RC-16-0083, Virgil C. Summer, Unit 1 - Updated Final Safety Analysis Report, Chapter 3, Design of Structures, Components, Equipment and Systems
Text
3.1-1 Reformatted July 2014 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS The chapter includes identif ication, description and discussion of the principal architectural and engineering design of those structures , components, equipment and systems important to safety.
3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA This section discusses briefly the extent to which the design crit eria for the plant structures, systems and components important to safety m eet the NRC "Gen eral Design Criteria for Nuclear Power Plants" specified in Appendix A to 10 CFR 50. A summary is provided to show how the principal design features meet each crit erion and identify any exceptions.
3.1.1
SUMMARY
DESCRIPTION The Virgil C. Summer Nuclear Station is designed, constructed and operated to comply with South Carolina Electric and Gas Company
's understanding of th e intent of the NRC's "General Design Criteria for Nuclear Power Plants," Appendix A to 10 CFR 50. This section outlines the philosophy that will be adopted in meeting the criteria. Detailed evaluations of compliance with t he various General Design Criteria are incorporated in applicable sections of the Final Safety Analysis Report as referenced herein.
3.1.2 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1.2.1 Overall Requirements Criterion 1 - Quality Standards and Records
Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to qualit y standards commensurate wit h the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or m odified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these
structures, systems, and compon ents will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit li censee throughout the life of the unit.
3.1-2 Reformatted July 2014 Discussion
South Carolina Electric and Gas Company an d Westinghouse, with it s subcontractors, maintains, either in their possession or under t heir control, a complete set of records of the design, fabrication, construction and te sting of safety components. Recognized codes and standards, when used, are ident ified and evaluated to assure their applicability, adequacy, and sufficiency in keepi ng with the required safety function.
The quality assurance program conforms with the requirements of 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Plants." This program is discussed in Chapter 17. Chapter 14 describes the init ial test program to assure performance of installed equipment commens urate with the impor tance of the safety function.
Criterion 2 - Design Bases for Prot ection Against Natural Phenomena
Structures, systems, and component s important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect:
(1) appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and su rrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena, and (3) the importance of the safety functions to be performed.
Discussion
The natural phenomena and thei r magnitude are selected in accordance with their probability of occurrence at this specific site. The criteria adopted in the design of affected structures, systems and component s, also depend on the likelihood of the natural phenomenon under consi deration. The designs are based upon the most severe of the natural phenomena recorded for the site , with an appropriate margin to account for uncertainties in the historical data. The natural phenomena postulated in the design are presented in Chapter 2. The design criteria for the structures, systems and components affected by each nat ural phenomenon are presented in the sections listed below. These sections also identify which combinations of natural and plant originated accidents are considered in the design.
The design criteria developed meet the requirements of Criterion 2.
3.1-3 Reformatted July 2014 For further discussion, s ee the following sections:
Section 1. Meteorology 2.3
- 2. Hydrologic Engineering 2.4 3. Geology, Seismology, and Geotechnical Engineering 2.5 4. Classification of Structures, Components and Systems 3.2 5. Wind and Tornado Loadings 3.3
- 6. Water Level (Flood) Design 3.4
- 7. Missile Protection 3.5
- 8. Seismic Design 3.7
- 9. Design of Category I Structures 3.8 10. Mechanical Systems and Components 3.9
- 11. Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment 3.10 12. Environmental Design of Mechanical and Electrical Equipment 3.11
Criterion 3 - Fire Protection
Structures, systems, and compon ents important to safety shall be designed and located to minimize, consistent with other safety requ irements, the probability and effect of fires and explosions. Noncombustible and heat resi stant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and Control Room. Fire detection and fight ing systems of appropriate capacity and
capability shall be provided and designed to mini mize the adverse effects of fires on structures, systems, and components important to safety.
Fire-fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of thes e structures, systems, and components.
Discussion
Fires in the plant are prevented or mitigat ed by the use of noncombustible and fire retardant materials. Redundant safety class equipment is separated by 1 or 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fire barriers or adequate clear space separation based upon the fire hazard. Safety-related ventilation systems are designed and arranged to ensure that no single fire occurrence will jeopardize plant safety. In addition, special attention is given to the safety-related electrical systems by provid ing items such as metal cabinets, metal wireways and fire retardant insulation.
Cabling within trays is suitably derated and c able tray loading is designed to minimize internal heat buildup. Cable trays are suit ably separated to avoid the loss of redundant channels of protection cabling should fire s occur. The arrangement of equipment in protection channels assigned to separate cabi nets provides physical separation and minimizes the effects of a possible fire.
RN 01-113 3.1-4 Reformatted July 2014 Combustible supplies such as logs, records, manuals, etc., are limited in such areas as the Control Room to amounts required for curr ent operation, thus minimizing the effect of a fire. No explosive gases or flammabl e liquids exist in these areas and therefore no explosion hazard exists.
The Plant Fire Protection System includes the following provisions:
- 1. Automatic fire detection equipment in t hose areas where fire danger is greatest.
- 2. Physical barriers and fire rated walls.
- 3. Automatic extinguishing systems for ar eas of highest fire loading as well as manually operated fire extinguishers for all areas.
- 4. Fire alarms and detection devices are connected to the Stat ion Communication System in the Control Room.
The Westinghouse supplied equipment is des igned to minimize the probability and effect of fires and explosions. Noncombusti ble and fire resistant materials are used in this equipment wherever practical.
The requirements of the National Fire Protection Association, the American Insurance Association, the Nuclear Energy-Liability Property Insurance Association (NE-LPIA), and the applicable local codes and regulat ions are observed in the design and installation of Westi nghouse supplied equipment.
The design of the Fire Protec tion System thus meets the r equirements of Criterion 3.
For further discussion, s ee the following sections:
Section 1. Instrumentation and Controls 7.0
- 2. Electric Power 8.0
- 3. Separation Cr iteria 8.3.1 4. Fire Protection System 9.5.1
- 5. Conduct of Operations 13.0
Criterion 4 - Environmental and Missile Design Bases
Structures, systems, and com ponents important to safety shall be designed to accommodate the effects of and to be compatib le with the environmental conditions associated with normal operation, maintenan ce, testing and postulated accidents, including loss-of-coolant accidents. These structures, systems and components shall be appropriately protected against dynamic effe cts, including the effects of missiles, pipe whipping, and discharging fl uids, that may result from equipment failures and from events and conditions outside the nuclear power unit.
RN 01-113 3.1-5 Reformatted July 2014 Discussion
Structures, systems and components important to safety are designed to accommodate the effects of, and to be compatible with, the environmental condi tions associated with normal operation, maintenance, testing, and postulated accidents, including loss of coolant accidents (LOCA).
These structures, systems and components are appropriately protected against dynamic effects including the effects of miss iles, pipe whipping and discharging fluids that may result from equipm ent failures and from events and conditions outside the nuclear power unit.
The electrical equipment, instrumentation and c ables for protection of engineered safety feature systems which are located inside the Reactor Building are discussed in the sections listed below indicating the design requirements in terms of the time which each must survive the extreme environmental conditions following a loss of coolant accidents or a main steam line break.
The design of these structures, systems, and components meets t he requirements of Criterion 4.
For further discussion, s ee the following sections:
Section 1. Meteorology 2.3
- 2. Hydrologic Engineering 2.4 3. Geology, Seismology, and Geotechnical Engineering 2.5 4. Classification of Structures, Components and Systems 3.2 5. Wind and Tornado Loadings 3.3
- 6. Water Level (Flood) Design 3.4
- 7. Missile Protection 3.5
- 8. Protection Against Dynamic Effects Associated with the Postulated Ruptur e of Piping 3.6 9. Seismic Design 3.7
- 10. Design of Category I Structures 3.8 11. Mechanical Systems and Components 3.9
- 12. Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment 3.10 13. Environmental Design of Mechanical and Electrical Equipment 3.11 14. Integrity of Reactor C oolant Pressure Boundary 5.2 15. Engineered Safety Features 6.0 16. Instrumentation and Controls 7.0 17. Electric Power 8.0 18. Main Steam System 10.3 RN 01-113 3.1-6 Reformatted July 2014 Criterion 5 - Sharing of Stru ctures, System s, and Components
Structures, systems, and com ponents important to safety shall not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform thei r safety functions, including, in t he event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
Discussion
No evaluation is required since this is a 1 unit installation and there are no shared facilities. This criterion does not apply.
3.1.2.2 Protection by Multiple Fission Product Barriers Criterion 10 - Reactor Design
The reactor core and associated coolant, control, and protection systems shall be
designed with appropriate margin to assure that specified accept able fuel design limits are not exceeded during any condition of norma l operation, including the effects of anticipated operational occurrences.
Discussion
The reactor core and associated coolant, c ontrol, and protection systems are designed with adequate margins to:
Preclude significant fuel damage during normal core operation and operational transients (Condition I)
[1] or any transient conditions arising from occurrences of moderate frequency (Condition II)[1]. Ensure return of the reactor to a safe state follo wing a Condition III
[1] event with only a small fraction of fuel rods damaged althoug h sufficient fuel damage might occur to preclude resumption of operation without considerable outage time.
Assure that the core is intact with acc eptable heat transfer geomet ry following transients arising from occurrences of limiting faults (Condition IV)
[1].
Chapter 4 discusses the design bases and design evaluation of reactor components including the fuel, reactor vessel internals, and reactivity control systems. Details of the
control and protection systems instrumentation design and logic are discussed in Chapter 7. This information supports the accident analyses of Chapter 15 which show that the acceptable fuel design limits are not exceeded for Condition I and II occurrences.
3.1-7 Reformatted July 2014 Criterion 11 - Reactor Inherent Protection
The reactor core and associated coolant system s shall be designed so that in the power operating range the net effect of the prompt inher ent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.
Discussion
Prompt compensatory reactivity feedback effects are assured wh en the reactor is critical by the negative fuel temperature effect (Doppler effect) and by the nonpositive operational limit on the moderator temperature coefficient of reactivity. The negative Doppler coefficient of reactivity is assur ed by the inherent design using low enrichment fuel; the nonpositive moderator temperature coefficient of reactivity is assured by administratively controlling the dissolved absorber concentration or by burnable poison.
These reactivity coefficients are discussed in Section 4.3.
Criterion 12 - Suppression of Reactor Power Oscillations
The reactor core and associated coolant, control, and protection systems shall be
designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible, or can be reliably and readily detected and suppressed.
Discussion
Power oscillations of the fundamental mode are inherently eliminated by the negative Doppler and nonpositive moderator temper ature coefficients of reactivity.
Oscillations, due to xenon spatial effects, in the radial, diametral, and azimuthal overtone modes are heavily damped due to t he inherent design and due to the negative Doppler and nonpositive moderator temper ature coefficients of reactivity.
Oscillations, due to xenon spatial effects, in the axial first overtone mode may occur.
Assurance that fuel design limits are not exceeded by xenon axial oscillations is provided by reactor trip functions usi ng the measured axial power imbalance as an input.
Oscillations, due to xenon spatial effects, in axial modes higher than the first overtone, are heavily damped due to the inherent design and due to the negative Doppler coefficient of reactivity.
Xenon stability control is discussed in Section 4.3.
3.1-8 Reformatted July 2014 Criterion 13 - Instru mentation and Control Instrumentation shall be provided to monitor variables and systems over their
anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety including those variables and systems that can affect the fission process, the integrit y of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
Discussion
Plant instrumentation and cont rol systems are provided to monitor significant variables in the reactor core, Reactor Coolant Syst em, and containment ov er their anticipated range for all conditions to the extent required. The instal led instrumentation provides continuous monitoring, warning, and initiati on of safety functions, by the use of instrumentation and control provided.
The following processes are controlled to ma intain key variables within their normal ranges:
- 1. Reactor power level (manual or auto by controlling thermal load).
- 2. Reactor coolant temperature (manual or auto by rod control cluster assembly (RCCA) motion, in sequential groups).
- 3. Reactor coolant pressure (manual or aut o by heaters and spray in the pressurizer).
- 4. Reactor coolant water inventory, as indi cated by the water level in the pressurizer (manual or auto charging flow).
- 5. Reactor axial power balance.
- 6. Reactor Coolant System boron concentra tion (manual or auto makeup of charging flow).
- 7. Steam generator water in ventory on secondary side (manual or auto feedpump flow through feedwater control valves).
The Reactor Control System is designed to automatically maintain a programmed average temperature in the reactor coolant dur ing steady state operation and to ensure that plant conditions do not reach reactor tr ip settings, as the result of a transient caused by a load change.
The Reactor Protection System trip setpoints are selected so that anticipated transients do not cause a departure from nucleate boiling ra tio (DNBR) of less than the safety limit.
3.1-9 Reformatted July 2014 Proper positioning of the control rods is monitored in the control room by bank arrangements of individual meters for each rod cluster control assembly (RCCA). A rod deviation alarm alerts the operator of a devia tion of 1 rod cluster control assembly from its bank position. There are also insertion limit monitors with visual and audible annunciation to avoid loss of shutdown margin. Each full length rod cluster control assembly is provided with an indication of posit ioning at the bottom of its travel. This condition is also alarmed in the control room. Four(4) excore long ion chambers also detect asymmetrical flux distribution indicative of rod misalignment.
Movable incore flux detectors and fixed incore thermo couples are provided as operational aids to the operator. Chapter 7 contains further details on instrumentation and controls. Section 7.5 details the info rmation available to the operator for the performance of required safety functi ons. Information regarding the Radiation Monitoring System provided to measure envir onmental activity and alarm high levels is contained in Section 11.4.
Overall reactivity control is achieved by the combination of soluble boron and rod cluster
control assemblies. Long term r egulation of core reactivity is accomplished by adjusting the concentration of boric acid in the reactor coolant. Short term reactivity control for power changes is accomplished by the Reac tor Control System which automatically moves rod cluster control assemblies. This system uses input signals including neutron flux, coolant temperat ure, and turbine load.
These systems are described in C hapters 6, 7, 8, 9, 11 and 12.
Criterion 14 - Reactor Coolant Pressure Boundary
The reactor coolant pressure boundary s hall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
Discussion
The Reactor Coolant System boundary is designed, fabricated and erected to accommodate the system pressures and temperatures attained under all expected
modes of plant operation, including all anticipated transients, and to maintain the stresses within applicable stress limits. See Sections 3.9 and 5.2 for details. Reactor Coolant System boundary materials selection and fabrication techniques ensure a low probability of gross rupture or significant leakage.
In addition to the loads imposed on the system under normal operating conditions, consideration is also given to abnormal l oading conditions, such as pipe rupture and seismic, as discussed in Sections 3.6 and 3.7, respectively. The system is protected from overpressure by means of pressure relieving devices as required by applicable codes.
3.1-10 Reformatted July 2014 Means are provided to detect significant unc ontrolled leakage from the reactor coolant pressure boundary with indication in the Cont rol Room. See Sect ion 5.2 for details.
The Reactor Coolant System boundary has provision for inspection, testing and surveillance of critical areas to assess the structural and leaktight integrity. See Section 5.2 for details. For the reactor vessel, a material surveillance program conforming to applicable codes is pr ovided. See Section 5.4 for details.
Criterion 15 - Reactor Coolant System Design
The Reactor Coolant System and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipat ed operational occurrences.
Discussion
The design pressure and temperature for each component in the reactor coolant and associated auxiliary, control and protecti on systems are selected to be above the maximum coolant pressure and temperature under all normal and anticipated transient load conditions.
Additionally, Reactor Coolant System boundary components achieve a large margin of safety by the use of proven ASME ma terials and design codes, use of proven fabrication techniques, nondestructive shop testing and integrated hydrostatic testing of assembled components.
The effect of radiation embrittlement is considered in reactor vessel design and surveillance samples monitor adherence to expected conditions throughout plant life.
Multiple safety and relief valves are provided for the Reactor Coolant System. These
valves and their setpoints sati sfy ASME criteria for overpre ssure protection. The ASME criteria are satisfactory based upon a long history of industry use. Chapter 5 discusses the reactor coolant system design.
Transient analyses are included in Reactor Coolant System design which conclude that design conditions are not exceeded during normal operation. Protection and control setpoints are based upon these transient analyses. The design margin includes the effects of thermal lag, coolant transport times, pressure drops, system relief valve
characteristics and instrumentation and control response characteristics.
3.1-11 Reformatted July 2014 Criterion 16 - Containment Design
Reactor containment and associated system s shall be provided to establish an essentially leaktight barrier against the unc ontrolled release of radioactivity to the
environment and to assure that the containm ent design conditions important to safety are not exceeded for as long as postulated accident conditions require.
Discussion
A steel lined prestressed concrete reactor building is provided. This structure encloses the entire Reactor Coolant System and is des igned to sustain, wit hout loss of required integrity, all effects of gr oss equipment failures up to and including the simultaneous occurrence of a double ended ruptur e of the largest pipe in the Reactor Coolant System and the safe shutdown earthquake (SSE). Should such an event occur, the engineered safety features (ESF) serve to cool the reactor core and re turn the containment to near atmospheric pressure. The containment systems and their associated engineered safety features are designed to ensure t he functional capability of preventing the uncontrolled release of radioactive material and that the design conditions important to safety remain inviolate for as long as postulated accident conditions require.
For further discussion, s ee the following sections:
Section 1. Design of Category I Structures 3.8 2. Engineered Safety Features 6.0 Criterion 17 - Electrical Power Systems
An Onsite Electric Power System and an Offsite Electric Power System shall be provided to permit functioning of structur es, systems, and compon ents important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.
The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.
Electric power from the transmission network to the Onsite Electric Distribution System shall be supplied by 2 physica lly independent circuits (not necessarily on separate rights-of-way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of RN 01-113 3.1-12 Reformatted July 2014 these circuits shall be designed to be availabl e in sufficient time following a loss of all onsite alternating current power supplies and the other offsit e electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One (1) of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment int egrity, and other vital safety functions are maintained.
Provisions shall be included to minimize the probability of losing elec tric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.
Discussion
Onsite and offsite power systems each independently provide the total power requirements to perform the functions required of the safety-related systems. The onsite power required to operat e Engineered Safety Featur es System equipment is supplied by two 100% capacity diesel generators. The offsit e power required to operate safety-related systems is supplied by 2 independent sources, one from the 230 kV system and 1 from the 115 kV system. Each source will supply total power requirements for either 1 or both of t he redundant and independent power distribution systems for the Engineered Sa fety Features Systems.
Each Engineered Safety Features power suppl y bus is normally connected to physically and electrically independent power supplies.
Electric power from the transmission network to the substation is provided by a sufficient number of independent lines to minimize the likelihood of simultaneous failure.
Two (2) onsite independent battery systems provide control power for the redundant and independent power distribut ion systems for Engineered Sa fety Features Systems as well as control power for the onsite power sources. The reactor protective instrumentation is powered from 4 independent 120 volt nomi nal ac vital buses which provide uninterrupted pow er from single phase inverters.
Each bus is supplied by its own associated inverter. Each Class 1E battery system supplies power to 2 static
inverters.
These systems are designed in accordance with IEEE-308
[2].
3.1-13 Reformatted July 2014 For further discussion, s ee the following sections:
Section 1. General Plant Description 1.2 2. Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment 3.10 3. Environmental Design of Mechanical and Electrical Equipment 3.11 4. Offsite Power Systems 8.2
- 5. Onsite Power Systems 8.3
Criterion 18 - Inspection and Testing of Electric Power Systems
Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of import ant areas and features , such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and func tional performance of t he components of the systems, such as onsite power sources, relays, switches and buses, and (2) the operability of the systems as whole and, under conditions as close to design as
practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.
Discussion
The ESF power supply buses and associated diesel generators are arranged for periodic, independent testing of each system. These tests, performed periodically in accordance with the Technical Specifications, prove the o perability of the Emergency Power Supply System under conditions as clos e to design as practical, thus permitting assessment of the continuity of the system and condition of the components.
The design of the standby power systems provi des testability in accordance with the
requirements of Criterion 18.
For further discussion, s ee the following sections: Section 1. Onsite Power System 8.3
- 2. Initial Test Program 14.0
- 3. Technical Specifications
Criterion 19 - Control Room
The Control Room shall be provided from wh ich actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit acce ss and occupancy of the control room under RN 01-113 RN 01-113 RN 99-136 3.1-14 Reformatted July 2014 accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.
Equipment at appropriate locati ons outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary
instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
Discussion
Safe occupancy of the Control Room under normal, abnormal, and accident conditions is assured by the design. The Control Room is located in a Seismic Category I
structure. Adequate shielding is provided to maintain toler able radiation levels in the Control Room in the event of a design basis accident. Redundant equipment is provided in the Control Room Ventilation System which permits recirculation of Control Room air through HEPA and charcoal filters. This equipment also permits control room
air to be drawn from outside through r oughing and HEPA filters and to be discharged outside or for the use of various combinatio ns of outside and recirculated air. Radiation and smoke detectors are provided for the C ontrol Room Ventilation System. Excessive concentrations of any one of t hese contaminants causes an al arm in the Control Room.
High radiation places the Control Room Ventilation System in the emergency filtration mode and causes an alarm in the Control Room.
The Control Room includes the following: substation control panel, electrical relay panels and control panels which contain those instruments and controls necessary for operation of the station functions, such as the reactor and its auxiliary systems, engineered safety features, turbine gener ator, steam and power conversion systems, station electrical distribution boards, and heating, ventilating and air conditioning systems.
The Control Room is continuously occupied by qualified operating personnel under all operating and accident conditions.
In the unlikely event that o ccupancy of the Control Room is restricted, a local control room evacuation panel and manual operation of critical com ponents are used to effect cold shutdown from outside the Control Room.
By use of appropriate procedures and equipmen t, the unit can also be brought to cold shutdown conditions. For this operation, it is assumed that offsite power will be available.
In accordance with the implementation of the alternative source te rm for the Virgil C.
Summer Nuclear Station, the above dose criteria is replaced by the 5 rem total effective dose equivalent (TEDE) acceptance criteri on provided in 10 CFR 50.67(b)(2) for a loss-of-coolant accident (LOCA), main steam line break (MSLB) accident, fuel handling RN 00-089 RN 12-034 RN 12-034 3.1-15 Reformatted July 2014 accident (FHA), steam generator tube rupt ure (SGTR), reactor coolant pump locked rotor accident (RCPLRA) and the cont rol rod ejection accident (CREA).
For further discussion, s ee the following sections:
Section 1. General Plant Description 1.2
- 2. Control Room Diffusion Estimates 2.3.4.3
- 3. Control Building Design 3.8
- 4. Habitability Systems 6.4
- 5. Instrument and Controls 7.0
- 6. Shutdown from Outside Control Room 7.4 7. Air Conditioning, Heating, Cooling and Ventilation Systems 9.4 8. Fire Protection System 9.5.1
- 9. Radiation Shielding 12.1
- 10. Ventilation 12.2
- 11. Control Room Dose 15.4
3.1.2.3 Protection and Reac tivity Control Systems Criterion 20 - Protection System Functions
The protection system shall be de signed: (1) to initiate aut omatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits ar e not exceeded as a result of anticipated operational occurrences; and (2) to sense accident cond itions and to initiate the operation of systems and components important to safety.
Discussion
A fully automatic protection system with appr opriate redundant channels is provided to
cope with transients where insufficient time is available for manual corrective action. The design basis for all protection systems is in accordance with the intent of
IEEE-279[3] and trial use guide IEEE-379
[4]. The Reactor Protection System automatically initiates reactor trip when any variable mo nitored by the system or combination of monitored variables exceeds the normal operating range. Setpoints are
designed to provide an envelope of safe operating conditions with adequate margin for uncertainties to ensure that fuel design limits are not exceeded.
Reactor trip is initiated by removing power to the rod drive mechanisms of all the full length rod cluster control assemblies. This caus es the rods to insert by gravity into the core, rapidly reducing the reactor power.
The response and adequacy of the protection system has been verified by analysis of all anticipated transients.
The Engineered Safety Featur es Actuation System automatically initiates emergency core cooling and other safeguards functions by sensing accident conditions, using redundant analog channels measuring divers e variables. Manual actuation of safeguards may be performed where time is available for operator action but is not RN 01-113 RN 12-034 RN 12-034 3.1-16 Reformatted July 2014 relied upon to satisfy this criterion. The Engineered Safety Features Actuation System automatically trips the reac tor upon manual or automatic sa fety injection (S) signal generation. See Section 7.5 and Secti on 7.1.2.1.5 for additional details.
The response and adequacy of the protection systems are analyzed for all environmental conditions s pecified by ANS N18.2
[1], through Condition IV.
Criterion 21 - Protection Syst em Reliability and Testability
The Protection System shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the Protection System shall be sufficient to assure that (1) no single failure results in loss of t he protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The Protection System shall be designed to permit periodic testing of its functioning when the r eactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
Discussion
The Protection System is in accordance with IEEE-279
[3]. It provides high functional reliability and adequate independence, redundan cy, and testability commensurate with the safety functions of the syst em. All actuation circuitry is provided with a capability of online testing. This extends to the final actuating device except where operational safety requirements prohibit ac tual operation of the device; e.g., turbine trip, steam line isolation, etc.
The Reactor Protection System is designed fo r high functional reliability by providing electrically isolated and physically separated, redundant anal og channels and two separate trip logic trains. This assures that no single failure will result in the loss of any protection function. Except for certain defined backup trip functions detailed in Section 7.2, the redundancy and independence provided in the Reactor Protection System allows individual channel test of channels required at power operation to be made during power oper ation without negating reactor pr otection or the single failure criterion. This testing will determine fa ilures and losses of redundancy that may have occurred. This arrangement also permits removal of a channel from service while still maintaining the high reliability of the prot ection function. Details of the Protection system design and testing provisions are contained in Chapter 7.
