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 reactor protection system, the scram discharge volume signal is overridden with a keylock override switch, and the scram discharge volume is drained and returned to atmospheric pressure.
Remote-manual switches in the pilot valve solenoid circuits allow the discharge volume vent and drain valves to be tested without disturbing the reactor protection system. Closing the scram discharge volume valves allows the outlet scram valve seats to be leak-tested by timing the accumulation of leakage inside the scram discharge volume.
Each instrument volume is monitored by level switches and by transmitter activated trip units <Figure 4.6-5>. One level switch and trip unit (contacts) in series constitutes one trip logic for input to the RPS.
Each RPS trip system receives two logic trip inputs both from one instrument volume. Two level switches and two transmitter/trip units in a one-out-of-two twice logic will provide redundant and diverse inputs to the RPS to initiate a reactor scram when water in each instrument volume exceeds that preset high water level. Furthermore, alarms and rod blocks will also provide warnings at lower water levels to control room operators if the instrument volume is not completely empty.
4.6.1.1.2.4.3 Hydraulic Control Units Each hydraulic control unit (HCU) furnishes pressurized water, on signal, to a drive unit. The drive then positions its control rod as
Revision 12 4.6-17 January, 2003 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>.
The basic components in each HCU are manual, pneumatic and electrical valves; an accumulator; related piping; electrical connections; filters; and instrumentation <Figure 4.6-5>, <Figure 4.6-7> and <Figure 4.6-8>.
The components and their functions are described in the following paragraphs.
4.6.1.1.2.4.3.1 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.
4.6.1.1.2.4.3.2 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.
4.6.1.1.2.4.3.3 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.
4.6.1.1.2.4.3.4 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.
Revision 12 4.6-18 January, 2003 4.6.1.1.2.4.3.5 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.
4.6.1.1.2.4.3.6 Scram Pilot Valve Assembly The scram pilot valve assembly is operated from the reactor protection system. The scram pilot valve assembly, with two solenoids, controls both the scram inlet valve and the scram exhaust valve. The scram pilot valve assembly is solenoid-operated and is 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 pilot valve assembly <Figure 4.6-5> 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 in the event of a failure of one of the pilot valve solenoids.
4.6.1.1.2.4.3.7 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 energizes a light in the control room as soon as the valve starts to open.
Revision 12 4.6-19 January, 2003 4.6.1.1.2.4.3.8 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.
4.6.1.1.2.4.3.9 Scram Accumulator The scram accumulator stores sufficient energy to fully insert a control rod at any 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 in the event 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 and sounds an alarm in the control room.
A float type level switch actuates an alarm if water leaks past the piston barrier and collects in the accumulator instrumentation block.
4.6.1.1.2.5 Control Rod Drive System Operation The control rod drive system performs rod insertion, rod withdrawal and scram. These operational functions are described in the sections that follow.
4.6.1.1.2.5.1 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
Revision 12 4.6-20 January, 2003 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 travel speed (nonscram operation) of 3 in./sec. During normal insertion, the pressure on the downstream side of the speed control valve is approximately 90 psi above reactor vessel pressure. However, if the drive slows for any reason, the flow through, and pressure drop across the insert speed control valve will decrease; the full differential pressure (260 psi) will then be available to cause continued insertion. With 260 psi differential pressure acting on the drive piston, the piston exerts an upward force of 1,040 lb.
4.6.1.1.2.5.2 Rod Withdrawal Rod withdrawal is, by design, more involved than insertion. The collet finger (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 1 second. The withdraw valves are then opened, applying driving pressure above the drive 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
Revision 12 4.6-21 January, 2003 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 the control rod a travel speed of 3 in./sec. The entire valving sequence is automatically controlled and is initiated by a single operation of the rod withdraw switch.
4.6.1.1.2.5.3 Scram During a scram the scram pilot valve assembly 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 scram discharge volume.
The large differential pressure (approximately 1,750 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 diminishing velocity. This characteristic provides a high initial rod insertion rate. As the drive piston nears the top of its stroke, the piston reaches the buffer and the driveline is brought to a stop at the full-in position.
