ML19298B456
| ML19298B456 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/04/2019 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19298B540 | List: |
| References | |
| Download: ML19298B456 (126) | |
Text
BFN-26 C.0-i APPENDIX C STRUCTURAL QUALIFICATION OF SUBSYSTEMS AND COMPONENTS TABLE OF CONTENTS Introduction................................................................................................................................................. C.0-1 C.1 Scope............................................................................................................................................. C.0-1 C.2 Loading Conditions, Definitions, and Overview.............................................................................. C.0-1 C.2.1 Seismic Classification...................................................................................................... C.0-1 C.2.2 Loading Conditions.......................................................................................................... C.0-2 C.2.3 Definitions........................................................................................................................ C.0-3 C.2.4 Seismic Design Input....................................................................................................... C.0-4 C.2.5 Overview of Structural Qualification................................................................................. C.0-5 C.2.6 Safety Margins................................................................................................................. C.0-6 C.3 Piping and Pipe Supports............................................................................................................... C.0-8 C.3.1 Large Bore Piping............................................................................................................ C.0-9 C.3.2 Small Bore Piping............................................................................................................ C.0-11 C.3.3 Instrument Tubing............................................................................................................ C.0-13 C.3.4 Buried Piping................................................................................................................... C.0-14 C.3.5 Torus Attached Piping Systems....................................................................................... C.0-14 C.3.6 Pipe and Tubing Supports............................................................................................... C.0-18 C.3.7 Computer Programs for Class I Piping System Analysis................................................. C.0-22 C.4 Major Components......................................................................................................................... C.0-25 C.4.1 Reactor Pressure Vessel (RPV), RPV Internals, and Supports....................................... C.0-25 C.4.2 Primary System Components Stress Analysis................................................................. C.0-29 C.5 Primary Containment System and Penetrations............................................................................. C.0-29 C.5.1 Primary Containment Vessels Stress Analysis................................................................ C.0-30 C.5.2 Primary Containment Bellows Stress Analysis................................................................ C.0-31 C.5.3 Long Term Torus Integrity Program (LTTIP).................................................................... C.0-31 C.5.4 Wetwell/Drywell Vacuum Breakers.................................................................................. C.0-33 C.6 Equipment...................................................................................................................................... C.0-33 C.6.1 Equipment Seismic Loading (Historical).......................................................................... C.0-34 C.6.2 Equipment Seismic Analysis (Historical).......................................................................... C.0-35 C.6.3 Seismic Qualification of New Equipment and Replacements for Existing Equipment from March 1988 until July 2007.................................................................... C.0-36 C.6.4 Equipment Seismic/Structural Qualification (ESQ) after July 2007.................................. C.0-37 C.6.5 Qualification of Equipment in Torus Attached Piping Systems........................................ C.0-38 C.6.6 Interface Loads from Class I Piping Analysis................................................................... C.0-38 C.7 Heating, Ventilation, and Air Conditioning (HVAC) Ductwork and Supports................................... C.0-38 C.7.1 Scope.............................................................................................................................. C.0-39 C.7.2 Ductwork System Seismic Analysis................................................................................. C.0-39 C.7.3 Ductwork Load Combinations and Allowable Stresses.................................................... C.0-40
BFN-26 C.0-ii C.8 Control of Heavy Loads.................................................................................................................. C.0-42 C.8.1 Introduction/Licensing Background................................................................................. C.0-42 C.8.2 Safety Basis.................................................................................................................... C.0-44 C.8.3 Scope of Heavy Load Handling Systems........................................................................ C.0-44 C.8.4 Control of Heavy Loads Program.................................................................................... C.0-45 C.8.5 Safety Evaluation............................................................................................................ C.0-50 C.9 References..................................................................................................................................... C.0-51
BFN-26 C.0-iii APPENDIX C LIST OF TABLES Table Title C.2-1 Deformation Limit C.2-2 Primary Stress Limit C.2-3 Buckling Stability Limit C.2-4 Fatigue Limit C.3-1A Load Combinations and Allowable Stress Criteria for Class I Piping and Tubing (Piping other than RRS, MS, FW, and CRDH Systems)
C.3.1B Load Combinations and Allowable Stress Criteria of Class I Piping for Reactor Recirculation (RRS), Main Steam (MS), and Feedwater (FW)
Systems C.3.1C Load Combinations and Allowable Stress Criteria for Control Rod Drive Hydraulic Piping C.3-2 Load Combinations and Allowable Stresses for Class I Pipe and Tubing Supports C.4-1 Reactor Vessel, Reactor Vessel Internals, and Supports - Critical Load Combinations, Locations, and Allowables.
C.4-2 Primary System Components - Critical Load Combinations, Locations, and Allowables.
C.5-1 Drywell Loading Conditions and Allowable Stresses
BFN-27 C.0-1 APPENDIX C STRUCTURAL QUALIFICATION OF SUBSYSTEMS AND COMPONENTS Introduction Appendix C received a general update in FSAR Amendment 10 to describe currently applicable criteria and methods for structural qualification (and verification) of subsystems and components for the operating BFN Units. The updated Appendix C (for Amendment 10 and later Amendments) is maintained in accordance with 10CFR50.71.
Prior to Amendment 10, Appendix C described basic structural loading criteria and qualification methods used in the original design of BFN components and piping subsystems. Those earlier versions of Appendix C are significant for historical purposes only.
C.1 Scope Appendix C presents the criteria and qualification methods for the following Seismic Class I (hereafter referred to as Class I) mechanical and electrical subsystems/components:
Piping and Pipe Supports Major Components Primary Containment System and Penetrations Equipment HVAC Ductwork and Supports Conduit and Supports Cable Tray Systems C.2 Loading Conditions, Definitions, and Overview C.2.1 Seismic Classification The design basis for Class I subsystems and components considers all applied loads such as pressure, temperature, deadweight, seismic, and hydrodynamic loads. Definitions of Class I, Class II, plant conditions, and seismic loading are as follows:
BFN-27 C.0-2 Class I This class includes those structures, systems, and components whose failure or malfunction might cause, or increase the severity of, an accident which would endanger the public health and safety. This category includes those structures, systems, and components required for safe shutdown and isolation of the reactor.
Class II This class includes those structures, systems, and components which are important to reactor operation, but are not essential for preventing an accident that would endanger the public health and safety, and are not essential for the mitigation of the consequences of these accidents. A Class II designated item shall not degrade the integrity of any item designated Class I.
C.2.2 Loading Conditions Normal Conditions A normal condition is any condition in the course of operation of the plant under planned and anticipated conditions, in the absence of upset, emergency, or faulted conditions.
Upset Conditions Upset conditions are any deviations from normal conditions anticipated to occur often enough that design should include a capability to withstand these conditions. The upset conditions include abnormal operational transients caused by a fault in a system component requiring its isolation from the system, transients due to loss of load or power, and any system upset not resulting in a forced outage. The upset conditions include the effect of the Operating Basis Earthquake.
Emergency Conditions Emergency conditions are any deviations from normal conditions that require shutdown for correction of the conditions or repair of damage in the system.
A fire event is an emergency condition, and its effects of weight and pressure on the components or systems are evaluated. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of specific damage developed in the system.
BFN-27 C.0-3 Faulted Conditions Faulted conditions are those combinations of conditions associated with extremely low-probability postulated events whose consequences are such that the integrity and operability of the nuclear system may be impaired to the extent where considerations of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.
Test Conditions A test condition is any condition, such as, when hydrostatic testing of component or system is conducted in the course of operation of the plant under planned and anticipated conditions and in the absence of upset, emergency, or faulted conditions.
C.2.3 Definitions Operating Basis Earthquake The Operating Basis Earthquake (OBE) is defined as that earthquake for which structures, systems, and components of the nuclear power plant that must continue operation without undue risk to the health and safety of the public are designed to remain functional.
Design Basis Earthquake [Also referred to as Safe Shutdown Earthquake (SSE)]
The Design Basis Earthquake (DBE) is defined as that earthquake for which the structures, systems, and components of the nuclear power plant that must be capable of safe shutdown and maintain the plant in a safe condition without undue risk to the health and safety of the public are designed to remain functional.
Amplified Response Spectra Amplified Response Spectra (ARS) are defined as plots of maximum seismic response versus frequency for single degree-of-freedom subsystem at various locations in Class I structures subjected to seismic loading. The generation of amplified response spectra, building accelerations, and displacements used for subsystem and component analyses are described in Section 12.2 of the UFSAR.
BFN-27 C.0-4 Zero-Period Acceleration Zero-period Acceleration (ZPA) is defined as the peak of the building floor seismic acceleration time history.
Primary Loads Those loads which produce stresses which are not self-limiting, such as dead weight, pressure, seismic inertia loads, and hydrodynamic loads.
Secondary Loads Those loads which produce stresses which are self-limiting, such as thermal effects and seismic anchor movements.
Torus Attached Piping Systems Piping and tubing systems attached directly or indirectly to the torus shell to the point where effects of torus motion are demonstrated to be insignificant and including piping up to at least the first containment isolation valve. These systems include the main steam relief valve discharge piping systems.
C.2.4 Seismic Design Input The input ground motions and seismic structural analysis methods used to develop design inputs for subsystem and component design and qualification are addressed in UFSAR Section 12.2 for each seismic Class I structure housing safety related subsystems and components. The seismic inputs used for subsystems and components addressed in Appendix C are summarized as follows:
(a)
The original design basis El Centro earthquake input motion, described in Section 2.5.4, is used for qualification of
- 1)
Major components (Section C.4),
- 2)
The primary containment system and penetrations (Section C.5),
- 3)
Torus attached piping systems (Section C.3.5).
Specifically, original design basis El Centro input is utilized in the following ways. Time history responses from the dynamic earthquake analyses described in Section 12.2.2.8 are used to seismically qualify the reactor pressure vessel (RPV), RPV internals, and drywell.
Amplified response spectra are used to seismically qualify torus attached piping systems. Equivalent static coefficients are used to
BFN-27 C.0-5 seismically qualify the torus and vent system as well as the primary system components addressed in Section C.4.2.
(b)
The alternate design basis artificial earthquake input motion, described in Section 2.5.4, is used for qualification of piping and pipe supports (except torus attached piping systems) and HVAC ductwork and supports. It will be used for verification of seismic adequacy of mechanical and electrical equipment and for long-term qualification of conduit and cable tray subsystems.
Specifically, the artificial earthquake amplified response spectra are used to qualify piping and pipe supports (except torus attached piping systems) and HVAC ductwork and supports. These response spectra will also be used in verification of seismic adequacy of equipment and long-term qualification of conduit and cable tray subsystems.
C.2.5 Overview of Structural Qualification An overview of the structural qualification of BFN Class I subsystems and components follows.
- 1.
Piping and Pipe Supports (Section C.3)
Qualification of Class I piping and tubing systems, including the large bore piping, small bore piping/CRDH piping, buried piping, and tubing is in accordance with USAS-B31.1.0-1967 (Reference 1). Plant conditions and associated stress limits not addressed in USAS-B31.1.0-1967 are delineated in Section C.3.
Qualification of torus-attached piping systems has been implemented within the scope of the Long Term Torus Integrity Program (LTTIP) as described in the Browns Ferry Plant Unique Analysis Report (PUAR)
(Reference 12) and NRC Safety Evaluation (Reference 22). Torus attached piping system stresses, nozzle loads, concrete anchor loads, and other interface loads are maintained below the allowable described in the PUAR and Section C.3.5.
Pipe support design criteria for all Class I piping and instrument tubing supports are based on the AISC Manual of Steel Construction (Reference 4), in conjunction with the Manufacturers Standardization Society (MSS) SP-58 (Reference 5). Plant conditions and associated stress limits not addressed in References 4 and 5 are delineated in Section C.3.
BFN-28 C.0-6
- 2.
Major Components (Section C.4)
Seismic design adequacy of reactor pressure vessel (RPV) and internals (RPV/Internals) have been reassessed (Reference 21). This assessment was based on the coupled seismic model described in Section 12.2.2.8.2. It qualified the RPV/Internals for the updated loads. Critical stresses and interface loads for the RPV/Internals and primary system components are maintained below the allowables described in Tables C.4-1 and C.4-2, respectively.
The steam dryers have been replaced to support EPU operating conditions. The replacement dryers (Ref. 55), support brackets, and vessel shell (Ref. 56) have been requalified for EPU operating conditions and the increased weight of the replacement dryers. All structural components are below the ASME Code allowable limits.
- 3.
Primary Containment System and Penetrations (Section C.5)
The torus (wetwell), connecting vent system between drywell and wetwell, penetrations, and equipment have been qualified to Long Term Torus Integrity Program (LTTIP) criteria as documented in the BFN-Plant Unique Analysis Report (PUAR)
(Reference 12). Stresses are maintained below allowables described in the PUAR. In addition, the drywell and its penetrations have been qualified for pressure, deadweight, and applicable loading components as described in Section C.5.1.
- 4.
Equipment (Section C.6)
Browns Ferry Nuclear Plant was identified as one of the operating plants to be reviewed for the NRC Unresolved Safety Issue (USI) A-46 requirements. The existing BFN safety-related equipment was originally designed to have sufficient margin of safety to withstand BFN design basis seismic loading.
Plant-specific verification of seismic adequacy of equipment has been conducted in accordance with References 24, 25, 44, and
- 45. Qualification of new equipment and replacements for existing equipment from March 1988 until July 2007 is described in Section C.6.3. Equipment Seismic/Structural Qualification (ESQ) after July 2007 is described in Section C.6.4. Qualification of equipment subjected to hydrodynamic loads of the Long Term Torus Integrity Program (LTTIP) is described in Section C.6.5.
BFN-27 C.0-7
- 5.
HVAC Ductwork and Supports (Section C.7)
Seismic qualification of Class I HVAC ductworks and supports is based on AISC allowable stress design methods (Reference 4) and SMACNA standards (References 26, 27,
- 28) as described in Section C.7.
C.2.6 Safety Margins This section describes and justifies the basic structural safety margins used in the original BFN design basis for Class I subsystems and components. In some cases BFN has committed to more specific safety margins or allowables as described in other Appendix C sections. This section remains applicable for those cases where more specific commitments have not been made.
In addition to the generic definitions in Section C.2.2, the meaning of these terms is expanded in quantitative probabilistic language. The purpose of this expansion is to clarify the classification of any hypothesized accident or sequence of loading events so that the appropriate structural safety margins for reactor vessel and reactor vessel internals are applied. Knowledge of the event probability is necessary to establish meaningful and adequate safety factors for structural design. The table in the next paragraph illustrates the quantitative event classifications.
P40 = 40-year event encounter probability Upset (likely) 1.0 > P40 > 10-1 Emergency (low probability) 10-1 > P40 > 10-3 Faulted (extremely low probability) 10-3 > P40 > 10-6 These probabilities have been assigned to establish the appropriate structural design safety margins for these loading criteria. A summary of these criteria is shown in the table listed below.
Deformation Limit Table C.2-1 Primary Stress Limit Table C.2-2 Buckling Stability Limit Table C.2-3 Fatigue Limit Table C.2-4
BFN-27 C.0-8 There are many places where, through the exercise of designer judgment, it is unnecessary to actually carry out a formal analysis for each of these limits. A simple example consists of the case where two pieces of pipe of differing wall thicknesses are joined at a butt weld. If they are both subjected to the same loading, only the thinner piece would require a formal analysis to demonstrate that the primary stress limit has been satisfied.
The term SFmin that appears in the tables is similar to the classical definition of a minimum safety factor on load or deflection. SFmin is related to the event probability by the following equation:
SFmin = 9 (Equation A) 3 - log10 P40 where 10-1
> P40 > 10-5 (Equation A applies) 10-5
> P40 > 10-6 (SFmin = 1.125) 1.0 > P40 > 10-1 (SFmin = 2.25)
These expressions show the probabilistic significance of the classical safety factor concept as applied to reactor safety. The SFmin values corresponding to the current governing accident event probabilities are summarized as follows:
Governing Load-P40 SFmin Item ing Conditions Upset N and AD 10-1 2.25 or N and U 10-1 2.25 Emergency N and R 10-3 1.5 or N and Am 10-3 1.5 or other 10-1 to 2.25 to 10-3 1.5 Faulted N and Am and R 1.5 x 10-6 1.125 or other 10-3 to 1.5 to 10-6 1.125 where:
BFN-27 C.0-9 N = normal loads U = upset loads excluding earthquake AD = Operating Basis Earthquake including any associated transients Am = Design Basis Earthquake including any associated transients R =
Any pipe rupture loading including any associated transients The minimum safety factor decreases as the event probability diminishes; and if the event is too improbable (incredible = P40 < 10-6),
no safety factor is appropriate or required.
C.3 Piping and Pipe Supports Piping design criteria for Class I piping and tubing, with the exception of torus attached piping, are in accordance with USAS-B31.1.0-1967 (Reference 1). Since this early code is incomplete relative to plant operating conditions and code equations, the later ASME Section III code has been used in the development of load combinations and allowable stress criteria.Section III of the 1971 ASME Boiler and Pressure Vessel Code, including the Summer 1973 addenda (Reference 2), are used as guidance. However, analysis parameters, such as material allowable stresses, Stress Intensification Factor (SIF), coefficient of thermal expansion, elastic modulus, etc., are in accordance with the USAS-B31.1.0-1967 Code (Reference 1).
The design criteria for torus attached piping systems are described in Section C.3.5.
Pipe support design criteria for Class I piping and tubing supports are based on the AISC Manual of Steel Construction (Reference 4), in conjunction with the Manufacturers Standardization Society (MSS)
SP-58 (Reference 5) as described in Section C.3.6.
C.3.1 Large Bore Piping Large bore piping is defined as all Class I, 21/2-inch Nominal Pipe Size (NPS) and larger, process piping which is not subject to Main Steam Relief Valve (MSRV) and postulated loss of coolant accident (LOCA) hydrodynamic loads due to torus excitation.
C.3.1.1 Analytical Models of Piping Systems Analytical models of rigorously analyzed large bore Seismic Class I Piping Systems are consistent with as-built configurations in accordance with the requirements of References 13 and 14.
BFN-27 C.0-10 Large bore Class I piping systems are analyzed by rigorous analysis, which is a detailed analysis of the piping system, generally computer-aided, to assure that the system or design and support locations meet all code requirements. Computer programs used are described in Section C.3.7.
The continuous piping system is mathematically idealized as an assembly of elastic structural members connecting discrete nodal points. Nodal points are placed in such a manner that force deformation characteristics for piping elements such as straight runs of pipe, elbows, etc., can be formulated.
System loads such as internal pressure, weight, thermal expansion, fluid transients, and inertia forces are applied at the nodal points. The stiffness matrix of the interconnecting members is computed and modified to account for flexibility characteristics specified in USAS-B31.1.0 (Reference 1).
Branch lines off the header piping are either modeled and analyzed as part of the header or decoupled from the header if its moment of inertia is less than 1/25 of that of the run pipe. For decoupled branch line analysis, header responses are considered. The header pipe analysis includes the applicable Stress Intensification Factor (SIF) due to branch pipe.
C.3.1.2 Seismic Analysis of Piping Systems The seismic analysis of large bore piping systems is performed by the response spectrum method. The damping is 0.5 percent of critical for both OBE and DBE. The seismic input for both horizontal and vertical directions is developed from structural analysis using artificial (Housner) time history described in Section 2.5.4.
The Amplified Response Spectra (ARS) developed from the artificial time history are broadened 10 percent for rock supported structures and 15 percent for soil supported structures. The seismic loading considers the dynamic inertia response of the system based on the ARS and the effects of differential Seismic Anchor Movement (SAM) of the structure to which the system is attached.
The rigorous analyses using ARS consider all modes of vibration below 20 Hz as flexural modes. Modes of frequencies 20 Hz and above are considered as rigid modes. Rigid mode responses are computed by using
BFN-27 C.0-11 the maximum Zero Period Acceleration (ZPA) as applicable to the building floor (structure) to which the system is attached.
Response spectra analysis is based on either uniform (enveloped) or Independent Support Motion (ISM) techniques. In case of uniform motion, the spectra input is developed by enveloping all applicable spectra of the system attachment points for each direction. When ISM technique is used, zonal responses are combined by absolute summation.
The flexural mode responses are combined by the Square Root of Sum of Squares (SRSS) method except for closely spaced modes (frequencies within 10 percent of each other) which are combined by absolute summation. Rigid mode is combined by SRSS method with flexural modes.
The spatial responses are combined by absolute summation of either East-West or North-South responses with Vertical responses. The net seismic responses are obtained by enveloping the East-West/Vertical and North-South/Vertical combinations.
SAM (including decoupled branch lines) effects are combined with either seismic inertia or thermal expansion loads. When combined with seismic inertia loads, absolute method with uniform technique and SRSS method with ISM technique is used. When combined with thermal expansion loads, the effects are combined by absolute summation.
C.3.1.3 Load Combinations and Acceptance Criteria The load combinations and allowable stress limits for the Class I piping are presented in Tables C.3-1A, C.3-1B, and C.3-1C. These load combinations are categorized in terms of Test, Normal, Upset, Emergency and Faulted conditions as defined in Section C.2.2. Additional definitions are listed below.
Sustained Loads: Effects of live weight (weight of fluid being handled or of fluid used for testing or cleaning) and dead weight (weight of piping, insulation, and fluid or other loads permanently imposed on piping).
Fire Event Loads: Effects of fire events. Loads include sustained loads and internal pressure.
Fluid Transients: Effects of fluid transients including steam hammer, water hammer (excluding check valve slam), and main steam relief valve actuation.
Differential Settlement: Effects of loads on piping due to movement caused by adjacent soil and building structures or by relative settlement of buildings.
BFN-27 C.0-12 C.3.2 Small Bore Piping Small bore piping is defined as all Class I, 2-inch Nominal Pipe Size (NPS) and smaller, process piping which is not subject to main steam relief valve actuation and LOCA hydrodynamic loads due to torus excitation.
The Control Rod Drive Hydraulic (CRDH) piping system, which consists of 185 1-inch NPS insert and 185 3/4-inch NPS withdrawal lines in generally bundled arrangements, is a unique group of small bore piping. Qualification of small bore piping other than the CRDH piping is described in Section C.3.2.1.
Qualification of CRDH piping system is performed separately as described in Section C.3.2.2.
Small bore piping lines which have been modeled and rigorously analyzed as part of the large bore piping system are excluded from the scope of this section. The details of the rigorous analysis for small bore and CRDH piping are consistent with the large bore piping analysis described in Section C.3.1.
C.3.2.1 General Small Bore Piping Class I general small bore piping is qualified by field verification, evaluation, and analysis using a generic attribute methodology to meet the same criteria of load combination and allowable stress as described in Section C.3.1.
Qualification of pipe supports for Class I small bore piping is described in Section C.3.6.2.
The Class I small bore piping qualification program performed rigorous analyses on a representative sample of problems in accordance with the methodology described in Section C.3.1. The sample size is approximately ten percent (10%) of the total program scope. The sample problems were selected from the as-built Class I small bore piping and pipe supports that are representative of the critical loading conditions and plant locations.