There are 2 series connected circuit breakers, 1 breaker for each trip logic train, which supply all power to the full length rod drive me chanisms. A reactor trip train supplies a signal to the undervoltage coil of its respective trip breaker and opening of either train
breaker will trip the reactor.
RN 99-101 3.1-17 Reformatted July 2014 The Engineered Safety Feat ures Actuation System is in accordance with IEEE-279
[3]. It also utilizes redundant analog channels meas uring the same parameter and redundant logic trains, either of which will actuate sa fety injection and/or reactor building spray.
The Engineered Safety Features Actuation System is test able at power with certain exceptions as detailed in Section 7.3. As with t he components of the Reactor Protection System, both physical and electrical separation is practiced for the Engineered Safety Features Ac tuation System to provide a high degree of availability for its safety function.
Criterion 22 - Protection System Independence
The Protection System shall be designed to assure that the effects of natural phenomena, and of normal operating, maint enance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
Discussion
Protection System components are designed and arranged so that the environment accompanying any emergency situation in which the components are required to function does not result in loss of the sa fety function. Various means are used to accomplish this. Functional diversity has been designed into the syst em. The extent of this functional diversity has been evaluated for a wide variety of postulated accidents. The general conclusion is that diverse protection functions would automatically terminate an accident before intolerable consequences could occur.
Automatic reactor trips occur as listed in Table 7.2-1.
Regarding the ESF actuation system, a safety injection signal can be obtained manually or by automatic initiation from any one of the following divers e sets of signals:
- 1. Low pressurizer pressure.
- 2. High Reactor Building pressure (Hi-1).
For a steam break accident, diversity of safety injection is provided by:
- 1. Low steam line pressure.
- 2. High steam line differential pressure.
- 3. For a steam break inside the Reactor Building, high Reactor Building pressure (Hi-1) provides an add itional parameter for generation of the safety injection signal.
RN 99-050 RN 99-050 RN 99-050 RN 99-050 RN 99-050 3.1-18 Reformatted July 2014 All of the above sets of signals are redundant and physically separated and meet the intent of the criterion.
High quality components, suit able derating and applicable qua lity control, inspection, calibration and tests are utilized to guard against common mode failure. Qualification testing is performed on the components of the various safety systems to demonstrate
functional operation at normal and post acci dent conditions of temp erature, humidity, pressure, and radiation for specified periods, if required. Typical protection system equipment is subjected to tests under simulated seismic conditions using conservatively large accelerations and applicable frequencies.
The test results indi cate no loss of the protection function.
The design criteria for instru mentation are given in Secti on 7.2 and qualification of the instrumentation is outlined in Sections 3.10 and 3.11.
Criterion 23 - Protecti on System Failure Modes
The Protection System shall be designed to fall into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as
disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extr eme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
Discussion
The Protection System is designed with due c onsideration of the most probable failure modes of the components under various perturbations of energy sources and the environment. Each reactor trip channel is des igned on the de-energize-to-trip principle so that a loss of power or disconnection of the channel causes that channel to go into its tripped mode. In addition, a loss of power to the full length Rod Cluster Control Assembly drive mechanisms causes the Rod Cluster Control Assembly to insert by
gravity into the core. S ee Section 7.2 for details.
With regard to engineered safety features, should a loss of the pr eferred offsite power source occur, onsite diesel generators are available to pow er emergency loads, with the station batteries being used to supply instrum entation power only for that period of time required for the diesel to start.
See Section 8.3 for details.
A loss of power to one train of safety injection equipment does not affect the ability of the other train to perform its function.
Criterion 24 - Separation of Protection and Control Systems
The Protection System shall be separated from control systems to the extent that failure of a single control system component or channel, or failure or removal from service of any single Protection System component or channel which is common to the control 3.1-19 Reformatted July 2014 and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.
Discussion
Protection and control channels in the facility protection system s will be designed in accordance with the IEEE-279[3].
The reactor protection system itself is designed to maintain separation between redundant protection channels and protection logic trains. Separation of redundant analog channels originates at the process sensors and continues along the wiring route and through reactor building penetrations to analog protection racks and terminating at
the Reactor Protection System logic racks.
Isolation of wiring is achieved using separate wireways, cable trays, conduit runs and reactor building penetrations for each
redundant channel. Analog equi pment is separated by loca ting components associated with redundant functions in different prot ection racks. Each redundant protection channel set is energized from a separate ac power feed.
The redundant (2) reactor trip logic trains are physicall y separated from one another. The Reactor Protection System is comprised of identifiable channels which are physically separated and el ectrically isolated.
Each trip circuit is designed so that the trip occurs upon de-energization of the circuit; an open circuit or loss of power to a channel will, therefore, result in that channel going into its trip mode. R edundant protection channels are provided to prevent a single failure from defeating a protection func tion. Redundancy provides reliability and independence of operation. Channel independence is carried throughout the system from the sensor to the logic interface. In some cases, however, it is advantageous to employ control signals derived from individual protection channels through isolation
amplifiers contained in the prot ection channel. As such, a failu re in the control circuitry does not adversely affect the protection channel.
The electrical supply and control conductors for redundant or backup circuits have such physical separation as is required to assure that no single credibl e event will prevent operation of the associated function by reason of electrical conductor damage. Critical circuits and functions include power, contro l and analog instrumentation associated with the operation of Reactor Protection, Engineered Safety Features, Reactor Shutdown and Residual Heat Removal Systems.
See Sections 7.1 and 8.3 for more details.
3.1-20 Reformatted July 2014 Criterion 25 - Protection S ystem Requirements for Reacti vity Control Malfunctions
The Protection System shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejec tion or dropout) of control rods.
Discussion
The Protection System is designed to limit reac tivity transients so that fuel design limits are not exceeded. Reactor shutdown by full length rod insertion is completely independent of the norma l control function since the trip breakers interrupt power to the rod mechanisms regardless of existing cont rol signals. Thus, in the postulated accidental withdrawal, (assumed to be initiated by a control malfunction) flux, temperature, pressure, level and flow signals would be indepen dently generated. Any of these signals (trip demands) would operat e the breakers to trip the reactor.
Analyses of the effects of possible malfunc tions are discussed in Chapter 15. These analyses show that for postulated dilution during refueling, star tup, or manual or automatic operation at power, the operator has ample time to determine the cause of dilution, terminate the source of dilution and initiate reboration before the shutdown
margin is lost. The analyses show that a cceptable fuel damage limits are not exceeded even in the event of a single malfunction of either system.
Criterion 26 - Reactivity Cont rol System Redundan cy and Capability Two (2) independent reactivity control systems of different design principles shall be provided. One (1) of the syst ems shall use control rods, preferably including a positive means for inserting the rods, and shall be ca pable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate ma rgin for malfunctions such as stuck rods, specified acceptable fuel design limit s are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits ar e not exceeded. One (1) of the systems shall be capable of holding the reactor core subc ritical under cold conditions.
RN 01-113 3.1-21 Reformatted July 2014 Discussion
Two (2) reactivity control systems are pr ovided. These are rod cluster control assemblies (RCCA's) and chemical shim (bor ic acid). The rod cluster control assemblies are inserted into the core by the force of gravity.
During operation the shutdown rod banks are fully withdrawn. The full length control rod system automatically maintains a programmed average reactor temperature
compensating for reactivity effects a ssociated with scheduled and transient load changes. The shutdown rod banks, along wit h the full length control banks, are designed to shut down the reactor with adequate margin under conditions of normal operation and anticipated operatio nal occurrences, thereby ensuring that specified fuel design limits are not exceeded. The most rest rictive period in core life is assumed in all analyses and the most reactive rod cluster is assumed to be in the fully withdrawn position.
The Boron System will maintain the reactor in the cold shutdown state independent of the position of the control rods and can compensate for all xenon burnout transients.
Details of the construction of the rod cluster control assembly are presented in
Chapter 4. Operation is discussed in Chapter 7. The means of controlling the boric acid concentration is described in Chapter 9. Performance analyses under accident
conditions are included in Chapter 15.
Criterion 27 - Combined Reactivity Control Systems Capability
The reactivity control systems shall be designed to have a combined capability, in conjunction with poison additi on by the Emergency Core Cooling System, of reliably controlling reactivity changes to assure t hat under postulated accident conditions and with appropriate margin for stuck rods the ca pability to cool the core is maintained.
Discussion
The plant is provided with means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. These means are discussed in detail in Sections 4.3 and 9.3.
Combined use of the Rod Cluster Control System and the Chemical Shim Control System permits the necessary shutdown margin to be maintained during long term xenon decay and plant cooldown. The single highest worth control cluster is assumed to be stuck full-out upon trip for this determination.
In the event of a loss of coolant accident, the Safety Injection Syst em is actuated and concentrated boric acid is injected into the cold legs of the Reactor Coolant System.
This is in addition to the boric acid content of the accumulators which is passively injected due to a decrease in system pressure. See Section 6.3 for further details.
3.1-22 Reformatted July 2014 Criterion 28 - Reactivity Limits
The reactivity control systems shall be desig ned with appropriate lim its on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither: (1) result in damage to the reactor coolant pressure boundary greater than limited local yiel ding; nor (2) sufficiently disturb the core, its support structures or other reactor pr essure vessel internals to impa ir significantly the capability to cool the core. These postu lated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means
), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.
Discussion
The maximum reactivity worth of control rods and the maximum rates of reactivity insertion employing control rods are limited to values that pr event rupture of the Reactor Coolant System boundary or disruptions of the core or vessel internals to a degree that could impair the effectiveness of emergency core cooling.
The maximum positive reactivity insertion rate s for the withdrawal of rod cluster control assemblies and the dilution of the boric acid in the reactor coolant system are limited by the physical design characteri stics of the rod cluster c ontrol assemblies and of the Chemical and Volume Control System. Technical Specifications on shutdown margin and on rod cluster control assembly inserti on limits and bank overlaps as functions of power provide additional assurance that the consequences of the postulated accidents are no more severe than those presented in the analyses of Chapter 15. Reactivity insertion rates, dilution, and withdrawal limits are also discussed in Section 4.3. The capability of the Chemical and Volume C ontrol System to avoid an inadvertent excessive rate of boron dilution is discussed in Section 9.3.
Assurance of core cooling capability follo wing Condition IV a ccidents, such as rod ejection, steam line break, etc., is obtained by keeping the reactor coolant pressure boundary stresses within faulted condition limi ts as specified by applicable ASME Codes. Structural deformations are checked also and limited to values that do not jeopardize the operation of nec essary safety features.
Criteria 29 - Protection Against An ticipated Operational Occurrences
The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
3.1-23 Reformatted July 2014 Discussion
The protection and reactivity control systems are designed to assure extremely high reliability in performing their required safety functions in any anticipated operational occurrence. Likely failure modes of system components are designed to be safe modes. Equipment used in these systems is designed, constructed, operated, and maintained with a high level of reliability. Loss of power to the Protection System results in a reactor trip. Details of system design are covered in Ch apter 7. Also refer to responses to General De sign Criteria 20 through 25.
3.1.2.4 Fluid Systems Criterion 30 - Quality of Reac tor Coolant Pressure Boundary
Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and test ed to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the
source of reactor coolant leakage.
Discussion
Reactor coolant pressure boundary component s are designed, fabricated, inspected and tested in conformance with the ASME Code, Section III. All components are classified in accordance with ANS N18.2
[1] and are accorded the quality measures appropriate to the classification. The desi gn bases and evaluations of reactor coolant pressure boundary components are discussed in Chapter 5.
Leakage is detected by an increas e in the amount of makeup wa ter required to maintain a normal level in the pressurizer. The reac tor vessel closure joint is provided with a temperature monitored leak off between double gaskets. Leakage inside the Reactor Building is drained to the Reactor Building sump where it is monitored.
Leakage is also detected by measuring the airborne activity within the Reactor Building and activity of manual samples of the condensate drained from the Reactor Building and recirculation units. Monitoring the inventory of reactor coolant in the system at the pressurizer, volume control tank and coolant drain collection tanks make available an accurate indication of integrated leakage. The Reactor Coolant Pressure Boundary
Leakage Detection System is discussed in Section 5.2.7.
Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary
The reactor coolant pressure boundary sha ll be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions: (1) the boundary behaves in a nonbrittle manner; and (2) the
probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the 3.1-24 Reformatted July 2014 uncertainties in determining: (1) material properties; (2) the effe cts of irradiation on material properties; (3) resi dual, steady-state and transient stresses; and (4) size of flaws.
Discussion
Close control is maintained over material selection and fabrication for the Reactor Coolant System to assure that the boundary behaves in a nonbrittle manner. The Reactor Coolant System materials which ar e exposed to the coolant are corrosion resistant stainless steel or inconel. The reference temperature (RTNDT) of the reactor
vessel structural steel is established by Charpy V-notch and dr op weight tests.
Materials testing is consistent with the intent of 10 CFR 50, Appendices G and H.
These tests ensure the selection of mate rials with adequate toughness properties and margins.
As part of the reactor vessel specification, certain requirements wh ich are not specified by the applicable ASME Codes are performed as follows:
- 1. Ultrasonic Testing
In addition to code requirements, the performanc e of a 100% volumetric ultrasonic test of reactor vessel plate for shear wave and a post hydrostatic test ultrasonic map of welds in the pressure vessel are required.
Also, Westinghouse requires cladding bond ultrasonic inspection to more restrict ive requirements than Code to preclude interpretation problems during inservice inspection.
- 2. Radiation Surveillance Program
In the surveillance programs , the evaluation of the r adiation damage is based on pre-irradiation and post-irradi ation testing of Charpy V-notch and tensile specimens. These programs are directed toward evaluation of the effect of radiat ion on the fracture toughness of reactor vessel steels based on the reference transition temperature approach and the fracture mechanics approach, and are in accordance with ASTM
E-185[5], and the requirements of 10 CFR 50, Appendix H.
The fabrication and quality control techniques used in the fabrication of the Reactor Coolant System are equivalent to those used for the reactor vessel. The inspections of reactor vessel, pressurizer, piping, pum ps, and steam generators are governed by ASME Code requirements.
See Chapter 5 for details.
Heatup and cooldown rates durin g plant life are predicted using conservative values for the change in ductility transition temperature due to irradiation.
3.1-25 Reformatted July 2014 Criterion 32 - Inspection of Reactor Coolant Pressure Boundary
Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of importa nt areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance
program for the reacto r pressure vessel.
Discussion
Provision has been made in the Reacto r Coolant System design for adequate inspection, testing and surveillance during t he service lifetime. Necessary accessibility has been factored into the design. The vessel inspection program will conform to ASTM
E-185[5]. These provisions are discussed in detail in Section 5.2.
Criterion 33 - Reactor Coolant Makeup
A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be prov ided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or ot her small components which are part of the boundary. The system shall be designed to assure that fo r Onsite Electric Power System operation (assuming offsite power is not available) and for Offsite Electric Power System operation (assuming onsite power is not avail able) the system safety function can be accomplished using the piping, pumps, and valv es used to maintain coolant inventory during normal reactor operation.
Discussion
The Chemical and Volume Control System provides a means of reactor coolant makeup
and adjustment of the boric acid concentration. Makeup is added automatically if the level in the volume control tank falls below a preset level. High pressure centrifugal charging pumps are provided which are capable of supplying the required makeup and reactor coolant pump seal injection flow when power is available from either onsite or offsite electric power systems. These pumps also serve as high head safety injection pumps. Details of system design are include d in Chapters 6 and 9. Details of the electric power systems are presented in Chapter 8.
Criterion 34 - Residual Heat Removal
A system to remove residual heat shall be pr ovided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.
3.1-26 Reformatted July 2014 Suitable redundancy in component s and features, and suitable interconnections, leak detection, and isolation capabilitie s shall be provided to assure that for Onsite Electric Power System operation (assuming offsite power is not available) and for Offsite Electric Power System operation (assuming ons ite power is not available) the system safety function can be accomplished, assuming a single failure.
Discussion
The Residual Heat Removal (RHR) System, in conjunction with the Steam and Power Conversion System, is designed to transfer the fission product decay heat and other
residual heat from the reactor core within acceptable limits.
Suitable redundancy is accomplished with the tw o residual heat removal pumps, located
in separate compartments with means available for draining and monitoring of leakage, the 2 heat exchangers and the associated piping , cabling, and electric power source. The residual heat removal system is able to operate on either onsite or offsite electrical
power systems.
The Residual Heat Removal System is able to accommodate a single failure (see Section 3.1.3). During the injection phase, no single active failure prevents the accomplishment of Residual Heat Removal Syst em objectives. During the recirculation phase, but not in the injection phase, the Residual Heat Removal System can accommodate one active or passive failure. One active or passive failure in the systems required for long term Residual Heat Removal System operation does not prevent the accomplishment of Residual Heat Removal System objectives.
Details of the system design c an be found in Se ction 5.5.7.
Criterion 35 - Emergency Core Cooling System
A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that: (1) fuel and clad damage that could interfere with continued effective core cooling is prevent ed; and (2) clad metal-water reaction is limited to negligible amounts.
Suitable redundancy in component s and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be pr ovided to assure that for Onsite Electric Power System operation (assu ming Offsite Power is not available) and for Offsite Electric Power System operation (assuming onsit e power is not available), the system safety function can be acco mplished, assuming a single failure.
3.1-27 Reformatted July 2014 Discussion
An Emergency Core Cooling System (ECCS) is provided to cope with any loss of coolant accident due to a pipe rupture. Abundant cooling water is available to the core at a rate sufficient to maintain the core geometry and to assure that clad metal-water reaction is limited to less than 1%. The design is adequate to ensure performance of
the required safety functions assuming a sing le failure and that electrical power is available from either the Offsite or Onsite Electrical Power System. Details of the
capability of the system are discussed in Section 6.3. An evaluation of the adequacy of
the system functions is presented in Chapter 15.
Criterion 36 - Inspection of Emergency Core Cooling System
The Emergency Core Cooling System shall be designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping, to assu re integrity and capabil ity of the system.
Discussion
Equipment design facilitates access to the crit ical parts of the reactor vessel internals, injection nozzles, pipes, and valves for visual inspection and for nondestructive inspection where such techniques are desirable and appropriate. The design enables compliance with ASME Code, Se ction XI requirements.
Components located outside the Reactor Building are accessible for leaktightness inspection during normal operation of the plant.
Details of the inspection program for the reactor vessel internals are included in Section 5.4. Inspection of the Emergency Core Cooling System is discussed in
Section 6.3.
Criterion 37 - Testing of Em ergency Core Cooling System
The Emergency Core Cooling System shall be designed to permit appropriate periodic pressure and functional testing to assure: (1) the structural and leaktight integrity of its components; (2) the operability and performance of the active components of the system; and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operati on of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.
3.1-28 Reformatted July 2014 Discussion
Components of the system are accessible for leaktightness inspection during periodic
tests.
Active components of the Emergency Core Cooling System may be individually actuated on the normal power so urce at any time during plant operation to demonstrate operability.
Tests may be performed during shutdown to demonstrate proper automatic operation of the Emergency Core Cooling System.
Active components are identified in Section 3.9.2. Inservice testing is discussed in
Section 3.9.4. The details of these tests are included in Section 6.3. Emergency power details are included in Chapter 8.
Criterion 38 - Containment Heat Removal
A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pr essure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.
Suitable redundancy in component s and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be pr ovided to assure that for Onsite Electric Power System operation (assuming offsite power is not available) and for Offsite Electric Power S ystem operation (assuming onsite pow er is not available) the system safety function can be accomplished, assuming a single failure.
Discussion
Two systems based on different principles ar e provided to remove heat from the Reactor Building following an accident to ma intain the pressure below the Reactor Building design pressure. The Reactor Bu ilding spray and the Reactor Building cooling units are each independently capable of removing sufficient energy to maintain the pressure below the Reactor Building design pr essure. Each of these systems consists of redundant components supplied from separate power buses. No single failure can cause a loss of more than half of the inst alled 200% cooling capac ity. These systems are described in Section 6.2.
Criterion 39 - Inspection of C ontainment Heat Removal System
The containment heat removal system shall be designed to permit appropriate periodic
inspection of important components, such as the torus, sumps, spray nozzles, and piping to assure the integrit y and capability of the system.
3.1-29 Reformatted July 2014 Discussion
The Reactor Building Heat Removal System s consist of the Reactor Building Spray System and the Reactor Buildi ng Cooling Units. The Reac tor Building Cooling Units, the Reactor Building sump and Reactor Building spray pumps, are located so that the
visual inspection of these items is possible during normal plant operation. The spray rings and nozzles of the Reactor Building Spray System are located under the dome of the Reactor Building. An air connection is provided on the supply piping to the spray
rings for testing the spray nozzles. Functional operability of each nozzle is tested by blowing air or smoke into the spray rings and observing tell-tale devices such as streamers or balloons.
For further discussion, see Section 6.2.2.
Criterion 40 - Testing of Cont ainment Heat Removal System
The containment heat removal system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leakti ght integrity of its components, (2) the operability and perform ance of the active components of the system, and (3) the operability of the system as a whole, and, under conditions as close to the design as practical, the performance of the full operational sequence that brings the system into operation, incl uding operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.
Discussion The Reactor Building Heat Removal System s have the capability of being periodically tested as follows:
- 1. Reactor Building Cooling Units
- a. The Reactor Building cooling units are used during normal operation and can be individually tested for emergency operation.
- b. The cooling coil service water valves can be operated through their full travel.
- c. The service water pumps can be tested for automatic operation.
- d. The service water booster pumps c an be tested for automatic operation.
3.1-30 Reformatted July 2014 2. Reactor Building Spray System
- a. The operation of the spray pumps c an be tested by recirculation to the refueling water storage t ank through a test line.
- b. The Reactor Building Spray System va lves can be operated through their full travel.
Active components are identified in Section 3.9.2. Inservice testing is discussed in
Section 3.9.4. The Reactor Building cool ing units and Reactor Building Spray System are discussed in Sections 6.2 and 9.2.
Criterion 41 - Contai nment Atmosphere Cleanup
Systems to control fission products, hydrogen, oxygen, and other substances which
may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, t he concentration and quantity of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.
Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for Onsite Electric Power System operation (assuming offsite power is not available) and for Offsite Electric Power System operation (assuming onsite power is not available) its safety function can be a ccomplished, assuming a single failure.
Discussion
The following systems are designed to clean up the reactor building atmosphere after a postulated loss of coolant accident (LOCA):
- 1. The Reactor Building Spray System spra ys a sodium hydroxi de (NaOH) solution into the Reactor Building to remove elem ental iodine. The power supply consists of 2 independent subsystems each supplied from separate buses. Either subsystem alone can provide the iodine removal capacity for which credit is taken
in Chapter 15. No single active failure will cause both subsystems to fail to operate. System design is discussed in Section 6.2.
- 2. The post accident hydrogen removal syst em is also designed with redundancy of vital components so that a single active fa ilure will not prevent timely operation of the system. This system is described in Section 6.2.5.
3.1-31 Reformatted July 2014 3. The recirculation system HEPA filters are capable of f iltering the full post loss of coolant accident recirculation air flow for wh ich credit is taken in Chapter 15. This system is discussed in Se ctions 6.2.2 and 6.5.1.
Criterion 42 - Inspection of Cont ainment Atmospher e Cleanup Systems
The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, su ch as filter frames , ducts, and piping to assure the integrity and capability of the systems.
Discussion
The Reactor Building atmosphere cleanup syst ems, with the exception of the spray headers and nozzles, are designed and located such that they can be inspected periodically as required. The spray headers and nozzles can be air tested as described in the discussion of Criterion 39.
Criterion 43 - Testing of Cont ainment Atmosphere Cleanup Systems
The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the oper ability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves, and (3) the operability of the system s as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency pow er sources, and the operation of associated systems.
Discussion
The Reactor Building Atmosphere Clean up System can be tested as follows:
- 1. Reactor Building Spray System
- a. The operation of the spray pumps can be tested by recirculation of borated refueling water to the refueling wa ter storage tank through a test line.
- b. The system valves can be oper ated through their full travel.
- c. The system is checked for leaktightness during testing.
3.1-32 Reformatted July 2014 2. Post Accident Hy drogen Removal System
- a. The post accident hydrogen recombiners are periodically tested to verify operation of the control system and functional performance of the heaters at the required temperature level. These tests are performed during normal plant operation from the reco mbiner control panels.
- b. The alternate purge line can be operated to test the full operational sequence.
- 3. Reactor Building Cooling Unit HEPA Filters
- b. The HEPA filters are tested for through leakage during system testing.
Identification of active components is presented in Section 3.9.2. A further discussion of inservice testing is provided in Section 3.9.4.
Criterion 44 - Cooling Water
A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat si nk shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.
Suitable redundancy in component s and features, and suitable interconnections, leak detection, and isolation capabilitie s shall be provided to assure that for Onsite Electric Power System operation (assuming offsite power is not available) and for Offsite Electric Power System operation (assuming ons ite power is not available) the system safety function can be accomplished, assuming a single failure.
Discussion
The systems provided to transfe r heat from items im portant to safety to the service water pond consist of systems identified as: Service Water, Component Cooling Water and Chilled Water.
Redundancy is provided by the creation of the Seismic Category I service water pond which functions as the ultimate heat sink in the unlikely event of main dam failure, by the installation of three 100% capacity pumps and by the provision of features of redundancy and isolation capability.
The cooling water systems provided to tr ansfer heat produced un der operating and accident conditions from the plant to the environment are as follows: service water pond, Service Water System, Chilled Water System, and Component Cooling Water System. The service water pond is the ulti mate heat sink. The Service Water System 3.1-33 Reformatted July 2014 transfers heat from the com ponent cooling heat exchangers to the service water pond. The Component Cooling Water System and the Chilled Water System are intermediate, closed cycle cooling systems. The Component Cooling Water System is used to isolate service water (water from the service wa ter pond) from normally radioactive streams that initially carry heat to be rejected to the environment.