Prior to a scram signal the accumulator in the Hydraulic Control Unit has 1,750-2,000 psig on the water side and approximately 1,750 psig on the nitrogen side. As the inlet scram valve opens, the full water side pressure is available at the control rod drive acting on a 4.1 sq inch area. As CRD motion begins, this pressure drops to the gas side pressure less line losses between the accumulator and the CRD. When the drive reaches the full-in position, the accumulator completely discharges with a resulting gas side pressure of approximately 1,200 psig.
Revision 12 4.6-22 January, 2003 The control rod drive accumulators are necessary to scram the control rods within the required time. Each drive, however, has an internal ballcheck valve which allows reactor pressure to be admitted under the drive piston. If the reactor is above 600 psi this valve ensures rod insertion in the event the accumulator is not charged or the inlet scram valve fails to open. The insertion time, however, will be slower than the scram time with a properly functioning scram system.
The maximum scram insertion time for each control rod from the fully withdrawn position, based on de-energization of the scram pilot valve solenoid as time zero is contained in the Technical Specifications.
4.6.1.1.2.5.4 Alternate Rod Insertion (ARI)
The Alternate Rod Insertion feature is designed to increase the reliability of the Control Rod Drive system scram function. ARI provides for insertion of reactor control rods by depressurizing the scram air header through valves which are redundant and diverse from the reactor protection system scram valves.
The Redundant Reactivity Control System (RRCS) <Section 7.6.1.12>,
signal to insert control rods results in energizing the eight ARI valves shown on <Figure 4.6-5>. Two valves in series with the backup scram valves assure venting of air from the air supply line in the event one or more of the ARI valves fails. Four valves provide for venting of the A and B HCU scram valve pilot air headers to atmosphere to depressurize the headers and scram all rods. Two additional valves vent the scram air header which serves the scram discharge volume drain and vent lines, closing those valves and isolating the SDV.
4.6.1.1.2.6 Instrumentation The instrumentation for both the control rods and control rod drives is defined by that given for the rod control and information system. The
Revision 12 4.6-23 January, 2003 objective of the rod control and information system 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 covered in <Chapter 7>,
Instrumentation and Control Systems.
4.6.1.2 Control Rod Drive Housing Supports 4.6.1.2.1 Safety Objective The control rod drive (CRD) housing supports prevent any significant power excursion in the event a drive housing breaks or separates from the bottom of the reactor vessel.
4.6.1.2.2 Safety Design Bases The CRD housing supports shall meet the following safety design bases:
- a.
Following a postulated CRD housing failure, control rod downward motion shall be limited so that any resulting power excursion could not be sufficient to cause fuel damage.
- b.
The clearance between the CRD housings and the supports shall be sufficient to prevent vertical contact stresses caused by thermal expansion during plant operation.
4.6.1.2.3 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
Revision 12 4.6-24 January, 2003 which are welded to the steel form liner of the drive room in the reactor support pedestal.
Hanger rods, approximately 10 ft long and 1-3/4 inch 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 control rod drives, position indicators and in-core instrumentation components be accessible for inspection and maintenance, each grid is designed for inplace 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 1 inch at room temperature (approximately 70F) is provided between the grid and the bottom contact surface of the control rod drive 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.
Revision 12 4.6-25 January, 2003 The American Institute of Steel Construction (AISC) Manual of Steel Construction, Specification for the Design, Fabrication and Erection of Structural Steel for Buildings, was 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 were 90 percent of yield and the shear stress used was 60 percent of yield. These design stresses are 1.5 times the AISC allowable stresses (60 percent and 40 percent 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 1,086 psig (at the bottom of the vessel) acting on the area of the separated housing. The weight of the separated housing, control rod drive and blade, plus the pressure of 1,086 psig acting on the area of the separated housing, gives a force of approximately 32,000 lb.
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 Category I (seismic) equipment in accordance with <Section 3.2>.
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6.6 REFERENCES
FOR SECTION 4.6
- 1.
Benecki, J. E., Impact Testing on Collet Assembly for Control Rod Drive Mechanism 7RD B144A, General Electric Company, Atomic Power Equipment Department, APED-5555, November 1967.
- 2.
C. H. Solanas, Fast Scram Control Rod Drive Qualification Program, October 1978, NEDO-24142