The 10 percent sample problem attributes were applied in evaluating the remaining 90 percent of small bore piping within the program scope. Small bore piping can also be qualified by rigorous analysis.
Design changes within the bounds of the LTTIP are qualified in accordance with Section C.3.5. Class I small bore piping design changes outside these bounds are qualified in accordance with large bore piping criteria (Section C.3.1) using rigorous analysis or equivalent static analysis techniques.
BFN-27 C.0-13 C.3.2.2 Control Rod Drive Hydraulic Piping The Control Rod Drive Hydraulic (CRDH) system consists of 185 1-inch diameter insert and 185 3/4-inch diameter withdrawal lines routed in generally bundled arrangements with up to 100 pipes per bundle. It was not necessary to explicitly analyze each line; however, rigorous analysis has been performed for each of the typical configuration groups of lines which represent the range of line configurations within each bundle.
Lines are arranged in groups based on similar geometry, size, and span length to enable justification of typical configurations for analysis. These typical configuration lines have been analyzed to assure that both maximum primary and primary plus secondary loads have been evaluated for pipe stress levels and for each CRDH support frame.
CRDH insert and withdrawal lines are attached to the CRD housings.
Analytical models and seismic input of CRDH lines are consistent with the TVA commitment to complete installation of seismic lateral restraints on the CRD housings prior to restart (References 3, 35), as described in Section C.4.1.6.
Support reaction forces for all CRDH pipes have been compiled based on the typical line analyses. These reaction forces are then combined for qualification of the CRDH support frames as described in Section C.3.6.2.2.
CRDH insert and withdrawal pipe guides have been removed from a few locations on the support frames in order to accommodate thermal expansion of the pipes and the drywell vessel. Seismic pipe stresses and support loads are calculated by omitting these unidirectional supports in the piping dynamic models. However, seismic support loads and local pipe stresses due to impact of the pipes at the unidirectional support points are included when the piping models indicate that impact will occur.
These unidirectional loads are included in the reaction forces applied to the CRDH support frames as described in C.3.6.2.2.
Load combinations and allowable stress criteria for CRDH piping are described in Table C.3-1C. CRDH piping stresses are maintained below these allowables. Interface loads with CRD housing, drywell penetrations, and CRD hydraulic control unit components are also maintained within established limits.
BFN-27 C.0-14 C.3.3 Instrument Tubing Class I tubing required for the safe shutdown is qualified by field verification, evaluation, and analysis using generic attributes to meet the same methodology and criteria as specified in Section C.3.1. Qualification of tubing supports is described in Section C.3.6.3.
The tubing qualification program performed rigorous analyses on a sample of representative tubing problems consisting of approximately 25 percent of the total program scope. The sample problems were selected from the as-built tubing and supports on a plant wide basis. The details of the rigorous analysis for tubing are consistent with the large bore piping analysis described in Section C.3.1. The 25 percent sample problem attributes were applied in evaluating the remaining 75 percent of the tubing within the program scope.
Design changes for tubing within the bounds of the LTTIP are qualified in accordance with Section C.3.5. Class I tubing design changes outside these bounds are qualified in accordance with large bore piping criteria (Section C.3.1) using rigorous analysis or equivalent static analysis techniques.
C.3.4 Buried Piping Buried Class I piping has been qualified by rigorous analysis of representative models from all the systems containing buried piping. Similar configurations and embedment depths exist in each system. Rigorous analyses have been performed for worst case models from each system containing critical components such as elbows and tees along various depths and pipe diameters.
Analysis Qualification of Class I buried piping is based on BFNP site specific geotechnical and seismic input data. Effects of surrounding soil on piping is simulated by horizontal and vertical soil springs. The spring rate and spacing requirements are according to References 9, 10 and 36.
Bounding analyses of all buried piping configurations are performed by using TPIPE computer program. Thermal, seismic, internal pressure, overburden pressure, and differential movement loading conditions are applied and qualified to satisfy the criteria depicted in Table C.3-1A.
Seismic analysis is done by statically applying the axial strain, which is calculated from the peak ground velocity and Raleigh wave velocity values
BFN-27 C.0-15 as established in References 10 and 11. The bending strain is ignored as it is negligible.
Class I buried piping at penetrations into the secondary containment, entry points into the intake structure, and penetrations into other structures are analyzed for the differential movements of the soil and the structure.
Typically, the analysis of piping within a structure includes a portion of the buried piping to a length sufficient enough to simulate the effects of an anchor. In some cases, the soil structure interface is protected from the effects of differential movements by using flexible couplings and/or guard boxes.
C.3.5 Torus Attached Piping Systems Torus attached piping systems, as defined in Section C.2.3, are within the scope of the BFN Long Term Torus Integrity Program (LTTIP) design criteria.
These systems are qualified to withstand the hydrodynamic loads associated with postulated loss of coolant accident (LOCA) loads and main steam relief valve discharges, seismic, static, and thermal loads defined in References 6 and 19. Structural qualifications, modifications, and design criteria for the torus attached piping systems, including the main steam relief valve discharge piping systems, are presented in sections 4, 7, 8, Appendix A, and Appendix B of the LTTIP Plant Unique Analysis Report (PUAR),(Reference 12). The NRC safety evaluation report for the LTTIP is Reference 22.
Some refinements and clarifications were made to the torus attached piping systems design criteria and associated methods after the LTTIP SER (Reference 22) was issued in May 1985. These changes neither increase allowable stresses nor reduce structural margins relative to the acceptance criteria considered in LTTIP SER. Significant changes include:
- 1.
Deflection limits were added for rigid pipe supports which are not attached to the torus and not included on the piping model. The deflection limit for supports existing on 7-31-87 is 5/32-inch, when normalized loads are applied. The corresponding limit for supports added after 7-31-87 is 1/8-inch. Normalized loads are obtained by dividing the piping analysis loads by the stress factors tabulated in LTTIP PUAR section 4.3.4.1.
- 2.
Friction loads due to piping deadweight and thermal expansion effects were added to the load combinations for qualification of rigid supports which act as pipe guides. A friction factor of 0.3 is used and the friction load is considered in support qualification if the pipe thermal movement in the unrestrained direction is 1/16-inch or more. Friction
BFN-27 C.0-16 loads are considered in deadweight plus thermal expansion (primary plus secondary) loading combinations which exclude dynamic loads.
- 3.
An additional loading condition for hydrostatic testing was added.
Normal (service level A) stress limits are applied for this testing condition.
- 4.
A fire event was added to the emergency condition (service level C) load combinations. This new load combination excludes any dynamic loading. It involves pressure plus deadweight effects only.
- 5.
Clarification was added that pipe support gaps up to 1/16-inch are considered in some thermal expansion/contraction analyses.
However, pipe support flexibility is not considered in these analyses unless the supports are directly attached to the torus and included in the piping models for all loading conditions.
- 6.
The active valve list was updated to add some new valves to the list.
The new active valves are evaluated by the same criteria as the existing active valves listed in PUAR Table 2-3 (i.e., the criteria in PUAR section 4.3.3).
- 7.
Interface requirements between torus attached piping and other piping (e.g., large bore piping) were clarified. When a torus attached piping model terminates in a lapping zone with other Class I piping, loads and stresses calculated from the separate analytical models are enveloped within the lapping zone and LTTIP criteria is satisfied for the enveloped loads and stresses.
BFN design criteria documents and engineering procedures control the qualification of torus attached piping systems for design changes.
Compliance with these documents ensures that the allowable stresses, deflection limits, nozzle load limits, and other interface limits described in LTTIP PUAR section 4, appendix A, appendix B, and the clarification/refinements described above are satisfied.
Discrepancies between the initial as-designed and as-built conditions of LTTIP torus attached piping system modifications, described in the PUAR, were identified by reinspections and corrected. Those modifications now comply with engineering requirements assuring compatibility with the LTTIP design criteria.
The Emergency Core Cooling System (ECCS) suction strainers were initially modified to satisfy LTTIP structural design criteria requirements as described in PUAR Section 8.5.2 and Figure 8-4. Later, in response to
BFN-27 C.0-17 NRC Bulletin 96-03 concerning potential plugging by debris (Reference 42),
those strainers were replaced by larger, more functionally efficient strainers designed by GE.
The replacement strainers are securely fastened to the previously existing 30-inch diameter flanges by twenty-four 3/4-inch bolts. Each bolting flange face is located approximately one foot inside its associated ECCS suction penetration.
The replacement ECCS suction strainers and associated header/piping and penetrations were structurally qualified to LTTIP design criteria and analytical methodology described in the PUAR with the following refinements and clarifications:
- 1. The ECCS suction header/piping models were modified to simulate the added mass of the new strainers and associated water mass. Strainer stiffness was simulated based on structural properties determined from a detailed model of the strainers. The effective water mass was determined based on GE Research & Development (R&D) test data for the strainer.
- 2. Applied hydrodynamic drag loads for the LOCA and Main Steam Relief Valve (MSRV) LTTIP load cases were defined by extrapolation of applied drag loads for the previously existing strainers. Hydrodynamic drag load factors were based on comparison of the size, location, hydrodynamic mass, and drag coefficients for the replacement and previously existing strainers. The effective hydrodynamic masses and drag coefficients for the replacement strainers were based on GE R&D test data.
- 3. LOCA and MSRV drag load responses for the new strainers were determined by multiplying the applied hydrodynamic drag loads by Dynamic Load Factors (DLFs) in three orthogonal directions. LOCA Pool Swell DLFs were maintained at 2.0. LOCA Condensation Oscillation (CO)/Chugging and MSRV DLFs were calculated based on characteristic frequencies of the submerged strainers mounted to the ECCS suction header/piping models. Load combination techniques for the harmonic source and fluid structure interaction CO and Chugging drag loads on the strainer were per LTTIP criteria requirements (PUAR Sections 4.2.3 and 4.2.4).
- 4. Load reduction ("knockdown") factors for the single and multiple main steam relief valve torus dynamic response effects were justified based on correlation of the test results from the LTTIP inplant MSRV tests described in Appendix C of the PUAR with the ECCS suction header/piping analyses for those conditions. These load reduction
BFN-28 C.0-18 factors conservatively account for increased MSRV flow rates due to 3%
setpoint tolerance and operation at 3952 MWt, plus a four-inch increase in maximum pressure suppression pool level. The MSRV load reduction factors were applied to the torus dynamic response MSRV inputs to the ECCS suction header/piping and associated torus penetrations analyses.
They were not applied to MSRV hydrodynamic drag loads.
- 5. Compliance of the ECCS header/piping systems and penetrations with LTTIP structural criteria (PUAR Section 4.3) was demonstrated for the updated models. No additional modifications were required to the header/piping systems or penetrations.
- 6. Structural integrity of the replacement strainers was demonstrated for enveloping loads. Strainer stresses comply with stresses from ASME Section III, 1989 Edition allowable stresses. Service levels for the various LTTIP load combinations were conservatively established based on the Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide (PUAR Reference 13) Table 5.1 "Class MC Components and Supports".
BFN LTTIP design criteria documents have been changed to: 1) Require the definition of MSRV load reduction factors for the ECCS suction header/piping systems, penetrations and strainers; 2) establish the allowable stress criteria for the new strainers; and permit use of the new strainer drag coefficients and hydrodynamic masses based on test data for the strainers.
These changes are justified by supporting test/analysis correlation and by compliance with the intent of NUREG 0661 (Reference 6) and the applicable Mark I Containment Program documents. The other methodology changes described above are permitted by the LTTIP design criteria and PUAR Section 4.
These design criteria changes are limited to the replacement strainers, installed in response to concerns identified by NRC Bulletin 96-03 (Reference 42), and the associated ECCS suction header/piping and penetrations. PUAR Section 4.2.2.1 included a provision to allow use of the MSRV inplant test results to "address future NRC concerns regarding the BFN containment system." Definition and application of the MSRV load reduction factors, as described above, is consistent with that PUAR provision which was considered in the LTTIP SER (Reference 22). It is also consistent with NUREG-0661, Appendix A, Section 2.13.9, "MSRV Load Assessment By In-Plant Tests." Therefore, the design criteria changes neither increase allowable stresses nor reduce structural margins relative to the acceptance criteria considered in the LTTIP SER.
BFN-27 C.0-19 C.3.6 Pipe and Tubing Supports Pipe/tubing supports, except LTTIP pipe supports, for Class I piping and tubing are designed based on the AISC manual of Steel Construction (Reference 4), in conjunction with the Manufacturer's Standardization Society SP-58 (Reference 5). Pipe/tubing supports are classified into the three general categories with respect to its load combination and stress allowables as following:
Linear Support - Any support which resists load essentially through a single component of direct stress. These supports provide resistance to movement of the pipe in a particular direction or directions from all load sources. A linear support is any variety of restraint configurations designed and fabricated from structural shapes and plates.
Dynamic Snubber - Provides resistance to dynamic movement without restricting gradually applied motion (e.g., piping thermal expansion) in the direction specified.
Component Standard Support - A support assembly consisting of one or more units which are catalog items and generally mass produced.
C.3.6.1 Large Bore Pipe Supports Class I pipe supports installed on Class I large bore piping limit deflections from any one or all of the applicable load and movement sources as follows:
DW -
Deadweight of sustained loads (includes applicable fluid weight in test condition or fire event)
E1 - Operational basis earthquake (OBE)
E2 - Safe shutdown earthquake (SSE)
Ti
- Thermal mode i=1, 2, --- (includes directional anchor movements)
VT - Valve thrust (relief forces)
WH - Waterhammer S1 - OBE anchor movements S2 - DBE (SSE) anchor movements PR - Pipe rupture C.3.6.1.1 Large Bore Pipe Supports Design Criteria Analysis The design load combinations and allowable stresses for the Class I large bore pipe supports are presented in Table C.3-2. These design load
BFN-27 C.0-20 combinations are categorized with respect to hydrotest, normal, upset, emergency, and faulted conditions as defined in Section C.2.2. The basic computer programs used for large bore pipe support analysis are described in Section 3.7.
Concrete Anchors Concrete Anchors for pipe supports are ductile type or expansion anchors.
Ductile anchors at BFN are predominantly welded studs or regular length undercut anchors. To ensure that the ductile anchor capacity is controlled by the anchor steel capacity, the allowable load is limited to one-fourth the anchor concrete pullout capacity.
Expansion anchors transfer loads to the concrete by expanding laterally against the side of a hole drilled in hardened concrete. Expansion anchors at BFN are predominantly wedge bolts and expansion shell anchors. The tensile allowable loads for all support loading conditions for the wedge bolts and expansion shell anchors are limited to one-fourth and one-fifth, respectively, of the anchor concrete pullout capacity.
C.3.6.2 Small Bore Pipe Supports C.3.6.2.1 General Small Bore Pipe Supports The design criteria for all Class I small bore pipe supports, other than LTTIP and the CRDH small bore pipe supports, is the same as that for Class I large bore pipe supports as described in Section C.3.6.1.1. (See Section C.3.2.1).
C.3.6.2.2 CRDH Pipe Supports For Control Rod Drive Hydraulic (CRDH) insert and withdrawal piping supports, in addition to the load combinations listed in Table C.3-2 the following load combinations are also evaluated.
Design Load Load Combination Direction Combination Upset
+
DW + T2(+)
DW + T2(-)
Emergency
+
DW + T3(+)
DW + T3(-)
BFN-27 C.0-21 where:
T2 is abnormal scram thermal mode, and T3 is normal scram with post-LOCA thermal mode.
Normal full scram thermal loads are combined with seismic loads in the upset, emergency, and faulted load combinations of Table C.3-2.
Since the CRDH piping are routed in generally bundled arrangements as described in Section C.3.2.2, special types of CRDH pipe supports are defined as follows.
Rack - A two-dimensional frame which provides support/restraint for a CRDH pipe bundle.
Bar - An individual member in a support rack which provides direct support/restraint for a row of pipes within a CRDH pipe bundle.
The following methodology is used to determine and evaluate CRDH piping support stresses and stress-related load effects, thereby accounting for the fact that the peak seismic forces from multiple individual pipes will not occur simultaneously.
- a. To calculate support stresses due to pipe seismic restraints on a pipe bundle support rack and on each individual support bar in that rack, the stresses caused by application of individual pipe peak seismic inertia forces in each separate orthogonal restraint direction are combined by a factored Absolute Sum (ABSUM) method. By this method, the support stresses due to individual pipe seismic inertia forces in an orthogonal restraint direction are multiplied by a factor and then combined by ABSUM. The multiplication factor is 0.5 for pipe seismic inertia (not impact) forces on racks which support 23 pipes or more; it is 0.75 for pipe seismic inertia forces on racks and individual bars which support from 9 to 22 pipes; and it is 1.0 for pipe seismic inertia forces on racks and individual bars which support from 1 to 8 pipes.
Pipe seismic impact force effects and the net (combined) pipe seismic inertia force effects are combined by the SRSS method.
The net (combined) pipe seismic inertial and impact force effects are combined with the support frame seismic self-weight excitation effects by the ABSUM method.
BFN-27 C.0-22 This process is repeated for each orthogonal restraint direction.
- b. Support frame stresses from seismic forces applied in each separate orthogonal restraint direction are combined by ABSUM. Similarly, seismic stresses from multiple racks in an overall support frame are combined by ABSUM. This is done either directly or by simultaneous application of the forces, for each rack and each orthogonal direction, to the overall frame in the directions which maximize overall frame seismic stresses and displacements. Each net (combined) seismic force effect is the larger of the North-South plus Vertical, or East-West plus Vertical seismic input effects. The seismic support stresses and displacements all have a
+ and - value for combination with static load effects.
- c. Individual pipe forces are applied to the support frames for each static load case (deadweight, normal scram thermal, abnormal scram thermal, and post-LOCA thermal), and stresses are calculated for each case by algebraic summation.
- d. Static and seismic frame stresses are combined and compared to the associated allowable stresses for each load combination defined by Table C.3-2 and this Section, considering the directional sense of each static stress and the + and - values of the seismic stress. Both general frame and local bar stresses are evaluated in this manner.
In addition, the combined deflection from seismic pipe forces, seismic self-weight excitation of the frames, and dead-weight of the pipes are limited to 1/8-inch at the point of pipe inertial load application for each direction of restraint.
- e. Each pipe clamp and guide is evaluated for the combined effects of the peak seismic and static forces for each load combination. The forces in each orthogonal direction of restraint are applied simultaneously and compared to the rated capacity of the clamp/guide.
The design criteria for CRDH pipe supports, with the exceptions that above additional load combinations are added to those specified in Table C.3-2, as well as the special evaluation methodology and provisions described above, are the same as that for Class I large bore pipe supports as described in Section C.3.6.1.1.
C.3.6.3 Tubing Supports The design criteria for Class I tubing supports is the same as that for Class I large bore pipe supports as described in Section C.3.6.1.1.
BFN-27 C.0-23 C.3.7 Computer Programs for Class I Piping System Analysis The following is a list of the principal computer programs used for dynamic and/or static analysis of Class I piping and pipe supports. Each program's scope, background, applicability, and method of validation is discussed in the program descriptions below. As required, additional computer programs are used to support these analyses.
Program Name Application TPIPE Static and Dynamic ANSYS Static, Dynamic and Non-linear FAPPS Static GTSTRUDL Static and Dynamic TPIPE--for the linear elastic structural analysis of arbitrary, 3-dimensional piping systems subject to static and dynamic loadings. Analyses are performed to requirements for ASME Classes 1, 2, 3 systems, and ANSI B31.1.0 Power Piping Code.
Piping system is idealized as a mathematical model consisting of lumped weights connected by weightless elastic members. The locations of the lumped weights are chosen to adequately represent the dynamic characteristics of the system for dynamic considerations. The direct stiffness method of structural analysis is used to form the stiffness matrix, including stiffness modifications for curved components.
Diagonal mass and damping matrices are assumed. The equations of equilibrium are solved to determine the system displacements, and hence member forces and moments for the applied loading and/or displacements, using a Gaussian elimination procedure.
TPIPE analyzes piping systems subject to applied static loading conditions using the method discussed in the preceding paragraph; however, the piping dead load analysis considers both distributed weight properties of the piping and any concentrated weights.
TPIPE analyzes piping systems for dynamic excitation using the analysis technique known as the response spectrum modal superposition method.
A direct integration or modal superposition time history capability is also available. Seismic options include a multiple support zone capability or
BFN-27 C.0-24 independent support motion (ISM) technique for response spectrum analysis. The dynamic properties of the system (periods of vibration and normal mode shapes) are determined using a modified subspace iteration technique, and the system response is then computed by the modal superposition procedure. The seismic analysis capability includes the contribution of rigid modes (ZPA effect).
TPIPE has been benchmarked against the NRC program EPIPE in accordance with the Standard Review Plans, NUREG-0800, Section 3.9.1.II and NUREG/CR-1677. TPIPE is verified and maintained by using formal software QA procedures.
ANSYS - The ANSYS computer program is a large-scale general purpose computer program for the solution of several classes of engineering analysis problems. ANSYS is capable of analyzing structures with static and dynamic loadings, elastic and plastic member properties, creep and swelling, buckling, and small and large deflections.
The matrix displacement method of analysis based upon finite element idealization is employed throughout the program.
FAPPS - The Frame Analysis Program for Pipe Supports (FAPPS) is an inter-active computer program specifically developed for the analysis and design of 18 standard frames as well as any non-standard frame for pipe support. It optimizes member sizes, welds, base plates, and embedments based upon various user specified design limitations.
The process of optimization of member sizes is controlled completely by the user to achieve an economical solution. In this way the user can limit the range of the member size selection process, subject to the available shapes, consideration of members connectivity and installation feasibility, and also the possible anticipation of future additional loads or revised design loads.
The FAPPS program also has the flexibility to perform normal load condition code checks for either the AISC, ASME Section III Subsection NF or AIJ codes.
The FAPPS also performs code check for upset, emergency, and faulted load conditions.
The FAPPS allows use of various types of load sets for simplification of input to allow algebraic, absolute, and/or SRSS combination of
BFN-28 C.0-25 results due to each load vector within a load set as well as each load set that is to be combined in one load set.
FAPPS is verified and maintained by using formal software QA procedures.
GTSTRUDL provides the ability to specify characteristics of structural problems, perform analyses, reduce and combine results, perform design, and output any part or all of the information stored in the structural problem data based on a selective basis.
Analytical procedures apply to any combination of framed structures and continuum mechanics problems of arbitrary configuration and composition. Force boundary conditions on member ends, and force and displacement boundary conditions at support joints, may be specified implicitly by means of structural type and orientation commands, or explicitly for a member or joint. Continuum mechanics problems are treated using the finite element method in which the domain of the problem consists of different shapes connected at a finite number of joints.
GTSTRUDL permits elements (members and finite elements) of different types to be mixed in the same problem solution whether they have the same or a different number of degrees-of-freedom per joint.
Properties of member elements may be specified by providing section properties of prismatic or variable section members, naming a section from a pre-established table of properties (such as "W14X237"), or specifying flexibility of stiffness matrices for special member elements.