The service water pond is an impounded portion of Monticello Reservoir and is enclosed by dams designed to withstand the effects of the SSE and by natural features.
For additional details, s ee Sections 2.4 and 9.2.
The Service Water System, Component Cooling Water System and Chilled Water System have 2 independent subsystems suppli ed with electric power from separate buses. These subsystems are operable from either Offsite or Onsite (emergency diesel generators) Electric Power S ystems. Therefore, for eac h system, the subsystems are totally redundant and the availability of the minimum engineered safety features requirements is ensured assuming a single failure.
Criterion 45 - Inspection of Cooling Water System
The Cooling Water System shall be designed to permit appropriate periodic inspection
of important components, such as heat exch angers and piping, to a ssure the integrity and capability of the system.
Discussion
Important Cooling Water S ystem components are accessi ble for required periodic inspection. These components have suitable manholes, handholes or inspection ports to allow for periodic inspection.
Criterion 46 - Testing of Cooling Water System
The Cooling Water System shall be designed to permit appropriate periodic pressure
and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and the perfo rmance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown a nd for loss-of-coolant accidents, including operation of applicable portions of the pr otection system and the transfer between normal and emergency power sources.
RN 99-155 3.1-34 Reformatted July 2014 Discussion
Redundancy and isolation are provided to allow periodic pressure and functional testing of the systems as a whole, including t he functional sequence that initiates system operation, and also including transfer between the normal and diesel power sources. At least one of the redundant systems is in service during normal operation.
Identification of active components is presented in Section 3.9.3. A further discussion of inservice testing is provided in Section 3.9.4.
3.1.2.5 Reactor Containment Criterion 50 - Containment Design Basis
The reactor containment structure, including access openings, penetrations and the
containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodat e, without exceeding the design leakage rate and, with sufficient marg in, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. Th is margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and energy
from metal-water and other chemical reac tions that may result from degraded
emergency core cooling functioning, (2) t he limited experience and experimental data available for defining a ccident phenomena and containm ent responses, and (3) the conservatism of the calculati onal model and input parameters.
Discussion The design of the Reactor Building is based upon the containment design basis accident which assumes the double ended rupt ure of a main steam line inside the Reactor Building and the worst single acti ve failure. The maximum pressure and temperature determined for the design basis accident are 53 psig and 372.7
°F. The super heated temperatures within the Reactor Building are limited in magnitude and duration via closure of the main steam isolation valves and spray actuation. Following spray actuation, the Reactor Building remain s saturated in the long term and below the Reactor Building design temperature of 283
°F. The pressure differential between 53 psig and the reactor building design pressure of 57 psig, and the long term temperature of less than 283
°F, provide ample margin to allow for increase energy sources as the result of degraded performance of em ergency core cooling systems.
3.1-35 Reformatted July 2014 For further discussion, s ee the following sections:
Section 1. Classification of Structures, Components and Systems 3.2 2. Wind Design Criteria 3.3 3. Missile Protection Criteria 3.5
- 4. Criteria for Protection Agains t Dynamic Effects Associated with a Loss of Coolant Accident 3.6 5. Seismic Design 3.7
- 6. Design of Reactor Building 3.8
- 7. Containment Functional Design 6.2.1
- 8. Reactor Building Heat Removal System 6.2.2 9. Accident Analyses 15.0
Criterion 51 - Fracture Prevention of Containment Pressure Boundary
The reactor containment boundary shall be designed with sufficient margin to assure
that under operating, maintenan ce, testing, and postulated a ccident conditions (1) its ferritic materials behave in a nonbrittle manner and (2) the probability of rapidly
propagating fracture is minimized. The desi gn shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, te sting, and postulated acci dent conditions, and the uncertainties in determining (1) material properties, (2) residual, steady state, and
transient stresses, and (3) size of flaws.
Discussion
The Reactor Building liner material has a ma ximum nil ductility transition temperature of at least 30
°F below the minimum service temperature.
Ferritic materials exposed to the external environment have been selected so that their temperatures under normal operati ng and testing conditions are 30
°F or more above nil ductility transition temper ature (see Section 3.8).
Criterion 52 - Capability for Containment Leakage Rate Testing
The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that per iodic integrated leakage rate testing can be conducted at containment design pressure.
Discussion
The Containment System is designed and cons tructed and the necessary equipment is provided to permit periodic integrated leak ra te tests during the pl ant lifetime. The testing program satisfies the require ments of Appendix J to 10 CFR 50.
The provisions for testing and the test program satisfy the requirements of Criterion 52.
RN 01-113 3.1-36 Reformatted July 2014 For further discussion, s ee the following sections:
Section 1. Concrete Reactor Building (Equipment Hatch and Personnel Airlocks) 3.8.1 2. Containment Functional Design 6.2.1
- 3. Containment Leakage Testing 6.2.6
- 4. Technical Specifications 16.0
Criterion 53 - Provisions for C ontainment Testing and Inspection
The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment des ign pressure of the leaktightness of penetrations which have resilient seals and expansion bellows.
Discussion
The program defining, and the means for per forming, individual leakage rate tests on applicable penetrations in accordance with Appendix J to 10 CFR 50 are presented in Section 6.2.6. The program defining the lifetime inse rvice tendon surveillance is presented in Chapter 16. This program provides physical evidence that the structural integrity of the Reactor Building has been maintained.
The provisions made for penetration testing satisfy the requirements of Criterion 53.
For further information, see the following sections: Section 1. Containment Functional Design 6.2.1 2. Containment Leakage Testing 6.2.6
- 3. Technical Specifications
Criterion 54 - Piping Systems Penetrating Containment
Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping shall be designed with a capability to test periodically the operability of the isolation va lves and associated apparatus and to determine if valve
leakage is within acceptable limits.
98-01 RN 01-113 RN 01-113 98-01 RN 99-136 3.1-37 Reformatted July 2014 Discussion
The required isolation and testing capabiliti es are provided in all piping systems penetrating containment. Test connections are provided as required to enable periodic leak rate determination for individual valv es and other isolation barriers. Means are provided for demonstrating the operability of remotely operated isolation valves or other isolation barriers. The Engi neered Safety Features Actuation System test circuitry provides the means for testing isolation valve operability. This is discussed in Section 6.2.
For further discussion, s ee the following sections:
Section 1. Containment Functional Design 6.2.1
- 2. Containment Isolation Systems 6.2.4
- 3. Containment Leakage Testing 6.2.6
Criterion 55 - Reactor Coolant Pressu re Boundary Penetr ating Containment
Each line that is part of t he reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that th e containment isolation provisions for a specific class of lines, such as instrum ent lines, are acceptable on some other defined basis:
- 1. One locked closed isolation valve in side and one locked closed isolation valve outside containment, or
- 2. One automatic isolation valve inside and one locked closed isolation valve outside containment, or
- 3. One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment, or
- 4. One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, aut omatic isolation valv es shall be designed to take the position that provides greater safety.
Other appropriate requirements to minimi ze the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Dete rmination of the appropriateness of these RN 01-113 3.1-38 Reformatted July 2014 requirements, such as high quality in desi gn, fabrication, and testing, additional provisions for inservice inspection, protec tion against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteri stics, and physical characterist ics of the site environs.
Discussion
The boundary for the Reactor Coolant System is defined in accordance with Section 4 of ANS N18.2
[1]. The entire Reactor Coolant System as defined above, is located within the Reactor Building. Thus, this criterion does not apply to Westinghouse pressurized water reactors. However, the specific va lving arrangements specif ied in Criterion 55 have been adhered to in the design of this plant with the specific exceptions listed and justified in Chapter 6.
Criterion 56 - Primary Containment Isolation
Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containm ent isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
- 1. One locked closed isolation valve in side and one locked closed isolation valve outside containment, or
- 2. One automatic isolation valve inside and one locked closed isolation valve outside containment, or
- 3. One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment, or
- 4. One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, aut omatic isolation valv es shall be designed to take the position that provides greater safety.
Discussion
Each line that connects directly to the Reactor Building atmos phere and penetrates containment is provided with containment isolation valves, except where it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable. Details are provided in Sections 5.5 and 6.2 and on the flow diagrams included in Chapters 6 and 9.
3.1-39 Reformatted July 2014 Criterion 57 - Closed System Isolation Valves
Each line that penetrates primar y reactor containment and is nei ther part of the reactor coolant pressure boundary nor connected direct ly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operat ion. This valve shall be outside the containment and located as close to the containment as practical. A simple check valve may not be used as the aut omatic isolation valve.
Discussion
Each line that penetrates containment and is not connected directly to the Reactor
Building atmosphere and is not part of the reactor coolant pressure boundary has at least one isolation valve located outside cont ainment near the penetration or is a closed system outside containment. Details are prov ided in Sections 5.5 and 6.2 and in the flow diagrams included in Chapters 6 and 9.
3.1.2.6 Fuel and Radioactivity Control Criterion 60 - Control of Release of Radioactive Materials to the Environment
The nuclear power unit design shall include m eans to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operat ion, including anticipated operational occurrences. Sufficient holdup capacity sha ll be provided for retention of gaseous and liquid effluents containing radioactive mate rials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such e ffluents to the environment.
Discussion
Waste handling systems have been in corporated in the plant design for retention and/or processing of radioactive wastes resulti ng from normal operati on. Controls and monitoring are provided to ensure that Appendi x I to 10 CFR 50 is satisfied. The plant is also designed such that radioactive rel eases during accidents will not exceed the limits of 10 CFR 100.11 or 10 CFR 50.67.
Chapter 11 describes the Radioactive Wast e Processing System , design criteria, and amounts of estimated releases of radioactive effluents to the environment. Chapter 5 and Sections 6.2, 12.1 and 12.2 describe the containment syst em which forms a barrier to the escape of fission products should a loss of coolant occur. Chapter 6 describes the engineered safety features for control of reactivity and Reactor Building pressure.
RN 12-034 3.1-40 Reformatted July 2014 Criterion 61 - Fuel Storage and Handling and Radioactivity Control
The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit
appropriate periodic inspection and testing of components im portant to safety, (2) with suitable shielding for radi ation protection, (3) wit h appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the im portance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduc tion in fuel storage coolant inventory under accident conditions.
Discussion
The Spent Fuel Cooling System, Fuel Handling System, Radioactive Waste Processing Systems and other systems that contain radioactivity are designed to assure adequate safety under normal and postulated accident conditions.
- 1. Components are designed and located such that appropriate periodic inspection and testing may be performed.
- 2. All areas of the plant ar e designed with suitable shield ing for radiation protection based on anticipated radiation dose rates and occupancy as discussed in Section 12.1.
- 3. Individual components which contain significant radioactivity are located in confined areas which are adequately vent ilated through appr opriate filtering systems or are vented to the Gaseous Wast e Processing System.
Details of the ventilation systems are pr esented in Section 9.4.
- 4. The Spent Fuel Cooling System provides cooling to remove residual heat from the fuel stored in the spent fuel pool. The system is designed with redundancy and testability to assure continued heat removal. The Spent Fuel Cooling System is described in Section 9.1.3.
- 5. The spent fuel pool is designed such that no postulated accident could cause excessive loss of coolant inventory.
Criterion 62 - Prevention of Crit icality in Fuel Storage and Handling
Criticality in the fuel storage and handling system sha ll be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
3.1-41 Reformatted July 2014 Discussion
Criticality in new and spent fuel storage areas is prevented by physical separation of fuel assemblies and the presence of borated water in the spent fuel pool. The fuel storage racks are constructed so that fuel assemblies may be inse rted in prescribed locations only. These have a minimum center to center spacing in both directions to ensure subcriticality even if assemblies are immersed in unborated water. Criticality prevention and criticality considerations are discussed in Sections 9.1 and 4.3, respectively.
Criterion 63 - Monitoring Fuel and Waste Storage
Appropriate systems shall be provided in f uel storage and radioa ctive waste systems and associated handling areas: (1) to detect conditions that may result in loss of residual heat removal capability and excessiv e radiation levels; and (2) to initiate appropriate safety actions.
Discussion
Monitoring systems are provided to cause alarms when excessive temperature or low water level occurs in the spent fuel pool.
Appropriate safety actions are initiated by operator action as outlined in Section 9.1.3.
Radiation monitors and alarms are provided as required to warn personnel of impending excessive levels of radiation or airborne activity. The radiation monitoring system is
described in Sections 11.
4, 12.1.4, and 12.2.4.
Criterion 64 - Monitoring Radioactivity Releases
Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipat ed operational occurrences, and from postulated accidents.
Discussion
The Reactor Building atmosphere is continually monitored during normal and transient plant operations, using the Reactor Building particulate, gaseous, and iodine radiation monitors. Under accident conditions, when temperature and pressure permit, samples
of the Reactor Building atmosphere can be obtained to provide data on existing airborne radioactivity concentrations within the Reactor Building. In addition, a Reactor Building high range area gamma monitor is used to m onitor the potential gamma radiation dose which may result from the postulated acciden
- t. Radioactivity levels contained in the plant effluent discharge paths and in the environs are monitored during normal and accident conditions by the pl ant radiation monitoring system, described in Sections 11.4 and 12.2.4, and the health physics pr ogram described in Section 12.3.
3.1-42 Reformatted July 2014 3.1.3 SINGLE FAILURE CRITERION The Engineered Safety Features systems are designed to tolerate a single failure (see General Design Criteria 34 and
- 35) during the period of recovery following an incident without loss of their protective function.
During the short term immediately following the in cident, this single failure is limited to a failure of an active component to complete it s function as required. Should the failure occur during the long term rather than the short term period following the incident, the failure definition is expanded such that the systems designs will tolerate either an active failure or a passive failure without loss of their protective function.
3.1.3.1 Definitions
- 1. Engineered Safety Features
Engineered safety features are provided for sensing the incident occurrence, initiating protective action and completing the necessary protective function to remain within NRC specified criteria for r adioactivity release from the plant site.
The engineered safety features are as follows:
- a. The containment
- b. The Reactor Building Heat Remova l Systems (Reactor Building Spray System and Reactor Buil ding Cooling Units)
- c. The Reactor Building Air Purification and Cleanup Systems
- d. The Containment Isolation System
- e. The Combustible Gas Control System
- g. Habitability systems
- h. Fission Product Remo val and Control Systems
3.1-43 Reformatted July 2014 Engineered Safety Features Support Syst ems include Component Cooling, Service Water, Chilled Water, related heating, v entilating and cooling equi pment, and electric power supply associated with the required flui d or steam system. Compressed air is not required. For systems failure analysis, the effect of a single failure is considered on the total of the engineered safe ty features and the related service system required for reactor protection following the incident. As an example, if the Em ergency Core Cooling System is required and a diesel generator failure is postulated, that would be the active failure for which the consequences are analyzed. The consequences would include other failures specifically caused by the dies el failure. No further active or passive failures of the systems are considered either for the short or long term.
- 2. Period of Recovery
The period of recovery is the time nece ssary to bring the plant to cold shutdown and regain access to faulted equipment. Th e recovery period is the sum of the short and long term periods defined below.
- 3. Incident
An incident is any natural or acci dental event of infrequent occurrence and the related consequences which affect pl ant operation and require the use of Engineered Safety Features systems.
Such events, which are analyzed independently, are not assumed to occur simultaneously and include loss of coolant accident, steam line ruptures, stea m generator tube ruptur es, etc. Loss of offsite power may be an isolated occurrenc e or may be concurr ent with any event requiring Engineered Safety Features systems use.
- 4. Short Term
The short term is the time immediat ely following the incident during which automatic actions are perform ed, system responses are che cked, type of incident is identified and preparations for long term recovery operation are made. In the event of a loss of coolant accident, the per iod of the injection mode of operation during Emergency Core Cooling System oper ation is the basis for the short term period.
- 5. Long Term
The long term is the remainder of the recovery period foll owing the short term. In comparison with the short term, where the main concern is to remain within NRC specified site criteria, the long term peri od of operation involves bringing the plant
to cold shutdown conditions where access to the Reactor Building can be gained and repair effected.
3.1-44 Reformatted July 2014 6. Active Failure
The failure of a powered component, su ch as a piece of mechanical equipment, component of the Electrical Supply S ystem or instrumentation and control equipment, to act on command to perform it s design function constitutes an active
failure. Examples include failure of a valve to move to its correct position, failure of an electrical circuit breaker or relay to re spond, failure of a pump, fan or diesel generator to start, etc.
Consideration of equipment moving s puriously from the proper safeguards position, such as a motor operated valve inadvertently shutting, is specifically excluded.
- 7. Passive Failure
The structural failure of a static component which limits the effectiveness of that component in carrying out its design function constitutes a passive failure. When
applied to a fluid system, this means a break in the pressure boundary resulting in abnormal leakage not exceeding 50 gpm. Su ch leak rates are consistent with limited cracks in pipes, sprung flanges, valve packing leaks or pump seal failures.
3.
1.4 REFERENCES
- 1. American Nuclear Society, "Nuclear Safe ty Criteria for the Design of Stationary Pressurized Water Reactor Plants," ANS N18.2, as discussed in Appendix 3A, Regulatory Guide 1.26.
- 2. Institute of Electrical and Electronics Engineers, "Criteria for Class 1E Electric Systems for Nuclear Power G enerating System s," IEEE-308-1971.
- 3. Institute of Electrical and Electronics Engineers, "IEEE Criteria for Nuclear Power Plant Generating Station Protection Systems," IEEE-279-1971.
- 4. Institute of Electrical and Electronics Engineers, "Trial Use Guide for the Application of the Single Failure Criterion to Nuclea r Power Generating Station Protection Systems," IEEE-379-1972.
- 5. American Society for Testing and Materials, "Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," ASTM E-185, 1970.
3.3-1Reformatted PerAmendment 02-013.3WIND AND TORNADO LOADINGS3.3.1WIND LOADINGS3.3.1.1Design Wind VelocityA design wind velocity of 100 mph was used in the analysis of Seismic Category Istructures. This value represents the maximum wind velocity at the site for an altitude of 30 feet above grade and for a 100 year recurrence interval. Sections 2.3.1 and 2.3.2 provide the basis for this selection of design wind velocity.The wind velocity profile for the Reactor Building was computed on the basis ofTable 1a of ASCE Paper No. 3269 [1]. The 140 mph velocity at the top of the ReactorBuilding, which is 167 feet above grade, was conservatively applied for the full height of the structure as shown by Figure 3.3-2.Reference [1] recommends a gust factor of 1.1 for structures similar in size to theReactor Building. However, due to the conservative application of the design windvelocity over the Reactor Building height, the design gust factor was taken as 1.0.The wind velocity profile for other Seismic Category I structures was computed basedupon Reference [1], using equation (2) with x equal to 0.20. These velocities weremultiplied by a gust factor of 1.1, based upon the recommendations of Reference [1].
The resulting wind velocity profiles are shown by Figure 3.3-1. These velocities are somewhat conservative since equation (2) of Reference [1] applies to "coastal areas" and the plant is located in an "inland area," as defined by Reference [1].The conservative wind velocity profiles result in "actual gust factors used" as givenbelow:REACTOR BUILDINGWind Velocities (mph)Required By ActualStatedHeight AbovePercent of TotalRef. [1]Used in GustDesignGrade (Feet)Exposed HeightTable 1aDesign Factor GustUsed Factor0-50 30 100 140 1.40 1.050-150 60 120 140 1.17 1.0150-167 10 140 140 1.0 1.0 02-01 3.3-2Reformatted PerAmendment 02-01Other Seismic Category 1 StructuresWind Velocities (mph)Required By ActualStatedHeight AbovePercent of TotalRef. [1]Used in GustDesignGrade (Feet)Exposed HeightTable 1aDesign*Factor GustUsed Factor0-30 30 100 110 1.10 1.130-50 20 100 122 1.22 1.150-99 50 120 140 1.17 1.1* V Z = 1.1 [V 30(Z/30)0.20]3.3.1.2Determination of Applied ForcesThe effective pressures resulting from the design winds were computed usingEquation (6) and the shape coefficients from Reference [1]. Resulting pressures for the Reactor Building and other Seismic Category I structures are shown by Figures 3.3-1 and 3.3-2.3.3.2TORNADO LOADINGS3.3.2.1Applicable Design ParametersThe design parameters applicable to the design basis tornado are as follows:1.Rotational wind speed of 290 mph.2.Translational wind speed of 70 mph.3.Atmospheric pressure drop of 3 psi at the rate of 2 psi/sec.
3.3.2.2Determination of Forces on StructuresForces upon structures resulting from tornado winds were determined as follows:
1.The effective pressures on the Reactor Building due to a 360 mph wind werecomputed using the same shape coefficients as were used for design windanalysis [1]. These pressures were combined with the 3 psi atmospheric pressureand are shown in Figure 3.3-3. For all other Seismic Category I structures, thedesign pressures are as follows:
02-01 3.3-3Reformatted PerAmendment 02-01a.Maximum positive pressure on each building wall was obtained by applying ashape coefficient of 0.9 to the 332 lb/ft 2 dynamic pressure, neglecting theatmospheric pressure drop. This resulted in a positive pressure ofapproximately 300 lb/ft 2.b.Maximum negative pressure (suction) on each building wall and roof wasobtained by applying a shape coefficient of -0.7 to the 332 lb/ft 2 dynamicpressure and superimposing on this value the 3 psi suction due to atmospheric pressure drop. This resulted in a negative pressure (suction) of approximately 664 lb/ft 2.The design tornado wind velocity of 360 mph was assumed to be constant with height.2.No reduction in tornado wind pressure due to venting was assumed.3.Structures were designed to resist the effects of tornado generated missilesdescribed in Section 3.5.1.4. In the design it is assumed that these missiles canoccur simultaneously with the tornado loads described in item 1, above.4.General stability of the entire structure was checked using net pressures of wind,tornado, and other loads as described in Section 3.8. However, individual elements are designed for the maximum effect in either direction.3.3.2.3Effect of Failure of Structures or Components not Designed for TornadoLoadsStructures not designed to resist tornado loadings are either located so that their failure does not affect structures designed for tornado loads or are designed not to collapse.
The metal siding and roofing of the Fuel Handling Building will blow off under tornado loading. The structural steel frames supporting the overhead traveling crane will resistthe tornado and wind loads. Non-Seismic Category I structures may lose parts or portions but such parts or portions are less serious missiles than those described in Section 3.5.1.4.3.3.3REFERENCE1."Wind Forces on Structures," Transactions of the American Society of CivilEngineers (ASCE), Paper No. 3269, Volum e 126, Part 2.
02-01
3.4-1AMENDMENT 97-01AUGUST 19973.4WATER LEVEL (FLOOD) DESIGNAll Seismic Category I structures are designed for a maximum flood water level elevation of 437'-6" at the berm and a high water table ground water level elevation of 420' +/- 3'.3.4.1FLOOD PROTECTION 3.4.1.1Safety-Related Systems and Components Protected Against FloodsAll safety-related systems and components are protected against surface flooding bygrading the site to carry surface water away from structures housing these systems and components. See Section 2.4 for details. Systems located below grade are protected as described in Sections 3.4.1.2 and 3.4.1.3.3.4.1.2Structures that House Safety-Related EquipmentThe portions of Seismic Category I structures located below finished grade areprotected on their outside surfaces by a continuous waterproofing membrane. Inaddition, the auxiliary building mat is protected by a waterproof membrane on the bottom surface.Below grade penetrations for conduit and piping are provided with waterproofingmembrane covers. No personnel or equipment hatches are located in outer wallsbelow grade.3.4.1.3Means of Providing Flood Protection for Vulnerable EquipmentIn the event that leakage occurs, additional flood protection is provided for safety classcomponents, equipment, and systems located below grade by sloping of floors to sumps and pumping of any water from these sumps.3.4.1.4Procedures Required and Implementation Times for Cold Shutdown forFlood ConditionsSpecial procedures for use in the event of flooding are not required. See Section 2.4.14.3.4.1.5Safety-Related Systems or Components Capable of Normal FunctionWhile FloodedSafety-related electrical cabling in underground duct runs is capable of fulfilling its normal function when completely or partially flooded.
3.4-2AMENDMENT 97-01AUGUST 19973.4.2ANALYSIS PROCEDURESSeismic Category structures are designed for buoyancy. No Seismic Category Istructures become unstable with respect to uplift or overturning due to loadcombinations including the design water level. Load combinations are presented in Section 3.8. The plant site is protected against potential floods up to elevation 438.0' (for details see Section 2.4.10). Therefore, dynamic effects of flooding are not applicable and were not considered.
3.6-1 Reformatted July 2014 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING Protection of structures, systems, and components important to safety from the dynamic
effects of piping failure is provided in accordance with the requi rements of 10 CFR 50, Appendix A, General Cr iterion 4. Subsequen t to the General Design Criterion 4 final rule change
[16], postulated breaks in the reactor coolant loop piping, except for branch line connections, have been elimin ated for V. C. Summer. T he dynamic effects of the postulated breaks at six terminal ends in t he cold, hot, and crossover legs, the steam generator inlet elbow, and the loop closure weld in the crossover leg were eliminated from the structural design basis by application of leak-before-break methodology, as presented in Reference [17] and updated by Refer ence [19]. Approval of the elimination of the V. C. Summer reactor coolant loop piping breaks is given in the Summer Safety Evaluation Report, dated January 11, 1993
[18]. To provide the high margins of safety required by General Design Criterion 4, the nonmechanistic pipe rupture design basis is maintained for containment design and ECCS analysis, and the postulated pipe ruptures are retained for electrical and mechanical equipment environmental qualification.
This section describes the design bases and design measures employed to protect the containment, reactor coolant pressure boundary, and other essential systems and equipment from the effe cts of postulated pipe ruptures.
This protection is provided inside and outside of containment and consider s jet blowdown from postulated breaks, as well as pipe whip and reactive forces.
3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS 3.6.1.1 Design Bases 3.6.1.1.1 Essential Syst ems Outside Containment Essential piping systems required for postulated piping failures outside of containment
are as follows:
- 1. Main Steam System up to and incl uding containment isolation valves.
- 2. Feedwater System up to and includ ing containment isolation valves.
- 3. Emergency Feedwater System.
- 4. Chemical and Volume Control System.
- 5. Residual Heat Removal System.
- 6. Safety Injection System.
02-01 3.6-2 Reformatted July 2014
- 7. Steam Generator Blowdown System up to and including cont ainment isolation valves.
- 8. Service Water System.
- 9. Chilled Water System.
- 10. Reactor Makeup Water System.
- 11. Component Cooling Water System.
3.6.1.1.2 Essential Systems Inside Containment Essential piping systems required for postulated piping failures inside containment are as follows:
- 2. Main Steam System.
- 3. Feedwater System.
- 4. Chemical and Volume Control System.
- 5. Residual Heat Removal System.
- 6. Safety Injection System.
- 7. Steam Generator Blowdown System.
- 8. Service Water System.
- 9. Reactor Building Spray System.
- 10. Steam Generator Sampling System.
3.6.1.1.3 Criteria for Pr otection Against Postulated Pipe Breaks in Reactor Coolant System Piping A loss of reactor coolant accident (LOCA) is assumed to occur for a Reactor Coolant System branch line break down to the restraint of the second normally open automatic isolation valve (Case II in Figure 3.6-1) on outgoing lines and down to and including the second check valve (Case III in Figure 3.6-1) on incoming lines normally with flow. A
pipe break beyond the restraint or second che ck valve does not result in an uncontrolled loss of reactor coolant if either of the two valves in the line close.
3.6-3 Reformatted July 2014 Accordingly, both of the automatic isolati on valves must be suitably protected and restrained as close to the valves as possibl e so that a pipe break beyond the restraint does not jeopardize the integrity and operability of the valves. This criterion takes credit for only one of the two valves performing its intended function. For normally closed isolation or incoming check valves (Cases I and IV in Figure 3.6-1), a LOCA is assumed to occur for pipe breaks on the reactor side of the valve.
Branch lines connected to the Reactor Cool ant System are defined as "large" for purposes of these criteria when the inside di ameter is greater than 4 inches up to the largest connecting line. Ruptur e of these lines results in a rapid blowdown from the Reactor Coolant System and basic protection is provided by the accumulators and the low head safety injection pumps (residual heat removal pumps).
Branch lines connected to the reactor coolant system are defined as "small" if they have an inside diameter equal to or less than 4 inc hes. This size is such that Emergency Core Cooling System analyses using realisti c assumptions show that no fuel cladding damage is expected for a br eak area of up to 12.5 in 2 which corresponds to a 4 inch inside diameter piping.
Engineered safety features provide for co re cooling and borati on, reactor building temperature and pressure reduction, and activity confinement in the event of a LOCA or steam or feedwater lin e break accident. This ensures that the public is protected in accordance with 10 CFR 50.67 guidelines. These safety systems are designed to provide protection for a Reactor Coolant System pipe rupture of a size up to and
including double ended severance of a Reactor Coolant System main loop.
To assure the continued integrity of the essential component s and the engineered safety features systems, consideration is gi ven to the consequentia l effects of the pipe break itself to the extent that:
- 1. Minimum performance capabilities of engi neered safety features systems are not reduced to less than those required to protect against the postulated break.
- 2. Containment leak tightness is not decr eased to less than the design value if the break leads to a loss of reactor coolant.
The containment is here defined as the Reactor Building liner and penetrations, t he steam generator sh ell, the steam generator steam side instrumentation connections and the steam, feedwater, blowdown, and steam generat or drain pipes within the Reactor Building.
- 3. Propagation of damage is limited in type and/or degree to the extent that:
- a. A pipe break which does not directly re sult in a loss of reactor coolant will not cause a loss of reactor coolant or steam or feedwater line break, through consequential damage.
RN 12-034 3.6-4 Reformatted July 2014
- b. A Reactor Coolant System pipe break does not cause a Steam or Feedwater System pipe break and vice versa.
Criteria relative to large and sma ll branch line breaks are as follows:
- 1. Large Branch Lines
Large branch line piping is restrained to satisfy the following criteria:
- a. Propagation of the break to unaffected loops is prevented to assure the delivery capacity of the accumulators and low head safety injection pumps.
- b. Propagation of the break in the affected loop is pe rmitted to occur but does not exceed 20% of the area of line which initially rupt ured. This criterion is voluntarily applied so as not to substantially increase the severity of the loss of coolant. Two exceptions to this criterion are permitted:
(1) The postulated rupture of the 12 in ch accumulator injection lines is permitted to result in rupture of the 6 inch safety injection lines.
(2) A postulated rupture of reactor coolant loop 2, 6 inch cold leg injection line is permitted to result in a rupture of the 3 inch normal charging line.
This results in an additional break ar ea equal to 25% of the area of the line which initially ruptured.
- c. Where restraints on the lines are necessary to prevent impact on and subsequent damage to neighboring equipm ent or piping, restraint type and spacing are chosen such that either a plastic hinge on the pipe at the two
support points closest to the break is not formed or, if a plastic hinge is formed, piping does not impact essential equipment.
- 2. Small Branch Lines
In the unlikely event that one of the small pressurized lines should fail and result in a LOCA, the piping is restrained or arr anged to satisfy the fo llowing criteria in addition to the criteria previously discussed for large branch lines:
- a. Break propagation is limited to the affected leg (i.e., propagation to the other leg of the affected loop and to other loops shall be prevented).
- b. Propagation of the break in the affected leg is permitted but is limited to a total break area of 12.5 in 2 (4 inch inside diameter).
The exception to this criterion is when the initia ting small break is in the high head safety injection line. Further propagation is not permitted for this case.
3.6-5 Reformatted July 2014
- c. Damage to the high head safety injecti on lines connected to the other leg of the affected loop or to other loops is prevented.
- d. Propagation of the break to the high head safety injection line connected to the affected leg is prevented if the line break results in a loss of core cooling capability due to a spilling injection line.
3.6.1.2 Description 3.6.1.2.1 High En ergy Systems High energy systems, systems with normal operating te mperatures in excess of 200
°F or normal operating pressures above 275 psig, are as follows:
- 2. Main Steam System.
- 3. Feedwater System.
- 4. Emergency Feedwater System.
- 5. Chemical and Volume Control System.
- 6. Safety Injection System.
- 7. Steam Generator Blowdown System.
- 8. Auxiliary Steam System.
- 9. Main Steam Dump System to atmosphere, up to control valves.
- 10. Main Steam Drains System.
- 11. High energy systems with lines all 1 inch and smaller.
No breaks are postulated in the following systems:
- a. Nuclear Sampling System.
RN 01-113 3.6-6 Reformatted July 2014
- 12. High energy systems remotely located in the Turbine Building and not considered for pipe rupture.
- a. Extraction Steam System.
- b. Condensate System.
- c. High Pressure Heater Dr ip, Vent and Relief System.
- d. Low Pressure Heater Dr ip, Vent and Relief System.
- e. Feedwater Pump St artup Drain System.
- f. Miscellaneous Steam Drain System.
- g. Auxiliary Boiler Chemical Feed System.
- h. Turbine Cycle Sampling System.
- i. Main Steam Dump System to condensers, up to control valves.
- j. Turbine Generator Electrohy draulic Fluid Control System.
3.6.1.2.2 Moderate Energy Systems Moderate energy systems, syste ms with normal operating tem peratures less than or equal to 200°F and normal operating pressures equal to or less than 275 psig, include all piping systems not list ed in Section 3.6.1.2.1.
3.6.1.2.3 High Energy Syst ems Enclosed in Compartment to Protect Nearby Essential Systems and Components Main steam and feedwater lines are enclos ed in the penetration access areas at floor elevation 436' and on the upper le vel of the Intermediate Building to protect nearby essential systems and components.
The high and moderate energy li nes outside of containment are enclosed in structures and compartments. The pressure buildups due to postulated pipe break within the enclosures are calculated to verify the adequacy of the structural design. The postulated break of the 32 inch main steam line produces the most adverse effects in the east and west penetration access areas, and Intermediate Building, while the 30 inch main steam line produces the most adverse effects in the Turbine Building. The
FLASH-2 [1] and FLASH-4
[2] computer codes, which perform mass and energy balance calculations for specified control volumes and flow paths on a time step basis, were used to determine the blowdown flow rates and enthalpies for these ruptures. These results were then used as input for FLASH-4
[2] and CONTEMPT
[4] calculations to 3.6-7 Reformatted July 2014 determine the pressure responses of the compartment. In calculating these pressure responses, the following assumptions were made:
- 1. As mass and energy are introduced into a control volume, thermodynamic equilibrium is assum ed throughout the node.
- 2. No heat transfer occurs between the accident environment and the surrounding structures and equipment.
- 3. Relief and vent areas are multiplied by a conservative 0.6 discharge coefficient.
- 4. The models used for subcompartment analysis in the Intermediate and Turbine Buildings are shown by Figures 3.6-1a through 3.6-1e and Tables 3.6-0 through 3.6-0g. Intermediate Building m odels were analyzed using the FLASH-4
[2] computer code. The Turbine Building was analyzed using the CONTEMPT [4] computer code to determine loadings on the Intermediate Building wall. Appropriate Tables and Figures are as follows:
- a. Intermediate Building - West Penetration Access Area
(1) Nodalization model - Fi gure 3.6-1a, Table 3.6-0.
(2) Mass/energy release data - Table 3.6-0b.
- b. Intermediate Building - East Penetration Access Area
(1) Nodalization model - Fi gure 3.6-1b, Table 3.6-0a.
(2) Mass/energy release data - Table 3.6-0b.
- c. Intermediate Building - Subcompa rtment Containing Main Steam and Feedwater Headers (1) Nodalization model - Fi gure 3.6-1c, Table 3.6-0c.
(2) Mass/energy release data - Table 3.6-0e.
- d. Intermediate Building - Outside Subcompartment Containing Main Steam and Feedwater Headers (1) Nodalization model - Fi gure 3.6-1d, Table 3.6-0d.
(2) Mass/energy release data - Table 3.6-0e.
3.6-8 Reformatted July 2014
- e. Turbine Building - Loading on Intermediate Building Wall
(1) Nodalization model - Fi gure 3.6-1e, Table 3.6-0f.
(2) Mass/energy release data - Table 3.6-0g.
- 5. Initial conditions:
A steam-water mixture with a density equal to dry air at 120°F was used as the initial condition in FLASH-4 pressure res ponse calculations. Initial conditions for CONTEMPT pressure response calculat ions are presented in Table 3.6-0f.
Since both the 30 inch and 32 inch main st eam lines are at an elevation between 436' and 463' elevation slabs, each structure is analyzed assuming that the mass and energy released are introduced into t he compartments between these two elevations. The results of these studies are shown by Figures 3.6-2 through 3.6-16. Table 3.6-0h pres ents a cross reference betwe en the results shown by Figures 3.6-2 through 3.6-16 and the models listed in Item 4, above. The peak differential pressures, design and calculated, are listed in Table 3.6-0i. The analytical methods used to determine s ubcompartment pressure and temperature responses are the same for subcompartments inside or outside containment.
Where piping systems have been enclosed in compartments to protect essential systems in other compartments, the pressure versus time curves, Figures 3.6-2 through 3.6-16, are used to determine the peak pressure to be applied as loads to the compartments walls or slabs.
This is done by considering the pressure versus time as a forcing function and calculating the resistance func tion of the structural element
- s. From this a dynamic load factor is determined and applied to the peak differential pressure to arrive at an equivalent static load. The equivalent static load is ap plied to the structure in the load combination descr ibed in Section 3.8.4.
For postulated ruptures of high and moder ate energy lines outside containment, there is the possibility that adverse environmental conditions might result. To ensure proper design of safety-rel ated equipment to withstand such an environment, it is necessary to calculate t he extent to which the environment would be effected by such ruptures. These calculations are performed using the
FLASH-2 [1] and FLASH-4
[2] computer code results for the mass and energy releases from these ruptur es as input to the MNODE
[3] and CONTEMPT
[4] computer codes. Where it is desirable to calculate the environmental response to a postulated rupture in a few interconnec ted volumes, the MNODE computer code is used. However, when a one-node study of the resulting environmental conditions is sufficient, the CONTEMPT progr am is used. Both of these programs give the pressure and temperature of each control volume directly. Humidity must be calculated. This is easily perfo rmed using the follo wing equation:
3.6-9 Reformatted July 2014 Relative humidity =
g v P P Where: P g = Saturation pressure corresponding to control volume temperature.
P v = Partial pressure of vapor which:
- a. For CONTEMPT results - can be taken directly from the output.
- b. For MNODE results - calculated using the control volume vapor mass and temperature in the ideal gas law equation.
The environmental conditions of the various compartments of the penetration access areas and the Intermediate Building are studi ed to determine how they are affected by postulated ruptures of the 32 in ch diameter main steam line, the 4 inch main steam line to the emergency feedwater pump turbine and the 3 inch steam generator blowdown line.
Results of these calculations for a postulat ed main steam line rupture in the penetration rooms and Intermediate Building ar e presented in Table 3.6-1.
Reactor Building temperature and pressure ve rsus time curves used to determine the effect on the structure of accident temperature during a pipe break are discussed in Section 6.2.1.
No postulated rupture of a high energy line will have any adverse effect on the control room.
Environmental qualification of equipm ent is discussed in Section 3.11.
3.6.1.3 Safety Evaluation Failures which could affect the ability to bri ng the plant to a safe shutdown condition are analyzed in Chapter 15. These analyses include consideration of the occurrence of a single active component failure in required systems concurrent with postulated pipe rupture except as noted below in Section 3.6.2.1.1.1, for APCSB 3-1, paragraphs B.3.b and B.3.d, for an environmentally-induced failure which would not of itself result in protective action. The pipe rupture analysis required for safe plant shutdown under the applicable criteria is rendered inoperable as a consequence of postulated pipe rupture.
RN 01-113 3.6-10 Reformatted July 2014 3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTUL ATED RUPTURE OF PIPING This section discusses the following:
- 1. The design bases for determining the loca tion of postulated cracks in piping inside and outside of containment.
- 2. Procedures used to define the jet thrust reaction at the break or crack location.
- 3. The jet impingement loadings on adjacent safety-related structures, systems, and components.
3.6.2.1 Criteria Used to Define Break and Crack Location and Configuration 3.6.2.1.1 High Energy System Piping Outside Containment Breaks are postulated to occur in ASME Code, Section III, Class 2 and 3 piping and branch runs at the following locations:
- 1. At terminal ends. The terminal end for pi ping which penetrates containment is the pipe to penetration weld (see Figure 3.8-15) outside the Reactor Building.
- 2. At intermediate locations selected by either one of the following criteria:
- a. At each pipe fitting.
- b. At each location where the stresses exceed 0.8 (1.2S h + S a). Stresses are determined under the combi nation of loadings associated with the OBE and the nominal and upset plant condition loadings.
Specific non-nuclear safety class piping has been classified as Quality Related piping for the purpose of minimizing postulated pi pe break locations. This Quality Related piping is designed in accordance with Code requirements. A rigorous analysis of this Quality Related piping is performed and breaks are then postulated based on the criteria noted above for Code Class piping.
Breaks in non-nuclear safety class piping are postulated to occur at the following locations in each piping or branch run:
- 1. At terminal ends.
- 2. At each intermediate pipe fitting, welded attachment, and valve.
Circumferential breaks are postulated to occu r in fluid system piping and branch runs with nominal pipe size in excess of 1 inch.
02-01 RN 01-113 3.6-11 Reformatted July 2014 Where a pipe elbow break location is select ed without benefit of stress calculations, the pipe-to-elbow weld that joins the elbow to the shorter, stra ight, piping run is considered as the location of the break. Where break locations are selected in full size branch connection tees without benefit of stress calculations, the two pi pe-to-tee welds that join the tee to the shorter, strai ght, piping runs are considered to be break locations. Where break locations are selected in reduced size branch connection t ees without benefit of stress calculations, the pipe to tee weld that joins the tee to the shorter, straight, main piping run and the pipe to tee weld that joins the tee to the branch piping run are considered to be break locations.
Longitudinal breaks are postulated to occur in high energy system piping at the location of each postulated circumferential break except at the terminal ends under conditions discussed below:
- 1. Longitudinal breaks are not postulated to occur in high energy system piping and branch runs of nominal 3 in ch pipe size and smaller.
- 2. Longitudinal breaks are postulated to occur in addition to, but not concurrently with, circumferential breaks.
- 3. Longitudinal breaks are not postulated to occur at terminal ends if the system piping at the terminal ends contains no longitudinal pipe welds.
- 4. Longitudinal breaks are assumed to result in an axial pipe split without pipe severance. Splits are oriented at two diametrically opposed points on the circumference of the pipe or fitting such that a jet reaction results that is normal to the plane formed by two of the applicable ort hogonal axes, x, y, and z, of the piping configuration.
3.6.2.1.1.1 Conformanc e to Branch Technical Positions APCSB 3-1
[13] and MEB 3-1 [14] An analysis has been performed which demonstrates that acceptable protection against the effects of piping failures outside c ontainment has been provided. This analysis satisfies the intent of t he guidelines of Branch Technical Positions (BTP) APCSB 3-1 and MEB 3-1. Since these positions were published a considerable period of time after the Virgil C. Summer Nuclear Station pipe rupture analysis had commenced, certain requirements could not be followed.
However, in certain respects, the design and analyses to cope with postulated pipe rupture for the Virgil C. Summe r Nuclear Station are more stringent than required by the previously referenced BTPs. Specific differences are as follows:
- 1. APCSB 3-1:
Paragraph B.3.b(1): Offsite power was assumed to be unavailable for all postulated piping failures.
3.6-12 Reformatted July 2014
Criteria stated in the previously referenced BTPs with which the analysis does not fully
comply and the alternative approaches are as follows:
- 1. APCSB 3-1:
- a. Paragraph B.2.c(1): The fluid system piping between containment isolation valves is not designed to the stress limits specified in Paragr aphs B.1.b or B.2.b of BTP MEB 3-1. Breaks or cracks, as appropriate, are postulated in these portions of the fluid system piping in accor dance with the criteria stated in Section 3.6.2.1.1.
- b. Paragraph B.2.d(2): For these portions of fluid system piping identified in Paragraph B.2.c, the inservice examination will be that required by the ASME Code, Section XI.
- c. Paragraph B.3.b and B.3.d: The effects of an environmentally - induced failure caused by a leak or rupture which would
not of itself result in protective action may include a loss of redundancy in the protective function, but not a loss of the protective function, as
permitted by BTP-APC SB 3-1, Appendix B, paragraph 11.b.
[15]. In these cases, plant shutdown is required. T he use of Appendix B in lieu of BTP-APCSB 3.1 is permitted by the implementation schedule of paragraph B.4.c, since the V. C. Summer construction permit is
dated March 1973. Other cr iteria of BTP-APCSB 3.1 for single failure a nalyses are met, including the paragraph B.3 criteria for evaluating effects of cracks in moderate energy lines.
3.6.2.1.2 High Energy System Piping Inside Containment As indicated in Section 3.6, dynamic effe cts resulting from pos tulated breaks in the reactor coolant loop piping (i.e., the six te rminal ends in the cold, hot, and crossover legs, a split in the steam generator inlet elbow, and the loop closure weld in the crossover leg) were eliminated from the st ructural design basis for V. C. Summer.
Reactor coolant loop branch line connection (i.e., accumulator connection, pressurizer surge line, residual heat removal, etc.) br eaks are postulated. These postulated break
locations and methods that are used to determine them are described in Reference [5].
RN 98-014RN 01-113RN 02-016RN 02-016 3.6-13 Reformatted July 2014 Breaks are postulated to occur in ASME Code , Section III, Class 1, piping, other than piping discussed in Reference [5], at the following locations in each piping or branch run:
- 1. At the terminal ends.
- 2. At any intermediate location between terminal ends where the primary plus secondary stress intensities (circumfer ential or longitudinal) derived on an elastically calculated basis under loadings associated with specific seismic events and normal and upset operational plant conditions exceed 2.4S
- m.
- 3. At any intermediate location between te rminal ends where the cumulative usage factor, U, derived from the piping fati gue analysis under the loadings associated with specified seismic events and normal and upset plant operational conditions exceeds 0.1.
Breaks are postulated to occur in ASME Code , Section III, Class 2 and 3 piping at the following locations in each piping or branch run:
- 1. At the terminal ends. The terminal end fo r piping which penetrates containment is the pipe to penetration weld (see Figure 3.8-15) inside the Reactor Building.
- 2. At intermediate locations selected by either one of the following criteria:
- a. Each pipe fitting.
- b. Any location where either the circum ferential or longitudinal stresses, derived on an elastically calculated basis under loadings associated with specified seismic events and normal and upset operational plant conditions exceeds
0.8 (1.2 S h + S a).
The following types of breaks are postulated to occur at locations previously identified for ASME Code, Section III, Class 1, 2, and 3 piping:
- 1. Circumferential breaks in piping runs and branch runs exceeding 1 inch nominal pipe size.
- 2. Longitudinal breaks in piping runs and branch runs of 4 inch nominal pipe size and large except as discussed in item 3, below.
- 3. Longitudinal breaks are not postulated to o ccur at terminal ends if system piping at the terminal ends dose not cont ain longitudinal pipe welds.
Where break locations are se lected without benefit of stre ss calculations, breaks are postulated to occur at the piping welds to each fitting or valve.
3.6-14 Reformatted July 2014 Longitudinal breaks are assumed to result in an axial split without pipe severance.
Splits are oriented at two diametrically opposed points on th e circumference of the pipe or fitting such that a jet reaction results that is normal to the plane formed by two of the applicable orthogonal axes, x, y, and z, of the piping configuration.
3.6.2.1.3 (Section Deleted) 3.6.2.1.4 Moderate Energy System Piping Leakage cracks are postulated to occur in the following moderate energy piping system locations where the maximum stress range exceeds 0.4 (1.2 S h + S a):
- 1. In system piping located within struct ures and compartments containing required systems and components. Cracks are postulated to occur individually at locations
appropriate to the formation of a basis for provision of maximum required protection against spray and flooding and re sultant hazard or environmental conditions.
- 2. In system piping and branch runs with nominal pipe size larger than 1 inch.
Crack openings are assumed to be a circular or ifice of cross sectional flow area equal to that of a rectangle with dimens ions of one-half pipe diameter in length and one-half pipe wall thickness in width.
3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Reactor Coolant Loop Piping Branch Line Connections Following is a summary of t he methods used to determine t he dynamic response of the reactor coolant loop associated with postulated pipe breaks at the loop piping branch nozzle. Detailed descriptions of the methods are given in Reference [5].
- 1. Time Functions of Jet Thrust Force on Ruptured and Intact Loop Piping To determine the thrust and reactive force loads to be applied to the reactor coolant loop during the postulated LOCA resulting from a break at the branch line nozzle, it is necessary to have a detailed description of the hydraulic transient.
Hydraulic forcing functions are calculated fo r the reactor coolant loops as a result of a postulated LOCA. These forces result from the transient flow and pressure histories in the Reactor Coolant System.
The calculation is performed in two steps. The first step is calculation of t he transient pressure, mass flow rates, and thermodynamic properties as a function of time. The second step uses the results obtained from the hydraulic analysis, along with input of areas and direction
coordinates, and calculates the time history of forces at appropriate locations in the reactor coolant loops.
3.6-15 Reformatted July 2014 The hydraulic model represents the behavior of the coolant fluid within the entire Reactor Coolant System. Key parameters calculated by the hy draulic model are pressure, mass flow rate, and density. These are supplied to the thrust calculation, together with appropriate plant layout inform ation, to determine the time dependent loads exerted by the fluid on the loops.
In evaluating the hydraulic forcing functions during a postulated LOCA, the pr essure and momentum flux terms are dominant. The inertia and gravitational terms are taken into account in evaluation of the local fluid conditions in the hydraulic model.
The blowdown hydraulic analysis was required to provide the basic information concerning the dynamic behavior of the r eactor core environment for the loop forces, reactor kinetics and core cooling ana lysis. This requires the ability to predict the flow, quality, and pressure of the fluid throughout the Reactor Coolant System. The MULTIFLEX computer code
[6] was developed with a capability to provide this information.
The MULTIFLEX computer code performs a comprehensive space-time dependent analysis of a LOCA and is designed to treat all phases of the blowdown. The stages are as follows:
- a. A subcooled stage where the rapidl y changing pressure gradients in the subcooled fluid exert the influence upon t he Reactor Coolant System internals and support structures.
- b. A two phase depressurization stage.
- c. The saturated stage.
The MULTIFLEX code employs a one dimensional analysis in which the entire Reactor Coolant System is divid ed into control volumes. The fluid properties are considered uniform and thermodynamic equilibrium is assumed in each element. Pump characteristi cs, pump coastdown and cavitation, core and steam generator heat trans fer, including the W-3 DNB correlation, in addition to the reactor kinetics are incorporated in the code.