GTSTRUDL analysis procedures perform linear small displacement static and dynamic analysis of structures composed of any combination of member and finite elements with the same or variable number of degrees-of-freedom per joint.
GTSTRUDL design procedures include steel design and code checking for member elements by the 1969 and 1978 AISC (American Institute of Steel Construction) Specifications for general steel structures.
GTSTRUDL is verified and maintained by using formal software QA procedures.
BFN-28 C.0-26 C.4 Major Components C.4.1 Reactor Pressure Vessel (RPV), RPV Internals, and Supports General Electric Company (GE) originally designed and qualified the BFN RPVs, RPV internals, and supports as described in Sections C.4.1.1 through C.4.1.5. In 1989, an upgraded seismic analysis was performed and a reassessment of the combined loading effects was made as described in Section C.4.1.6. In 1997, an additional reassessment was made for operation at 3458 MWt which included the current seismic loads. The steam dryers have been replaced to support EPU operating conditions. The replacement dryers (Ref. 55), support brackets, and vessel shell (Ref. 56) have been requalified for EPU operating conditions and the increased weight of the replacement dryer. Stresses and loads remain within design basis allowables, as indicated by Table C.4-1. In 2015, a reassessment was performed for operation at 3952 MWt and is presented in References 50, 51, 52, and 54.
C.4.1.1 RPV Stress Analysis The RPVs were designed, fabricated, inspected, and tested in accordance with the ASME Boiler and Pressure Vessel Code,Section III, its interpretations, and applicable requirements for Class A vessels as defined therein, as of the date that the reactor vessel order was placed (Reference Appendices J, K, and L for BFN 1, 2, and 3, respectively).
Stress analysis requirements and load combinations for the RPVs were evaluated for the cyclic conditions expected throughout the 40-year life, with the conclusion that ASME code limits are satisfied.
The 60 year operating life has been evaluated as a Time Limited Aging Analyses (TLAA). The summary of these evaluations are provided in Appendix O, Sections O.3.2.1 and O.2.4.
The RPV design report (original Appendix J that was redocketed by Reference 23 on June 23, 1989) provides the results of the original detailed design stress analyses performed for the RPV to meet the code requirements. Selected RPV components considered to have possibly higher than code design primary stresses, as a result of rare events or a combination of rare events, were also analyzed in accordance with the requirements of the loading criteria in this section and the safety margins in Section C.2.6.
BFN-28 C.0-27 Analyses were performed for the replacement steam dryer bracket load combinations. General and local primary membrane and bending stress intensities for the RPV Shell and Bracket were evaluated using a finite element model (Ref. 56).
The reactor vessel stress and fatigue evaluation for operation at 3952 MWt is presented in Reference 55, with the conclusion that the code for affected components continues to be satisfied.
Results (critical load combinations, locations, and allowables) for the most critical of those original analyses are included in Table C.4-1. The allowables were met in all cases.
C.4.1.2 RPV Fatigue Analysis The analysis of the RPV shows that all components are adequate for cyclic operation by the rules of Section III of the ASME Boiler and Pressure Vessel Code. The critical components of the vessel are evaluated on a fatigue basis, calculating cumulative usage factors (ratios of required cycles to allowed cycles-to-failure) for all operating cycle conditions. The cumulative usage factors for the critical components of the RPV are below the code allowable of 1.0.
Reference 50 documents the fatigue analysis for affected RPV components for operation at 3952 MWt. The analysis considers that effects of environmentally-assisted fatigue for a 60-year unit life. The cumulative usage factors for the critical components are below the code allowables.
C.4.1.3 RPV Internals and Supports Stress Analysis The RPV internals are designed using Section III of the ASME Boiler and Pressure Vessel Code as a guide. The material used for fabrication of most of the internals is solution heat-treated, unstabilized, Type 304 austenitic stainless steel conforming to ASTM specifications.
Allowable stresses for the internals materials under normal operating conditions are taken directly from Section III.
For rare events or a combination of rare events, the RPV internals and supports were analyzed in accordance with the requirements of the loading criteria in this section and the safety margins in Section C.2.6, and results (critical load combinations, locations, and allowables) for the most critical of those analyses are included in Table C.4-1. The allowables were met in all cases.
BFN-28 C.0-28 Analysis of the affected RPV internals and supports for operation at 3952 MWt is documented in References 51 and 52, with the conclusion that the code allowables are met for the affected components.
C.4.1.4 RPV Internals Deformation Analysis C.4.1.4.1 Control Rod System If there were to be excessive deformation of the Control Rod System, made up of the control rod drive, control rod drive housing, control rod, control rod guide tube and fuel channels, and the core structural elements which support them (top guide, core support, and shroud and shroud support), it could possibly impede control rod insertion. The maximum loading condition that would tend to deform these long, slender components is the Design Basis Earthquake. The highest calculated stresses occur where the Design Basis Earthquake and loads resulting from the DBA pipe break are considered to occur simultaneously. Even in these cases the general stress levels are relatively low and for operation at 3458 MWt, control rod guide tube buckling potential is within allowable criteria. No significant deformation is associated with these calculated stresses, therefore, rod insertion would not be impeded after an assumed simultaneous Design Basis Earthquake and pipe break accident. The addition of control rod drive housing lateral restraints (reference Section C.4.1.6) provides added assurance in this regard.
Assessment of the Control Rod System performed for operation at 3952 MWt continues to support the conclusion that rod insertion would not be negatively affected (Reference 51).
C.4.1.4.2 Core Support The core support sustains the pressure drop across the fuel. This pressure drop is the only load which causes significant deflection of the core support. Excessive core support deflection could lift the control rod guide tubes off their seats on the control rod drive housings and hereby increase core bypass leakage. This upward deflection would have to be 1/2-inch to begin to lift guide tubes. The maximum deflections under normal operating conditions and pipe rupture differential pressures for the core support are calculated to be very small as compared to 1/2-inch.
The guide tubes will, therefore, not be lifted off. For operation at 3458 MWt, the core support beams have been evaluated for buckling as a result of the pressure drop and have been determined to have margin within allowable criteria. Analysis performed for operation at 3952 MWt continues to support this conclusion for the core support (Ref. 51).
BFN-28 C.0-29 C.4.1.5 RPV Internals Fatigue Analysis The fatigue analysis was performed using as a guide the ASME Boiler and Pressure Vessel Code,Section III. The method of analysis used to determine the cumulative fatigue usage is described in APED-5460, 'Design and Performance of GE-BWR Jet Pumps,' September 1968. The most significant fatigue loading occurs in the jet-pump shroud support area of the internals.
The analysis was performed for a plant where the configuration (leg-type shroud support) was almost identical to Browns Ferry.
Therefore, the calculated fatigue usage is expected to be a reasonable approximation for BFN.
Fatigue analyses for the replacement steam dryers use the GE Hitachi Nuclear Energy Plant Based Load Evaluation methodology to develop the fluctuating pressure design loads from main steam line acoustic pressure measurements taken at the three BFN units (Ref. 55). The steam dryer fatigue evaluation consists of calculating the alternating stress intensity from flow induced vibration loading at all locations in the steam dryer structure and comparing it with the allowable design fatigue threshold stress intensity limits.
The following loading combinations and transients were considered.
- 1. Normal startup and shutdown;
- 2. Operating Basis and Design Basis Earthquakes;
- 3. Ten-minute blowdown from a stuck main steam relief valve;
- 4. HPCI operation;
- 6. Improper start of a recirculation loop.
Calculated Cumulative Fatigue Usage was less than the allowable of 1.0.
Based on the analysis performed for operation at 3952 MWt, the cumulative fatigue usage for the affected internals is less than the allowable of 1.0 (Reference 56).
BFN-28 C.0-30 Remarks The location of maximum calculated fatigue usage is at the bottom side of the baffle plate at the point where the baffle plate attaches to the shroud in the vicinity of the minimum ligament C.4.1.6 RPV, RPV Internals, and Supports Seismic Analysis GE originally performed a detailed seismic analysis of the RPV, RPV internals, and supports as described in redocketed Appendix J. Loads from other load sources were combined with the seismic loads from that analysis, with the results described in Sections C.4.1.1 through C.4.1.5.
The seismic loads on the RPV, RPV internals, and supports are now based on a dynamic analysis of an upgraded model of the RPV and RPV internals coupled with the Reactor Building as described in Section 12.2.2.8.2. The results of the upgraded seismic analyses were combined with the results of other loads for the various loading conditions in the reassessment documented by Reference 21. All stresses and loads remain within the allowables given in Table C.4-1.For operation at 3952 MWt, stresses and loads for the RPV internals are within allowables (Reference 51) given in Table C.4-1.
The upgraded RPV, RPV internals, and supports seismic model includes consideration of the Control Rod Drive (CRD) housing lateral restraints which were added (References 3 and 35) to limit CRD housing and attached CRD hydraulic piping seismic stresses and displacements.
C.4.2 Primary System Components Stress Analysis GE supplied, specified the allowables, and performed the original qualifications for the primary system components listed in Table C.4-2. Seismic loads used in these qualifications were based on El Centro input motion as described in Section C.2.4.
The extent of GE's stress analyses performed on equipment/components were dependent upon the type of equipment/components and the type of fabrication.
Fabricated shapes are generally made from plate or rolled shapes with uniform thickness and shapes with regular geometric configurations. Cast shapes are generally made with nonuniform material thickness in complicated shapes that are not regular geometric configurations. Manufacturers have traditionally designed cast shapes conservatively since they do not lend themselves to rational analysis. Usually a design is developed based on extensive tests and experience. Selected components considered to have possibly higher than code design primary stress as a result of rare events or a combination of rare events were analyzed in accordance with the requirements of the loading
BFN-27 C.0-31 criteria in this section and the safety margins in Section C.2.6. Results (critical load combinations, locations, and allowables) for the most critical of those analyses are included in Table C.4-2.
Class I large bore and torus attached piping systems have been qualified to satisfy requirements described in Section C.3. As part of this qualification the piping nozzle loads on primary system components are maintained within the allowable nozzle loads described in Table C.4-2.
In addition, primary system component stresses are within the allowable stresses and component thicknesses are greater than the required thicknesses described in Table C.4-2.
C.5 Primary Containment System and Penetrations Each Browns Ferry unit employs a pressure suppression primary containment system which houses the reactor vessel, the reactor coolant recirculation loops, and other branch connections of the Reactor Primary System (RPS). The main functions of the primary containment system are:
- 1) to withstand the pressures resulting from a loss-of-coolant accident (LOCA) and/or main steam relief valve discharge,
- 2) to provide enclosure for the decay of any radioactive material which may ultimately be released,
- 3) to store sufficient water to condense steam released as a result of a LOCA and/or MSRV discharge, and
- 4) to supply water to the Emergency Core Cooling Systems (ECCS).
The primary containment system consists of a drywell, a pressure suppression chamber (wetwell or torus) which stores a large volume of water, a connecting vent system between the drywell and the wetwell, isolation valves, a vacuum relief system, containment cooling systems, equipment for establishing and maintaining a pressure differential between the drywell and the wetwell, and other service equipment. Section C.5.2 provides a more complete description of the primary containment system.
The following C.5 sections describe the structural qualification of the primary containment system except for the torus attached piping systems.
Qualification of torus attached piping systems, including MSRV discharge piping systems, is described in Section C.3.5.
BFN-28 C.0-32 C.5.1 Primary Containment Vessels Stress Analysis The primary containment vessels (consisting of the drywell, torus, vent system, and associated integral penetrations) were originally designed, fabricated, inspected, and tested in accordance with the ASME Boiler and Pressure Vessel Code,Section III, its interpretations, and applicable requirements for Class B vessels as defined therein as of the date that the vessel order was placed.
The Browns Ferry Containment vessels for Units 1 and 2 were built to Section III of the ASME Boiler and Pressure Vessel Code, 1965 edition, and Addenda through Winter 1966, inclusive (Reference 17). The Unit 3 containment vessel was built to Section III of the ASME code, 1965 edition, and Addenda through Summer 1967, inclusive. The ASME code design condition categories presently defined as Normal, Upset, and Emergency were not defined in these editions.
The loading conditions and allowable stresses for the drywell are presented in Table C.5-1. The accident condition allowables are based on 1968 ASME Section III Code with Addenda through Summer of 1969. Stresses in the drywell vessel including its integral penetrations are maintained within these allowables. These primary containment loadings were validated to be acceptable for operation at 3458 MWt (Reference 43) and operation at 3952 MWt (References 52 and 53).
See Sections C.5.3 and C.5.4 for qualification of the torus vent system, non-safety related internal structures, and wetwell/drywell vacuum breakers.
C.5.2 Primary Containment Bellows Stress Analysis The vent pipes bellows were designed, fabricated, and tested in accordance with ASME Boiler and Pressure Vessel Code, 1965 edition to Winter 1966 Addenda, and including Code Case Interpretations 1177-5 and 1330-1.
In accordance with Code Case 1177-5, the membrane stress in the bellows is limited to the tabulated maximum allowable stress value of 15,500 psi for the material (SA-240-T304) at 300F in accordance with the ASME Boiler and Pressure Vessel Code,Section VIII, Table UHA-23.
BFN-27 C.0-33 Pressure tests were conducted on each bellows by sealing the annulus between the bellows and protective guard plate and pressurizing and monitoring the pressure decay for several hours. All bellows were found to be leak-tight.
The bellows-type expansion joints for containment penetrations and vent pipes are designed for an internal pressure of 56 psig at 281F and an external pressure of 2 psig (7 psig in the steam vault room) at 281F. The joints are also designed to permit an axial extension of 0.6 in., an axial compression of 2.0 in., and a lateral offset of from 0.3 to 1.55 in. depending on the elevation of the penetration. The design fatigue life of the joints is 7000 cycles.
The containment penetration expansion joints are designed, fabricated, and tested to meet Interpretations of the ASME Boiler and Pressure Vessel Code, Cases 1177 and 1330, and to meet the standards of the Expansion Joint Manufacturer's Association, Inc. The primary stresses are limited in accordance with Code Cases 1177 and 1330 to ASME Boiler and Pressure Vessel Code,Section VIII (Reference 18) allowable stress intensities. The secondary stresses are limited to the design fatigue life stress values for the expansion joint at 7000 cycles.
The longitudinal butt welds in the expansion joints are radiographed in accordance with the ASME Boiler and Pressure Vessel Code prior to forming.
The welds attaching the bellows elements to the transition elements are made and inspected in accordance with Code Case 1330. The vent pipe bellows were evaluated further for the effects of postulated hydrodynamic loads in the LTTIP (Section C.5.3).
C.5.3 Long Term Torus Integrity Program (LTTIP)
In July 1980, the Nuclear Regulatory Commission (NRC) issued NUREG-0661, "Safety Evaluation Report, Mark I Containment Long-Term Program" (Reference 6) to address the NRC acceptance criteria for the Mark I nuclear plant containment system evaluation on the identified loss-of-coolant accident (LOCA) and main steam relief valve hydrodynamic loads.
The BFN Long Term Torus Integrity Program (LTTIP) reevaluated the plant-specific responses, in compliance with the intent of NUREG-0661. Results of the reevaluation, qualification, and implemented modifications are documented in the BFN-LTTIP Plant Unique Analysis Report (PUAR)
(Reference 12) and the NRC safety evaluation report (Reference 22). This program addressed the torus (wetwell), vent system, torus attached piping
BFN-28 C.0-34 systems, and non-safety related internal structures portions of the BFN primary containment system.
LTTIP structural qualification criteria, methods, and modifications for the torus and torus penetrations, vent system and vent pipe bellows, and non-safety related internal structures are described in Sections 4, 5, 6, 9, and Appendix B of the LTTIP PUAR. There have been no significant changes in the criteria or methods described therein since the LTTIP SER was issued in May 1985, except as follows:
- a. Analyses have been performed to account for a maximum analytical pressure suppression pool water level elevation of 536'-10", which is 4 inches higher than the previously analyzed value. The new analytical water level includes consideration of potential instrument error and a margin for future use.
- b. Applied pool swell and vent thrust loads were generated based on a more refined Design Basis Accident blowdown analysis performed using the LAMB vessel blowdown model described in NEDO-20566 (Reference 41).
Structural analysis methods were not changed.
- c. Stresses and loads remain below the acceptance limits considered in the LTTIP SER.
- d. The LTTIP requirements are also demonstrated to be satisfied for operation at 3458 MWt (Reference 43) and operation at 3952 MWt (References 52 and 53).
BFN design criteria documents and engineering procedures control the qualification of these primary containment system components for design changes. Compliance with these documents ensures that the allowable stresses and interface limits described in PUAR Section 4 and Appendix B are satisfied.
Discrepancies between the initial as-designed and as-built conditions of the LTTIP modifications for these components were identified by re-inspection and corrected. Those modifications now comply with engineering requirements assuring compatibility with the design criteria.
See Section C.3.5 for structural qualification of the torus attached piping system.
Changes in LTTIP design criteria and methodology for structural qualification of replacement ECCS suction strainers and associated header/piping systems and penetrations are described and justified in Section C.3.5.
BFN-28 C.0-35 C.5.4 Wetwell/Drywell Vacuum Breakers Wetwell/drywell vacuum breaker dynamic loads associated with the LOCA chugging phenomena were identified during full scale tests for the Mark I Containment Program. Those loads were not included in NUREG-0661 and, consequently, were not addressed by the LTTIP PUAR. They were the basis of NRC Generic Letter 83-08 (Reference 37). The BFN wetwell/drywell vacuum breakers were evaluated and modified for these chugging dynamic loads in response to GL 83-08, as described by Reference 38. The LOCA hydrodynamic loads are also demonstrated to be satisfied for operation at 3458 MWt (Reference 43) and operation at 3952 MWt (References 52 and 53).
C.6 Equipment The general question regarding the adequacy of seismic qualification of safety-related equipment in operating plants has been recognized as an industry wide concern by the nuclear industry. The NRC established this concern in December 1980 as Unresolved Safety Issue (USI) A-46, "Seismic Qualification of Equipment in Operating Plants", (NUREG-0606 and -0705).
It is important to note, however, that the A-46 statement of the issue recognized the industry consensus position that the application of the rules and procedures in existence at the time of operating plant design served to ensure that conservative margins were incorporated into safety equipment design.
In 1987, Browns Ferry Nuclear Plant was identified as one of the operating plants applicable to A-46 requirements. The existing Browns Ferry safety equipment was qualified originally to have sufficient margin of safety to withstand seismic loading. Plant-specific verification of seismic adequacy of equipment was implemented in accordance with References 24, 25, 43, and
- 44.
For historical purposes, the original equipment seismic loading and analysis methods are described in Sections C.6.1 and C.6.2. Qualification of replacements for existing equipment and new equipment from March 1988 to July 2007, while the A-46 Generic Implementation Procedure (GIP) was being developed by the Seismic Qualification Utilities Group (SQUG),
approved by the NRC, and implemented at BFN, is described in Section C.6.3. Seismic/Structural Qualification (S/SQ) of new and replacement equipment after July 2007 is described in Section C.6.4.
BFN-27 C.0-36 C.6.1 Equipment Seismic Loading (Historical)
For GE-supplied equipment, seismic design conditions were included in the purchase specification of seismic Class I equipment. These were in the form of equivalent static seismic coefficients both in the horizontal and vertical directions. Vendor design submittals in this area (as well as all other functional areas) were reviewed for adequacy and applicability by the design engineer. Qualifications of GE-supplied mechanical equipment, which are primary system components, are described in Section C.4.2.
For TVA purchased Class I mechanical equipment, static seismic coefficients were specified for pumps, motors, etc., that were known to have a natural frequency greater than 20 Hz. The DBE building response in the vertical and horizontal directions at the equipment location was specified.
Valves, relays, etc., were specified to withstand seismic loads equal to or greater than resonance response at the equipment location. For equipment whose natural frequency was felt to be less than 20 Hz, TVA required the vendor to determine the natural frequency. Appropriate spectral curves were included in the specifications for equipment suspected of having a low natural frequency ( 20 Hz), and the vendor was required to make a thorough dynamic analysis or test of this equipment.
The support structure of all Class I equipment was designed to adequately protect the equipment it supports.
For TVA-purchased Class I electrical equipment, the following requirements were included in the purchase specifications to assure adequate design and functional integrity under the seismic design conditions.
- 1. The equipment, the devices mounted on it, and its supports shall be designed to withstand seismic forces determined from floor accelerations of (A) g horizontal, and (B) g vertical, provided that all of the natural frequencies of vibration of the equipment are greater than 20 Hz.
- 2. Devices located in equipment or in areas of the equipment which have natural frequencies of vibration less than 20 Hz shall be designed to withstand seismic forces determined from resonant acceleration of (C) g horizontal and (B) g vertical.
- 3. The stresses in the supports and the anchor bolts due to seismic loads shall be combined with the stresses due to other live and dead loads
BFN-27 C.0-37 and operating loads. The allowable stress for this combination of loads shall be based on the ordinary allowable stresses set forth in the applicable codes.
- 4. All equipment shall be anchored or fastened in such a way that it will remain in place when friction is considered nonexistent with the following accelerations:
(A) specified from 0.20g to 1.70g, dependent upon location, (B) specified as 0.134g, and (C) specified from 1.24g to 29.80g, dependent upon location.
Provisions were included in the purchase specification of electrical equipment to ensure that the seismic requirements were satisfied.
The specifications required the vendor to certify that his equipment meets the seismic requirements and to submit a verification report giving the following information.
- 1. Rigid equipment (has no natural frequencies of vibration less than 20 Hz): Show that the lowest natural frequency of more than 20 Hz and that all components on, or in, the equipment will continue to function properly when subjected to the seismic forces as determined from the floor accelerations specified.
- 2. Nonrigid equipment (has natural frequencies of vibration less than 20 Hz): Show that all components on, or in, the equipment will continue to function properly when subjected to the seismic forces as determined from the resonant accelerations specified.
C.6.2 Equipment Seismic Analysis (Historical)
Equipment was typically analyzed statically to determine its response to earthquake loads. The equivalent static coefficients for the equipment were obtained from the floor response spectra corresponding to the support elevations of the equipment. In lieu of determining the natural frequency of the equipment, the peak value of the applicable floor response spectrum was used in calculating the earthquake-induced loads. In cases where the allowable stress limits were exceeded, the corresponding input acceleration was obtained from the appropriate floor response spectra.
BFN-27 C.0-38 Equipment which was outside the reactor coolant pressure boundary (RCPB) used the design conditions of pressure, thermal, and deadweight plus Operating Basis Earthquake (OBE), which was equivalent to Normal plus Upset. Substituting the Design Basis Earthquake for the OBE was equivalent to the Emergency condition. The Faulted condition was not applicable.
C.6.3 Seismic Qualification of New Equipment and Replacements for Existing Equipment From March 1988 Until July 2007 NRC Generic Letter 87-02 (Reference 8) specified that replacements should be "verified for seismic adequacy either by using A-46 criteria and methods or, as an option, qualifying by current licensing criteria." In the interim time period from March 1988 until July 2007, BFN optional equipment seismic qualification criteria for new items and replacements, which were plant modifications, were generally in accordance with NRC Regulatory Guide 1.100 (Reference 20) and IEEE 344-1975 (Reference 32). This optional, current criteria was used for a majority of BFN design changes for Class I equipment. This practice was continued until the A-46 Generic Implementation Procedure (GIP) methodology was included in a licensing basis revision per Section C.6.4.
Alternately, during this interim time period, qualification of Class I replacements and new items were addressed by the following approaches:
- 1. Qualification of a new or replacement item of equipment was accomplished by similarity to an existing installed item. Any dissimilarities were evaluated to show that the new or replacement item was no less capable of withstanding seismic loading than the installed item upon which the qualification by similarity was based.
- 2. Qualification was accomplished by comparison to an identical existing component. The two components were required to meet all of the following conditions.
- a.
The form, fit, function, weight and weight distribution, and materials of construction were identical.
- b.
The parts were from the same manufacturer and had the same model number.
BFN-27 C.0-39
- c.
It was positively shown that the mounting configuration and location of the existing item created a seismic loading condition which was equal to or greater than the seismic loading condition predicted for the new or replacement item.
- 3.
Qualification was accomplished by use of the draft guidelines and criteria of USI A-46 and the associated Seismic Experience Data Base (See References 8, 33, 34) and then verified by A-46 GIP implementation.
In each of these alternative approaches the objective was to assure that the required seismic adequacy was maintained or achieved so that it could be verified later by NRC-approved A-46 criteria. Results of the BFN A-46 GIP implementation activities (References 24, 25, 44, and 45) confirmed that these approaches were successful.
C.6.4 Equipment Seismic/Structural Qualification (ESQ) After July 2007 After July 2007, S/SQ of BFN Class I and Class II electrical and mechanical equipment of all types, including electrical assemblies and devices, electrical conduit and cable tray raceway systems, and mechanical equipment and fluid system components (pump, tank and vessel assemblies, valves, and other in-line fluid system components) is performed in accordance with BFN ESQ design criteria, which replaced the interim design criteria described in Section C.6.3.
The ESQ design criteria requires new Class I equipment to be seismically qualified either by compliance with current criteria (based on IEEE 344-1975, NRC Regulatory Guide 1.100 R1, and applicable ASME codes) or SQUG GIP 3A methods (Reference 46). It also requires modifications to existing Class I equipment to comply with current criteria when there are specific BFN licensing commitments for the existing equipment to comply with IEEE 344-1975 (or 1971). In addition, it requires modifications to existing Class I equipment to comply with current criteria if the existing equipment S/SQ documentation is in accordance with current criteria. Otherwise, GIP 3A methods (Reference 46) are identified as an alternative approach for Seismic Qualification of Class I equipment when applied in accordance with SQUG Implementation Guidelines for Seismic Qualification of New and Replacement Equipment/Parts (NARE) (Reference 47).
Seismic Qualification of Class I equipment includes compliance with applicable criteria for normal operating loads plus seismic loads. However, S/SQ also entails qualification of Class I equipment in some locations for other BFN design basis structural loading conditions. For example, the ESQ design
BFN-27 C.0-40 criteria also requires Class I equipment within the LTTIP boundaries to be qualified for load combinations and acceptance criteria described in the LTTIP PUAR (Reference 12). In addition, Class I equipment, which is exposed to the outside environment, is required to be qualified for BFN design basis wind, tornado missiles, snow, and ice as applicable at the equipment location. The requirements for these additional loading conditions are considered current criteria because those conditions are not addressed by GIP 3A.
The ESQ design criteria requires new Class I electrical conduit and cable tray raceway systems and modifications to existing Class I electrical raceway systems to be seismically qualified in accordance with GIP 3A methods and the SQUG Implementation Guidelines for NARE (References 46 and 47).
Per the ESQ design criteria, new Class II equipment and modifications to existing Class II equipment are seismically qualified by demonstration of structural and pressure boundary integrity by current criteria or by GIP 3A methods.
Replacement items (e.g. parts) for Class I and Class II equipment for plant maintenance (not plant modifications) are verified to ensure that the S/SQ of the replacement item or its host equipment is not degraded. This is accomplished by application of a TVA design standard, in accordance with the ESQ design criteria. The design standard implements the Seismic Technical Evaluation of Replacement Items (STERI) process per References 48 and 49.
It ensures that S/SQ of the equipment is maintained (not degraded) in accordance with current criteria and the SQUG Implementation Guidelines for NARE (Reference 47).
C.6.5 Qualification of Equipment in Torus Attached Piping Systems Equipment in torus attached piping systems is also qualified to the requirements defined in section 4.3 of the LTTIP PUAR (Reference 12).
Section C.3.5 of this Appendix describes the qualification of torus attached piping systems.
C.6.6 Interface Loads from Class I Piping Analysis Interface loads between Class I equipment and piping which is rigorously analyzed, as described in Section C.3, are maintained within acceptable limits justified by TVA, TVA-contractors, or equipment vendors. For example, nozzle loads on primary system components are maintained within allowables described in Section C.4.2 and Table C.4-2.
BFN-27 C.0-41 C.7 Heating, Ventilation, and Air Conditioning (HVAC) Ductwork and Supports C.7.1 Scope Seismic qualification of the Class I HVAC ductwork and associated supports is described in this section. The BFN Class I HVAC ductwork consists of rectangular and round sheet metal ducts (References 26, 27, 28), as well as scheduled pipe used as ductwork. For rectangular ducts, the Companion Angle (CA) and Pocket Lock (PL) type transverse joint constructions as specified by the Sheet Metal and Air Conditioning Contractors' National Association (SMACNA) are used. Additionally, welded joint constructions are also used.
The analytical methods are described in Section C.7.2 and C.7.3. When scheduled pipe is used as ductwork, qualification may be done according to methods and stress limits described in Section C.3.1 and C.3.6. The buried HVAC ductwork for Standby Gas Treatment (SGT), constructed of scheduled pipe, is evaluated in accordance with Section C.3.4.
C.7.2 Ductwork System Seismic Analysis The ductwork system consists of a duct run and a series of supports. The original ductwork design was based on Amplified Response Spectra (ARS) using the El Centro earthquake input ground motion as described in Section 2.5.4. Subsequently, all Class I ductwork systems have been reevaluated using ARS developed from artificial time history (Housner) and impact assessments have been made. The impact assessments include combining the two directional responses absolutely, as opposed to the SRSS method used with the El Centro response spectra. The results of the impact assessments have been used to determine whether the ductwork previously qualified using the El Centro response spectra (SRSS combination) meets the allowables specified in Section C.7.3. For any ductwork system not meeting the specified allowables, modifications have been made for compliance with the requirements of Section C.7.3.
For the seismic evaluation, computer aided modal response method is used.
The ducts are modeled as beam elements with effective bending and shear properties. For rectangular ducts with Companion Angle and Pocket Lock type transverse joints, frequency correction factors of 0.87 and 0.59, respectively, have been applied to more accurately predict the frequency.
However, for round ducts (welded joints), the frequency correction factor used is 1.0. Modes of frequencies 20 Hz and above are considered as rigid.
The effects of Zero Period Acceleration (ZPA) are also considered, which is
BFN-27 C.0-42 applied to the rigid mode with the effective values as the maximum building floor accelerations at or above 20 Hz.
The flexural mode responses are combined by SRSS method for each direction, except for closely spaced modes (frequencies within 10 percent of each other), which are combined by absolute summation. The rigid mode is combined by SRSS method with flexural modes. Two sets (xy and zy) of resultant seismic responses are generated, where x and z represent two horizontal directions and y represents the vertical direction. A set is formed by absolute summation of responses of the two directions. The controlling response is the larger of the two sets of responses.
Alternatively, an equivalent static method with the peak acceleration values corresponding to the system fundamental frequency has been used. A multimode correction factor of 1.5 is applied to the peak acceleration in an equivalent static analysis method.
The ARS analysis considers all effective concentrated weights lumped along the ductwork. Differential building seismic movements are also considered.
In addition, weather induced loads (as applicable) are considered for ductwork exposed to exterior conditions.
The damping ratio of DBE (SSE) response spectra used on each type of the ductwork are as follow:
Critical Ductwork Damping %
Rectangular, companion angle or pocket lock 7
Rectangular, all welded 2
Round duct, all types except scheduled pipe 2
Scheduled pipe 1
C.7.3 Ductwork Load Combinations and Allowable Stresses Ductwork systems are designed for Normal and Emergency loading conditions. Normal condition consists of Deadweight loads. Emergency condition is a combination of Deadweight and DBE (SSE) seismic loads.
BFN-27 C.0-43 C.7.3.1 Duct Allowable Stresses Bending Bending stresses are limited to the following allowables:
Rectangular Duct: 8000 psi (Normal) 12000 psi (Emergency)
- Round Duct:
10000 psi (Normal) 15000 psi (Emergency)
- Not applicable to scheduled pipe analyzed in accordance with Sections C.3.1 and C.3.4.
Shear
- 1)
Rectangular Duct:
Va = 5.5 w (6.4) (Ie/w).25 (Emergency)
Where, Va = Allowable shear capacity of a rectangular duct cross section, 1b.
w = Uniform weight per foot length of ductwork, lb/ft.
Ie = Effective moment of inertia (bending) of cross section, in.4 (based on approach in Reference 26).
Notes a) For evaluating the shear load at a section with an unreinforced opening, Va is reduced in proportion to the reduction in gross shear area.
b) For ductwork with heavier gauge steel than the SMACNA construction, an alternative method of qualification by AISI Specification (Reference 39) is adopted. If this method is used, a 1/3 increase is allowed for the Emergency load combination.
BFN-27 C.0-44
- 2)
Round Duct Maximum shear stress: 0.53 Fy (Emergency) where, Fy = Minimum specified yield stress of duct section, psi.
Buckling Maximum allowable = 0.9 x critical buckling for axial compression (Emergency)
C.7.3.2 Ductwork Supports Allowable Stresses The allowables for Class I ductwork supports, connecting bolts, and welds are as follows:
Normal:
AISC Manual allowables (Reference 4)
Emergency: Tension and Bending :
1.5 x Normal
- Compression :
1.5 x Normal
- Shear
- 0.9 Fy / 1.7321 Bolt stress in tension : 1.5 x Normal Bolt stress in shear : 1.4 x Normal Weld stress
- 1.5 x Normal
- Buckling 0.9 x Critical
- Maximum allowable: 0.9 Fy.
C.8 Control of Heavy Loads C.8.1 Introduction/Licensing Background Generic Letter (GL) 81-07, Control of Heavy Loads, was initiated for all Licensees of Operating units to review the controls of handling heavy loads in regard to the requirements of NUREG-0612. GL 81-07 was in response to the occurrence of load drops in the industry.
The implementation of Phase I of NUREG-0612 included the development and evaluation of critical lift zones; improved designation and inspection of rigging equipment; improved crane operator and rigging training; the development of station procedures to control heavy lifts; review of reactor building or containment crane systems for single failure proof capabilities; and for stations without single failure proof capabilities, the performance of load drop analyses.
BFN-27 C.0-45 Twelve sites (twenty reactors in all) including both PWRs and BWRs were selected as a pilot program to review the effectiveness of the implementation of Phase I of NUREG-0612. BFN Units 1, 2, and 3 were included in this pilot assessment by the NRC. Due to the improvements made with the incorporation of NUREG-0612 at the pilot plants, the conclusion by the NRC was to not proceed with the implementation of Phase II of NUREG-0612.
Generic Letter 85-11 cancelled Phase II of NUREG-0612.
The Control of Heavy Loads program at BFN was established through a number of letters submitted to the NRC. The submittals and letters include:
H. G. Parris (TVA) letter to A. Schwencer (NRC), Comparison of the Browns Ferry Nuclear plant reactor building crane design, testing and maintenance requirements with the positions given in the document entitled Branch Technical Position APCSB 9-1, Overhead Handling Systems for Nuclear Power Plants, dated June 30, 1976 L.M. Mills (TVA) letter to T. A. Ippolito (NRC), Requested information on the Browns Ferry Reactor Building crane, dated February 10, 1981 L.M. Mills (TVA) letter to D. B. Vassallo (NRC), Requested information regarding NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, dated April 12, 1982 L.M. Mills (TVA) letter to D. B. Vassallo (NRC), Requested information regarding NUREG-0612, Control of Heavy Loads at Nuclear Power Plants - Enclosed is information regarding section 2.1 of NUREG-0612, dated June 3, 1982 L.M. Mills (TVA) letter to D. B. Vassallo (NRC), Requested information regarding NUREG-0612, Control of Heavy Loads at Nuclear Power Plants - Enclosed is information regarding section 2.2 and 2.3 of NUREG-0612, dated September 28, 1982 L.M. Mills (TVA) letter to D. B. Vassallo (NRC), NRC / TVA telecon review of the technical evaluation report for the Browns Ferry entitled Control of Heavy Loads - NUREG-0554, dated December 14, 1982
BFN-27 C.0-46 In response to these submittals, the NRC issued a Safety Evaluation Report (SER) on the Control of Heavy Loads (Phase I) for BFN. This SER was transmitted to TVA by a letter from D. B. Vassallo (NRC) to H. G. Parris Control of Heavy Loads - (Phase 1) {review and acceptance of BFNs response to Control of Heavy Loads (Phase 1), 1980 Generic Letter Control of Heavy Loads}, dated June 6, 1984.
C.8.2 Safety Basis The safety basis for the Control of Heavy Loads is provided by assuring the risks associated with load-handling failures is acceptably low. This assurance is provided by meeting the requirements of NUREG-0612, Section 5.1.1, the use of an equivalent single-failure-proof crane for the reactor head lift, drywell head lift, lifts associated with reactor disassembly and inspections; and for dry cask lifts on the refuel floor. Critical lifts performed in plant areas other than the refuel floor are conducted in accordance with program documents and procedures.
C.8.3 Scope of Heavy Load Handling Systems A heavy load for BFN is defined as any load weighing in excess of 1,000 lbs that is lifted in an area designated as a critical lift zone. Critical Lift Zones (CLZ) are defined as zones in designated strategic regions of the plant where a load drop impact could potentially release radioactive material into the environment or prevent equipment from functioning that may be required to achieve safe shutdown and continued decay heat removal. Overhead handling systems, including mobile, lifting lugs and monorail devices, that meet these criteria are:
Reactor Building overhead crane Mobile crane lifts over the Units 1 & 2 and Unit 3 Standby Diesel Generator Buildings Mobile crane lifts over Intake Pumping Station Lifting lugs or temporary rigging over the Core Spray Pumps Monorail and hoist assembly over the Recirculation Pumps In addition, other overhead handling systems were reviewed and excluded from this list on the basis that a load drop would not result in damage to any system required for plant shutdown or decay heat removal for one of the following reasons:
- 1. There is sufficient physical separation of the overhead handling system from any system or component required for safe shutdown or decay heat removal.
BFN-27 C.0-47
- 2. The system or component over which the load is carried is out of service while the load handling system is used.
- 3. The load weighs less than 1,000 lbs and is not considered to be a heavy load.
C.8.4 Control of Heavy Loads Program The Control of Heavy Loads Program consists of the following:
- 1. BFN commitments in response to NUREG-0612, Section 5.1.1 elements
- 2. For reactor disassembly/reassembly and inspections and refuel associated lifts, a single-failure-proof crane
- 3. For spent fuel cask lifts over the spent fuel pool, a single-failure-proof crane
- 4. Performance of critical lifts in accordance with program documents and procedures
- 5. Performance of preventive maintenance, inspection, and testing of cranes, special lifting devices, and rigging equipment in accordance with program documents and procedures
- 6. Use of task qualified crane operators and rigging personnel C.8.4.1 BFN Commitments in Response to NUREG-0612, Section 5.1.1 The control of heavy loads is performed by compliance with the seven guidelines outlined in NUREG-0612, Section 5.1.1.
BFN-27 C.0-48 These guidelines are met through the following:
Guideline Compliance Method 1
Safe load paths - Safe load paths are contained in maintenance instructions. Directions contained within these instructions provide requirements for control of any lift greater than 1,000 pounds, lifts on the refuel floor of the Reactor Building in the regions adjacent to and over the spent fuel pool and over the reactor cavity, lifts in the Reactor Building inside the drywell over the Recirculation system pumps, lifts in the Reactor Building over the Core Spray system pumps, lifts over the Units 1 & 2 and Unit 3 Standby Diesel Generator Buildings, and lifts at the Intake Pumping Station over the Residual Heat Removal Service Water (RHRSW) pumps in those areas designated as critical lift zones (CLZ). The critical lifting zones are defined as follows:
Reactor building refuel floor CLZ - The region defined as the Spent Fuel Pool CLZ, within 15 feet of the spent fuel pool. Also, the region defined as the Reactor Well CLZ, within 15 feet of the reactor well when at least one of the lower horizontal reactor well shield blocks has been removed and when spent fuel is in the reactor vessel.
Also, when Alternate Decay Heat Removal (ADHR) is being used as the primary heat removal source for a particular units fuel pool, the CLZ shall extend over the affected ADHR piping service when it is in service.
Standby Diesel Generator Building CLZ - The region, over the roof of the Unit 3 Standby Diesel Generator Building, when any one of the four diesel generators is inoperable; and the Unit 1 & 2 Standby Diesel Generator Building when Diesel Generator Auxiliary Board A is inoperable.
Intake Pumping Station CLZ - The region over the remaining operable RHRSW pumps when limiting conditions per Technical Specification exist and/or the region over the remaining operable High Pressure Fire pumps.
BFN-27 C.0-49 Core Spray CLZ - The region, when irradiated fuel is in the reactor vessel, with the potential of associated Core Spray pumps and piping being impacted by equipment transported through the hatches on El. 565.0.
Recirculation Pump Motor CLZ - The region, with irradiated fuel in the reactor vessel, when removing or replacing a recirculation pump motor. The Recirculation pump piping CLZ is not considered an impact target area for NUREG-0612 Heavy Loads when all of the following conditions are in place: 1) affected unit is in cold shutdown; 2) reactor is depressurized with the reactor head removed (irradiated fuel may be in the reactor), 3) recirculation nozzle and jet pump plugs are installed, and
- 4) the recirculation system is tagged out of service.
To control load movement, maintenance instructions direct the crane operator to raise and transfer the load to its destination following safe load paths which have been designated in the instructions. To ensure that the established load paths are followed, all lifts performed per these instructions are done under the supervision of a designated individual (person-in-charge) who will verify the load path is clear prior to load movement. Deviations from approved load paths require prior approval of the Plant Operations Review Committee (PORC).
2 Procedures - Load handling procedures for the reactor building crane, placement of mobile cranes adjacent to the Units 1 & 2 and Unit 3 Standby Diesel Generator Building and the Intake Pumping Structure; and rigging over the Core Spray system pumps and the Recirculation system pumps are contained in maintenance instructions. These instructions contain sections covering scope of control, references, prerequisites, precautions and limitations, acceptance criteria, performance, inspections, tables of approved heavy load lifts, and drawings identifying safe load paths. Tables of the various approved heavy load lifts identify the crane to be used, approved rigging or lifting devices, component weights, and reference drawings and procedures.
3 Crane Operators - Programs for crane operator training, qualification, and conduct are contained in TVA Safety
BFN-27 C.0-50 Procedures. The training programs include:
Operating Practices and Functional Characteristics Rigging Fundamentals Electrical Maintenance Certification Skills for Overhead cab-operated Cranes Crane Operator Medical These training programs incorporate all of Chapter 2.2 of ANSI (ASME) B30.2 - 1976.
4 Special lifting devices - BFN Special Lifting Devices are the reactor pressure vessel carousel head strong back, the drywell head strong back, the Preferred Engineering reactor cavity inspection platform lifting device, the dryer separator slings, recirculation pump motor lifting beam, and hardware to support dry cask storage lifts (Hi-Trac Lift Yoke, Hi-Storm Lifting Bracket, MPC Lift Cleats, Hi-Trac Lift Link Assembly and Mating Device-125D). Qualification of these devices was performed in accordance with NUREG-0612 and is documented in design documents. Inspection of these lift devices is performed pre-use and on a periodic bases in accordance with plant procedures. Refer to Chapter 10.4.4 and the Holtec International FSAR Report for the Hi-Storm 100 Cask System and for a description of Special Lifting Devices for the Hi-Storm 100 Cask System used for the dry cask storage of spent fuel.
5 Lifting devices that are not specially designed - All slings and other lifting devices not specially designed and used with cranes subject to NUREG-0612, Section 5.1, are designed, inspected, and tested in accordance with ANSI B30.9 - 1971.
6 Cranes are inspected, tested, and maintained in accordance with Chapter 2-2 of ANSI B30.2 - 1976 - Cranes and hoists at BFN are inspected, tested, and maintained in accordance with specific site maintenance (MI) and preventative maintenance (PM) instructions which implement the requirements of the applicable ANSI (ASME) standard. Each handling system as listed below has its own unique instruction or procedure to control inspection and testing. The load handling system and applicable standard are as follows:
Handling Procedure Reference System Standard
BFN-27 C.0-51 Reactor Building Crane MI ANSI B30.2 - 1976 Mobile Cranes PM ANSI B30.5 - 1976 7
The reactor building crane was designed to meet the applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976 and CMAA 1975. Refer to FSAR Chapter 12.2.2.4, 12.2.2.5, 12.2.2.8, and the Holtec International FSAR Report for the Hi-Storm 100 Cask System for additional details on the design requirements for the reactor building overhead crane and the supporting structural steel features.
Analysis results for the reactor building overhead crane, supporting structural steel features, and special lifting devices are documented in design calculations, vendor technical reports, and / or other design documents.
C.8.4.2 Reactor Pressure Vessel Head (RPVH) Lifting Procedures BFN operations and maintenance instructions are used to control lifts associated with reactor disassembly and reassembly, refuel activities, and dry cask storage activities. These instructions and TVA Safety Procedures contain requirements to ensure the single-failure-proof equivalency of the reactor building crane is maintained. Additionally sections 3.9.4, 3.9.5, 3.9.6, and 3.9.7 of the BFN Units 1, 2, and 3 Technical Requirements Manual contain requirements for the establishment of operability of the reactor building overhead crane and requirements for the performance of lifts and equipment staging in conjunction with reactor disassembly / reassembly and refueling activities. These requirements include:
Crane ambient temperature operating limits.
Preventive maintenance, inspection, and functional testing of the reactor building crane.
Crane safety functions verification requirements.
The BFN reactor building crane was evaluated against NUREG-0554, Single Failure-Proof Cranes for Nuclear Power Plants, as part of the station response to NUREG-0612, Section 5.1.3 (1) (and thus Section 5.1.6) compliance. This evaluation indicated that the crane is equipped with numerous single-failure-proof features. These features, also as described in greater detail in FSAR Chapter 12.2.2.5 and the Holtec International FSAR Report for the Hi-Storm 100 Cask System, include:
BFN-27 C.0-52 The Dry Cask Storage spent fuel lift yoke is designed to higher factors of safety than specified in ANSI N14.6 - 1993 Cab Mounted Emergency Stop Button Floor Mounted Emergency Stop Buttons Overload Protection Overspeed Detection Main hoist is equipped with a single-failure proof hoist system designed in accordance with Ederers Generic Licensing Topical Report EDR-1 to provide compliance with NUREG-0554.