The MULTIFLEX computer program
[7] was developed to compute the transient (blowdown) hydraulic loads resulting from a LOCA. RN 09-022RN 09-022RN 09-022RN 09-022RN 09-022 3.6-16 Reformatted July 2014 The blowdown hydraulic loads on primary loop compo nents are computed from the following equation which includes both static and dynamic effects:
F = 144A ()+144gA m7.14P 2 m 2 Where:
F = Force, lb f ,
A = Aperture area, ft 2 ,
P = System pressure, lb f/in 2 , m = Mass flow rate, lb m/sec, = Density, lb m/ft 3 ,
g = Gravitational constant = 32.174 ft-lb m/lb f - sec 2 ,
A m = Mass flow area, ft 2 In the model to compute forcing functi ons, the reactor coolant loop system is
represented by a model similar to that employed in the blowdown analysis.
The entire loop layout is described in a global coordinate system. Each node is fully described by the following:
- a. Blowdown hydraulic information.
- b. The orientation of the streamlines of the force nodes in the system which includes flow areas and projection coeffi cients along the three axes of the global coordinate system.
Each node is modeled as a separate control volume with one or two flow apertures associated with it. Two apertures are used to simulate a change in flow direction and area. Each force is divided into its x, y, and z components using the projection
coefficients. The force components ar e then summed over the total number of apertures in any one node to give a total x forc e, total y force, and total z force.
These thrust forces serve as input to the piping/restraint dynamic analysis.
- 2. Dynamic Analysis of the Reactor C oolant Loop Piping E quipment Supports and Pipe Whip Restraints.
The dynamic analysis of the reactor coolant loop piping for the LOCA loadings is described in Section 5.2.1.10.
3.6-17 Reformatted July 2014
3.6.2.2.2 Balance of Plant Piping By using the transient response of a piping system to a postulated r upture, the resulting blowdown thrust and jet forces are calculated as functions of time. For the secondary systems this information is ca lculated using the FLASH-2
[1] and FLASH-4
[2] computer codes. In performing these calculations, t he following assumptions are made for both circumferential and lo ngitudinal ruptures:
- 1. For feedwater line ruptures, it is assu med that the break area takes 1 millisecond to be fully opened, while for the main steam li nes, it is assumed that the rupture is instantaneous.
- 2. A discharge coefficient of 1.0 for the escaping fluid.
- 3. For Reactor Coolant System and Feedwater System piping ruptures, the plant is at 100% power.
- 4. For Main Steam System piping breaks, the plant is in the hot standby condition.
Using these results, the blowdown forc es after the wave propagation period are calculated using the JIP computer code
[8]. The techniques used by this program depend upon the fluid stagnation c onditions present in the pi ping system. However, in general, the thrust is calculated from the following equation as presented by Moody
[9], c ME 2 E E E gVGPP A T+= (3.6-1) Where:
T = Thrust, A E = Exit (break) area, P E = Exit pressure, P = Environmental pressure, G E = Exit mass flow rate per unit area, V ME = Exit momentum specific volume, defined below, g c = Gravitational constant
3.6-18 Reformatted July 2014 The momentum specific volume, V M is defined by the following equation:
[]++=KX1XKV)X1(XVV f g M (3.6-2) Rfg)V/V(K= (3.6-3) Where:
X = Quality, V g = Vapor specific volume, V f = Liquid specific volume, K = Vapor to liquid velocity ratio, R = Velocity ratio exponent
Evaluation of P E , G E , and V ME is discussed below for the various stagnation conditions:
- 1. Cold Water (i.e., T < 212°F)
The thrust is calculated a ssuming the exit pressure, P E , is equal to the environmental pressure, P. The mass flow rate, G E , is then calculated from the Bernoulli equation.
GgPP V EcoE f=2 12 ()/ (3.6-4) Where:
V f = Liquid specific volume at calculated exit pressure, P o = Stagnation pressure
The thrust is then calculated using E quation (3.6-1) above, assuming that V ME is equal to V
- f.
- 2. Subcooled Water
Thrust for subcooled water at stagnation conditions is calculated by the use of the Henry-Fauske model
[10]. Solution of the transcendental expressions of the Henry-Fauske model gives predict ions for the exit pressure, P E , and mass flow rate, G E. The quantity V ME is calculated assuming that the exit quality is equal to the stagnation quality. The thrust is calculated usi ng Equation (3.6-1) where a unity velocity ratio at the exit is assumed, in Equation (3.6-2).
3.6-19 Reformatted July 2014
- 3. Water-Steam Mixtures
Water-steam mixtures, or two phase flow , are handled by a combination of the Moody model
[9,11] and the Henry-Fauske model. For stagnation qualities greater than 2%, the Moody model is used and the mass flow rate and exit conditions are calculated with the velocity ratio exponen t, R, equal to 1/3. The thrust is then calculated using Equation (3.6-1). Fo r stagnation quality less than or equal to
2% [12] the Henry-Fauske model is used and the thrust is calculated using a unity velocity ratio.
- 4. Steam Saturated and superheated steam is analyzed as a perfect gas with k, the ratio of specific heats, nominally equal to 1.3 and the gas constant equal to 85.76 ft-lb f/°F-lb m. For frictionless flow, the thrust reduces to a theoretical maximum as shown in Equations (3.6-5) and (3.6-6) below:
T A k k k k PP E o=++()()1 2 11 (3.6-5) PP26.1 o (3.6-6) Where:
k = C P/C v = 1.3 for steam
Typical results of thrusts versus time for postulated main steam and feedwater breaks are shown by Fi gures 3.6-17 and 3.6-18.
In certain cases, for small lines inst ead of the above, the blowdown force is represented by a steady-state function equal to KpA.
Where: K = Thrust coefficient, p = System pressure prior to pipe break, A = Pipe break area The K values used are 1.26 for steam-s aturated water and 2.0 for subcooled, nonflashing water.
3.6-20 Reformatted July 2014 3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability In addition to pipe restraints, barriers and layout are used to provide protection from
pipe whip, blowdown jet, and reactive forces.
One of the barriers utilized for protection against pipe wh ip is the steam generator compartment wall which serves as a barrier between the reactor coolant loops and the Reactor Building liner. In addition, the ref ueling cavity walls and the steam generator compartment wall enclose each reactor c oolant loop in a separate compartment, thereby preventing an accident, which may o ccur in any loop, from affecting another loop or the Reactor Building liner. The porti on of the steam and feedwater lines within the containment are r outed behind barriers which separate these lines from all reactor coolant piping. These barriers are designed to withstand loadings caused by jet forces and pipe whip impact forces.
Engineered safety features, except Emergency Core Cooli ng System lines which must circulate cooling water to the reactor vesse l and engineered safety features instrument lines, are located outside the steam generator compartment wall. Emergency Core Cooling System lines which penetrate the steam generator compartment wall are routed around and outside the wall so that they penetrate the wall in the vicinity of the loop to which they connect.
Each individual postulated break is review ed to determine whic h essential components and piping, if any, are in t he projected path of either the pipe whip or the jet.
In each case where essential components are in the path of either the pipe whip or the jet from a postulated break, such components are moved, where practical, to unaffected locations. Components that cannot be relo cated are protected from pipe whip by restraining the pipe in which a break is postu lated to occur and from jet impingement by installation of shields.
In reviewing the mechanical aspects of these lines, it has been demonstrated, through tests performed by the nuclear steam supply system manufacturer [5], that lines hitting equal or larger size lines of the same schedule do not cause failure of the line being hit; e.g., a 1 inch line, should it fail, does not c ause subsequent failure of a 1 inch or larger size line. The converse, however, is assume d to be probable; i.e., a 4 inch line, should it fail and whip as a result of the fluid discharged through the line, could break smaller size lines, such as neighboring 3 inch or 2 inch lines.
Piping, that can be damaged and is determined to be in the projected path of a postulated whipping pipe and that cannot be reloca ted, is protected by restraint of the postulated whipping pipe. Pip ing in the projected path of a jet resulting from a postulated break is either rerouted or shie lded. When rerouting or shielding is not practical, the piping system design specificat ion identifies jet impingement loadings of
safety-related piping, as well as the allo wable stress limits and loading combination requirements.
3.6-21 Reformatted July 2014 The piping system is then analyzed using the equivalent static load method or, in some cases, a detailed time history method of analysis to determine the jet impingement
effect upon the piping. The resulting lo ads are then combined and compared to the allowable stress limits as def ined in the design specificat ion. In the event that compliance with the specified stress limit s cannot be demonstrated, shielding of the piping is provided.
3.6.2.3.1 Restra int Criteria Where pipe restraints are employed they are designed usi ng the principles of the equations of motion (dynamic analysis) or those of energy balance. Since the forces due to dead, live, seismic, and thermal loads ar e considered negligible or self relieving, only the dynamic effect of the whipping pipe is considered in the design.
The equation of motion methods (dynamic anal ysis) employ the use of the computer programs DYREC (S061), DYNAL (S085), and nonlinear numeric al integration techniques. See Section 3.8.4.4 for a description of computer codes.
The program uses the thrust versus time data described previously. The pipe and the restraint are modeled as lump mass syst ems. Nodal masses and element spring properties are determined and gaps are input. Using direct num erical integration of the equations of motion, the dynamic response of the pipe and restraint are calculated at specific time points.
The following criteria were used for material properties:
- 1. Minimum yield strength of pipe steel is reduced in accordance with operating temperature.
- 2. Minimum tensile strength of the pipe material, as listed in the material specification, is used as the ultimate strength of the pipe. Refinement of assumed ultimate strength for changes due to operating tem peratures would not result in a substansive change in pipe restraint design or reactions.
- 3. Ultimate tensile strain of both piping and restraint material is one half of guaranteed minimum percent el ongation. Ultimate shear strain is equal to guaranteed percent elongation.
- 4. Minimum values of yield strength, ultima te strength, and modulus of elasticity for pipe are taken from the ASME Code. Values for restrain t material are taken from the applicable ASTM specification.
- 5. A 10% increase in material properties is applied to allow for strain rate effect.
3.6-22 Reformatted July 2014 Acceptability of the restraint design was based upon the results of the dynamic analysis.
Neither the pipe nor the re straint stresses and strains exceed the following limits:
- 1. Tensile strains are limited to 50% of the assumed ultimate tensile strain. This is equal to 0.25 times percent elongation.
- 2. Bending and axial tensile st resses are limited to the values at the above strain limit as determined appropriate from the stre ss/strain, moment/curvature, or P/ curves.
- 3. Shearing strains are limited to 50% of the assumed ultima te shear strain. This is equal to 0.5 times percent elongation.
- 4. Shearing stresses are limit ed to the value at the above strain limit as determined from the shear/shear strain curve.
Restraints, connections, anchorages and the s upporting structure ar e designed for the maximum reactions obtained from the dynamic analysis.
In the energy balance method, an amplification factor of 1.2 is applied to the peak thrust force to determine the force on the pipe.
Initial gap and the kinetic ener gy of the pipe are consider ed in balancing the internal work with the external.
Pipe and restraint properties as described fo r the dynamic analysis, are used with the exception that no strain rate effect is assumed.
Acceptable restraints design limits are the same as for the dynamic analysis.
Restraint connections, anchorages and the supporting structure are designed for 15%
more than the maximum reactions from the energy balance analysis.
3.6.2.3.2 Jet Impingement 1. Balance of Plant Piping Blowdown forces obtained as described in Section 3.6.2.2 are used in the jet
impingement analysis. It is assumed that the total jet impingement force is equal to the thrust calculated at the break
[9] and is uniformly distributed across the cross sectional area of the jet at any particular location.
For both circumferential and longitudinal ruptures, the configur ation of the break areas is assumed to be circular in nat ure. These break areas, i.e., for the longitudinal rupture and for each end of the circumferentia l rupture, are assumed to be equal to the flow area of the pipe in the vicinity of the postulated rupture
[12].
3.6-23 Reformatted July 2014 Calculation of the jet expa nsion profile is accomplished assuming a constant 10 degree half angle of expansion for the escapi ng fluid. Due to these assumptions, the area of the jet as a f unction of distance from the break point can be expressed by the following equation:
2 E E j10tanDx21A)x(A+= (3.6-7) Where: A j(x) = The jet area as a functi on of distance from the break, x = The distance from the break, A E = Exit or break area, D E = Equivalent diameter of break ar ea (equals the pipe diameter in these cases) In some instances where the postulated break is very close to structures, a more detailed blowdown analysis may be performed to evaluate jet impact forces. This detailed analysis may prove beneficial if the fluid properties support the jet expansion profile as presented by Moody
[9]. In such cases, calculation of the jet profile may be performed by the computer code JIP
[8], which evaluates the fluid properties and determined jet profiles and pressures in accordance with the following expansion model.
The Moody expansion model estimate s fully expanded, one-dimensional asymptotic jet properties for jet expansion calculations.
Diffusion, friction, turbulent momentum, and energy exchange, and heat tr ansfer effects with the environment are not considered. Current usage of this model conservatively assumes that the expansion takes place over the first five equivalent diamet ers from the break
[12]. The area ratio existing at five diameters can be calculated using the following
equation [9]: )A/T(W g G A A E M c 2 E E= (3.6-8) Where:
A = Area at 5 diameters, A E = Exit or break areas, G E = Exit mass flow rate per unit area, g c = Gravitational constant, V M = Environmental momentum s pecific volume, discussed below, T = Total thrust at the break RN 01-113 3.6-24 Reformatted July 2014 The environmental momentum specific volume, V M, is defined by Equation (3.6-2) of Section 3.6.2.2.
However, the velocity ratio used in evaluating V M is different for jet impingement than for thrust. As concluded in Reference [9], a unity velocity ratio predicts the two-phase jet forc e better than a velocity ratio, (V g/V f)1/3 , as used in the Moody thrust evaluat ion. Therefore, a unity velocity ratio is used for determination of the Moody area ratio at five diameters. For ca lculation of areas before the full expansion at five diameters is reached, an additional assumption is made, i.e., that the jet area increases uniformally fr om zero to five diameters
[12] . After the Moody area ratio is determined and the break geometry specified, the equivalent Moody angle can be determined.
The jet expansion profile calculations used by JIP
[8] are discussed for cold water, subcooled water, and water-steam and st eam conditions in the following paragraphs:
- a. Cold and Subcooled water Subcooled water is treated wi th a 10 degree half angle and a Moody expansion angle, if applicable. If the s ubcooled thrust is calculated from the Bernoulli equation, as discussed in Section 3.6.2.2, a jet expansion half angle of 10 degrees is used. However, if the thrust is calculated using the Henry-Fauske theory [10], as explained in Section 3.6.2.2, this uniform expansion is not always used. Henry-Fauske gives predictions for exit pressure and mass flow ra te. In addition, an equili brium exit quality is calculated. If the equilibrium exit qual ity is greater than or equal to 0.1% (approximately equal to a volume breakdown of 10% steam and 90%
saturated water at the ex it), the Moody expansion model, as described earlier, is used. Subsequent to jet expansion out to five diameters, the area is held constant until the 10 degree half angle is ap plicable. The overall jet geometry combining the Moody expansion profile with the 10 degree half angle is shown in Figure 3.6-19.
If the calculated exit qual ity is less than 0.1%, the uniform 10 degree half angle expansion is used. the application of this jet profile is limited, however, since t he pressure will rapidly drop to the saturation value.
- b. Water-Steam and Steam
For water-steam and steam conditions , the Moody expansion model is directly applicable. Ther efore, the jet geometry fo r water-steam, saturated steam, and superheated steam is depicted in Figure 3.6-19.
3.6-25 Reformatted July 2014 NUREG/CR-2913
[20] contains a model that has been developed for predicting two phase, water jet loadings on target
- s. The model was developed using advanced two dimensional computational techniques to solve the governing equations of mass, momentum, and energy. The application of this model results in a series of charts of the target load and pressure distributions.
These charts were developed for a wide range of pressure and temperatures that completely cover the range of intere st in pressurizer wa ter reactors. In lieu of the methodology described previously, the charts contained in
NUREG/CR-2913 are used to determine the e ffective jet pressure at varying distances from the jet origination.
- 2. Reactor Coolant Loop Piping
The methods described below are used in the design and verification of the adequacy of reactor coolant loop components and supports.
The design basis postulated pipe rupture locations for the main reactor coolant loop piping are determined us ing the criteria given in Section 3.6.2.1.2 and Reference [5]. These design basis ruptur es are used as the rupture locations for consideration of jet impingement effe cts on primary equipm ent and supports.
The dynamic analysis as discussed in Sect ion 5.2, is used to determine maximum piping displacements at each design basis rupture location. These maximum piping displacements are used to compute t he effective rupture flow area at each location. These areas and rupture orient ations are then used to determine the jet
flow patterns and to identify any primary components and supports which are potential targets for jet impingement.
The jet thrust at the point of rupture is based on the fluid pressure and temperature conditions occurring during normal (100%) st eady-state operating conditions of the plant. At the point of rupt ure, the jet force is equal an d opposite to the jet thrust.
The force of the jet is conservatively assumed to be constant throughout the jet
flow distance. The subcool ed jet is assumed to expand uniformly at a half angle of 10 degrees, from which the area of the jet at the target and the fraction of the jet intercepted by the target structure can be determined.
The shape of the target af fects the amount of mom entum change in the jet and thus affects the impingement force on the ta rget. The target shape factor is used to account for target shapes which do not deflect the flow 90 degrees away from the jet axis.
RN 03-022 3.6-26 Reformatted July 2014 The method used to compute the jet impi ngement load on a tar get is one of the following:
- a. The dynamic effect of jet impingement on the target structure is evaluated by applying a step load whose magnitude is given by:
F j = K o P o A mB RS Where:
F j = jet impingement load on target, K o = dimensionless jet thru st coefficient based on initial fluid conditions in broken loop, P o = initial system pressure, A mB = calculated maximum break flow area, R = fraction of jet intercepted by target, S = target shape factor
Discharge flow areas for limited flow area circumferential breaks are obtained
from reactor coolant loop analyses per formed to determine the axial and lateral displacements of the brok en ends as a function of time. A mB is the maximum break flow area occurring during the transient, and is calculated as
the total surface area through which the fluid must pass to emerge from the broken pipe. Using geometrical formulat ions, this surface area is determined to be a function of the pipe separat ion (axial and transverse) and the dimensions of the pipe (i nside and outside diameter).
If a simplified static analysis is perfor med instead of a dynamic analysis, the above jet load (F j) is multiplied by an appropriate dynamic load factor. For an equivalent static analysis at the target structure, t he jet impingement force is multiplied by a dynamic load factor of 2.0.
This factor assumes the target can be represented as essentially a one-degree of freedom system and the impingement force is conservatively applied as a step load.
- b. The dynamic effect of jet impingem ent is evaluated by applying the following time-dependent load to t he target structure:
F j = KPA B RS where the system pressure P is a function of time; the jet thrust coefficient K
is evaluated as a function of a syst em pressure and enthalpy, and the break flow area A B is a function of time.
3.6-27 Reformatted July 2014 3.6.2.4 Guard Pipe Assembly Design Criteria Guard pipes are used as shields as describ ed in sections 3.6.
2.1.4 and 3.
6.2.5.1 and are designed consistent wi th the moderate energy pipi ng postulated to crack.
3.6.2.5 Material to be Submitted for the Operating License Review 3.6.2.5.1 Location and Orientation of Design Basis Breaks The locations and orientations of the postulated design basis breaks are determined by stress analysis. Break locations and orientat ions for postulated reactor coolant piping branch line connection breaks are discussed bel ow. Break locations for all other high energy system piping are postulated at terminal ends accord ing to Sections 3.6.2.1.1 and 3.6.2.1.2. Break orientations for lines outsi de containment are discussed in Sections 3.6.2.1.
1 and 3.6.2.1.2.
Cracks are postulated in moder ate energy system piping in the vicinity of essential components. Components located in the vicinity of moderate energy system piping were relocated, if practical, or were specif ied as spray proof. In instances where neither relocation nor specification as spray proof is possible, shields are provided.
Due to the elimination of postulated breaks in the reactor coolant loop piping, as described in Section 3.6, jets are no longer considered in the analysis of the reactor coolant loop piping supports.
Table 3.6-2 and Figure 3.6-51 i dentify the design basis break locations and orientations for the main reactor coolant loop.
3.6.2.5.2 Restraint and Shield Location and Design Information The E-303-300 series drawings includes a su mmary of the protection of essential equipment, located outside containment, from pipe whip and jet impingement resulting from postulated rupture of high energy system piping.
Reference should be made to E-303-300 series drawings for protection of essential equipment for the plant. Pipe whip restraints associated with the reactor coolant loop are discussed in Section 5.5.14.
3.6.2.5.3 Analytical Results The analytical results show that no e ssential system or component is rendered incapable of performing its necessary func tions as a result of any postulated pipe rupture. 99-01 3.6-28 Reformatted July 2014 The methods and analysis procedures us ed to determine jet impingement loads
associated with the rupture of the main reactor coolant loop are discussed in Section 3.6.2.3.2.
3.6.2.5.4 Interface Responsibilities The interface responsibility between Westinghouse and the Balance of Plant Supplier for the design of the component supports and the reactor coolant pressure boundary, which is not part of the reacto r coolant loop is as follows:
Westinghouse
- 1. Design and evaluation of all primary equipment supports,
- 2. Design of the pressurizer surge line piping and fittings,
- 3. Analysis of all Class 1 branch lines attached to the reactor coolant loop piping (except for the pressurizer spray line) fo r deadweight, pressure, thermal, seismic, and effects of pipe rupture in the reactor coolant loop piping,
- 4. Design and analysis of All Class 1 reactor coolant loop branch line nozzles, including the pressurizer spray nozzle,
- 5. Provide the Balance of Plant supplier with loop displacements due to all loading conditions at branch line nozzle locations.
Balance of Plant Supplier
- 1. Design of all Class 1 branch lines except for pressu rizer surge line,
- 2. Analysis of all Class 1 branch lines not within Westinghouse's scope for deadweight, pressure, thermal, seismic, and effects of pi pe rupture in the reactor coolant loop piping,
- 3. Evaluation of all Class 1 br anch lines for effects of pipe rupture in auxiliary Class 1 branch lines,
- 4. Supply Westinghouse with loads at the reactor coolant loop and line branch nozzles.
3.6-29 Reformatted July 2014 3.
6.3 REFERENCES
- 1. Redfield, J. A., et al., "FLASH-2: A Fortran-IV Program for the Digital Simulation of a Multinode Reactor Plant During Loss of Coolant," WAPD-TM-666, April 1967.
- 2. Porsching, T. A., et al., "FLASH-4: A Fully Implicit Fortran-IV Program for the Digital Simulation of Transients in a R eactor Plant," WAPD-TM-840, March 1969.
- 3. Huang, Y. S., Kowal, G. M., and Savery, C. W., "Subcompartment Analysis of High-Energy Pipe Ruptures," Nuclear Technology, Vol. 27, November 1975.
- 4. Richardson, L. C., et al., "CONTEMP T A Computer Program for Predicting the Containment Pressure-Temperature Response to a Loss of Coolant Accident,"
IDO-17220, June 1976.
- 5. "Pipe Breaks for the LOCA Analysis of the Westinghouse Pr imary Coolant Loop," WCAP-8082-P-A (Proprietary) and WCAP-8172-A (Non-Proprietary), January 1975.
- 6. Takeuchi, K. et at., "MULTIFLEX, A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structu re System Dynamics, "WCAP-8708-PA, WCAP-8709-A (Non-Proprieta ry), September, 1977.
- 7. "Documentation of Sele cted Westinghouse Structural Analysis Computer Codes," WCAP-8252 Rev. 1 (Non-Pr oprietary), May 1977.
- 8. Webb, S. W., Gilbert Associates, Inc., "JIP Thrust and Jet Impingement Analysis Code," Revision 2, December 1975.
- 9. Moody, F. J., "Prediction of Blowdown and Jet Thrust Forces," ASME Paper 69-HT-31, August 1969.
- 10. Henry, R. E. and Fauske, H. K., "The Two-Phase Crit ical Flow of One-Component Mixtures in Nozzles, Orifices, and Shor t Tubes," ASME Paper 70-WA/HT-5, May 1971.
- 11. Moody, F. J., "Maximum Flow Rate of a Single Component, Two-Phase Mixture," ASME Paper 64-HT-35, February 1965.
- 12. "Design for Pipe Break E ffects," Bechtel Report BN-T OP-2, Revision 2, May 1974.
- 13. Branch Technical Position APCSB 3-1, "Protection Against Postulated Piping Failures in Fluid Systems Outside C ontainment," attached to USNRC Standard Review Plan 3.6.1, "Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Cont ainment," November 24, 1975.
RN 09-022 3.6-30 Reformatted July 2014
- 14. Branch Technical Position MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," Revision 2, June 1987, attached to
Generic Letter 87-11, "Relaxation in Arbitrary Intermediate Pipe Rupture Requirements," June 19, 1987.
- 15. Letter of A. Giambusso, USNRC, to Applicants and Licensees, "General Information Required for Consideration of the Effects of a Piping System Break
Outside Containment," December 12, 1972; attached, with Errata of January 1973, as Appendix B to BTP-APCSB 3-1 (Reference [13], above).
- 16. Modification of General Design Criter ion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Rupt ures - Final Rule (Broad Scope), 52 FR 41288, October 27, 1987.
- 17. Schmertz, J.C., Swamy, S. A., and Lee Y. S., "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Virgil C. Summer Nuclear Power Pl ant," Westinghouse Report WCAP-13206, April 1992.
- 18. Nuclear Regulatory Commission Safety Evaluation Report, "Request to Use Leak-Before-Break Technology for Reactor Coolant System Piping - Virgil C.
Summer Nuclear Station, Unit No. 1 (TAC No. M83971)," January 11, 1993.
- 19. D. C. Bhowmick, J. F. Petsche, and S. A. Swamy, "Technical Justification For Eliminating Large Primary Loop Pipe Rupture As The Structural Design Basis For The Virgil C. Summer Nuclear Power Plant," Westinghouse Report WCAP-13206, Revision 1, June 2000.