All the crane controls are spring-returned to off.
Undervoltage protection is provided on all motions to sense low, or loss of, control voltage and cause the driven equipment to stop.
Two overhoist limit switches and one down-travel limit switch are provided on each hoist.
A torque-proving circuit checks that current is actually flowing in the main and auxiliary hoist drive motors armatures before the motor brakes are permitted to be released.
Designed for Safe Shutdown Earthquake with the Maximum Critical Load NEI 08-05, Industry Initiative on Control of Heavy Loads, defines the requirements for an equivalent single-failure-proof crane for the purposes of lifting of the reactor head. In addition to having the required safety features, the following measures are provided for the reactor head/reactor internals lifts associated with refueling activities over the reactor, drywell head lifts, and dry cask storage associated lift over the spent fuel pool:
All safety functions of the crane are verified to be operational prior to performing the lift Direct communications are provided between the Crane Operator, Person-In-Charge and Signal Person via headsets Emergency stop buttons are manned during lift Backup Emergency Stop Signal is provided Pre-job brief performed that includes identification of supervisory oversight, establishment of lift management protocol, acceptable travel limits of crane, verification of load travel path Maintenance rule (a)(4) measures addressed in outage safety plan
BFN-27 C.0-53 C.8.5 Safety Evaluation Heavy load lifts at BFN are performed safely and in accordance with NUREG-0612. Basis is provided by:
Controls implemented by NUREG-0612, Section 5.1.1, make the risk of a load drop very unlikely.
The use of a single-failure-proof crane makes the risk of a reactor head load drop, drywell head load drop, or a load drop associated with refueling activities over the reactor extremely unlikely and acceptably low.
The risk associated with the movement of heavy loads is evaluated and controlled by station maintenance instructions, operations procedures, and site procedures.
C.9 References
- 1. USAS B31.1.0 - 1967 Code for Power Piping, published by the American Society of Mechanical Engineers.
- 2. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Division 1, 1971 Edition through the 1973 Summer addenda, Subsection NC.
- 3. TVA Letter from J. R. Bynum to U.S. Nuclear Regulatory Commission, Licensee Event Report (LER) Seismic Reanalysis of Reactor Pressure Vessel, November 16, 1989.
- 4. American Institute of Steel Construction - AISC, "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings",
November 1, 1978 (eighth edition).
- 5. Manufacturers Standardization Society - MSS - SP-58, 1967 Edition "Pipe Hangers and Supports".
- 6. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report, Mark I Long Term Program, Resolution of Generic Technical Activity A-7",
NUREG-0661, July 1980.
- 7. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Code Case N-397.
BFN-27 C.0-54
- 8. U.S. NRC Generic Letter 87-02 "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors Unresolved Safety Issue (USI) A-46, 1987.
- 9. J.M.E. Audibert and K. J. Nyman, Soil Restraint Against Horizontal Motion of Pipes. Journal of the Geotechnical Engineering Division, pp 1119 to 1142. October 1977.
- 10. E. C. Goodling, Flexibility Analysis of Buried Pipe, ASME Publication 78-PVP-82 Joint ASME/CSME Pressure Vessel and Piping Conference, Montreal, Canada, June 1978.
- 11. ASCE, Seismic Response of Buried Pipes and Structural Components, Report by Committee on Seismic Analysis of the ASCE Structural Division Committee on Nuclear Structures and Materials, 1983.
- 12. Browns Ferry Nuclear Plant, Torus Integrity Long Term Program, Plant Unique Analysis Report (PUAR), TVA Report No. CEB 34, Revision 2, December 10, 1984 (CEB 841210-008).
- 13. Nuclear Regulatory Commission NRC, Office of Inspection and Enforcement Bulletin (IEB) No. 79-14 Seismic Analysis for As-Built Safety-Related Piping Systems, Revision 1, July 1979.
- 14. Nuclear Regulatory Commission, Office of Inspection and Enforcement Bulletin (IEB) No. 79-02, Revision 1, Supplement 1, dated August 20, 1979, Pipe Support Base Plate Design Using Concrete Anchor Bolts.
- 15. TVA BFNP Nuclear Performance Plan - Volume 3, Browns Ferry Nuclear Plant, Revision 2, October 24, 1988.
- 16. U.S. Nuclear Regulatory Commission NUREG 1232, Supplements 1 and 2, Volume 3, Safety Evaluation Report on TVA Browns Ferry Nuclear Performance Plan, Browns Ferry Unit 2 Restart, October 1989 and January 1991, respectively.
- 17. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, Class B, 1965 through Winter 1966 Addenda and code case interpretations, including Code Cases 1177 and 1330.
- 18. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section VIII, Winter 1965 Addenda.
BFN-27 C.0-55
- 19. General Electric Company, Mark I Containment Program - Plant Unique Load Definition - Browns Ferry Nuclear Plant Units 1, 2, and 3, Rev. 2, NEDO-24580, January 1982.
- 20. U.S. NRC Regulatory Guide 1.100, Revision 1, Seismic Qualification of Electrical Equipment for Nuclear Power Plants.
- 21. General Electric Company, Seismic Assessment of Browns Ferry 2 Reactor Vessel and Internals, DRF-B11-00457, September 5, 1989.
- 22. U.S. Nuclear Regulatory Commission, Safety Evaluation of Browns Ferry Nuclear Plant, Units 1, 2, and 3, Mark I Containment Long-Term Program, Pool Dynamic Loads Review, May 6, 1985.
- 23. TVA Letter from M. J. Ray to U.S. Nuclear Regulatory Commission Unit 1, 2, 3 Original FSAR Appendix J, K, and L submittals, June 23, 1989.
- 24. TVA Letter from O. J. Zeringue to U.S. Nuclear Regulatory Commission, BFN-Supplement 1 to Generic Letter 87-02, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, USI-A46, and Supplement 4 to Generic Letter 88-20, IPEEE for Severe Accident Vulnerabilities, September 21, 1992.
- 25. TVA Letter from O. J. Zeringue to U.S. Nuclear Regulatory Commission, BFN-GL 87-02, Supplement 1, 120-Day Response, Request for Additional Information, January 19, 1993.
- 26. SMACNA, Rectangular Industrial Duct Construction Standards, 1980, Section 9.
- 27. SMACNA, Round Industrial Duct Construction Standards, 1977 Section 7.
- 28. SMACNA, High Velocity Duct Construction Standards, Second Edition - 1969.
- 29. Biggs, John M., Introduction to Structural Dynamics, McGraw-Hill, 1964, page 153.
- 30. TVA Summary Report for HVAC Ducts Seismic Qualification and Verification/Improvement Program, Report MA2-79-1, June 16, 1979.
BFN-27 C.0-56
- 32. IEEE 344-1975, Institute of Electrical and Electronic Engineering (IEEE) Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Station.
- 33. NUREG 1030, February 1987, Seismic Qualification of Equipment in Operating Nuclear Power Plants.
- 34. NUREG 1211, February 1987, Regulatory Analysis for Resolution of Unresolved Safety Issue A-46, Seismic Qualification of Equipment in Operating Plants.
- 35. TVA Letter from M. O. Medford to U. S. Nuclear Regulatory Commission, NRC Inspection Report No. 50-260/89 Response to Notice of Violation, November 27, 1989.
- 36. M. Ayub Iqbal and E. C. Goodling, Seismic Design of Buried Piping, Second ASCE Specialty Conference of Structural Design of Nuclear Plant Facilities, Vol 1-A, December 8-10, 1975.
- 37. U.S. NRC Generic Letter 83-08, Modification of Vacuum Breakers on Mark I Containments, February 2, 1983.
- 38. TVA Letter from J. A. Domer to U.S. Nuclear Regulatory Commission responding to Generic Letter 83-08, November 5, 1984.
- 39. Specification of the AISI Cold-Formed Steel Design Manual, 1983 Edition.
- 40. U.S. NRC letter: Evaluation of Seismic Design Criteria for HVAC - BFNP dated July 16, 1992.
- 41. NEDO-20566, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, January 1976.
- 42. US-NRC, "NRC Bulletin 96-03: Potential Plugging of Emergency Core Cooling System Strainers By Debris In Boiling Water Reactors,"
May 6, 1996.
- 43. NEDC-32751P, Power Uprate Safety Analysis for BFN Units 2 and 3, September 1997.
BFN-28 C.0-57
- 44. TVA Letter from T. E. Abney to U. S. Nuclear Regulatory Commission, Browns Ferry Nuclear Plant (BFN) Unit 1 - Response To NRC Generic 45.
Letter (GL) 87-02, Supplement 1 That Transmits Supplemental Safety Evaluation Report No. 2 (SSER No. 2) On SQUG Generic Implementation Procedure, Revision 2, As Corrected On February 14, 1992 (GIP-2),
October 7, 2004.
- 45. TVA Letter from W. D. Crouch to U. S. Nuclear Regulatory Commission, Browns Ferry Nuclear Plant (BFN) - Unit 1 - Request For Additional Information (RAI) For Response to Generic Letter 87-02, Supplement 1 (TAC No.
MC4796), August 29, 2006.
- 46. Seismic Qualification Utilities Group - Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment - Revision 3A, December 2001.
- 47. Seismic Qualification Utilities Group - Implementation Guidelines for Seismic Qualification of New and Replacement Equipment/Parts (NARE) Using the Generic Implementation Procedure (GIP) - Revision 5, October 2002.
- 48. Electric Power Research Institute - NP-7484 Guideline for the Seismic Technical Evaluation of Replacement Items for Nuclear Power Plants, February 1993.
- 49. Electric Power Research Institute - TR-105849 Generic Technical Evaluation of Replacement Items for Nuclear Power Plants - Item Specific Evaluations, March 1996 and Supplement 1 September 1997.
- 50. GE-Hitachi - Task T0302, Reactor Vessel Integrity and Fatigue Evaluation, Browns Ferry Nuclear 1, 2, 3 Extended Power Uprate, Revision 0, May 2015.
- 51. GE-Hitachi - Task T0303, RPV Internals Structural Integrity Evaluation, Browns Ferry Nuclear 1, 2, 3 Extended Power Uprate, Revision 0, April 2015.
- 52. GE-Hitachi - NEDC-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, Revision 1, October 2016.
- 53. GE-Hitachi - Task T0400, Containment System Response, Browns Ferry Units 1, 2, and 3, Revision 0, May 2015.
- 54. GE-Hitachi - Task 0305, RPV Flow Induced Vibration, Browns Ferry Nuclear 1, 2, 3, Revision 1, May 2015.
BFN-27 Table C.2-1 DEFORMATION LIMIT Either One of (Not Both)
General Limit
- a.
Permissible Deformation, DP Analyzed Deformation Causing Loss of Function, DL
0 9.
min SF
- b.
Permissible Deformation, DP Experimental Deformation Causing Loss of Function, DE
1.0 SFmin where DP = permissible deformation under stated conditions of normal, upset, emergency, or faulted DL = analyzed deformation which would cause a system loss of function(1)
DE = experimentally determined formation which would cause a system loss of function(1)
(1)
"Loss of Function" can only be defined quite generally until attention is focused on the component of interest. In cases of interest, where deformation limits can affect the function of equipment and components, they will be specifically delineated. From a practical viewpoint, it is convenient to interchange some deformation condition at which function is assured with the loss of function condition if the required safety margins from the functioning condition can be achieved. Therefore, it is often unnecessary to determine the actual loss of function condition because this interchange procedure produces conservative and safe designs. Examples where deformation limits apply are: control rod drive alignment and clearances for proper insertion, core support deformation causing fuel disarrangement, or excess leakage of any component.
BFN-27 Sheet 1 Table C.2-2 PRIMARY STRESS LIMIT Any One of (No More than One Required)
General Limit
- a.
Elastic Evaluated Primary Stresses, PE Permissible Primary Stresses, PN
2.25 min SF
- b.
Permissible Load, LP Largest Lower Bound Limit Load, CL
1.5 SFmin
- c.
Elastic Evaluated Primary Stress, PE Conventional ultimate strength at Temperature, US
0 75 min SF
- d.
Elastic Plastic Evaluated Nominal Primary Stress, PE Conventional ultimate strength at Temperature, US
0 9 min SF
- e.
Permissible Load, LP Plastic Instability Load, PL
0 9 min SF
- f.
Permissible Load, LP Ultimate Load From Fracture Analysis, UF
0 9 min SF
- g.
Permissible Load, LP Ultimate Load or Loss of Function Load from Test, LE
1.0 SFmin
BFN-27 Sheet 2 Table C.2-2 (continued)
PRIMARY STRESS LIMIT where PE = Primary stresses evaluated on an elastic basis. The effective membrane stresses are to be averaged through the load carrying section of interest.
The simplest average bending, shear or torsion stress distribution which will support the external loading will be added to membrane stresses at the section of interest.
PN = Permissible primary stress levels under normal or upset conditions under applicable industry code.
LP = Permissible load under stated conditions of emergency or faulted.
CL = Lower bound limit load with yield point equal to 1.5 Sm, where Sm is the tabulated value of allowable stress at temperature of the ASME III code or its equivalent. The "lower bound limit load" is here defined as that produced from the analysis of an ideally plastic (nonstrain hardening) material where deformations increase with no further increase in applied load. The lower bound load is one in which the material everywhere satisfies equilibrium and nowhere exceeds the defined material yield strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding to the uniaxial case.
US = Conventional ultimate strength at temperature or loading that would cause a system malfunction, whichever is more limiting.
EP = Elastic-plastic evaluated nominal primary stress. Strain hardening of the material may be used for the actual monotonic stress strain curve at the temperature of loading or any approximation to the actual stress strain curve which everywhere has a lower stress for the same strain as the actual monotonic curve may be used. Either the shear or strain energy of distortion flow rule may be used.
PL = Plastic instability load. The "plastic instability load" is defined here as the load at which any load bearing section begins to diminish its cross-sectional area at a faster rate than the strain hardening can accommodate the loss in area. This type analysis requires a true stress-true strain curve or a close approximation based on monotonic loading at the temperature of loading.
BFN-27 Sheet 3 Table C.2-2 (continued)
PRIMARY STRESS LIMIT UF = Ultimate load from fracture analyses. For components that involve sharp discontinuities (local theoretical stress concentration > 3) the use of a "fracture mechanics" analysis where applicable, utilizing measurements of plain strain fracture toughness may be applied to compute fracture loads. Correction for finite plastic zones and thickness effects as well as gross yielding may be necessary. The methods of linear elastic stress analysis may be used in the fracture analysis where its use is clearly conservative or supported by experimental evidence. Examples where "fracture mechanics" may be applied are for fillet welds or end of fatigue life crack propagation.
LE = Ultimate load or loss of function load as determined from experiment.
In using this method account shall be taken of the dimensional tolerances which may exist between the actual part and the tested part or parts as well as differences which may exist in the ultimate tensile strength of the actual part and the tested parts. The guide to be used in each of these areas is that the experimentally determined load shall use adjusted values to account for material properties and dimension variations, each of which has no greater probability than 0.1 of being exceeded in the actual part.
BFN-27 Table C.2-3 BUCKLING STABILITY LIMIT Any One of (no more than one required)
General Limit
- a.
Permissible Load, LP Code Normal Event Permissible Load, PN
2.25 min SF
- b.
Permissible Load, LP Stability Analysis Load, SL
0 9.
min SF
- c.
Permissible Load, LP Ultimate Buckling Collapse Load from Test, SE
1.0 SFmin where:
LP = Permissible load under stated conditions of emergency or faulted.
PN = Applicable code normal event permissible load.
SL = Stability analysis load. The ideal buckling analysis is often sensitive to otherwise minor deviations from ideal geometry and boundary conditions. These effects shall be accounted for in the analysis of the buckling stability loads. Examples of this are ovality in externally pressurized shells or eccentricity of column members.
SE = Ultimate buckling collapse load as determined from experiment. In using this method, account shall be taken of the dimensional tolerances which may exist between the actual part and the tested part. The guide to be used in each of these areas is that the experimentally determined load shall be adjusted to account for material property and dimension variations, each of which has no greater probability than 0.1 of being exceeded in the actual part.
BFN-27 Table C.2-4 FATIGUE LIMIT General Limit Summation of mean fatigue(1)
- a. Fatigue cycle usage usage including emergency or from analysis 0.05 faulted events with design and operation loads following
- b. Fatigue cycle usage Miner hypotheses....
from test 0.33 either one (not both)
(1)
Fatigue failure is defined here as a 25% area reduction for a load carrying member which is required to function or excess leakage causing loss of function, whichever is more limiting. In the fatigue evaluation, the methods of linear elastic stress analysis may be used when the 3Sm range limit of ASME Code,Section III has been met. If 3Sm is not met, account will be taken of (a) increases in local strain concentration, (b) strain ratcheting, and (c) redistribution of strain due to elastic-plastic effects. The January 1969 draft of the USAS B31.7 Piping Code may be used where applicable, or detailed elastic-plastic methods may be used. With elastic-plastic methods, strain hardening may be used not to exceed in stress for the same strain the steady-state cyclic strain hardening measured in a smooth low cycle fatigue specimen at the average temperature of interest.
BFN-27 Sheet 1 of 8 TABLE C.3-1A LOAD COMBINATIONS AND ALLOWABLE STRESS CRITERIA FOR CLASS I PIPING AND TUBING (PIPING OTHER THAN RRS, MS, FW AND CRDH SYSTEMS)9 Plant Conditions Moment Constituents2 NC-36521 Concurrent Loads From Load Sources Equations and Stress Limits Eq. No.
Design and Normal Design Pressure + Sustained MA = M(DW)10 (8)
Upset Max (Peak) Pressure +
MBU = M(E1,VT,WH)3,6 Sustained + OBE + Fluid (9U)
Transient Emergency Max (Peak) Pressure +
MBE = M(E2,VT,WH,JI)5,6,8,11 Sustained + Fluid Transient (9E)
+ (DBE or Jet Impingement)
P D
D D
iM Z
S i
o i
A h
2 2
2 075
P D
D D
i M
M Z
S m
i o
i A
BU h
2 2
2 0 75 12
P D
D D
i M
M Z
S m
i o
i A
BE h
2 2
2 0 75 18
BFN-27 Sheet 2 of 8 TABLE C.3-1A LOAD COMBINATIONS AND ALLOWABLE STRESS CRITERIA FOR CLASS I PIPING AND TUBING (PIPING OTHER THAN RRS, MS, FW AND CRDH SYSTEMS)9 Plant Conditions Moment Constituents2 NC-36521 Concurrent Loads From Load Sources Equations and Stress Limits Eq. No.
Faulted (Max (Peak) Pressure +
MBF = M(E2,VT,WH,JI)6,8 (9F)
Sustained + DBE + Fluid Transient + Jet Impingement)
Normal and Upset (Secondary)
Thermal Expansion +
MC = M(Ti,SD,S1)3,4,7 (10)
Thermal Anchor Movement +
Seismic Anchor Movement OR Design Pressure + Sustained +
(11)
Thermal Expansion + Thermal Anchor Movement + Seismic Anchor Movement Differential Settlement Differential Settlement MD = M(BS)
P D
D D
i M
M Z
S m
i o
i A
BF h
2 2
2 0 75 2
.4 iM Z
S c
A
P D
D D
iM Z
iM Z
S S
i o
i A
C A
h 2
2 2
0 75
iM Z
S D
C
3
BFN-27 Sheet 3 of 8 TABLE C.3-1B LOAD COMBINATIONS AND ALLOWABLE STRESS CRITERIA OF CLASS I PIPING FOR REACTOR RECIRCULATION (RRS)
MAIN STEAM (MS) AND FEEDWATER (FW) SYSTEMS9 Plant Conditions Moment Constituents2 NC-36521 Concurrent Loads From Load Sources Equations and Stress Limits Eq. No.
Design and Normal (Primary)
Design Pressure +
MA = M(DW)10 (8)
Sustained Upset (Primary)
Design Pressure +
MBU = M(E1,VT,WH)3,6 (9U)
Sustained + Occasional Normal (Primary + Secondary)
Design Pressure +
M'C = M(Ti,SD)
(11)
Sustained + Thermal Expansion + Thermal Anchor Movement P
D D
D iM Z
S i
o i
A h
2 2
2 0 75
P D
D D
i M
M Z
S i
o i
A BU h
2 2
2 0 75 12
P D
D D
i M iM Z
S S
i o
i A
C A
h 2
2 2
0 75
BFN-27 Sheet 4 of 8 TABLE C.3-1B LOAD COMBINATIONS AND ALLOWABLE STRESS CRITERIA OF CLASS I PIPING FOR REACTOR RECIRCULATION (RRS)
MAIN STEAM (MS) AND FEEDWATER (FW) SYSTEMS9 Plant Conditions Moment Constituents2 NC-36521 Concurrent Loads From Load Sources Equations and Stress Limits Eq. No.
Upset (Primary + Secondary)
Design Pressure +
MC = M(Ti,SD,S1)3,4,7 (9U+10)
Sustained + Thermal Expansion & Thermal Anchor Movement + OBE + SAM Emergency (Primary)
Design Pressure +
MBE = M(E2,VT,WH,JI)5,6,8,11 (9E)
Sustained + Fluid Transient
+ (DBE or Jet Impingement)
Max. (Peak) Pressure +
MBE' = M(E1,VT,WH)6,8 (9E)
Sustained + OBE + Fluid Transient Max. (Peak) Pressure +
Sustained + Fluid Transient (9E)
+ (DBE or Jet Impingement)
Faulted Primary Max (Peak) Pressure +
MBF = M(VT,E2,WH,JI)6,8 (9F)
Sustained + Fluid Transient
+ DBE + Jet Impingement
PD D
D i
M M
iM Z
S S
i o
i A
BU C
h A
2 2
2 0 75 12
PD D
D i
M M
Z S
i o
i A
BE h
2 2
2 0 75 18
P D
D D
i M
M Z
S m
i o
i A
BE h
2 2
2 0 75 15
P D
D D
i M
M Z
S m
i o
i A
BE h
2 2
2 0 75 2 0
P D
D D
i M
M Z
S m
i o
i A
BF h
2 2
2 0 75 2
.4
BFN-27 Sheet 5 of 8 TABLE C.3-1C LOAD COMBINATIONS AND ALLOWABLE STRESS CRITERIA FOR CONTROL ROD DRIVE HYDRAULIC PIPING Plant Conditions Moment Constituents2 NC-36521 Concurrent Loads From Load Sources Equations and Stress Limits Eq. No.