- 20. NUREG/CR-2913, "Two-Phase Jet Loads," G. G. Weigand, S. L. Thompson, and L. Tomasko, Sandia National Laboratories, January 1983.
02-01 RN 03-022 RN 02-016 3.6-31Reformatted PerAmendment 02-01TABLE 3.6-0INTERMEDIATE BUILDING - WEST PENETRATION ACCESS AREANODALIZATION MODELControl VolumeNumberControl VolumeIdentificationVolume(ft 3)Area(ft 2)Height(ft)Bottom(ft)1Intermediate Building134,0005570 24 436 2Intermediate Building50,0002070 24 436 3Intermediate Building168,0007650 22 412 4Environment1. + 81. + 7 10 458 5East PenetrationAccess Area72,2002887 25 436 6 West PenetrationAccess Area63,5002900 22 412 JunctionNumbe r FromC.V.ToC.V.Elevation(ft)Flow Area(ft 2)Inertia(ft-1)Friction(K)1 1 2 448 550.013 1.5 2 1 3 448 525.0074 1.5 3 1 4 448 350.00787 1.5 4 2 4 448 350.0079 1.5 5 2 4 448 350.0079 1.5 6 2 4 448 350.0079 1.5 7 2 4 448 350.0079 1.5 8 5 1 448 12.5.25 1.5 9 5 1 448 335.096 1.5 10 5 6 448 50.048 1.5 3.6-32Reformatted PerAmendment 02-01TABLE 3.6-0aINTERMEDIATE BUILDING - EAST PENETRATION ACCESS AREANODALIZATION MODELControl VolumeNumberControl VolumeIdentificationVolume(ft 3)Area(ft 3)Height(ft)Bottom(ft)1Intermediate Building134,0005570 24 436 2Intermediate Building50,0002070 24 436 3Intermediate Building168,0007650 22 412 4Environment1. + 81. + 7 10 458 5East PenetrationAccess Area63,4002643 24 436 6 East PenetrationAccess Area58,1002643 22 412 JunctionNumbe r FromC.V.ToC.V.Elevation(ft)Flow Area(ft 2)Inertia(ft-1)Friction(K)1 1 2 448 550.013 1.5 2 1 3 448 525.0074 1.5 3 1 4 448 350.00787 1.5 4 2 4 448 350.0079 1.5 5 2 4 448 350.0079 1.5 6 2 4 448 350.0079 1.5 7 2 4 448 350.0079 1.5 8 5 1 448 12.23 1.5 9 5 1 448 410.067 1.5 10 5 6 448 45.05 1.5 11 6 3 430 11.24 1.5 02-01 02-01 02-01 3.6-33Reformatted PerAmendment 02-01TABLE 3.6-0bINTERMEDIATE BUILDING - PENETRATION ACCESS AREASMASS/ENERGY RELEASE DATATime (sec)Mass Release Rate (lbm/sec)Enthalpy (Btu/lbm) 0.02.202 + 41186.003751.586 + 41198.01131.079 + 41203.02389.592 + 31204.04631.184 + 41202.1089.358 + 31204.4137.009 + 31203.8758.312 + 31203 1.957.838 + 31203 3.6-34Reformatted PerAmendment 02-01TABLE 3.6-0cINTERMEDIATE BUILDING - SUBCOMPARTMENT CONTAININGMAIN STEAM AND FEEDWATER HEADERSNODALIZATION MODELControlVolumeNumberControl VolumeIdentificationVolume(ft 3)Area(ft 2)Height(ft)Bottom(ft)1Intermediate Building134,0005570 24 436 2Intermediate Building50,0002070 24 436 3Intermediate Building168,0007650 22 412 4Environment1. + 81. + 7 10 458JunctionNumbe r FromC.V.ToC.V.Elevation(ft)Flow Area(ft 2)Inertia(ft-1)Friction(K)1 1 2 448 550.013 1.5 2 1 3 448 525.0074 1.5 3 1 4 448 350.00787 1.5 4 2 4 448 372.0079 1.5 02-01 02-01 02-01 3.6-35Reformatted PerAmendment 02-01TABLE 3.6-0dINTERMEDIATE BUILDING - OUTSIDE SUBCOMPARTMENTCONTAINING MAIN STEAM AND FEEDWATER HEADERSNODALIZATION MODELControl VolumeNumberControl VolumeIdentificationVolume(ft 3)Area(ft 2)Height(ft)Bottom(ft)1Intermediate Building134,0005570 24 436 2Intermediate Building60,5002520 24 436 3Intermediate Building168,0007650 22 412 4Environment1. + 81. + 7 10 458JunctionNumbe r FromC.V.ToC.V.Elevation(ft)Flow Area(ft 2)Inertia(ft-1)Friction(K)1 1 2 448 580.0107 1.5 2 1 3 448 525.0074 1.5 3 1 4 448 350.00787 1.5 4 2 4 4481372.0060 1.5 02-01 02-01 3.6-36Reformatted PerAmendment 02-01TABLE 3.6-0eINTERMEDIATE BUILDING - MAIN STEAM AND FEEDWATER HEADERSINSIDE AND OUTSIDE HEADER SUBCOMPARTMENTMASS/ENERGY RELEASE DATATime (sec)Mass Release Rate (lbm/sec)Enthalpy (Btu/lbm) 0.02.202 + 41186.006251.900 + 41193.01631.538 + 41198.03131.354 + 41200.04131.409 + 41199.05631.555 + 41197.1881.404 + 41198.3631.248 + 41200.4881.151 + 41201.7251.034 + 41201.9759.880 + 31201 1.439.232 + 31202 3.6-37Reformatted PerAmendment 02-01TABLE 3.6-0fTURBINE BUILDING NODALIZATION MODELVolume, ft 33 x 10 6Initial Temperature, F 80Initial Relative Humidity, %
60Relief Area, ft 2 570Discharge Coefficient 0.6 3.6-38Reformatted PerAmendment 02-01TABLE 3.6-0gTURBINE BUILDINGMASS/ENERGY RELEASE DATATime (sec)Mass Release Rate (lbm/sec)Enthalpy (Btu/lbm) 0.01.65 + 41197.0051.65 + 41197.041.06 + 41203.19.71 + 31204.29.09 + 31204.48.32 + 31203.67.52 + 31203.87.15 + 31203 1.6.87 + 31203 1.66.55 + 31203 2.16.36 + 31203 3.11.04 + 4 717 5.11.57 + 4 559 7.11.63 + 4 574 10.11.57 + 4 594 15.08.59 + 3 623 20.05.36 + 3 675 3.6-39Reformatted PerAmendment 02-01TABLE 3.6-0hINTERMEDIATE AND TURBINE BUILDING SUBCOMPARTMENTPRESSURE RESPONSE CROSS REFERENCEAnalysisApplicableControl VolumesFSAR FigureWest Penetration Access Area 5; 63.6-25-63.6-35-1; 5-33.6-41-6; 3-63.6-5East Penetration Access Area 5; 63.6-65-63.6-75-1; 5-33.6-81-6; 3-63.6-9Intermediate Building (Inside compartmentcontaining main steam and feedwater headers) 2; 12-12-3; 2-1 3.6-103.6-13 3.6-15Intermediate Building (Outside compartment containing main steam and feedwater headers) 11-31-4; 1-2 3.6-113.6-123.6-14Turbine Building (Loading on IntermediateBuilding wall)13.6-16 02-01 3.6-40Reformatted PerAmendment 02-01TABLE 3.6-0iINTERMEDIATE AND TURBINE BUILDING DIFFERENTIAL PRESSURESSubcompartmentCalculated Pressure (psi)Design Pressure (psi)(Including Dynamics Effect)West Penetration Access Area 5.0 6.5East Penetration Access Area 4.8 5.9Intermediate Building (Insidecompartment main steam andfeedwater headers) 3.0 4.0Intermediate Building (Outside compartment containing mainsteam and feedwater headers) 2.2 4.0Turbine Building (Loading on intermediate building wall) 2.3 4.0 3.6-41Reformatted PerAmendment 99-01TABLE 3.6-1ENVIRONMENTAL CONDITIONS RESULTING FROM POSTULATEDPIPE RUPTURES OUTSIDE OF CONTAINMENTLarge Line Break Conditions (1)AreaTemperature (F)Relative Humidity
(%)East and West Penetration Access Areas,Floor Elevation 412' 220 100East and West Penetration Access Areas, Floor Elevation 436' 220 (2)100West Penetration Access Area, Floor Elevation 463' 220 100Intermediate Building, Floor Elevation 412' 212 100Intermediate Building, Floor Elevation 436' 212 (3)100Small Line Break Conditions (4)AreaTemperature (F)Relative Humidity
(%)East and West Penetration Access Areas and Intermediate Building, All Elevations 200 100 (1)Conditions last for 3 minutes, then return to 100F in about 30 minutes.(2)Temperature of 320F for 30 seconds.(3)Temperature of 283F for 4 seconds
.(4)Conditions last for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, then return to 100F in about 30 minutes.NOTE:Details of the main steam line break HELB/SBOC (High Energy Line Break /Superheated Blowdown Outside Containment) environmental conditions areprovided in Section 3.11.2.2.2.2.
99-01 3.6-42Reformatted PerAmendment 02-01TABLE 3.6-2POSTULATED BREAK LOCATIONS FOR THE LOCA ANALYSIS OF THE PRIMARY COOLANT LOOP (1) (3)Location of Postulated RuptureTypeBreak Opening Area (2) 1.Residual Heat Removal (RHR)Line/Primary Coolant LoopConnectionGuillotine (viewed from the RHR line)Cross-sectional flow area of the RHR line 2.Accumulator Line/Primary CoolantLoop ConnectionGuillotine (viewed from the accumulator line)Cross-sectional flow area of the accumulator line 3.Pressurizer Surge Line/Primary Coolant Loop ConnectionGuillotine (viewed from the pressurizer surge line)Cross-sectional flow area of the pressurizer surge line (1)Refer to Figure 3.6-51 for location of postulated breaks in reactor coolant loop.(2)Less break opening area will be used if justified by analysis, experiments or considerations of physical restraintssuch as concrete walls or structural steel.(3)Elimination of the dynamic effects of postulated pipe ruptures in the reactor coolant loop piping at the terminal ends,steam generator inlet elbows, and crossover leg closure welds have been eliminated as allowed by the revisedGeneral Design Criterion 4 (Section 3.6).
02-01
Amendment 02-01May 2002Figures 3.6-20 through 3.6-50 were deletedby Amendment 1 dated August, 1985(RN 01-071)
3.7-1 Reformatted July 2014 3.7 SEISMIC DESIGN In addition to the steady-state loads impos ed on the system under normal operating conditions, the design of equipment, equipment supports, and Seismic Category I structures requires that consideration also be given to abnormal loading conditions such as earthquakes. Seismic loading, are considered for earthquakes of two magnitudes: safe shutdown earthquake (SSE) and operating basis earthquake (OBE). The SSE is defined as the maximum vibratory ground motion at the plant site that can reasonably
be predicted from geologic and seismic eviden ce. The OBE is that earthquake which, considering the local geology and seismolo gy, can be reasonably expected to occur during the plant life.
Aside from these two major earthquakes, the effects of reservoir induced seismicity were reviewed in a separate study which is summarized in Section 3.7.1.5. Specific details on input, analysis methods and procedures, etc. for reservoir induced seismicity are not included within Section 3.7 but are contained in Reference [26].
For the OBE loading condition, the Nuclear Steam Supply System and safety-related balance of plant equipment and st ructures are designed to be capable of continued safe operation. The design for the SSE is intended to assure:
- 1. That the integrity of the reactor coolant pressure boundary is not compromised;
- 2. That the capability to shutdow n the reactor and maintain it in a safe condition is not compromised; and
- 3. That the capability to pr event or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100.11 and 10 CFR 50.67 is not compromised.
It is necessary to ensure that required Seismic Category I st ructures, systems, and components do not lose their capability to per form their safety function. Not all components have the same functional safety requirements. For example, a charging pump must retain its capability to function normally during the SSE. Therefore, the deformation in the pump must be restricted to appropriate limits to a ssure its ability to function. On the other hand, many com ponents can experience significant permanent deformation without loss of func tion. Piping and vessels ar e examples of the latter where the principal requirement is that they retain thei r contents and allow fluid flow.
RN 12-034 3.7-2 Reformatted July 2014 This Section is presented in the following subsections:
3.7.1 Seismic Input
3.7.2 Seismic System Analysis
3.7.3 Seismic Subsystem Analysis
3.7.4 Seismic Instru mentation Program
3.7.5 Seismic Design Control
The seismic requirements for safety-related instrumentation and electrical equipment are covered in Section 3.10. The safety class definitions and classifications are given in Section 3.2.
3.7.1 SEISMIC INPUT 3.7.1.1 Design Response Spectra Design response spectra for the horizontal component of the SSE, formerly the design basis earthquake, and OBE are applied in the design of Seismic Category I structures, systems, and components according to preliminary results of studies conducted by N.
M. Newmark and J. A. Blume, consultants to the AEC. These studies were performed prior to and the response spectra were dev eloped before issuance of Regulatory Guide 1.60. Regulatory Guide 1.60 is discussed in Appendix 3A.
Separate design response spectra are specif ied at rock and soil foundation elevations. Horizontal spectra for stru ctures founded upon rock like medi a are presented as Figures 3.7-1 and 3.7-2 for the SSE and OBE, respectively. Design response spectra for structures founded upon soil ar e presented as Figures 3.7-3 and 3.7-4 for the SSE and OBE, respectively.
The vertical component spectr a used are two thirds of the horizontal components in all frequency ranges and occur simultaneously.
The maximum horizontal ground acceleration for the SSE is 0.15g at the competent rock foundation elevation and 0.25g for the so il foundation. For t he OBE, the maximum horizontal ground accelerations used are 0.1g (rock) and 0.15g (soil). Seismological and geological background data pertaining to the plant site are presented in Section 2.5.
The design response spectra are similar to the previous NRC minimum criteria (see Figure 17 of Reference [20]). The differ ences between the design response spectra and the current NRC recommended spectra of Regulatory Guide 1.60 have been investigated for all damping values considered. It was found that the design response spectra are, in general, enveloped by the recommended spectra throughout the entire frequency range and are apparently less conservati ve in the range of frequencies below 3.7-3 Reformatted July 2014 about 0.55 Hz. In the latter region, the di fference is of no concern, in general, for Seismic Category I structures , systems, and components.
For the region from about 2 to 9 Hz, where the predominant frequencies of the Seismic Category I structures lie, the differences between the spectra are relatively small. Specifically, in the frequency range from about 4.6 to 6.8 Hz, the tw o sets of spectra are practica lly the same for the 2, 5, and 7% curves. In the frequency ranges from about 2 to 4.6 Hz and from about 6.8 to 9 Hz, the maximum differences with respect to the design spectra are about 22, 20, 18, 18, and 23%, respectively, for 0.5, 2, 5, 7, and 10% critical damping values. In the ranges of frequencies from about 0.55 to 2 Hz and from about 9 to 33 Hz, the differences between the spectra are again rela tively small, except possibly for the spectrum curves of 7 and 10% critical damping.
The differences between the two sets of response spectra stated above might result in an underestimate of the responses of Seismi c Category I structures should the damping values of Regulatory Guide 1.61 be used. Howe ver, with the use of more conservative (i.e., lower) damping values than those of Regulatory Guide 1.61, the two sets of spectra would lead to very nearly the same results, as is shown in Section 3.7.1.3.
As indicated in Section 3.7.1.
4, some of the Seismic Catego ry I structures are founded upon competent soils. Others are founded upon rock by means of caissons or lean
concrete. The foundation elevation for each of the Seismic Category I structures is given in Tables 3.7-5 and 3.7-6. In the seismic analyses, with soil structure interaction effects included, the design spectra are applied at the various foundation locations of Seismic Category I structures. For the Reactor Building, Auxiliary Building, and Control Building, which are founded upon lean concrete bearing on rock, the design spectra in Figures 3.7-1 and 3.7-2 are applied in the re spective lumped-mass-spring soil-structure interaction system at the foundation, as listed in Table 3.7-5. For the Diesel Generator Building, Fuel Handling Building, and Intermediate Building, which are founded upon caissons, the effective design spectra are applied throughout the elevation of the caissons in the lumped-mass-spring soil-structure interaction systems. Such effective design spectra are determined based upon equivalent energy input accounting for the design spectra at rock elevation (Figures 3.7-1 and 3.7-2) and the design spectra at soil elevation (Figures 3.7-3 and 3.7- 4), as described in Section 3.7.2.4. For the Service Water Pumphouse, which is founded on competent soils, the design spectra (Figures 3.7-1 and 3.7-2) are applied at the base rock elevation, at 350 feet, in the finite element soil-structure interaction system (see Section 3.7.2.1.1).
Since no structural response amplification was expected from the servic e water intake structure and service water discharge structure, design spectra were not used for these two st ructures. Instead, effective seismic loads, based upon the maximum ground accelerations, were
conservatively used for design as is descr ibed in Section 3.8.
4.4.8 and 3.8.4.4.9, respectively.
3.7-4 Reformatted July 2014 3.7.1.2 Design Time History A synthetic earthquake ground motion time history was developed to simulate the SSE
and OBE, based upon the design response spectr a discussed in Section 3.7.1.1, for the time history analysis of structures, components, and equipment.
The time history is compatible with the corresponding response spectra. The "valleys" in the unsmoothed response spectra generated from the time history do not fall below the corresponding smoothed response spectra presented as Figures 3.7-1 through 3.7-4 for
rock and soil foundation elevations. Figures 3.7-5 through 3.7-8 present comparisons between the smoothed response spectra and spectra derived from the synthetic earthquake time history for 2%, 5%, 7%, and 10% of critical damping, respectively.
The synthetic earthquake time history was deve loped with an equal time interval of 0.01 second. The period intervals at which t he unsmoothed response spectra values were calculated ranged from 0.001 sec ond to 0.2 second. A total of 108 points were used in generating each spectrum. The cl osest period intervals were used in the most critical
portions of each spectrum. In Table 3.7-0, the frequency intervals used for calculating the unsmoothed response spectra (see Figures 3.7-5 and 3.7-8) fr om the synthetic earthquake time history are given.
3.7.1.3 Critical Damping Values 3.7.1.3.1 Balance of Plant Scope The specific percentage of critical damping values used for Seismic Category I structures, systems, and com ponents are presented in Tabl e 3.7-1 and 3.7-2. The values were obtained from Reference [1]. T hese values are equivalent to or provide less damping than the current values in Regula tory Guide 1.61 (see Appendix 3A). For
example, the damping values provided in Tabl e 3.7-1 for reinforced concrete structures are 2 and 5% critical for stress levels no mo re than about half the yi eld point and at or just below the yield point, respectively.
The corresponding values in Regulatory Guide 1.61 are 4 and 7% critical for OBE and SSE, respectively. The damping values for stress levels beyond the yield point, as shown in Table 3.7-1, were not used in the analysis. A comparison of the design spec trum of 2% (Section 3.7.1.1) and the recommended spectrum of 4% in Regulatory Guide 1.60 indicates that, except in the range of frequencies below about 0.55 Hz, the design spectrum envelops the
recommended spectrum with significant margin , especially in the frequency range from about 0.55 to 9 Hz. The 5% design spec trum is also compared with the 7%
recommended spectrum with similar conclusion obtained.
In the case of prestressed concrete structures, such as the Reactor Building shell, the corresponding damping values required in Table 3.7-1 are equiva lent to those in Regulatory Guide 1.61 which are 2 and 5% cr itical for OBE and SSE respectively. In actual design of the Reactor Building shell, however, a damping value of 2% was used for both OBE and SSE. This, coupled with the only slight unconserva tism of the design 3.7-5 Reformatted July 2014 spectra in comparison to the recommended spectra (see Section 3.7.1.1), suffices to assure the adequacy and conservatism of the structural responses under both OBE and SSE.
Based upon the above, the use of more conservative damping values, together with the design spectra, has provided a measure of t he seismic responses of Seismic Category I structures, systems, and components which is equivalent to or more conservative than would be provided by using the damping values in Regulatory Guide 1.61 combined
with the recommended spectra in Regulatory Guide 1.60.
For the soil-structure interaction system, the hysteretic damping values for the foundation media were taken to be 5% for the rocking mode and 10% for the swaying mode, regardless of the type of foundation media and frequency range. For conservatism, the radiation (viscous) dampi ng due to wave propagati on from a structure into the half-space, which is frequency dependent, was not considered.
In a typical classical modal analysis, different damping has been specified for each mode of the vibratory system. The method of weighted modal damping as described in Reference [2] was used to calculate the damping for each mode in the dynamic analysis
of the soil-structur e interaction system.
3.7.1.3.2 Components and Equipment Provided by the NSSS Vendor The damping values given in Table 3.7-3 are used for the systems analysis of Westinghouse equipment. These are c onsistent with the damping values
recommended in Regulatory Guide 1.61 except in the case of the primary coolant loop system components and large piping (excluding reactor pressure vessel internals) for which the damping values of 2% and 4% are used as established in testing programs reported in Reference [3]. The damping va lues for control rod drive mechanisms (CRDM) and the fuel assemblies of the Nu clear Steam Supply System are in conformance with the values for welded and/o r bolted steel structures as listed in Regulatory Guide 1.61.
Tests on fuel assembly bundles justified conservative component damping values of 7%
for OBE and 10% for SSE to be used in the fuel assembly component qualification.
Documentation of the fuel assembly tests is presented in Reference [4].
The damping values used in component analysi s of CRDM and their seismic supports were developed by testing programs performed by Westinghouse. These tests were performed during the design of the CRDM support; the support was designed so that the damping in Table 3.7-3 could be conservatively used in the seismic analysis. The CRDM support system is designed with plates at the top of the mechanism and gaps between mechanisms. These are encircled by a box section frame which is attached by
tie-rods to the refueling cavity wall.
The test conducted was on a full size CRDM complete with rod position indicator coils, attachment to a simulated vessel head, and
variable gap between the top of the pressure housing s upport plate and a rigid bumper 3.7-6 Reformatted July 2014 representing the support. The internal pressure of the CRDM was 2250 psi and the temperature on the outside of the pressure housing was 400°F.
The program consisted of trans ient vibration tests in wh ich the CRDM was deflected a specified initial amount and suddenly released. A logarithmic decrement analysis of the
decaying transient provides the effective damping of the assembly. The effect on damping of variations in the drive shaft ax ial position, upper se ismic support clearance, and initial deflection amplitude was investigated.
The upper support clearance had the larges t effect on the CRDM damping with the damping increasing with increasing clearance. With an upper clearance of 0.06 inches, the measured damping was approximately 8%.
The clearances in a typical upper seismic CRDM support are a minimum of 0.10 inches. The increasing damping with
increasing clearances trend from the test results indicated that the damping would be greater than 8% for both the OBE and the SSE based on a co mparison between typical deflections during these seismi c events to the initial deflect ions of the mechanisms in the test. Component damping values of 5% ar e, therefore, conser vative for both OBE and SSE.
These damping values are used and applied to CRDM component analyses by response spectra techniques. Where time history analyses are used, damping in the pressure housing is set at the levels stated in Table 3.7-3 for welded steel structures and impact damping is applied at the contact points in the upper seismic supports.
3.7.1.4 Supporting Media for Se ismic Category I Structures The Seismic Category I service water intake structure, Service Water Pumphouse, and service water discharge structure are suppor ted by mats founded upon competent soil.
Other Seismic Category I structures are founded upon rock by means of caissons or mats. For the Reactor Buildi ng, Auxiliary Building, and C ontrol Building, soils below planned foundation grade have been excavat ed and replaced with lean concrete bearing on rock.
Dynamic engineering properties of the soil and bedrock at the site have been evaluated for use in foundation interaction analyses. T hese properties, present ed in Table 3.7-4, were developed using field geophysical and geologic data and static and dynamic
laboratory test data. These values are appropriate for design at strain levels corresponding to the SSE and OBE. The shear module and subgrade module
developed from laboratory and geophysical data have been reduced using factors related to rock quality designation (RQD).
More detailed information concerning the foundation media is prov ided in Section 2.5.
Descriptions of the supporting media for Seismic Category I structures are presented in Table 3.7-5.
3.7-7 Reformatted July 2014 The foundation elevation and the foundation type for the Seis mic Category I structures are presented in Table 3.7-5.
Foundation material data fo r those structures supported on caissons is presented in Table 3.7-6.
Cross sections illustrating building foundation elevations, general subsurface conditions, and surface topography are pres ented in Figures 2.5-77 thr ough 2.5-82. The locations of the sections are shown on Figures 2.5-46 and 2.5-47. The approximate dimensions, including height and top elevation, as well as the embedment depth of the Seismic
Category I structures, as previously discussed, are presented in Table 3.7-7.
3.7.1.5 Effects of Reservoir Induced Seismicity In addition to the seismic design data and procedures for the normal tectonic
earthquake as presented within Section 3.7 herein, the effects of Reservoir Induced Seismicity were fully investigated in a separate study completed in 1983.
Shortly after filling of Monticello Reservoir in late 1977, a series of small earthquakes began occurring in the vicinity of the Virgil C.
Summer Nuclear Station. These events, termed Reservoir Induced Seismicity (RIS), are still continuing to date but much less frequently than during the initial two years after filling the reservoir. The largest of the RIS events which have occu rred are local magnitude M L 2.8 microearthquakes and are characterized by free field motions of ve ry short duration, high frequency, and high peak
accelerations.
As a result of this RI S phenomena, both the Adviso ry Committee on Reactor Safeguards (ACRS) and the Atomic Safety and Licensing Board (ASLB) expressed concerns on the impact that these small earthquakes would have on plant equipment and components required for the shutdown and continued re moval of residual heat.