Design and Normal (Primary)
Design Pressure +
MA = M(DW)10 Sustained (8)
Upset (Primary)
Max Operating Pressure +
MBU = M(E1,VT,WH)3,6 (9U)
Sustained + Occasional (9U)
Upset (Primary + Secondary)
Max Operating Pressure +
MC1 = M(Ti,SD,S1)3,7 OR Sustained + Normal Scram (10)
Thermal Expansion and Anchor Movement + SAM (OBE)
(11)
PD D
D i M Z
S i
o i
A h
2 2
2 0 75
P D D
D 0.75i M M
Z 1.2S n
i 2
o 2
i 2
A BU h
iM Z
S c
A 1
P D
D D
i M iM Z
S S
n i
o i
A C
A h
2 2
2 1
075
BFN-27 Sheet 6 of 8 TABLE C.3-1C LOAD COMBINATIONS AND ALLOWABLE STRESS CRITERIA FOR CONTROL ROD DRIVE HYDRAULIC PIPING Plant Conditions Moment Constituents2 NC-36521 Concurrent Loads From Load Sources Equations and Stress Limits Eq. No.
Max Operating Pressure +
MC2 = M(Ti,SD)7 Sustained + Abnormal Scram OR (10)
Thermal Expansion and Anchor Movement (11)
Emergency (Primary)
Max Operating Pressure +
MDE = M(E2,VT,WH,JI)6,8,11 Sustained + Fluid Transient (9E)
+ (SSE or Jet Impingement)5 Faulted (Primary)
Max Operating Pressure +
MDF = M(E2,VT,WH,JI)6,8 Sustained + Fluid Transient (9F)
+ SSE + Jet Impingement iM Z
S C
A 2
P D
D D
i M iM Z
S S
n i
o i
A C
A h
2 2
2 2
0 75
P D
D D
i M
M Z
S n
i o
i A
DE h
2 2
2 0 75 18
P D
D D
i M
M Z
S n
i o
i A
DF h
2 2
2 0 75 2
.4
BFN-27 Sheet 7 of 8 TABLE C.3-1A, 1B, 1C (Cont'd)
Nomenclature P
=
Design Pressure, psi.
Pm
=
Max (Peak) Pressure, psi.
Pn
=
Maximum operational or scram pressure for the Hydraulic System Pump Pressure for CRDH System only.
Do
=
Outside Pipe Diameter, in.
Di
=
Nominal Inside Pipe Diameter, in.
i
=
Stress Intensification Factor from B31.1.0 - 1967.
Sh
=
Basic material allowable stress at maximum operating temperature.
Sc
=
Basic Material Allowable Stress at Ambient Temperature.
=
Allowable expansion stress defined in B31.1.0 - 1967.
U,E,F =
Added Suffixes for differentiation between Upset, Emergency, and Faulted.
Z
=
Pipe section modulus (in3).
=
Deadweight.
E1
=
Operating Basis Earthquake (OBE) Inertia Effect.
E2
=
Design Basis Earthquake (DBE) Inertia Effect.
WH
=
Steam/Water Hammer.
Ti
=
Thermal mode i (i = mode number).
=
Thermal Anchor Movements.
S1
=
OBE Seismic Anchor Movements.
BS
=
Differential movement between the soil and building structure for buried piping or relative differential building settlement for piping attached to two buildings.
=
Valve Thrust (Main Steam Relief Valve Actuation).
JI
=
Jet Impingement.
BFN-27 Sheet 8 of 8 TABLE C.3-1A, 1B, 1C (Cont'd)
Notes
- 1.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, 1971 edition, through Summer 1973 Addenda and Code Case 1606-1. Material allowables and SIFs from USAS B31.1.0 -
1967
- 2.
The sequence of events, consistent with the system operational requirements, is considered in establishing which load sources are taken as acting concurrently.
- 3.
Seismic anchor movements are included in the evaluation of either equation (9) or equation (10), but need not be included in both.
- 4.
All secondary load sources resulting from plant normal or upset conditions are identified and evaluated for the limiting operating modes of the system. The effects of these load sources are used in evaluating equipment loading, support loading, and type.
- 5.
The largest loads from either DBE or Jet Impingement are used. Jet impingement loading requirements for piping inside and outside of containment are described in Appendix M.
- 6.
If more than one dynamic load source is involved, such as earthquake, valve thrust, and water hammer, the SRSS method will be used to combine resultant moments from individual load sources. In the event that the dynamic load sources are determined to act nonconcurrently, then they can be considered independently.
- 7.
For Mc, the effects of Ti and corresponding SD are combined algebraically first, and then combined absolutely with S1.
- 8.
Only inertia term of earthquake effect to be considered.
- 9.
Exceptions from the requirements in Table C.3-1A, -1B, and -1C may be allowed with proper justification and NRC concurrence.
- 10.
Additional stresses caused by hydrostatic testing weight are evaluated when applicable.
- 11.
Fire events are evaluated as separate emergency loading conditions. No dynamic loads are postulated to occur simultaneously with these events. Piping is evaluated for pressure plus deadweight effects of the events.
BFN-27 TABLE C.3-2 Sheet 1 of 5 LOAD COMBINATIONS AND ALLOWABLE STRESSES FOR CLASS I PIPE AND TUBING SUPPORTS Support Category Load Condition Direction Design Load Combinations1,2,9 Allowable3 Stresses Linear Type Support Normal
+
DW + Ti+
DW + Ti-1.0S AISC Hydrotest DW 1.0S AISC Upset
+
DW + Ti+ + SRSS[VT+, WH+,
E1, S1]
1.33S AISC4 DW + Ti- - SRSS [VT-, WH-,
-E1, -S1]
Emergency
+
DW + Ti+ + SRSS [VT+, WH+,
E2, S2]
1.5S AISC4 or DW + Ti+ + SRSS [VT+, WH+]
+ PR+
or DW + Ti+ (fire event)
DW + Ti- - SRSS [VT-, WH-,
-E2, -S2]
or DW + Ti- - SRSS [VT-, WH-] +
PR-or DW + Ti- (fire event)
Faulted
+
DW + Ti+ + SRSS [VT+, WH+,
E2, S2] + PR+
1.5S AISC4 DW + Ti- - SRSS [VT-, WH-,
-E2, -S2] +PR-
BFN-27 TABLE C.3-2 (CONTINUED)
Sheet 2 of 5 Support Category Load Condition Direction Design Load Combinations1,2,9 Allowable3 Stresses Snubbers Hydraulic Upset Same as Linear VLR Emergency Same as Linear 1.2 VLR Faulted Same as Linear 1.2 VLR Mechanical Pre-NF Upset Same as Linear VLR Emergency Same as Linear The lesser of 1.33 VLR or LCD Level 'C' Faulted Same as Linear The lesser of 1.33 VLR or LCD Level 'C' Post-NF Upset Same as Linear LCD Level 'B' Emergency Same as Linear LCD Level 'C' Faulted Same as Linear LCD Level 'C'
BFN-27 TABLE C.3-2 (CONTINUED)
Sheet 3 of 5 Support Category Load Condition Direction Design Load Combinations1,2,9 Allowable Stresses3,5,6 Standard Support Components Normal Same as Linear S58 Hydrotest Same as Linear 2.0S588 Upset Same as Linear 1.2S58 Emergency Same as Linear (See Note 7)
Faulted Same as Linear (See Note 7)
BFN-27 TABLE C.3-2 (CONTINUED)
Sheet 4 of 5 Notes:
- 1.
Signs for Load Evaluation DW - Carries the actual analysis signs.
Ti - Thermal load shall be evaluated for both hot and cold conditions.
- 2.
Design value for (+) direction is the larger of zero and the value calculated; (-) direction is the smaller of zero and the value calculated.
- 3.
S AISC =
The basic allowable stresses defined in Part I of the AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, November 1978. (Excluding the 1.33 factor).
S58 =
The basic allowable load as defined by the vendor in accordance with MSS SP-58, 1967 edition, Pipe Hangers and Supports.
Fy =
The minimum yield stress of support member at elevated sustained temperature (i.e., normal operating temperature exceeds 150F).
VLR =
The basic load rating supplied by the vendor.
LCD =
Load capacity data sheet as levels supplied by the vendor.
- 4.
Linear Allowables shall not exceed 0.9Fy for tension or 0.9Fy/3 = 0.52Fy for shear.
- 5.
Load rated allowables established according to ASME section III subsection NF are acceptable using the appropriate load level.
- 6.
Linear support allowables may be used for detailed analysis of standard support components.
BFN-27 TABLE C.3-2 (CONTINUED)
Sheet 5 of 5 Notes:
- 7.
Allowable stress shall not exceed the lesser of 2.0558 or the linear support allowance. However, the lesser shall not exceed available LCD Level 'D' limits.
- 8.
Maximum allowable stress for hydrotest condition shall not exceed 0.8Fy.
- 9.
SRSS combinations shall be consistent with the provisions of Section C.3.1.2.
BFN-27 Sheet 1 Table C.4-1 REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Primary Stress Type Allowable Stress (psi)
Stabilizer Bracket and Adjacent Shell Primary Stress Limit - ASME Boiler Normal and upset condition loads Membrane and bending 40,000 and Pressure Vessel Code, Sect. III
- 1. Operating Basis Earthquake defines primary membrane plus
- 2. Design pressure primary bending stress intensity limit for SA 302 - Gr. B Emergency condition loads Membrane and bending 60,000
- 1. Design Basis Earthquake For normal and upset condition
- 2. Design pressure Stress limit = 1.5 X 26,700 = 40,000 psi Faulted condition loads Membrane and bending 80,000 For emergency condition
- 1. Design Basis Earthquake Stress limit = 1.5 X 40,000 = 60,000 psi
- 2. Jet reaction forces
- 3. Design pressure For faulted condition Stress limit = 2.0 X 40,000 = 80,000 psi Vessel Support Skirt Primary Stress Limit - ASME Boiler Normal and upset condition loads General membrane 26,700 and Pressure Vessel Code, Sect. III
- 1. Dead weight defines stress limit for SA 302
- 2. Operating Basis Earthquake Gr. B Emergency condition loads General membrane 40,000 For normal and upset condition
- 1. Dead weight SM = 26,700 psi
- 2. Design Basis Earthquake For emergency condition Faulted condition loads General membrane 53,400 Slimit = 1.5 SM = 1.5 X 26,700 =
- 1. Dead weight 40,000 psi
- 3. Jet reaction forces For faulted condition Slimit = 2.0 SM = 20 X 26,700 = 53,400 psi
BFN-27 Sheet 2 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Primary Stress Type Allowable Stress (psi)
Shroud leg Support Primary Stress Limit - ASME Boiler Normal and upset condition loads Tensile 23,300 and Pressure Vessel Code, Sect. III
- 1.
Operating Basis Earthquake defines allowable primary membrane
- 2.
Pressure drop across shroud stress SB-168 material.
(normal)
- 3.
Subtract dead weight
- 1.
Tensile Loads For normal and upset condition Emergency condition loads Tensile 35,000 SM = 23,300 psi
- 1.
- 2.
Pressure drop across shroud For emergency condition (normal)
Slimit = 1.5 SM
- 3.
Subtract dead weight
= 1.5 X 23,300 = 35,000 psi Faulted condition loads Tensile 46,600 For faulted condition
- 1.
Design Basis Earthquake Slimit = 2.0 SM
- 2.
Pressure drop across shroud
= 2.0 X 23,300 = 46,600 psi during faulted condition
- 3.
Subtract dead weight
- 2.
Compressive Loads For normal and upset condition Normal and upset condition loads Compressive 14,000 SA = 0.4 Sy
- 1.
= 0.4 X 35,000 = 14,000 psi
- 2.
Zero pressure drop across shroud For emergency condition
- 3.
Dead weight Slimit = 0.6 Sy
= 0.6 X 35,000 = 21,000 psi Emergency condition loads Compressive 21,000
- 1.
Design Basis Earthquake For faulted condition
- 2.
Subtract operating pressure Slimit = 0.8 Sy drop across shroud
= 0.8 X 35,000 = 28,000 psi
- 3.
Dead weight Faulted condition loads Compressive 28,000
- 1.
- 2.
Zero pressure drop across shroud
- 3.
Dead weight
BFN-27 Sheet 3 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Primary Stress Type Allowable Stress (psi)
Top Guide Longest Beam Primary Stress Limit - The allowable Normal and upset condition loads*
General membrane plus 25,388 primary membrane stress plus bending
- 1.
Operating Basis Earthquake bending stress is based on ASME Boiler and
- 2.
Weight of structure Pressure Vessel Code, Sect. III for Type 304 stainless steel plate.
For normal and upset condition Emergency condition loads*
General membrane plus 38,081 Stress Intensity
- 1.
Design Basis Earthquake bending SA = 1.5 Sm = 1.5 X 16.925 = 25,388 psi
- 2.
Weight of structure For emergency condition Slimit = 1.5 SA = 1.5 X 25,388
= 38,081 psi Faulted condition loads*
General membrane plus 50,775 (Same as emergency condition) bending For faulted condition Slimit = 2SA = 2 X 25,388 = 50,775 psi Top Guide Beam End Connections Primary Stress Limit - ASME Boiler Normal and upset condition loads*
Pure shear 10,155 and Pressure Vessel Code, Sect. III
- 1.
Operating Basis Earthquake defines material stress limit for
- 2.
Weight of structure Type 304 stainless steel For normal and upset condition Stress Intensity Emergency condition loads*
Pure shear 15,232 SA = 06 Sm = 0.6 X 16,925 = 10,155 psi
- 1.
- 2.
Weight of structure For emergency condition Slimit = 1.5 SA
= 1.5 X 10,155 = 15,232 psi Faulted condition loads*
Pure shear 20,310 (Same as emergency condition)
For faulted condition Slimit = 2SA = 2 X 10,155 = 20,310 psi
- Note: Normal, upset, and accident top guide hydraulic loads are upward. These are not included in the stress analysis since they counteract the effect of the structure weight.
BFN-28 Sheet 4 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Primary Stress Type Allowable Stress (psi)
Core support Allowable pressure For power uprate the allowable differential differential (psid) loading is based on the ratio of applied pressure to buckling pressure.
For normal and upset:
Normal and Upset condition loads Buckling 28.0 allowable ratio = 0.40
- 1. Normal operation pressure drop
- 2. Operating Basis Earthquake For emergency:
Emergency condition loads Buckling 42.0 allowable ratio = 0.60
- 1. Normal operation pressure drop
- 2. Design Basis Earthquake For faulted:
Faulted condition loads Buckling 56.0 allowable ratio = 0.80
- 1. Pressure drop after main steam line rupture.
- 2. Design Basis Earthquake Allowable Stress (psi)
Core Support Aligners Primary Stress Limit - ASME Boiler Normal and upset condition load Pure shear 10,155 and Pressure Vessel Code, Sect. III
- 1. Operating Basis Earthquake defines material stress limit for Type 304 stainless steel Emergency condition load Pure shear 15,232
- 1. Design Basis Earthquake For allowable shear stresses, see top guide beam end connections Faulted condition load Pure shear 20,310 above
BFN-27 Sheet 5 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Moment Limit Accounting Criteria Loading Primary Stress Type for Pressure Loads (in-lb)
Fuel Channels Primary Stress Limit - The allowable Normal and Upset condition loads Membrane and bending 28,230 Sm for Zircaloy determined according
- 1. Operating Basis Earthquake to methods recommended by ASME
- 2. Normal pressure load Boiler and Pressure Vessel Code, Sect. III. Allowable moment Emergency condition loads Membrane and bending 42,350 determined by calculating limit
- 1. Design Basis Earthquake moment using Table C.2-2
- 2. Normal pressure load equation (b), then applying SFmin for applicable loading conditions.
Faulted condition loads Membrane and bending 56,500
- 2. Loss-of-coolant accident (Sm = 9,270 psi, 1.5 Sm = 13,900 psi) pressure Emergency limit load = 1.5 X Normal limit load calculated using 1.5 Sm = yield
BFN-27 Sheet 6 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Location Allowable Stress (psi)
RPV Stabilizer Primary Stress Limit - AISC specification Upset condition Rod 130,000 for the construction, fabrication
- 1. Spring preload Bracket 22,000 and erection of structural steel for
- 2. Operating Basis Earthquake 14,000 buildings Emergency condition Bracket 33,000 For normal and upset conditions
- 1. Spring preload 21,000 AISC allowable stresses, but without
- 2. Design Basis Earthquake the usual increase for earthquake loads Faulted condition Bracket 36,000 For emergency conditions
- 1. Spring preload 21,500 1.5 X AISC allowable stresses
- 3. Jet reaction load For faulted conditions Material yield strength RPV Support (Ring Girder)
Primary Stress Limit - AISC specification Normal and upset condition Top flange 27,000 for the design, fabrication and erection
- 1. Dead loads of structural steel for buildings
- 2. Operating Basis Earthquake Bottom Flange 27,000
- 3. Loads due to scram Vessel to girder bolts 60,000 For normal and upset conditions 22,500 AISC allowable stresses, but without the usual increase for earthquake loads For faulted conditions Faulted condition Top flange 45,000 1.67 X AISC allowable stresses for
- 1. Dead loads Bottom flange 45,000 structural steel members
- 2. Design Basis Earthquake Vessel to girder bolts 125,000 Yield strength for high strength
- 3. Jet reaction load 75,000 bolts (vessel to ring girder)
BFN-27 Sheet 7 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Location Allowable Stress (psi)
CRD Housing Support "Shootout Steel" Primary Stress Limit - AISC specification Faulted Condition loads Beams (top cord) 33,000 for the design, fabrication
- 1. Dead weight 33,000 and erection of structural steel
- 2. Impact force from failure Beams (bottom cord) 33,000 for buildings of a CRD housing 33,000 For normal and upset condition (Dead weights and earthquake Grid structure 41,500 Fa = 0.60 Fy (tension) loads are very small as 27,500 Fb = 0.60 Fy (bending) compared to jet force.)
Fv = 0.40 Fy (shear)
For faulted conditions Fa limit = 1.5 Fa (tension)
Fb limit = 1.5 Fb (bending)
Fv limit = 1.5 Fb (shear)
Fy = Material yield strength Recirculating Pipe and Pump Pipe Rupture Restraints Primary Stress Limit - Structural Faulted condition loads Brackets on 28 in. pipe 33,000 Steel: AISC specification for the
- 1. Jet force from a complete design, fabrication and erection circumferential failure Cable on pump restraints 99,000 of structural steel for buildings.
(break) of recirculation line For normal or upset conditions Fa = 0.60 Fy (tension)
For faulted conditions Fa limit = 1.5 Fa (tension)
Fy = yield strength Cable (wire rope)
For faulted conditions Fa = 0.80 Fu (tension)
Fu = ultimate strength
BFN-27 Sheet 8 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Location Allowable Stress (psi)
Control Rod Drive Housing Primary Stress Limit - The allowable Normal and upset condition loads Maximum membrane 16,925 primary membrane stress is based on
- 1. Design pressure stress intensity occurs the ASME Boiler and Pressure Vessel
- 2. Stuck rod scram loads at the tube to tube Code Sect. III, for Class A vessels
- 3. Operating Basis Earthquake weld near the center of for Type 304 stainless steel the housing for normal upset and emergency For normal and upset condition conditions Sm = 16,925 psi at 575oF For emergency conditions Emergency condition loads 25,100 Slimit = 1.5 Sm = 1.5 X 16,925=25,400 psi
- 1. Design pressure
- 2. Stuck rod scram loads
- 3. Design Basis Earthquake Control Rod Drive Primary Stress Limit - The allowable Normal and upset condition loads Maximum stress intensity 26,060 primary membrane stress plus Maximum hydraulic pressure occurs at a point on the bending stress is based on ASME from the control rod drive Y-Y axis of the indicator Boiler and Pressure Vessel Code Supply pump.
tube Sect. III for SA-212 TP 316 NOTE - Accident conditions tubing do not increase this loading Earthquake loads are negligible For normal and upset condition SA = 1.5 Sm = 1.5 X 17.375 = 26,060 psi
BFN-28 Sheet 9 Table C.4-1 (Continued)
REACTOR VESSEL, REACTOR VESSEL INTERNALS AND SUPPORTS CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Criteria Loading Location Allowable Stress (psi)
Control Rod Guide Tube Allowable loads (lbs) Pressure differential (psi)
(vertical)
The allowable loading is based on Faulted condition loads The maximum loading 35,200 84 the ratio of applied load to bucklling
- 1. Dead weight conditions occur at the load
- 2. Pressure drop across guide center of the guide tube tube due to failure of length For normal and upset:
steam line allowable ratio = 0.40
- 3. Design Basis Earthquake For faulted:
allowable ratio = 0.80 Incore Housing Allowable Stress (psi)
Primary Stress Limit - The allowable Emergency condition loads Maximum membrane 25,400 primary membrane stress is based on
- 1. Design pressure stress intensity occurs ASME Boiler and Pressure Vessel
- 2. Design Basis Earthquake at the outer surface of Code, Sect. III, for Class A vessels the vessel penetration for Type 304 stainless steel For normal and upset conditions Sm = 16,925 psi at 575oF For emergency condition (N + AM)
Slimit = 1.5 Sm = 1.5 X 16,925 = 25,400 psi
BFN-27 Sheet 1 Table C.4-2 PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES MAIN STEAM ISOLATION VALVES Criteria Method of Analysis Minimum Dimension Required
- 1.
Body Minimum Wall Thickness Minimum wall thicknesses in the cylindrical Body wall thickness portions of the valve shall be calculated Loads:
using the following formula:
t = 1.83 in. at 23-in. diameter Design pressure and temperature Primary Membrane Stress Limit:
S = 7,000 lb/in.2 per ASA B16.5 where:
S = allowable stress of 7000 psi P = primary service pressure, 655 psi d = Inside diameter of valve at section being considered, in.
C = corrosion allowance of 0.12 in.
- 2.
Cover Minimum Thickness Valve cover thickness Loads:
where:
t = 4.888 in.
t = minimum thickness, inches Design pressure and temperature d = diameter or short span, in.
Design bolting load C = attachment factor Gasket load S = allowable stress, psi W = total, bolt load, lb hG = gasket moment arm, in.
Ci = corrosion allowance, in.
Primary Stress Limit:
Allowable working stress per ASME Section VIII t
Pd S
P C
15 2 12 t
d CP S
Wh Sd C
G
178 3
1 2 1
/
BFN-27 Sheet 2 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Isolation Valves (Continued)
Allowable Stress or Criteria Method of Analysis Actual Dimension
- 3. Cover Flange Bolt Area Loads:
Total, bolting loads and stresses shall be Flange Bolt Stress calculated in accordance with "Rules for Loads:
Bolted Flange Connections" - ASME Boiler S = 30,900 lb/in.2 and Pressure Vessel Code,Section VIII, at 575F Design pressure and temperature Appendix II, except that the stem operational Gasket load load and seismic loads shall be included in Stem operational load the total load carried by bolts. The Seismic load-Design Basis Earthquake horizontal and vertical seismic forces shall be applied at the mass center of the valve Bolting Stress Limit:
operator assuming that the valve body is rigid and anchored.
Allowable working stress per ASME Nuclear Pump & Valve Code, Class I
- 4. Body Flange Thickness and Stress Flange thickness and stress shall be calcu-Body Flange Stress lated in accordance with "Rules for Bolted Loads:
Flange Connections" = ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II, except Design pressure and temperature that the stem operational load and seismic SH = 26,700 lb/in.2 Gasket load loads shall be included in the total load SR = 26,700 lb/in.2 Stem operational load carried by the flange. The horizontal and ST = 26,700 lb/in.2 Seismic load - Design Basis vertical seismic forces shall be applied at Earthquake the mass center of the valve operator assum-ing that the valve body is rigid and anchored.