The ACRS concern was with the largest pos tulated earthquake which might occur from
Reservoir Induced Seismicity. The ASLB concern dealt with the shallow RIS events recorded at Monticello Reservoir. To address the ACRS and ASLB concerns, programs were set up to develop and establish envelope spectra for both the ACRS and ASLB
type of events.
For the ACRS it was determined that t he earthquake could be adequately represented by a 4.5 M L event of normal tectonic depth anchored to a zero period acceleration of .22g. This response spectrum was considered by the NRC as conservative for the necessary evaluation. Comparing the ACRS response spectrum with the SSE response spectrum and considering the domi nant frequencies of the structures, it was concluded that only buildings on rock had to be evaluated for the ACRS commitment.
Buildings on rock include the Reactor Building, Auxiliary Building, Control Building, and Intermediate Building. The Inte rmediate Building, although on caissons, is so restrained that it was considered as a building on rock. Four Oroville aftershocks of magnitude 4.5 were used at nine varied time increment s to generate 36 separ ate time history components for the ACRS. These components were applied to the seismic models of
the appropriate plant structures to generate the necessary floor response spectra. This 3.7-8 Reformatted July 2014 floor response spectra was required for use in the subsequent ACRS equipment margin study for safe shutdown equipment and components.
For the ASLB program, the envelope spectra for RIS events, as recorded by the USGS accelerograph located near the site, was used as a basis to further determine the response spectra which should be applied to t he specific foundation ty pes of the nuclear plant structures (i.e.
rock, caissons, soil). The develop ment of this specific response spectra included a program of explosion test experiments and resulted in a reduced
envelope spectra for the three individual fo undation types on rock, caissons, or soil.
This reduced enveloped spectra, as approv ed by the NRC, was compared to the SSE response spectra at the foundatio n level. It was concluded that for buildings on rock, the SSE foundation response spectra exceeded the ASLB reduced envelope spectra, except for the frequency region hi gher than 16 Hz. For buildings on caissons or on soil, the SSE response spectra exceeded the ASLB reduced response spectra essentially at all frequencies. Therefore only the structures on rock were required to be evaluated for the ASLB criteria.
The ASLB reduced envelope spectra was applied to the seismic models of the four buildings on rock to generate floor respons e spectra curves. These floor response spectra were the bases for performing the subsequent ASLB equipment margin study
for the safe shutdown equipment and components.
Because of the similarities of the ACRS and the ASLB review procedures, it was determined that the equipment margin study could be concurrently performed for both the ACRS and ASLB criteria. This study was divided into two categories; one for active components and one for passive components.
Active components are defined as those com ponents which require actuation or moving
parts to perform their func tion for shutdown and residual heat removal. Passive components are defined as those components which do not require actuation or moving
parts to perform this same function, but are requi red to maintain their structural integrity.
For the active equipment margin study, separate design margins were determined for both the ACRS and ASLB. The active component s which were originally qualified by tests were evaluated for margin by comparison of the ACRS and ASLB floor response spectra and the original test response spectra (TRS).
For those active components originally qual ified by analyses, the original design spectrum or ZPA value was compared to the ACRS and ASLB floor response spectra to
determine the resulting margin. Results of the active equipment margin study showed that all active equipment and components possess ed sufficient margin in original design to equal or exceed that required by the ACRS and ASLB criteria.
Passive components were generically addre ssed and qualified by using inelastic response spectra and ductility demand criteria as a more appropriate and accurate 3.7-9 Reformatted July 2014 technique of measuring real damage potential. Results of the analysis for passive components showed that the SSE is a relative ly more severe design criterion than either the ACRS and ASLB spectrum. Maxi mum ductility demands were well below the minimum capacity of the passive components, proving that neither the ACRS or the ASLB criteria pose a significant seismic risk to the passive component s in the station.
Thus the equipment margin studies showed that both active and pa ssive equipment and components possessed adequate design margins to withstand the affects of any
imposed reservoir induced seismicity.
The overall conclusion of the program is t hat the ACRS and ASLB cr iteria do not affect the safe shutdown and residual heat removal equipment in the plant and that no plant modifications are required. Based on these positive results, both the ACRS requirement and the ASLB License Condition 2.C
[25] are satisfied with no further actions required.
Detailed input, data, and conclusions for the ACRS and ASLB seismic programs are presented in the report, "Seismic Confirmato ry Program, Equipment Margin Study, V. C.
Summer Nuclear Station Unit 1, OL No.
NPF-12, November 1983". (Reference [26])
3.7.2 SEISMIC SYSTEM ANALYSIS 3.7.2.1 Seismic Analysis Methods 3.7.2.1.1 Balance of Plant Scope Seismic analyses of the Seismic Category I structures are perform ed using the method of normal mode. The design response spectra or synthetic time hi story, described in Section 3.7.1.2, are us ed as the input motion.
For the Reactor Building, includin g the interior concrete structure, the flexibility matrix of the structure system is firs t formulated using the STRUDL
[5] computer program (see Section 3.8.4.4). This flexib ility matrix is then combi ned with the mass matrix of the structure system to solve for the natural frequencies and mode shapes by the Jacobian diagonalization routine.
Each individual modal response is obtained by the numerical integration method of Nigam and Jennings
[6]. The total structural response is taken as the superposition of the responses of all significant modes. Sign ificant modes for structural response are defined using the dominant m odal participation factor as a base. Any mode with a modal participation factor gr eater than 10% of the domin ant mode is considered a significant mode. Since all other modes lie within frequencies well above the rigid frequency and have small participation factors with response spectrum values much
less than that of the domi nant mode, they contribute less than 10% of the total response. For example, the participation factors of the R eactor Building, as shown in Table 3.7-7c, indicate that the significant modes are first through seventh mode. The 3.7-10 Reformatted July 2014 frequency for the seventh mode is 45.87 cps.
beyond which there is very little energy input in the time history.
Floor response spectra are calculated from the resulting floor acceleration time history using the method of Nigam and Jennings
[6].
For the remaining Seismic Category I structures, except the service water intake structure, Service Water Pumphouse, and service water discharge structure, the seismic analysis is performed using the computer program DYNAL
[7] (see Section 3.8.4.4). The Householder-O rtega-Wielandt method, described in Reference [7], is used in the modal analysis to obtain the natural frequencies and mode shapes for the structure system. Struct ural responses are obtained by superposition of modal responses of all significant modes. The time histories of structural responses at mass points of interest are used to generate the floor response spectra.
Because of the embedment condition and the underlying foundation char acteristics, the Service Water Pumphouse and associated supporting foundation m edia are analyzed for the soil-structure interaction effect (see Section 3.7.2.4) using the computer program FLUSH[8] (see Section 3.8.4.4). This program uses the plane-strain finite element method and is capable of taking into account the three-dimensional characteristics of the soil-structure system by use of viscous boundaries on planar sides of the plane-strain model. Additional features of FLUSH are discussed in Reference [8] and Section 3.7.2.4. The analytical procedure used is two-dimensional. Material damping is included in stiffness matrices formed from complex modules of foundation media. The structural response is obtained by applying the method of complex response. This includes solving the displacements in the frequency domain through Gaussian elimination, followed by the inverse Fast Fourier Transform for the displacements in the time domain.
Typical lumped mass mathematical models used for seismic analysis are shown by
Figure 3.7-9 for the Reactor Building wh ich is founded upon rock.
Figure 3.7-10 depicts a typical lumped mass mathematical model for the Intermediate Building which is supported by caissons seated in rock. The finite element model of the Service Water Pumphouse and its supporting medium is presented by Figure 3.7-11.
The mass points of a building are, as a rule, chosen at the major floor elevations of the building. In the case of the Reactor Buildi ng interior concrete st ructure, typical mass points are also selected at t he supports of safety class equ ipment, such as the reactor vessel, steam generators, pressurizer, and reactor coolant pumps. The number of degrees of freedom is the same as the total number of modes. Since, for the Virgil C. Summer Nuclear Station Seismic Category I st ructure, the total number of modes is always greater than the number of significant modes and the significant modes
contribute to more than 90% of the total response, the number of degrees of freedom selected is adequate.
3.7-11 Reformatted July 2014 For Seismic Category I structures founded upon rock (see Figure 3.7-9), the effects of foundation torsion, rocking, and translation with respect to the supporting media are represented by the effective linear foundation springs derived from the vibration of the rigid disc sitting upon an elastic half- space as described in Sect ion 3.7.2.4. The stiffness members between floors are connec ted at stiffness c enters, and the mass center and the stiffness center of each floor are connected by a rigid link to simulate torsional effects. The torsi onal spring is also attached to the bottom of the mat at the same point as the rocking and translational springs. The induced torsions, transverse shears, and bending moments at every section of the model are us ed in the design of the buildings. A typical torsional model is illustrated by Figure 3.7-11a. The torsional response is presented by Table 3.7-7b.
For Seismic Category I structures support ed on caissons (see Figure 3.7-10), the effective stiffness of the surrounding soil medi um is represented by lateral soil springs attached to the caissons in addition to the previously discussed foundation springs. The
lateral soil springs are calculated in accordance with the method proposed by Penzien[9] (see Section 3.7.2.4).
The concrete structure element is a ssumed to have linear elastic properties.
3.7.2.1.2 Components and Equipment Provided by the NSSS Vendor Those Seismic Category I components and systems that must remain functional in the event of the SSE are identified by applying the criteria of Section 3.2.1.
In general, the dynamic analyses are performed using a modal analysis plus either the
response spectrum analysis or integrat ion of the uncoupled modal equations as described in Sections 3.7.2.1.2.
3 and 3.7.2.1.2.4 respectively, or by direct integration of the coupled differential equations of mo tion described in Section 3.7.2.1.2.5.
3.7.2.1.2.1 Dynamic Analysi s - Mathematical Model The first step in any dynamic analysis is to model the structure or component, i.e., convert the real structure or component in to a system of masses, springs, and dashpots suitable for mathematical analysis. The essence of this step is to select a model so that the displacements obtained will be a good repres entation of the moti on of the structure or component. Stated differently , the true inertia forces should not be altered so as to appreciably affect the internal stresses in the structure or com ponent. Some typical
modeling techniques are pr esented in Reference [10].
Equations of motion consider the multi-degree-of-freedom system shown in Figure 3.7-12. Making a force balance on each mass point r, the equations of motion can be written in the form:
0ukucym i irii i rirr=++ (3.7-1) RN 01-113 3.7-12 Reformatted July 2014 Where:
m r = the value of the mass or mass moment of rotational inertia at mass point r.
y r = absolute translational or angular acceleration of mass point r.
c ri = damping coefficient - external force or moment required at mass point r to produce a unit translational or angular veloci ty at mass point i, maintaining zero translational or angular velocity at all other mass points. Fo rce or moment is positive in the direction of positiv e translational or angular velocity.
i u = translational or angular velocity of mass point i relative to the base.
k ri = stiffness coefficient - t he external force (moment) r equired at mass point r to produce a unit deflection (rotation) at mass point i, maintaining zero displacement (rotation) at all other mass points.
Force (moment) is positive in the di rection of the displacement (rotation).
u i = displacement (rotation) of mass point i relative to the base.
Note that Figure 3.7-12 does not attempt to show all of the sp rings (and none of the dashpots) which are represent ed in Equation (3.7-1).
Since:
srryuy+= (3.7-2) Where:
y s = absolute translational (angul ar) acceleration of the base.
u r = translational (angular) acceleration of mass point r relative to the base.
Equation (3.7-1) can be written as:
sr iiri iirirrymukucum=++ (3.7-3) For a single degree of freedom system with displacement u, ma ss m, damping c, and stiffness k, the correspondi ng equation of motion is:
mucukumy s.....++= (3.7-4) RN 01-113RN 01-113 3.7-13 Reformatted July 2014 3.7.2.1.2.2 Modal Analysis 1. Natural Frequencies and Mode Shapes
The first step in the modal analysis method is to establish the normal modes, which
were determined by eigen solution of Equat ion (3.7-3). The right hand side and the damping term are set equal to zero for this purpose as illustrated in Reference [11] (pp. 83 through 111). Thus, Equation (3.7-3) becomes:
0.ukum i irirr=+ (3.7-5) The equation given for each mass point r in Equation (3.7-5) can be written as a system of equations in matrix form as:
[]{..}[]{}MK+=0 (3.7-6) Where:
[M] = mass and rotational inertia matrix.
{} = column matrix of the general disp lacement and rotation at each mass point relative to the base.
[K] = square stiffness matrix.
{} = column matrix of general translational and angular accelerations at each mass point relative to the base, d 2 {}/dt 2. RN 01-113 3.7-14 Reformatted July 2014 Harmonic motion is assumed and the {} is expressed as:
{} = {} sin t (3.7-7)
Where:
{} = column matrix of the spatial displacement and rotation at each mass point relative to the base.
= natural frequency of harmonic motion in radians per second.
The displacement function and its second derivative are substituted into Equation (3.7-6) and yield:
[K] {} = 2 [M] {} (3.7-8)
The determinant [K] - 2 [M] is set equal to zero and is then solved for the natural frequencies. The associated mode shapes ar e then obtained from Equation (3.7-8). This yields n natural frequencies and mode shapes where n equals the number of dynamic degrees of freedom of the system.
The mode shapes are all orthogonal to each other and are sometimes referred to as normal mode vibrations. For a single degree of freedom system, the stiffness matrix and mass matrix are single terms and the determinant [K] - 2 [M] when set equal to zero yields simply:
k - 2 m = 0 or:
m k= (3.7-9) where is the natural angular frequency in r adians per second. The natural frequency in cycles per second is therefore m k 2 1 f= (3.7-10)
To find the mode shapes, the natural fr equency corresponding to a particular mode, n , can be substituted in Equation (3.7-8), however, only n-1 of these equations are independent. This means that the elements of {} can be expressed only as multiples of one another. Normalizing {} such that the maximum disp lacement (rotation) of any element is unity gives:
rn = displacement (rotation) of mass point r in mode n relative to the base.
RN 01-113RN 01-113RN 01-113RN 01-113RN 01-113RN 01-113 RN 01-113RN 01-113 3.7-15 Reformatted July 2014 2. Modal Equations
The response of a structur e or component is always some combination of its normal modes. However, good accuracy can be obtained by using only the first few modes of
vibration. In the normal mode method, the mode shapes are used as principal coordinates to reduce the equations of motion to a set of uncoupled differential equations that describe the motion of each mode n. These equations may be written as (Reference [11], pp. 116-125):
sn n 2nnnnnyAWAp2A=++ (3.7-11)
where the modal displacement or rotation, A n , is related to the displacement or rotation of mass point r in mode n, u rn , by the equation:
rnnrnAu= (3.7-12)
Where:
n = natural frequency of mode n in radians per second.
p n = critical damping ratio of mode n.
n = modal participation factor of mode n given by: =n 2rnr nrnr n m m (3.7-13) and:
rn= value of rn in the direction of the earthquake.
The essence of the modal analysis lies in the fa ct that Equation (3.7-
- 11) is analogous to the equation of motion for a single degree of freedom system that will be developed from Equation (3.7-4). Dividing Equation (3.7-4) by m gives:
syu m k u m c u=++ (3.7-14) RN 01-113RN 01-113RN 01-113RN 01-113RN 01-113RN 01-113 3.7-16 Reformatted July 2014 The critical damping ratio of a single degree of freedom system, p, is defined by the equation:
c c c p= (3.7-15)
where the critical damping coeffici ent is given by the expression:
=m2c c (3.7-16)
Substituting Equation (3.7-16) into Equat ion (3.7-15) and solv ing for c/m gives:
p2 m c= (3.7-17)
Substituting this expression and the expression for k/m giv en by Equation (3.7-9) into Equation (3.7-14) gives:
s 2yuup2u=++ (3.7-18)
Note the similarity of Equations (3.7
-11) and (3.7-18). Thus each mode may be analyzed as though it were a single degr ee of freedom system and all modes are independent of each other.
By this method a fraction of cr itical damping, i.e., c/cc, may be assigned to each mode and it is not necessary to identify or evaluate individual damping coefficients, i.e., c. However, a ssigning only a single damping ratio to each mode has a drawback. Normally, there are two ways used to overcome this limitation when considering a slightly damped struct ure supported by a massive moderately damped structure.
The first method is to develop and analyze separate mathematical models for both structures using their respective damping values. The massive moderately damped support structure is analyzed first. The calculated response at the support points for the slightly damped structures is used as a forcing function for the subsequent detailed analysis. The second method is to inspect the mode shapes to determine which modes
correspond to the slightly damped structure and then use the damping associated with the structure having predominant motion.
3.7.2.1.2.3 Response Spectrum Analysis The response spectrum is a plot showi ng the variation in the maximum response (Reference [12], pp. 24-51) (displacement, velocity, and acceleration) of a single-degree-of-freedom system versus its natural frequency of vibration when subjected to a time history motion of its base. Examples of response spectra are shown in Figures 3.7-13 and 3.7-14.
RN 01-113RN 01-113RN 01-113RN 01-113 3.7-17 Reformatted July 2014 The response spectrum concept can best be ex plained by outlining t he steps involved in developing a spectrum curve. Determination of a single point on the curve requires that the response (displacement, velocity, and a cceleration) of a single degree of freedom system with a given damping and natural frequency be calculated for a given base motion. The variations in response are es tablished and the maximu m absolute value of each response is plotted as an ordinat e with the natural frequency used as the abscissa. The process is repeated for other assumed values of frequency in sufficient detail to establish the complete curve. Other curves corresponding to different fractions of critical damping are obtained in a similar fashion. Thus, the determination of each point of the curve requires a complete dynamic response analysis, and the
determination of a complete spectrum may involve hundreds of such analyses. However, once a response spectrum plot is generated for the particular base motion, it may be used to analyze each structure and component with that base motion. The
spectral acceleration, velocity, and displacement are related by the equation:
n n n d 2 nvnaSSS== (3.7-19)
There are two types of response spectra that must be considered. If a given building is shown to be rigid and to have a hard foundat ion, the ground response spectrum or ground time history is used. It is referr ed to as a ground response spectrum. If the building is flexible and/or has a soft foundation, the ground response spectra are modified to include these effects. The res ponse spectrum at various support points must be developed. This is called a floor response spectrum. The specific response spectra used are discussed in Sections 3.7.1 and 3.7.2.5.
3.7.2.1.2.4 Integration of Modal Equations This method can be separated into the following two basic steps:
- 1. Integration procedure for the uncoupled modal Equation (3.7-11) to obtain the modal displacements and accelerations as a function of time.
- 2. These modal displacement s and accelerations are combined to obtain the total displacements, accelerati ons, forces, and stresses.
Integration of these uncoupled modal equations is done by step-by-step numerical integration. The step-by-step numerical integration procedure
[13] consists of selecting a suitable time interval, t, and calculating modal acceleration, n A, modal velocity, n A, and modal displacement, A n , at discrete time stations t apart, starting at t = 0 and continuing through the range of interest for a given time history of base acceleration. RN 01-113RN 01-113 3.7-18 Reformatted July 2014 From the modal displacements and accelerations, the total displacements, accelerations, forces, and stresse s can be determined as follows:
- 1. Displacement of mass point r in mode n as a function of time is given by Equation (3.7-12) as:
uArnnrn= (3.7-20) with the corresponding acceleration of mass point r in mode n as:
uArnnrn (3.7-21)
- 2. The displacement and acceleration values obtained for the various modes are superimposed algebraically to give the total displacement and acceleration at each time interval.
- 3. The total acceleration at each time interv al is multiplied by the mass to give an equivalent static force. Stresses are calculated by applying these forces to the model. Alternatively these stresses may be determined from the deflections at each time interval.
3.7.2.1.2.5 Integr ation of Coupled Equations of Motion The dynamic transient analysis is a time hi story solution of the response of a given structure to known forces and/or displacement forcing functions. The structure may
include linear or nonlinear elements, gaps, in terfaces, plastic elements, and viscous and Coulomb dampers. Nodal disp lacements, nodal forces, pr essure, and/or temperatures may be considered as forcing functions. Nodal displacements and elemental stresses for the complete structure are calculated as functions of time.
The basic equations for the dynamic analysis are as follows:
)}t(F{}x{]K[}x{]C[}x{]M[=++ (3.7-22)
where the terms are as defined earlier and {F(t
)} and may include the effects of applied displacements, forces, pressure s, temperatures, or nonlinear effects such as plasticity and dynamic elements with gaps. Options of translational accelerations input to a structural system and the inclusion of static defor mation and/or preload may be considered in the nonlinear dy namic transient analysis. The option of translational input such as uniform base motion to a structur al system is considered by introducing an inertia force term of -[M]
{}z to the right hand side of the ba sic Equation (3.7-22), i.e., }z{]M[}F{}x{]K[}x{]C[}x{]M[=++ (3.7-23) RN 01-113RN 01-113 3.7-19 Reformatted July 2014 The vector }z{ is defined by its components }z{i where i refers to each degree of freedom of the system. }z{i is equal to a 1 , a 2, or a 3 if the i(th) degree of freedom is aligned with the direction of the system translational acceleration a 1 , a 2, or a 3 , respectively. }z{i = 0 if the i(th) degree of freedom is not aligned with any direction of the system translational acceleration. Typica l application of this op tion is a structural system subjected to a seismic excitation of a given ground acceleration record. The displacement {x} obtained from the solution of Equation (3.7-23) is the displacement relative to the ground.
The option of the inclusion of initial static deformation or preload in a nonlinear transient dynamic structural analysis is considered by solving the static problem prior to the dynamic analysis. At each stage of integr ation in transient analysis, the portion of internal forces due to static deformation is always balanced by the portion of the forces which are statically applied. Hence, only the portion of the forces which deviate from the static loads will produce dynamic effects. The output of this analysis is the total
result due to static and dynamic applied loads.
One available method for the num erical integration of Equati ons (3.7-22) and (3.7-23) is the third order (cubic) integrat ion scheme. In the third-or der (cubic) integration scheme values of {x} are assumed to be a cubic func tion of time over a small time increment, i.e.,
3 2 tt}d{t}c{t}b{}a{}x{+++= (3.7-24)
The velocity and acceleration vectors are found by differentiating Equation (3.7-24) with respect to time:
2 tt}d{3t}c{2}b{}x{++= (3.7-25) t}d{6}c{2}x{
t+= (3.7-26)
The unknown coefficients {a}, {b}, {c}, and {d} can be obtained from Equation (3.7-24) in terms of the displacements at time t, t-1, t-2, and t- 3 (i.e., present and three previous values of displacements). Thus for each time interval, the velocity and acceleration arrays may be expressed by:
})x{},x{},x{},x({f}x{3t2t1tt1t= (3.7-27)
})x{},x{},x{},x({f}x{3t2t1tt2t= (3.7-28)
3.7-20 Reformatted July 2014 Substituting Equations (3.7-27) and (3.7-28) into Equation (3.7-22) and solving for the present value of displa cement vector gives:
+=++}F{}x{]K[]C[)t(C]M[)t(Ctt 2 2 1 })x{},x{},x{],C[],M([f3t2t1t (3.7-29)
The above set of simultaneous linear equations is solved to obtain the present values of nodal displacements {x t} in terms of the previous (known) values of the nodal displacements. Since {M}, {C}, and {K} are included in the equation, they can also be time or displacement dependent.
3.7.2.2 Natural Frequencies and Response Loads A summary of significant frequencies and m ode shapes for the representative Seismic Category I structures is presented by Fi gures 3.7-15 through 3.7-20 for the Reactor Building, Control Building, Auxiliary Building, Intermediate Building, Fuel Handling Building, and Diesel Generator Building, respectively.
Floor response spectra at critical Seismic Category I structure el evations and points of support are presented for the OBE by Figur es 3.7-21 through 3.7-27 for the Reactor Building, Figures 3.7-28 and 3.7-29 for the Control Buil ding, Figures 3.7-30 through 3.7-32 for the Auxiliary Buildi ng, Figures 3.7-33 and 3.7-34 fo r the Intermediate Building, Figures 3.7-35 and 3.7-36 for t he Fuel Handling Building, Figures 3.7-37 through 3.7-39 for the Diesel Generator Bu ilding and Figures 3.7-40 and 3.
7-41 for the Service Water Pumphouse. These floor response spectra are used for the seismic qualification of Seismic Category I safety class equipment and components as described in Section
3.7.3. Response loads (displacements, accelerations and masses) are presented by Table 3.7-7a for the OBE.
In Table 3.7-7a and Figures 3.7-21 through 3.
7-41, the x component corresponds to the plant East-West direction, the y component corresponds to the plant North-South direction and the z component corresponds to the plant vertical direction. One exception occurs in the ca se of the service water pumphouse where the x component is perpendicular to the shoreline and the y component is parallel to the shoreline. Both the actual, narrow and artificially broadened response spectra are drawn.
Since the artificial time history has frequency content around 20 cps and between 24 and 30 cps, as shown by Figure 3.7-5, and the building has a horizontal natural frequency at 29.2 cps, as shown by Figure 3.7-15, the secondary peak at 29 cps is higher than the spectrum value at 20 cps. However, the spectrum value at 20 cps is still
higher than the maximum floor acceleration of 0.18g. For t he vertical earthquake, there is a natural frequency at 22.5 cps. This natural frequen cy and the frequency content of the artificial time history caused the secondary peaks at 21 and 26 cps. RN 01-113 3.7-21 Reformatted July 2014 The SSE response spectrum envelope accelerations have been calculated by scaling the OBE response spectrum envelope accelerations. Since the SSE structured damping is conservatively assumed to be the same as the OBE structural damping, the SSE and OBE response ratio is proportional to the input acceleration value. For buildings on rock, the ratio of SSE to OBE acceleration is 1.5. For buildings on soil, the ratio is 1.67. For buildings on caissons t he ratio varies from 1.62 to 1.55 as described by Figure 3.7-43. Scale factors for each of the buildings are as follows:
- 1. Reactor Building, 1.5 OBE.