Flange Stress Limits:
SH, SR, ST 1.5 Sm per ASME Nuclear Pump and Valve Code, Class I.
BFN-27 Sheet 3 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Isolation Valves(Continued)
Criteria Method of Analysis Allowable Stress
- 5. Valve Disc Thickness Loads:
where:
S = 17,800 lb/in2 Sr = radial stress, psi Design pressure and temperature St = tangential stress Primary bending stress limit:
v = Poisson's ratio P = design pressure, psi Allowable working stress per R = radius of disc, inches ASME Section VIII t = thickness of disc, inches
- 6. Valve Operator Supports The valve assembly shall be analyzed assuming that the rigid mass and that the valve body Loads:
is an anchored, rigid mass and that the specified vertical and horizontal seismic Design pressure and temperature forces are applied at the mass center of the S = 18,000 lb/in2 Stem operational load operator assembly, simultaneously with Equipment dead weight operating pressure plus dead weight plus Seismic load-Design Basis operational loads. Using these loads, stresses and deflections shall be determined Support Rod Stress Limit:
for the operator support components.
Allowable working stress per ASME ASME Section VIII
S S
3 3 v PR 8t r
t 2
2
BFN-27 Sheet 4 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Safety Valves Criteria Method of Analysis Allowable Stress Minimum Dimension Required
- 1. Inlet Nozzle Wall Thickness Loads:
t = 0.183 in.
where:
1.1 X Design pressure at 600F T = min. required thickness, in.
S = allowable stress, lb/in.2 Primary Membrane Stress Limit:
P = 1.1 X design pressure, lb/in.2 R = internal radius, in.
Allowable stress intensity as defined E = joint efficiency by ASME Standard Code for Pumps and C = corrosion allowable, in.
Valves for Nuclear Power
- 2. Valve Disc Thickness Loads:
where:
Ss= 20,190 lb/in.2 1.1 X Design pressure at 600F W = shear load, lb A = shear area, in.2 Diagonal Shear Stress Limit:
P = 1.1 X design pressure, lb/in.2 A1 = disc area, in.2 0.6 x allowable stress intensity and:
as defined by ASME Standard Code A = S (R + R1) for Pumps and Valves for Nuclear S = slope of frustrum of shear cone, in.
Power R1 = radius at base of cone, in.
R = radius at top of cone, in.
- 3. Inlet Flange Bolt Area Total bolting loads and stresses shall be calculated in accordance with procedures of Loads:
Para. 1-704.5.1 Flanged Joints, of B31.7 Sb = 27,700 lb/in.2 Nuclear Piping Code.
Design pressure and temperature Gasket load Operational load Design Basis Earthquake Bolting Stress Limit:
Allowable stress intensity, Sm, as defined by ASME Standard Code for Pumps and Valves for Nuclear Power t
PR SE P
C
0.6 Ss W
A PA A
1
BFN-27 Sheet 5 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Safety Valves (Continued)
Criteria Method of Analysis Allowable Stress
- 4. Inlet Flange Thickness Flange thickness and stresses shall be SH= 27,300 lb/in.2 calculated in accordance with procedures of SR= 27,300 lb/in.2 Loads:
Para. 1-704.5.1 Flanged Joints, of B31.7 ST= 27,300 lb/in.2 Nuclear Piping Code.
Design pressure and temperature Gasket load Operational load Seismic load-Design Basis Earthquake Flange Stress Limits:
SH, SR, ST 1.5 Sm per ASME Nuclear Pump and Valve Code Set Point
- 5. Valve Spring-Torsional Stress S = 82,500 lb/in2 Loads:
where:
Smax = torsional stress, lb/in2 Maximum Lift W1 = Set point load P = W1 or W2 = spring load, W2 = Spring load at maximum D = means diameter of coil, in.
S = 112,500 lb/in.2 lift, lb d = diameter of wire, in.
C = D = correction factor d
Torsional Stress Limit 0.67 X torsional elastic limit when subjected to a load of W1.
0.90 X torsional elastic limit when subjected to a load of W2.
S PD d
C C
C max
8 4
1 4
4 0615 3
BFN-27 Sheet 6 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Safety Valves (Continued)
Criteria Method of Analysis Allowable Stress Minimum Dimension Required
- 6. Yoke Rod Area Loads:
where:
Spring load at maximum lift A = required area per rod, in2 A = 0.852 in.2 F = total spring load, lb Primary Stress Limit:
Sm = allowable stress, lb/in.2 Allowable stress intensity, Sm, as defined by ASME Standard Code for Pumps and Valves for Nuclear Power.
- 7. Yoke Bending and Shear Stresses Sb = 18,200 lb/in.2 Loads:
where:
Ss = 10,900 lb/in.2 Spring load at maximum lift Sb = bending stress, lb/in.2 Ss = shear stress, lb/in.2 Bending and Shear Stress Limits:
M = bending moment, in.-lb Z = section modulus, in.3 Bending-allowable stress intensity, V = vertical shear, lb Sm, per ASME Nuclear Pump and Valve A = shear area, in.2 Code Shear - 0.6 X allowable stress intensity, 0.6 Sm, per ASME Nuclear Pump and Valve Code.
- 8.
Body Minimum Wall Thickness Loads:
where:
Body Bowl t = required thickness, in t = 0.3312 in Primary service pressure S = allowable stress, 7,000 lb/in.2 P = primary service pressure, 150 lb/in2 Inlet Nozzle Primary Stress Limit:
d = inside diameter of valve at t = 0.231 in.
section being considered, in.
Allowable stress, 7,000 lb/in2, Outlet Nozzle in accordance with USAS B16.5.
t = 0.2823 in.
A F
Sm
2 S
M Z
S V
A b
s
t Pd S
P C
15 2 1
.2
BFN-27 Sheet 7 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Safety Valves Criteria Method of Analysis Allowable Stress Load Limit
- 9. Inlet Nozzle Combined Stress S = 27,300 lb/in.2 Loads:
where:
S = combined bending and tensile Spring load at maximum lift stress, lb/in.2 Operational load F1 = maximum spring load, lb Seismic load-Design Basis Earthquake F2 =
vertical component of reaction thrust, lb Combined Stress Limit:
A = cross section area of nozzle, in.2 1.5 X allowable stress intensity, M1 = moment resulting from horizontal 1.5 Sm, per ASME Code for Pumps component of reaction, lb-in.
and Valves for Nuclear Power.
M2 = moment resulting from horizontal seismic force, in.-lb
- 10. Spindle Diameter Load limit (0.2Fc)
Loads:
where:
F = 30,210 lb Spring load at Maximum lift Fc = critical buckling load, lb E = modulus of elasticity, lb/in.2 Spindle Column Load Limit:
I = moment of inertia, in.4 L = length of spindle in compression, in.
0.2 X critical buckling load
- 11. Spring Washer Shear Area Ss = 15,960 lb/in.2 Loads where:
Spring load at maximum lift Ss = shear stress, lb/in.2 F = spring load, lb Shear Stress Limit:
A = shear area, in.2 0.6 X allowable stress intensity, 0.6Sm, per ASME Nuclear Pump and Valve Code.
S F
F A
M M
Z
1 2
1 2
F EI L
c
2 2
S F
A s
BFN-27 Sheet 8 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Relief Valves Criteria Method of Analysis Minimum Dimension Required
- 1. Body Minimum Wall Thickness Main Body:
Loads:
where:
t = 0.625 in.
Design pressure and temperature t = minimum required thickness, in.
Bonnet:
S = allowable stress, 7,000 lb/in.2 Primary Membrane Stress Limit:
P = primary service pressure, 655 t = 0.287 in.
d = inside diameter of valve at section Allowable working stress as being considered, in.
defined by USAS B16.5 (7,000 C = corrosion allowance, 0.12 in.
psi at primary service pressure).
Minimum Thickness t = 0.612 in.
Loads:
where:
t = minimum required thickness, in.
Pilot Base:
Design pressure and temperature d = diameter or short span, in.
Gasket load C = attachment factor, ASME t = 2.117 in.
Section VIII Primary Stress Limit:
P = design pressure, lb/in.2 Sm = allowable stress, lb/in.2 Allowable stress intensity, Sm, W = total bolt load, lb as defined by ASME Standard hg = gasket moment arm, in.
Code for Pumps and Valves C1 = corrosion allowance, 0.12 in.
for Nuclear Power.
t 1.5 PD 2S 1 2P C
t d
CP S
WhG S d C
m m
178 3
1 2 1
/
BFN-27 Sheet 9 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Relief Valves (Continued)
Criteria Method of Analysis Allowable Stress Minimum Dimension Required
- 3.
Flange Bolt Area - Inlet Flange, Total bolting loads and stresses shall be Body to Base:
Outlet Flange, Body to Bonnet, calculated in accordance with procedures Bonnet to Base of Para. 1-704.5.1 Flanged Joints, of Ab = 10.26 in2 Ab = 2.854 in.2 B31.7 Nuclear Piping Code Loads:
Bonnet to Cap:
Design pressure and temperature Ab = 1.452 in.2 Ab = 0.995 in.2 Gasket load Operational load Inlet Flange Design Basis Earthquake Ab = 13.9 in.2 Ab = 6.25 in.2 Bolting Stress Limit:
Outlet Flange:
Allowable stress intensity, Sm as Ab = 12.2 in2 defined by ASME Standard Code for Ab = 5.5 in.2 Pumps and Valves for Nuclear Power.
- 4.
Flange Thickness - Inlet, Outlet, Flange thickness and stresses shall be Bonnet Flanges calculated in accordance with procedures SH = 26,250 lb/in.2 of Para. 1-704.5.1 Flanged Joints, of SR = 26,250 lb/in.2 Loads:
B31.7 Nuclear Piping Code ST = 26,250 lb/in.2 Design pressure and temperature Gasket load Operational load Design Basis Earthquake Flange Stress Limits:
SH, SR, ST 1.5 Sm per ASME Nuclear Pumps and Valve Code.
BFN-27 Sheet 10 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Main Steam Relief Valves (Continued)
Criteria Method of Analysis Allowable Stress
- 5.
Valve Disc. Thickness and Stress Disc Stress:
Loads:
where:
Sm = 15,800 lb/in2 Design pressure and temperature Sr = radial stress, lb/in2 St = tangential stress, lb/in2 Primary Stress Limit:
v = Poisson's ratio P = design pressure, lb/in2 Allowable stress intensity, Sm R = radius of disc, in.
as defined by ASME Standard Code for t = thickness of disc, in.
Pumps and Valve for Nuclear Power.
Inlet Nozzle Diameter Thickness and Stress Inlet Nozzle Stress:
Loads:
where:
S = 26,250 lb/in2 S = combined bending and tensile Design pressure and temperature stress, lb/in2 Operational load F1 = vertical load due to design pressure, lb Design Basis Earthquake F2 = vertical component of reaction thrust, lb Primary Stress Limit:
A = cross section area of nozzle, in2 M1 = moment resulting from horizontal 1.5 X allowable stress intensity, reaction, in.-lb 1.5 Sm as defined by ASME M2 = moment resulting from horizontal Standard Code for Pumps and seismic force at mass center of Valves for Nuclear Power.
valve, in.-lb
S S
v PR t
r t
3 3 8
2 2
S F
F A
M M
Z
1 2
1 2
BFN-27 Sheet 11 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Pumps Criteria Method of Analysis Allowable Stress Minimum Dimension Required
- 1.
Casing Minimum Wall Thickness t = 2.68 in.
Loads: Normal and Upset Condition where:
Design pressure and temperature t = minimum required thickness, in.
P = design pressure, psig Primary Membrane Stress Limit:
R = maximum internal radius, in.
S = allowable working stress, psi Allowable working stress per E = joint efficiency ASME Section III, Class C C = corrosion allowance, in.
- 2.
Casing Cover Minimum Thickness Loads: Normal and Upset Condition Design pressure and temperature Sr = 15,075 psi Primary Bending Stress Limit:
1.5 Sm per ASME code for Pumps and Valves for St = 15,075 psi Nuclear Power Class I where:
Sr = radial stress at outer edge, psi St = tangential stress at inner edge, psi w = pressure load, psi W =
uniform load along inner edge, lb t = disc thickness, in.
m = reciprocal of Poisson's ratio a = radius of disc, in.
b =
radius of disc hole, in.
Sr 3W 4t2 a2 2b2 b4 m 1
4b4 m 1 ln a b a2 b2 m 1
a2 m 1
b2 m 1
/
3W 2pt 2 1
2mb 2 2b 2 m
1 ln a b a 2 m
1 b 2 m 1
/
S t 3W m 2 1
4mt 2 a 4 b 4 4a 2 b 2 ln a b a 2 m
1 b 2 m
1
/
3W 2pmt 2 1
ma 2 m
1 mb 2 m
1 2 m 2 1 a 2 ln a b a 2 m
1 b 2 m
1
/
t PR SE 06P C
BFN-27 Sheet 12 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Pumps (Continued)
Criteria Method of Analysis Allowable Stress Minimum Dimension Required
- 3.
Cover and Seal Flange Bolt Areas Bolting loads, areas and stresses shall be calculated in accordance with "Rules for Loads: Normal and upset conditions Bolted Flange Connections" - ASME 20,000 psi Section VIII, Appendix II Design pressure and temperature Design gasket load 20,000 psi Bolting Stress Limit:
Allowable working stress per ASME Section III, Class C
- 4.
Cover Clamp Flange Thickness Flange thickness and stress shall be Flange Thickness calculated in accordance with "Rules 8.9 in.
Loads: Normal and upset condition for Bolted Flange Connections" -ASME Section VIII, Appendix II Design pressure and temperature Design gasket load Design bolting load Tangential Flange Stress Limit:
Allowable working stress per ASME Section III, Class C
- 5.
Pump Nozzle Stress Pipe Stress is compared to allowable 21,708 psi of 0.9 (Yield stress of pump nozzle)
Loads: Normal, Upset and Faulted Condition Sheet 13 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Pumps (Continued)
Criteria Method of Analysis Allowable Stress
- 6.
Mounting Bracket Combined Stress Bracket vertical loads shall be determined summing the equipment and fluid weights
BFN-27 Loads:
and vertical seismic forces.
Pump Lug Bracket horizontal loads shall be determined Flood weight by applying the specified seismic force at 17,280 psi Design Basis Earthquake mass center of pump-motor assembly (flooded).
Combined Stress Limit:
Horizontal and vertical loads shall be applied simultaneously to determine Yield Stress tensile, shear and bending stresses in Motor Lug the brackets. Tensile shear, and bending stress shall be combined to determine 21,000 psi maximum combined stresses.
BFN-27 Sheet 14 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Pumps (Continued)
Criteria Method of Analysis Allowable Stress
- 7.
Stresses Due to Seismic Loads The flooded pump-motor assembly shall Motor Bolt Tensile Stress:
be analyzed as a free body supported by Loads:
constant support hangers from the pump 11,200 psi brackets. Horizontal and vertical seismic Operating pressure and forces shall be applied at mass center of Pump Cover Bolt Tensile Stress:
temperature assembly and equilibrium reactions shall Design Basis Earthquake be determined for the motor and pump 32,000 psi brackets. Load, shear, and moment Combined Stress Limit:
diagrams shall be constructed using live Motor Support Barrel loads, dead loads, and calculated snubber Combined Stress:
Yield stress reactions. Combined bending, tension and shear stresses shall be determined 22,400 psi for each major component of the assembly including motor, motor support barrel, bolting and pump casing. The maximum combined tensile stress in the cover bolting shall be calculated using tensile stresses determined from loading diagram plus tensile stress from operating pressure.
BFN-27 Sheet 15 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Fuel Storage Racks Criteria Loading Location Allowable Stress Stresses due to normal, upset, or emergency Emergency condition At column to base welds 11,000 psi(1) loading shall not cause the racks to fail "A" loads so as to result in a critical fuel array
- 1. Dead loads At base hold down lug 20,000 psi(2)
- 2. Full fuel load in rack (casting)
- 3. Design Basis Earthquake Primary Stress Limit-Paper numbers 3341 and 3342, Proceedings of the ASCE, Journal Emergency condition of the Structural Division, December 1962 "B" loads (see below)
(task committee on lightweight alloys)
(Aluminum)
Emergency Conditions Stress limit = yield strength at 0.2% offset.
(1) Load testing shows that the structure will not yield when subjected to simulated emergency condition "A" loads.
Strain gages mounted on the welds show that calculated stresses are conservative.
(2) Calculated stresses compare very well with test results.
Emergency Condition "B" Loading In addition to the loading conditions given above, the racks are tested and analyzed to determine their capability to safely withstand the accidental, uncontrolled drop of the fuel grapple from its full retracted position into the weakest portion of the rack.
Method of Analysis The displacement of the vertical columns at the ends of the racks is determined by considering the effect of the grapple kinetic energy on the upper structure. The energy absorbed shearing the rack longitudinal structural member welds is determined.
The effect of the remaining energy on the vertical columns is analyzed. Equivalent static load tests are made on the structure to assure that the criteria are met.
BFN-27 Sheet 16 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RHR Pumps Criteria Method of Analysis Allowable Stress
- 1. Closure bolting shall be designed to
- 1. Bolting loads and stresses shall be 25,000 psi contain the internal design pressure calculated in accordance with the "Rules of the pump casing without exceeding for Bolted Flange Connections," ASME the allowable stress of the bolting Boiler and Pressure Vessel Code, material. Allowable stresses at Section VIII, Appendix II.
design temperature shall be in accordance with ASME Boiler and Pump Design Pressure 450 psig pressure Vessel Code,Section VIII.
Maximum Design Temperature 350F
- 2. The minimum wall thickness of the
- 2. Stress in the pump casing shall be 14,000 psi pump shall limit stress to the calculated at the point of maximum allowable stress when subjected to internal pump diameter by the formula design pressure and temperature.
Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
where Sc = calculated stress, psi P = pump design pressure, psi D = maximum pump internal diameter t =
actual minimum metal thickness less corrosion allowance, 0.080 in.
S P D t
t c
0 2
.2
BFN-27 Sheet 17 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RHR Pumps (Continued)
Criteria Method of Analysis and Allowable Nozzle Loads
- 3. Application of forces and moments by
- 3. Stresses will not be excessive if the attaching pipe on pump nozzles under maximum resultant force when taken with combined maximum thermal expansion the maximum resultant moment falls below and Operating Basis Earthquake the line.
loading reaction plus load due to internal pressure shall not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME Boiler and Pressure Vessel Code,Section VIII.
Suction OBE DBE Fintercept 88,000 lb 146,000 lb (M=0)
For Design Basis Earthquake stress Mintercept 1,200,000 in.-lb 1,800,000 in.-lb shall be less than 1.5 of allowable (F=0) stress.
Discharge Fintercept 68,000 lb 126,000 lb (M=0)
Mintercept 760,000 in.-lb 1,300,000 in.-lb (F=0)
Pipe Design Pressure Suction = 150 psig Discharge
= 450 psig
BFN-27 Sheet 18 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Core Spray Pumps Criteria Method of Analysis Allowable Stress
- 1.
Closure bolting shall be designed to
- 1. Bolting loads and stresses shall be 20,000 psi contain the internal design pressure calculated in accordance with the "Rules of the pump casing without exceeding for Bolted Flange Connections," ASME the allowable stress of the bolting Boiler and Pressure Vessel Code, Section material. Allowable stresses at VIII, Appendix II.
design temperature shall be in accordance with ASME Boiler and Pump Design Pressure 500 psig Pressure Vessel Code,Section VIII.
Maximum Design Temperature 210F
- 2.
The minimum wall thickness of the
- 2. Stress in the pump casing shall be 14,000 psi pump shall limit stress to the allow-calculated at the point of maximum able stress when subjected to design internal pump diameter by the formula pressure and temperature. Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
where Sc = calculated stress, psi 17,500 psi allowable for 216 WCB X P = pump design pressure, psi 0.8 (quality factor) = 14,000 psi D = maximum pump internal diameter t = actual minimum metal thickness less corrosion allowance, 0.080 in.
S P D t
t c
0 2
.2
BFN-27 Sheet 19 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Core Spray Pumps (Continued)
Criteria Method of Analysis and Allowable Nozzle Loads Representative Results
- 3. Application of forces and moments by
- 3. Stresses will not be excessive if the attaching pipe on pump nozzles under maximum resultant force when taken with the combined maximum thermal expansion maximum resultant moment falls below the line.
Operating Basis Earthquake loading reaction plus load due to internal pressure shall not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME Boiler and Pressure Vessel Code,Section VIII.
Suction OBE DBE Fintercept 66,686 lb 104,955 lb (M=0)
For Design Basis Earthquake stress Mintercept 564,193 in.-lb 880,105 in.-lb shall be less than 1.5 of allowable (F=0) stress.
Discharge Fintercept 35,105 lb 65,982 lb (M=0)
Mintercept 266,479 in.-lb 463,492 in.-lb (F=0)
Pipe Design Pressure Suction = 125 psig Discharge = 500 psig
BFN-27 Sheet 20 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES HPCI Pumps Criteria Method of Analysis Allowable Stress
- 1.
Closure bolting shall be designed to
- 1. Bolting loads and stresses shall be Main Pump contain the internal design pressure calculated in accordance with the "Rules of the pump casing without exceeding for Bolted Flange Connections," ASME 20,000 psi the allowable stress of the bolting Boiler and Pressure Vessel Code, Section material. Allowable stresses at VIII, Appendix II.
Boost Pump design temperature shall be in accordance with ASME Boiler and Main Pump Design Pressure 1500 psig 20,000 psi Pressure Vessel Code,Section VIII.
Boost Pump Design Pressure 450 psig
- 2.
The minimum wall thickness of the
- 2. Stress in the pump casing shall be Main Pump pump shall limit stress to the allow-calculated at the point of maximum able stress when subjected to design internal pump diameter by the formula 14,000 psi pressure and temperature. Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
Volute stress shall be calculated by the Boost Pump following formula 14,000 psi The maximum stress in the pump Roark casing when subjected to design
- p. 307 Case 26 pressure shall not exceed the allow-able working stress of the material.
The allowable stress shall be in and R = a - 0.5b accordance with ASME Boiler and Pressure Vessel Code,Section III.
S P D t
ET h
0 2
.2 S
Pb R
a R
v t
2
BFN-27 Sheet 21 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES HPCI Pumps (Continued)
Criteria Method of Analysis and Allowable Nozzle Loads
- 3. Application of forces and moments by
- 3. Stresses will not be excessive if the attaching pipe on pump nozzles under maximum resultant force when taken with the combined maximum thermal expansion maximum resultant moment falls below the line.
and Operating Basis Earthquake loading reaction plus load due to internal pressure shall not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME Boiler and Pressure Vessel Code,Section VIII.
Suction OBE DBE Fintercept 33,000 lb 43,000 lb (M=0)
For Design Basis Earthquake stress Mintercept 500,000 in.-lb 700,000 in.-lb shall be less than 1.5 of allowable (F=0) stress.
Discharge Fintercept 32,000 lb 47,000 lb (M=0)
Mintercept 250,000 in.-lb 460,000 in.-lb (F=0)
Pipe Design Pressure Suction = 150 psig Discharge
= 1500 psig
BFN-27 Sheet 22 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RCIC Pump Criteria Method of Analysis Allowable Stress
- 1. Closure bolting shall be designed to
- 1. Bolting loads and stresses shall be contain the internal design pressure calculated in accordance with the "Rules of the pump casing without exceeding for Bolted Flange Connections," ASME 20,000 psi the allowable stress of the bolting Boiler and Pressure Vessel Code, Section material. Allowable stresses at VIII, Appendix II.
design temperature shall be in accordance with ASME Boiler and Pump Design Pressure 1500 psig Pressure Vessel Code,Section VIII.
- 2. The minimum wall thickness of the
- 2. Stress in the pump casing shall be 14,000 psi pump shall limit stress to the allow-calculated at the point of maximum able stress when subjected to design internal pump diameter by the formula pressure and temperature. Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
SC = 0.8Sa The maximum stress in the pump Volute stress shall be computed by the 14,000 psi casing when subjected to design following formula:
pressure shall not exceed the allowable working stress of the Roark p.
material. The allowable stress 225 Case No. 36 shall be in accordance with ASME Boiler and Pressure Vessel Code,Section III.
= factor from Roark a = volute length b = volute width
S P D t
tE c
.02 2
S P
t b
b
2 2
BFN-27 Sheet 23 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RCIC Pump (Continued)
Criteria Method of Analysis and Allowable Nozzle Loads
- 3.
Application of forces and moments by
- 3. Stresses will not be excessive if the attaching pipe on pump nozzles under maximum resultant force when taken with the combined maximum thermal expansion maximum resultant moment falls below the line.
and Operating Basis Earthquake loading reaction plus load due to internal pressure shall not produce an equivalent bending and torsional stress in the nozzles in excess of the allowable stress as defined by the ASME Boiler and Pressure Vessel Code,Section VIII.
Suction OBE DBE Fintercept 9,000 lb 13,500 lb (M=0)
For Design Basis Earthquake stress Mintercept 54,000 in.-lb 69,000 in.-lb shall be less than 1.5 of allowable (F=0) stress.
Discharge Fintercept 9,000 lb 13,500 lb (M=0)
Mintercept 54,000 in.-lb 69,000 in.-lb (F=0)
Pipe Design Pressure Suction = 150 psig Discharge = 1500 psig
BFN-27 Sheet 24 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Standby Liquid Control Pumps Criteria Method of Analysis Allowable Stress
- 1. Closure bolting shall be designed to
- 1. Bolting loads and stresses shall be Stuffing Box Bolts contain the internal design pressure calculated in accordance with the "Rules of the pump without exceeding the for Bolted Flange Connections," ACME 25,000 psi allowable working stress of the Boiler and Pressure Vessel Code, Section bolting material. Allowable stresses VIII, Appendix II.
Cylinder Head Bolts shall be in accordance with ASME Boiler and Pressure Vessel Code.
25,000 psi
- 2. The maximum stress in the pump
- 2. Stress in the pump fluid cylinder shall be 16,500 psi fluid cylinder when subjected to calculated at the point of maximum stress design pressure shall not exceed by the pressure area method.
the allowable working stress of the material. The allowable stress Pump Design Pressure 1400 psig shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
- 3. The stresses in the motor mounting
- 3. The seismic forces acting on the motor to Tension bolts when the motor is subjected subject the bolting to shear or tension to the Design Basis Earthquake shall are considered. The motor is isolated 16,500 psi not exceed 0.9 of yield stress and from the pump and nozzle forces by the twice the allowable shear stress for flexible coupling.
Shear bolting material in accordance with the ASME Boiler and Pressure Vessel 10,000 psi Code,Section VIII.
BFN-27 Sheet 25 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Standby Liquid Control Pumps (Continued)
Criteria Method of Analysis and Allowable Nozzle Loads
- 4. The stresses in the pump mounting bolts
- 4. The maximum force taken with the maximum due to the combination of Operating resultant moment shall fall below the line on the Basis Earthquake acting on the flooded force-moment diagram:
pump plus the attaching pipe reactions under combined maximum thermal expan-sion plus Operating Basis Earthquake shall not exceed the allowable shear and tensile stresses for the bolting material in accordance with the ASME Boiler and Pressure Vessel code,Section VIII. The attaching pipe reaction plus the load due to internal pressure shall not produce an equivalent bending and torsional stress in OBE nozzles in excess of the allowable Discharge M = 2.3 (342-F) stress.
not to exceed 283 ft-lb The stresses in the pump mounting bolts Suction M = 4.59 (711-F) due to the combination of the Design not to exceed 1385 ft-lb Basis Earthquake acting on the flooded DBE pump plus the attaching pipe reactions Discharge M = 2.3 (684-F) under combined maximum thermal expan-not to exceed 444 ft-lb sion plus Design Basis Earthquake shall Suction M = 4.59 (1422-F) not exceed 0.9 times the yield stress not to exceed 2060 ft.lb in tension and twice the allowable shear stress for the bolting material Where M is maximum moment (ft-lb) in in accordance with the ASME Boiler and any direction and F is maximum force Pressure vessel Code,Section VIII.
(lb) in any direction.
The attaching pipe reaction plus the load due to internal pressure shall not produce an equivalent bending and tor-sional stress in nozzles in excess of 1.5 times allowable stress.
BFN-28 Sheet 26 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RHR Service Water Pumps A2, A3, B2, B3, C1, C2, C3 Pipe Design Pressure Discharge = 185 psig Stresses will not be excessive if the loads on the pump nozzles do not exceed the following values:
Condition F(Axial) F(Vertical) F(Lateral) M(Torsion) M(Vertical) M(Lateral)
Normal 6,211 lb 6,888 lb 3,882 lb 5,552 ft-lb 17,499 ft-lb 10,419 ft-lb Upset 9,110 lb 8,970 lb 5,103lb 8,790 ft-lb 19,218 ft-lb 13,006 ft-lb Emergency 12,010 lb 11,052 lb 6,984 lb 12,047 ft-lb 30,527 ft-lb 15,593 ft-lb
BFN-27 Sheet 27 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RCIC Turbine Criteria Method of Analysis Allowable Stress
- 1.
Closure bolting shall be designed to
- 1. Bolting loads and stresses shall be 20,000 psi contain the internal design pressure calculated in accordance with the "Rules of the turbine casing without for Bolted Flange Connections," ACME exceeding the allowable working Boiler and Pressure Vessel Code, Section stress of the bolting material.
VIII, Appendix II.
Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
- 2.
The maximum wall thickness of the
- 2. Stresses in the various pressure contain-17,500 psi turbine casing shall be based on ing portions of the turbine casing shall that to limit stress to the allowable be calculated according to the rules of working stress when subjected to Part UG,Section VIII, of the ASME Boiler design pressure plus corrosion and Pressure Vessel Code.
allowance. Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
BFN-27 Sheet 28 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RCIC Turbine (Continued)
Criteria Method of Analysis and Allowable Nozzle Loads
- 3.
The forces and moments imposed by the
- 3. The total resultant of the forces and the total attached piping on the turbine inlet resultant of the moments on both the inlet and and exhaust connections shall satisfy the exhaust connections of the turbine shall the following conditions:
satisfy the following equations:
- a. The resultant force and moment For the combination of dead weight and maximum from the combination of dead thermal expansion, weight, and thermal expansion shall be less than that stipulated Inlet F = (2620-M)/3.0 by the equipment vendor.
Exhaust F = (6000-M)/3.0
- b. The resultant force and moment For the combination of dead weight, maximum from the combination of dead thermal expansion, and Operating Basis Earth-weight, thermal expansion, and quake.
Operating (or Design) Basis Inlet F = (3000-M)/2.5 Earthquake shall be less than Exhaust F = 3.0 (6000-M), but not that demonstrated acceptable to exceed 8,370 lb by detailed seismic analysis of the equipment.
For the combination of dead weight, maximum thermal expansion, and Design Basis Earthquake Inlet F = (4500-M)/2.5 Exhaust F = 3.0 (9000-M), but not to exceed 12,555 lb Where "F" is the resultant force in lb and "M" is the resultant moment in ft-lb Typical acceptable area on the force-moment diagram is indicated below:
BFN-27 Sheet 29 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES RCIC Turbine (Continued)
Criteria Method of Analysis
- 4.
The stresses in the turbine anchor
- 4.
Vertical forces on the anchor bolts shall be bolts (turbine to baseplate) due to the sum of the following:
the combination of the Operating Basis Earthquake acting on the
- a. Weight of the turbine assembly times the turbine while running plus the total vertical component of acceleration, piping loads (weight, thermal & OBE)
- b. The vertical pipe force reactions, shall not exceed the allowable tensile
- c. The pipe moment reactions tending to tip the stress nor the allowable shear stress turbine and subject the bolting to tension.
for the bolting materials as specified in the ASME Boiler and Pressure Horizontal forces on the anchor bolts, Vessel Code,Section VIII.
subjecting them to shear, shall be the sum of the following:
- a. Weight of the turbine assembly times the horizontal component of acceleration,
- b. The horizontal pipe force reactions,
- c. The effect of pipe moment reactions causing horizontal loading at the anchor bolts NOTE: Friction shall not be considered to be restrictive
- 5.
The stresses in the turbine anchor
- 5. Same as analysis under 4, above.
bolts (turbine to baseplate) due to the combination of Design Basis Earthquake acting on the turbine in standby plus the total piping loads (weight, thermal, and DBE) shall not exceed 0.9 times the yield stress in tension, nor twice the allowable shear stress for the bolting materials as specified in the ASME Boiler and Pressure Vessel Code,Section VIII.
BFN-27 Sheet 30 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES HPCI Turbine Criteria Method of Analysis Allowable Stress
- 1.
Closure bolting shall be designed to
- 1. Bolting loads and stresses shall be 20,000 psi contain the internal design pressure calculated in accordance with the "Rules of the turbine casing without for Bolted Flange Connections," ASME exceeding the allowable working Boiler and Pressure Vessel Code, Section stress of the bolting material.
VIII, Appendix II.
Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel code,Section VIII.
- 2.
The minimum wall thickness of the
- 2. Stresses in the various pressure 17,500 psi turbine casing shall be based on that containing portions of the turbine casing to limit stress to the allowable work-shall be calculated according to the rules ing stress when subjected to design of Part UG,Section VIII, of the ASME pressure plus corrosion allowance.
Boiler and Pressure Vessel Code.
Allowable stresses shall be in accordance with ASME Boiler and Pressure Vessel Code,Section VIII.
BFN-27 Sheet 31 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES HPCI Turbine (Continued)
Criteria Method of Analysis and Allowable Nozzle Loads
- 3.
The forces and moments imposed by the
- 3. The total resultant of the forces and the total attached piping on the turbine inlet of the moments on both the inlet and and exhaust connections shall satisfy the connections of the turbine shall the following conditions:
satisfy the following equations:
- a. The resultant force and moment For the combination of dead weight and from the combination of dead maximum thermal expansion, weight and thermal expansion shall be less than that stipulated Inlet F = (7570-M)/3.0 by the equipment vendor.
Exhaust F = (9930-M)/3.0
- b.
The resultant force and moment For the combination of dead weight, maximum from the combination of dead thermal expansion, and Operating Basis Earthquake weight, thermal expansion, and Inlet F = (20,000-M)/2.5 but not Operating (or Design) Basis to exceed 5000 lb Earthquake shall be less than Exhaust F = (20,000-M)/0.8, but not that demonstrated acceptable to exceed 11,500 lb by detailed seismic analysis of the equipment For the combination of dead weight, maximum thermal expansion, and Design Basis Earthquake, Inlet F = (30,000-M)/2.5, but not to exceed 17,250 lb Exhaust F = (30,000-M)/0.8, but not to exceed 17,250 lb Where "F" is the resultant force in lb and "M" is the resultant moment in ft-lb Typical acceptable area on the force-moment diagram is indicated below:
BFN-27 Sheet 32 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES HPCI Turbine (Continued)
Criteria Method of Analysis
- 4.
The stresses in the turbine anchor
- 4.
Vertical forces on the anchor bolts shall be the bolts (turbine to baseplate) due to sum of the following:
the combination of the Operating Basis Earthquake acting on the turbine while
- a. Weight of the turbine assembly times the running plus the total piping loads vertical component of acceleration, (weight, thermal and OBE) shall not
- b. The vertical pipe force reactions, exceed the allowable tensile stress
- c. The pipe moment reactions tending to tip the nor the allowable shear stress for turbine and subject the bolting to tension.
the bolting materials as specified in the ASME Boiler and Pressure Horizontal forces on the anchor bolts, subjecting Vessel Code,Section VIII.
them to shear, shall be the sum of the following:
- a. Weight of the turbine assembly times the horizontal component of acceleration,
- b. The horizontal pipe force reactions,
- c. The effect of pipe moment reactions causing horizontal loading at the anchor bolts NOTE: Friction shall not be considered to be restrictive
- 5.
The stresses in the turbine anchor
- 5.
Same as analysis under 4, above.
bolts (turbine to baseplate) due to the combination of Design Basis Earthquake acting on the turbine in standby plus the total piping loads (weight, thermal and OBE) shall not exceed 0.9 times the yield stress in tension, nor twice the allowable shear stress for the bolting materials as specified in the ASME Boiler and Pressure Vessel Code,Section VIII.
BFN-27 Sheet 33 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Valves - Units 1 and 2 Criteria Method of Analysis Allowable Stress Minimum Dimension Required
- 1.
Body Minimum Wall In Pipe Run Codes and Standards 2 in. (Equalizer Bypass Valve)
- 1.
USAS B31.1.0 1967 t = 0.253 in.
2 in. Equalizer Bypass Valve
- 2. Manufacturers Standards 4 in. (Discharge Bypass Valve) 4 in. Discharge Bypass Valve Society MSS-SP.66 t = 0.405 in.
22 in. Equalizer Valve 28 in. Suction Valve 22 in. (Equalizer Valve) 28 in. Discharge Valve t = 1.520 in.
Loads:
where:
Design Pressure t = minimum wall thickness, in.
28 in. (Suction Valve)
Design Temperature P = design pressure, psig t = 1.938 in.
d = minimum diameter of flow passage, but not less than 28 in (Discharge Valve)
Primary Membrane Wall 90% of inside diameter at t = 1.938 in.
Thickness welding end, in.
S = allowable working stress, psi y = plastic stress distribution factor, 0.4
- 2.
Body-to-Bonnet Bolt Area Loads ASME Boiler and Pressure Vessel 2 in. (Equalizer Bypass Valve)
Code,Section VIII, Appendix II, 2 in. Equalizer Bypass Valve 1968 Edition.
Sallow = 29,000 lb/in.2 4 in. Discharge Bypass Valve Loads:
Total bolting loads and stresses 4 in. (Discharge Bypass Valve) shall be calculated in accordance Design pressure and temperature with "Rules for Bolted Flange Con-Sallow = 29,000 lb/in.2 Gasket load nections," ASME Boiler and Pressure Stem operational load Vessel Code,Section VIII, Appendix Design Basis II, except that the stem operation-Earthquake al load and seismic loads shall be included in the total load carried by bolts. The horizontal and vertical seismic forces shall be applied at the mass center of the valve operator assuming that the valve body is rigid and anchored.
t P
S P
y d
15 2
2 1
01
BFN-27 Sheet 34 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Valves - Units 1 and 2 (Continued)
Criteria Method of Analysis Allowable Stress
Code,Section VIII, Appendix II, 2 in. Equalizer Bypass Valve 1968 Edition.
SH SR ST 4 in. Discharge Bypass Valve 20,100 3,426 13,426 Loads:
Flange thickness and stress shall be calculated in accordance with 4 in. (Discharge Bypass)
Design pressure and temperature "Rules for Bolted Flange Connec-Gasket load tions," ASME Boiler and Pressure 20,100 13,426 13,426 Stem operational load Vessel Code,Section VIII, Appen-Seismic load -
dix II, except that the stem Design Basis operational load and seismic loads Earthquake shall be included in the total load carried by the flange. The horizontal and vertical seismic forces shall be applied at the mass center of the valve operator assum-ing that the valve body is rigid.
- 4. (A) Body and Bonnet Flange ASME Boiler and Pressure Vessel Primary Stresses Stress Code,Section III, Article 4 Membrane Stress Allowable =
(B) Body Neck Wall Stress Primary, secondary, and peak 15,800 psi stresses were analyzed in accordance 22 in. Equalizer Valves with ASME Section III using finite Local Membrane Stress Allowable =
28 in. Suction Valves element computer analysis. The 23,700 psi 28 in. Discharge Valves model was verified by strain gage Primary Plus Secondary Stresses tests Loads:
Code Allowable - 3Sm =
Design pressure and 47,400 psi Design temperature
BFN-27 Sheet 35 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Valves - Units 1 and 2 Criteria Method of Analysis Allowable Stress
- 5. Body to Bonnet Bolting Under operating conditions Loads:
67,000 psi Design Pressure Maximum conditions Design Temperature 100,500 psi
- 6. Valve Operator Support Bolting The valve assembly is analyzed Sb allowable = 20,000 lb/in.2 assuming that the valve body is an 2 in Equalizer Bypass Valve anchored, rigid mass and that the 4 in. Discharge Bypass Valve specified vertical and horizontal 22 in. Equalizer Valve seismic forces are applied at the 28 in. Suction Valve mass center of the operator assembly, 28 in. Discharge Valve simultaneously with operating pres-sure plus dead weight plus opera-Loads:
tional loads. Using these loads, stresses and deflections are deter-Design Pressure and Temperature mined for the operator support Stem operational load components.
Equipment dead weight Seismic load Design Basis Earthquake
BFN-27 Sheet 36 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Valves - Unit 3 Criteria Method of Analysis Allowable Stress Minimum Required Dimension
- 1. Body Minimum Wall In Pipe Run 22 in. Valve - t = 1.52 in.
Loads:
4 in. Valve - t = 0.405 in.
Design pressure and temperature where:
2 in. Valve - t = 0.253 in.
t = minimum wall thickness, in.
Primary Membrane Stress Limit:
P = design pressure, psig 28 X 24 X 28 in. Valve -
d = minimum diameter of flow t = 1.677 in. (Suction)
Allowable working stress per passage but not less than 90%
ASME Section 1 of inside diameter at welding 28 X 24 X 28 in. Valve -
end, in.
t = 1.938 in. (Discharge)
S = allowable working stress, psi y = plastic stress distribution factor, 0.4
- 2. Body-to-Bonnet Bolt Area Total bolting loads and stresses Flanged Bolt Stress shall be calculated in accordance Loads:
with "Rules for Bolted Flange Sallow = 29,000 lb/in.2 Connections," ASME Boiler and Design pressure and temperature Pressure Vessel Code,Section VIII, Gasket load Appendix II, except that the stem Stem operational load operational load and seismic loads Seismic load -
shall be included in the total load Design Basis Earthquake carried by bolts. The horizontal and vertical seismic forces shall Bolting Stress Limit:
be applied at the mass center of the valve operator assuming that Allowable working stress per the valve body is rigid and anchored.
ASME Boiler and Pressure Vessel Code,Section VIII, Appendix II, 1968 Edition.
t 1.5 Pd 2S 2P 1 y
0.1
BFN-27 Sheet 37 Table C.4-2 (Continued)
PRIMARY SYSTEM COMPONENTS - CRITICAL LOAD COMBINATIONS, LOCATIONS, AND ALLOWABLES Recirculation Valves - Unit 3 (Continued)
Criteria Method of Analysis Allowable Stress
- 3. Flange Stress Flange thickness, and stress shall be SH: 20,100 lb/in.2 (Hub Stress) calculated in accordance with "Rules SR: 13,426 lb/in.2 (Radial Stress)
Loads:
for Bolted Flange Connections"-ASME ST: 13,426 lb/in.2 (Tangential Stress)
Boiler and Pressure Vessel Code, Design pressure and temperature Section VIII, Appendix II, except Gasket load that the stem operational load and Stem operational load seismic loads shall be included in Seismic Loads -
the total load carried by the flange.
Design Basis The horizontal and vertical seismic Earthquake forces shall be applied at the mass center of the valve operator as-Flange Stress Limits; suming that the valve body is rigid.
SH,SR,ST:
Sm per ASME Boiler and Pres-sure Vessel Code,Section VIII, Appendix II, 1968 Edition.
- 4. Valve Operator Support Bolts The valve assembly is analyzed assum-Sb allowable = 20,000 lb/in.2 ing that the valve body is an anchored, Loads:
rigid mass and that the specified vertical and horizontal seismic forces Design pressure and temperature are applied at the mass center of the Stem operational load operator assembly, simultaneously with Equipment dead weight operating pressure plus dead weight Seismic load -
plus operational loads. Using these Design Basis loads, stresses and deflections are Earthquake determined for the operator support components.
Yoke and Yoke Bolt Stress Limits:
Allowable working stress per ASME Section VIII.
BFN-27 Sheet 1 of 1 TABLE C.5-1 DRYWELL-LOADING CONDITIONS AND ALLOWABLE STRESSES Loading Allowable Stress Intensity (ksi)
Condition Loading Components (Notes 1 and 2)
Initial and Final Dead Loads Pm < Sm = 17.5 Test Condition Test Pressure PL < 1.5 Sm = 26.3 Vent Thrusts PL + Pb < 1.5 Sm = 26.3 OBE PL + Pb + Q < 3.0 Sm = 52.5 Normal and Upset Dead Loads Pm < Sm = 17.5 Operating Condition Vent Thrusts PL < 1.5 Sm = 26.3 OBE PL + Pb < 1.5 Sm = 26.3 Accident Temperature PL + Pb + Q < 3.0 Sm = 52.5 Accident Pressure Emergency Condition Dead Loads Region not Backed by Concrete (Note 3)
Accident Pressure Pm < 0.9 Sy = 30.3 Accident Temperature PL < 0.9 Sy = 30.3 Vent Thrusts OBE Region Backed by Concrete Jet Loads Pm < Sy = 33.7 PL < 1.5Sy = 50.6 Flooded Condition Dead Loads Pm < Sy = 38.0 Hydrostatic Pressure PL < Sy = 38.0 From Flooded DryWell PL + Pb < Su = 70.0 DBE PL + Pb + Q < Su = 70.0 NOTE:
- 1.
Stress intensities are based on ASME Boiler and Pressure Vessel Code,Section III, Subsection B of Reference 17.
- 2.
Definition of symbols are as follows:
Pm = Primary membrane stress, PL = Primary local membrane stress, Pb = Primary bending stress, Q = secondary stress.
- 3.
The 1965 ASME Code does not address accident conditions. Therefore, this design criteria utilizes the 1968 ASME Code with addenda through the summer of 1969 to establish design allowables for the accident condition for that portion of the vessel backed by concrete.