- 2. Auxiliary Build ing, 1.55 OBE.
- 3. Control Building, 1.55 OBE.
- 4. Fuel Handling Bu ilding, 1.62 OBE.
- 5. Intermediate Bu ilding, 1.55 OBE.
- 6. Diesel Generator Building, 1.62 OBE.
- 7. Service Water Pumphouse, 1.67 OBE.
To properly account for amplification due to flexure of floor slabs, a scaling factor, , has been applied to the response spectrum envel opes for the vertical direction. The magnitude of the scaling fact or depends upon the locati on of equipment, components, or systems on a particular building floor and is discussed in Section 3.7.3.
The maximum floor accelerations at the ma jor building floor elev ations and equipment supports correspond to the acceleration at the high frequency ends of the associated floor response spectra.
3.7.2.3 Procedure Used for Modeling In the seismic analysis, Seismic Category I structures which fo rm a soil-structure interaction system with foundation media are defined as "seismic systems." Except for
the Service Water Pumphouse, as mentioned in Section 3.7.2.2, t hese structures are simulated by lumped mass mathematical model
- s. Other Seismic Category I structures, components, and equipment, defined as "seism ic subsystems," are decoupled from the "seismic system" in the mathematical model
- s. The general uncou pling criterion is based upon the ratio of mass of the supported subsystem to that of the supporting system. If the total building mass is used as the supporting mass, the mass ratio is always small. However, if the modal ma ss is used as the supporting mass, the mass ratio can be greater than 0.01. Based upon the theory of random vibration, the effect of a large mass ratio is a decrease in the mean square responses of the supported
subsystem and the supporting system (see Figure 2.12 and 2.13 of Reference [25]). In the uncoupled analysis, the s ubsystem is analyzed using a floor response spectrum which is obtained by assuming that the subs ystem is rigidly attached to the supporting 3.7-22 Reformatted July 2014 system. This assumption of rigid attachment is equivalent to the zero mass ratio case discussed in Reference [25]. Since the zero mass ratio case yields the highest subsystem response, the uncoupled analysis is always conservative.
The effects of such decoupling upon the dy namic response of the Reactor Building have been investigated in detail. Both the decoupled case and the case where the heavy equipment is incorporated into the "sei smic system" mathematical model have been examined. It has been found t hat, for the interior concrete structure considered, decoupling of the heavy equipment from the supporting stru cture adds only slight additional conservatism to the structural res ponse and is therefore just ified. The criteria used for lumping masses is as described in Section 3.7.2.1.
Dynamic lumped mass models are constructed so that three-dimensional responses can be obtained. For the finite element model of the Service Water Pumphouse supporting medium, two separat e, othogonal plane strain m odels are constructed.
Three directions of earthquake ground motion are input to the dynamic models, one at a time, for the seismic analysis, including floor response spec tra generation. The spatial combination of such responses is discussed in Section 3.7.2.6.
For analysis of NSSS supplied equipment, prim ary importance is given to the Reactor Coolant System. The analysis of this system can be performed using several different methods depending upon the level of seismic ac tivity at the plant site. The possible methods include:
- 1. Linear modal analysis of the primary loop piping and components.
- 2. Coupled building/loop linear modal analysis.
- 3. Coupled building/loop non-linear time-history analysis.
Methods (2) and (3) are considered to be "seismic systems" in accordance with the
guidance noted in Standard Review Plan 3.
7.3. Method (1) and all other systems and components within the NSSS Vendor's scope F, except primary loop piping, are analyzed independently and are classified as "sei smic subsystems." Examples of these are the primary system components, auxilia ry pumps, branch piping, tanks, etc.
Method (3) was used for the seismic analysis of the Virgil C. Su mmer Nuclear Station Reactor Coolant Loop Piping System. All other analyses of NSSS vendor supplied equipment were performed as "s eismic subsystem" analyses. 02-01 02-01 3.7-23 Reformatted July 2014 3.7.2.4 Soil-Structure Interaction The effect of soil-structur e interaction on the seismic response of lumped mass models is represented by six equiva lent linear foundation springs: three translational; two rocking; and one torsional (see Section 3.7.2.1).
The validity of the foundat ion spring method used for the foundation-structure configuration has been investigated. This investigation is summarized as follows:
- 1. The Seismic Category I structures are supported either on rock supported fill concrete or caissons seated in rock. Therefore, the layering effect is insignificant.
- 2. Since the plant site has suffici ently deep and uniform overburden, frequency independent foundation springs are consid ered to be adequate for simulation of the soil-structure interaction phenomenon[14].
- 3. The effect of embedment depth (Section 3.7.1.4) on the foundation spring values has been considered and, for reasons of conservatism, is ignored.
- 4. For the Seismic Category I structur es supported on caissons, the following considerations have been further investigated through modeling:
- a. The interaction between the caiss on (and, if any, t he underground portion of the structure) and the surr ounding soil medium is accounted for by connecting effective lateral soil springs to the lu mped mass of the caisson (and, if any, the underground portion of t he structure). The soil springs are calculated based upon the work by Penzien
[9] on pile foundations in which Midlin's half-space formulas for the force-disp lacement relationship are utilized.
The amplified free-field seismic motion at each elevation of interest in the soil
medium has been considered. As show n by Figure 3.7-42, the acceleration time histories, A 1 (t), A 2 (t), . . . A i (t), represent the amplified motions at the elevations of interest. However, in the lumped mass approach considered, a single acceleration time history wi th equivalent energy input has been conservatively estimated and applied thr oughout the elevation of the caisson as shown by Figure 3.7-43.
- b. Settlement of the soil underneath the mat is considered.
- c. The stress levels in the soil at t he junction of the under ground structures and the lateral soil medium are compared with the corresponding bearing capacity to check the stability of t he entire soil-st ructure system.
- d. Caissons are drilled into compet ent rock for end bearin g and frictional load transfer. Rock properties relative to caisson design and construction are described in Section 2.5.4.
3.7-24 Reformatted July 2014 e. The pertinent parameters used in t he parametric study of soil-structure interaction are as follows:
(1) Soil shear modulus (G) is 1.0 x 10 4 to 3.5 x 10 4 psi. This range is based upon both in-situ dynamic wave test s and laboratory triaxial tests.
(2) Dynamic concrete Young's modulus (Edynamic) is 1.0 Estatic to 2.0 Estatic.
(3) Settlement of soil under neath the mat: soil takes 0 to 50% of the vertical load.
- 5. The foundation mat of the caisson s upported Intermediate Building is rigidly connected to the Control Building foundation mat for purposes of lateral stability. For such a structure-foundation configur ation, both foundation soil springs and caisson soil springs are used in the combined dynamic lumped mass model of the two buildings.
As mentioned in Section 3.
7.2.1, the Service Water Pumphouse is analyzed for the soil-structure interaction effect by us ing the finite element program, FLUSH
[8]. Approximate strain dependent shear modules of the foundation medium are first estimated for the expected level of ground motion and surface discharge pressure using the one-dimensional wave pr opagation theory. Final strain and frequency dependent soil properties and damping ratio are obtai ned through an iterative sequence of dynamic response calculations for the soil-structure system. The so il property curves presented in Section 2.5 are used.
It is considered that motions in the vicinity of the structure are due to vertical
propagation of body waves from underlying, stiffer formations. Accordingly, the basic method for the dynamic response calculation can be illustrated as follows. A control motion compatible with the design response spec tra, as presented in Section 3.7.2.1, is first specified. This control motion is then us ed as input to a finite element model of the soil-structure system and the re sponse is computed at elevat ions and points of interest.
The computer program FLUSH[8] is also equipped with a transmitting boundary, simulating sufficient extent of the half-space. Furthermore, with use of a viscous boundary on the planar side of the plane-strain finite element model, three dimensional
wave propagation phenomena of the foundation medium ar e taken into account.
3.7-25 Reformatted July 2014 3.7.2.5 Development of Floor Response Spectra Floor response spectra are developed by applyi ng the time history earthquake motion to the multi-degree of freedom m odels of the Seismic Categor y I structures with the various mass points representi ng the described elevation in each structure (see Section 3.7.2.1). Such floor response spectra take into consideration the effect of the three components of earthquake motion and reflect predominant response near the dominant frequencies of t he structure.
Due to the general asymmetry of the configuration of the Seismic Category I structures, the responses of the structures are norma lly three-directional in nature even when subjected to only one component of earthquake motion.
To properly estimate floor response spectrum values, including the three-directional effects of both the earthquake (which would not be likely to produce maximum responses in all three directions simu ltaneously) and the structural behavior, the following equation is used:
2/1 23x 22x 21x x)S S S(S++=
Where:
S x = the final spectrum value at any frequency point in the x direction.
S xk = the spectrum value at the same frequency point in the x direction due to earthquake component in the k(th) direction (k = 1, 2, 3).
In the development of vertical floor respons e spectra, additional effects of vertical amplification other than thos e due to the overall structural response are considered. These are the amplification due to floor flex ibility and rocking motion of the structure.
With the inclusion of the floor flexibility in the vertical response calculation, additional response spectrum peaks are det ermined at the characteristic frequencies of the floor.
To account for the rocking mo tion of the Auxiliary, Contro l, and Intermediate Buildings, several nodal points are select ed at the corners of the floor with which walls usually intersect. These corner points are connected to the lumped mass of the floor by rigid links as shown in the torsional model in Fig.
3.7-11a. Additional amplifications are observed in the floor response spectrum.
The final floor response spectrum is constructed to envelop the spectrum peaks attri butable to the overall vertical structural response and the floor flexibil ity, as well as to the rocki ng motion of the structure.
The effect of torsional response on horizontal floor response spectra, due primarily to the eccentricity of the stru ctural configuration, is di scussed in Section 3.7.2.11. 98-01 RN 01-113 3.7-26 Reformatted July 2014 3.7.2.6 Three Components of Earthquake Motion 3.7.2.6.1 Balance of Plant Scope The time history method is employed in the seismic analysis of Seismic Category I structures. The maximum responses (e.g., accelerations) due to each of the three components of earthquake motion are first calc ulated separately at a particular point of a structure or of t he corresponding mathematical model. These component responses are then combined by taking the square root of the sum of the squares of maximum codirectional responses caused by each of the three components of earthquake motion (see Section 3.7.2.5). This procedure is in conformance with Regulatory Guide 1.92, Revision 1.
3.7.2.6.2 Components and Equipment Provided by the NSSS Vendor The seismic design of the pi ping and equipment includes t he effect of the seismic response of the supports, equipment, structures, and components. The system and equipment response is determined using th ree earthquake component s, two horizontal and one vertical. The design gro und response spectra, specif ied in Section 3.7.1, are the bases for generating these three input components. Floor response spectra are generated for two perpendicular horizontal directions, (i.e., N-S and E-W) and the
vertical direction. System and equipment analysis is perform ed with these input components applied in the N-S, E-W, and vertical directions. The damping values used in the analysis are those given in Table 3.7-3.
In computing the system and equipment re sponse by response spectrum modal analysis the methods of Sect ion 3.7.3.7 are used to comb ine all significant modal responses to obtain the combined unidirectional responses.
The combined total response is then calculated using the square root of the sum of the
squares formula applied to t he resultant unidirectional re sponses. For instance, for each item of interest such as displacement, force, stresses, etc., the total response is obtained by applying the above described method. The mathematical expression for this method (with R as t he item of interest) is:
2/1 2 T 31T C]R[R== (3.7-30)
Where:
2/1 2 Ti N1i T]R[R== (3.7-31) and:
3.7-27 Reformatted July 2014 R C = total combined response at a point.
R T = value of combined response of direction T.
R Ti = absolute value of response for direction T, mode i.
N = total number of modes considered.
The subscripts can be reversed without c hanging the results of the combination.
For the case of closely spaced modes, R T in Equation (3.7-31) above is replaced with R T as given by Equation (3.7-32) in Section 3.7.3.7, where the criteria and justification for meeting the intent of Regulatory Guide 1.92, Revision 1 are presented.
3.7.2.7 Combination of Modal Responses 3.7.2.7.1 Balance of Plant Scope Since only the time history method is em ployed in the seismic (system) analysis of Seismic Category I structures, modal response s are algebraically combined, in the time domain, in the solution following the principl e of superposition. Regulatory Guide 1.92, Revision 1, is, therefore, not applicable.
3.7.2.7.2 Components and Equipment Provided by the NSSS Vendor Conformance with the recommendations of Regulatory Guide 1.92, Revision 1 for combination of modal responses is presented in Section 3.7.3.7.
3.7.2.8 Interaction of Nonc ategory I Structures with Seismic Category I Structures Complete separation of Seismic Category I st ructures from adjacent Seismic Category I or non-Seismic Category I structures prevents interaction with or impact from adjacent structures. Seismic effects on non-Seismic Category I structures are investigated to prevent the possibility of damage to Seismi c Category I structures due to possible collapse of non-Seismic Category I structures.
3.7.2.9 Effects of Parameter Vari ations on Floor Response Spectra The peak width and period coordinates of the floor response spectra (see Section 3.7.2.5) are determined by parametric study based upon t he variations of the soil springs and/or material proper ties of the structure and foundation. In any case, the peak width is assigned a minimum range of
+/-10% of the center frequency. For cases where the variation of calc ulated periods due to the va rious assumptions regarding material properties and soil-structure inte raction change the spectrum values, the parametric study, coupled with the envelopin g process, is used to conservatively estimate the floor resp onse spectrum values.
RN 01-113 3.7-28 Reformatted July 2014 3.7.2.10 Use of Constant Vertical Static Factors Dynamic analyses with vertical earthquake motion are performed instead of using constant vertical load factors.
3.7.2.11 Methods Used to Account for Torsional Effects A mathematical model with a rigid link connecting the center of mass and center of rigidity at each floor elevation is used to ca lculate the actual torsional responses. For typical asymmetric structures, such as the Intermediate Building, eac h floor in the model is taken as a rigid diaphragm with th ree translational and one torsional degree of freedom. Responses, including the amplification effect of the corner nodal points, as presented in Section 3.7.2.5, are obtained separately for the vertical and horizontal excitations. The combined response is obtained in the manner discussed in
Section 3.7.2.6.
3.7.2.12 Comparison of Responses Since the time history method is used throughout for seismic system analyses, no comparison of responses is needed.
3.7.2.13 Methods for Seismic Analysis of Dams Methods for seismic analysis of dams are presented in Section 2.5.
3.7.2.14 Determination of Seismic Cat egory I Structure Overturning Moments Overturning moments for Seismic Category I structures are deter mined at the base of the Seismic Category I structures. Ea ch of the three components of earthquake excitation is considered separately. The resultant overturning moment is obtained by combining the three earthquake components in accordance with Section 3.7.2.6. The vertical earthquake component is viewed as r educing the dead weight of the structure in counteracting the overturning moment. So il reaction is calculated by adding to, or subtracting (whichever controls) the vertical earthquake co mponent from, the dead weight and other loads on the structure. Safety factor s of 1.1 and 1.5 are provided against overturning of Seismic Category I structures due to the SSE or OBE, respectively, combined with other appropriate design loads.
3.7-29 Reformatted July 2014 3.7.2.15 Analysis Procedure for Damping 3.7.2.15.1 Balance of Plant Scope A classical modal analysis is applied to t he lumped mass soil-structure interaction system. In the modal analysis, different modal damping is s pecified for each mode according to the concept of weighted average damping
[15, 16]. In this approach a higher damping value is specified for a mode in which soil deformation is predominant compared to structural deformation.
The basic equation for the weighted average damping is as follows:
+===N1in,in,ii,v i n N1ii,H n EE)DD(D Where:
D n = Weighted modal damping for the n(th) mode.
i = 1, 2, 3, . . . , S, S+1, . . . . , N.
S = The degree of freedom of the structure for wh ich damping values are assigned.
N-S = The degree of freedom r epresenting the soil springs.
D H,i = Hysteretic damping in the i(th) degree of freedom of the soil-structure interaction system.
D v,i = Viscous damping in the i(th) degree of freedom of the system.
n = Frequency of n(th) mode.
i = Frequency at which D v,i is defined.
E i,n = Energy stored in i(th) degree of freedom of the system in n(th) mode.
Damping in structural elements (of superstru cture and caissons) genera lly is hysteretic in nature which shows frequen cy independent energy loss per cycle. The internal damping in soil is also hyst eretic. Radiation damping d ue to wave propagation from a structure into the soil medium is viscous damping which shows frequency dependent energy loss per cycle. When a structure is founded upon ca issons, the overall damping in rocking is primarily hysteretic. Theor etically, radiation damping in swaying is RN 01-113 RN 01-113 3.7-30 Reformatted July 2014 generally very large. However, such radiation damping is se t to zero for conservatism.
Typical choices for the damping ratios (see Section 3.7.1.3) are as follows:
- 1. For i s a. D H,i = (see Table 3.7-1)
- b. D v,i = 0 2. For Rocking Mode Soil Spring
- a. D H,i = 5%
- b. D v,i = 0
- 3. For Swaying Mode Soil Spring
- a. D H,i = 10%
- b. D v,i = 0 3.7.2.15.2 Components and Equipment Provided by the NSSS Vendor For components and equipment provided by We stinghouse, either the lowest damping value associated with the elements of the sy stem is used for all modes, or an equivalent modal damping value is determined according to the energy distribution in each mode.
Testing programs for damping were done for the reactor coolant loop
[3]. 3.7.3 SEISMIC SUBSYSTEM ANALYSIS 3.7.3.1 Seismic Analysis Methods 3.7.3.1.1 Balance of Plant Scope Seismic analysis is performed for those subsystems that can be modeled to correctly predict the seismic response. A component is modeled as a multi-degree-of-freedom, lumped mass system with mass free interconnections and sufficient mass points to ensure adequate representation. The resulting system is analyzed using the response spectrum modal analysis technique. An alte rnative time history method may also be applied. The time history method, when us ed, conservatively simulates the response spectrum envelope of interest. A stre ss analysis is then perform ed using the inertia forces or equivalent static loads obtained from the dynamic analysis.
Moments, shears, accelerations, deflections and stresses ar e calculated on a mode by mode basis. The total seismic response is obtained by combining each modal response using the square
root of the sum of the squares method. The absolute sum of the responses is considered for closely spaced, in phase modes as set forth in Section 3.7.3.7. In cases 02-01 RN 01-113 3.7-31 Reformatted July 2014 for which some dynamic degrees of freedom do not contribute to the total response, kinematic condensation is employed in the analysis.
3.7.3.1.2 Components and Equipment Provided by the NSSS Vendor Seismic analysis methods for subsystems wi thin Westinghouse scope of responsibility are given in Section 3.7.2.1.
3.7.3.2 Determination of Number of Earthquake Cycles 3.7.3.2.1 Balance of Plant Scope One SSE plus five OBE events are assumed to occur during the life of the plant.
Ten (10) maximum stress cycles are assumed to occur during each event.
3.7.3.2.2 Components and Equipment Provided by the NSSS Vendor For each OBE the system and component will have a maximum response
corresponding to the maximum induced stre sses. The effect of these maximum stresses for the total number of OBE's must be evaluated to assure resistance to cyclic loading.
The OBE is conservatively assumed to occu r 20 times over the life of the plant. The number of maximum stress cycles for each occurrence depends on the system and component damping values, complexity of the system and component, duration, and frequency content of the input earthquake. A precise determi nation of the number of maximum stress cycles can only be made usi ng time history analysis for each item.
Instead, a time history study has been conducted to arrive at a realistic number of maximum stress cycles.
To determine the conservative equivalent number of cycles of maximum stress associated with each occurrence, an evaluation was performed considering both equipment and its supporting build ing structure as single-degr ee-of-freedom systems.
The natural frequencies of the building and the equipment are conservatively chosen to coincide. The damping in the equipment and building are equiva lent to the damping values in Table 3.7-3.
The results of this study indicate that the total number of maximum stress cycles in the equipment having peak acceleration above 90% of the maximum absolute acceleration did not exceed eight cycles. If the equipmen t was assumed to be rigid in a flexible building, the number of cycles exceeding 90
% of the maximum st ress was not greater than three cycles.
This study was conservative since it was performed with si ngle-degree-of-freedom models which tend to produce a more uniform and unattenuated response than a complex interacted system. The conclusions indicate that 10 maxi mum stress cycles for flexible equipment (nat ural frequencies less than 33 Hz) and 5 maximum stress 3.7-32 Reformatted July 2014 cycles for rigid equipment (natural frequenc ies greater than 33 Hz) for each of 20 OBE occurrences should be used for fatigue evaluation.
3.7.3.3 Procedure Used for Modeling Equipment within the balance of plant sc ope is modeled as a series of discrete mass points, connected by mass free members, having sufficient mass points to ensure adequate representation of dynam ic behavior. Detailed modeling of piping systems is described in Section 3.7.3.8.
Procedures used for modeling the equipment and components provided by the NSSS vendor are described in Section 3.7.2.1.2.
3.7.3.4 Basis for Selection of Frequencies 3.7.3.4.1 Balance of Plant Scope Where applicable for analysis of equipment, modes selected as significant include the following:
- 1. Natural frequencies less than 30 Hz.
- 2. Frequencies above 30 Hz with modal partici pation factors greater than 10% of the fundamental mode part icipation factor.
For piping systems where a detailed seismic analysis is performed, all modes with natural frequencies of less than 30 Hz are included in the response calculation.
For piping systems where the simplified seis mic support spacing method is employed, the limits of seismic support spans are established under the condition that the fundamental frequencies are hig her than those associated wi th the dominating peak of the floor response spectrum.
3.7.3.4.2 Equipment and Components Provided by the NSSS Vendor The analysis of the equipment subjected to seismic loading involves several basic steps, the first of which is the establishment of the intensity of the seismic loading.
Considering that t he seismic input originates at the point of support, the response of the equipment and its associated suppor ts based upon the mass and stiffness characteristics of the system, will determine the seismic accelerations which the equipment must withstand.
Three (3) ranges of equipment/support behavio r which affect the magnitude of the seismic acceleration are possible:
- 1. If the equipment is rigid relative to the structure, the maximum acceleration of the equipment mass approaches that of the st ructure at the poi nt of equipment 3.7-33 Reformatted July 2014 support. The equipment acceleration val ue in this case corresponds to the low-period region of the floor response spectra.
- 2. If the equipment is very flexible relative to the structure, the equipment will show very little response.
- 3. If the periods of the equipment and supporting structure are nearly equal, resonance occurs and must be taken into account.
In the above cases, equipment under earthquake loadings is designed to be within code allowable stresses.
Also, as noted in Section 3.7.3.2.2, rigid equipment/support systems have natural frequencies greater than 33 Hz.
3.7.3.5 Use of Equivalent Static Load Method of Analysis 3.7.3.5.1 Balance of Plant Scope If the fundamental frequency of the component is greater than 30 Hz, the component is analyzed statically. The equivalent static fo rces are obtained by multiplying the lumped mass of each mass point by the appropriate maximum floor acceleration. The maximum floor acceleration is obtained from the response spectra envelope at the high frequencies.
3.7.3.5.2 Equipment and Components Provided by the NSSS Vendor The static load equivalent or static analysis method involves the mu ltiplication of the total weight of the equipm ent or component member by the specified seismic acceleration coeffici ent. The magnitude of the seismi c acceleration coefficient is established on the basis of the expected dynamic response characteristics of the component. Components which can be adequately characterized as single-degree-of-freedom system s are considered to have a modal participation factor of one. Seismic acceleration coefficient s for multi-degree-of-f reedom systems which may be in the resonance region of the amplified response spectra curves are increased by 50% to account conservatively for the increased m odal participation.
3.7.3.6 Three Components of Earthquake Motion For the balance of plant scope the responses to the two horizontal and the vertical component seismic inputs are calculated separately for the entir e subsystem. The maximum value of a particular response due to simultaneous action of three components of earthquake were obtained by ta king the square root of the sum of the squares of corresponding maximum response values to each of the three components calculated separately. This procedure is in conformance with Regulatory Guide 1.92.
3.7-34 Reformatted July 2014 For components and equipment provided by the NSSS vendor, methods used to account for three components of earthquake mo tion are given in Section 3.7.2.6.2.
3.7.3.7 Combination of Modal Responses 3.7.3.7.1 Balance of Plant Scope The combination of modal responses is limited to the respons e spectrum modal analysis technique.
Two consecutive modes are defined as clos ely spaced if their frequencies differ from each other by 10% or less of the lower fr equency. For modes that are not closely spaced, the maximum value of the response of a giv en element of a system or component, subjected to a single independent spatial co mponent (response spectrum) of a three-component earthquake, is obtained by taking the s quare root of the sum of the squares of corresponding maximum values of the response of t he element attributed to individual significant m odes of the system or component.
If some or all of the modes are closely spac ed, they are divided into groups that include all modes having frequencies between t he lowest frequency in the group and a frequency 10% higher. The representative maxi mum value of a particular response of a given element of a system or component attri buted to each such group of modes is first obtained by taking the sum of the absolute values of the corresponding peak values of the response of the element attributed to individual modes in that group. The representative maximum value of this particular response attributed to all the significant modes of the system or com ponent is then obtained by taki ng the square root of the sum of the squares of corres ponding representative maximum values of the response of the element attributed to each closely spac ed group of modes and the remaining modal responses for the modes that are not closely spaced.
This procedure is in conformanc e with Regulatory Guide 1.92.
3.7.3.7.2 Components and Equipment Supplied by the NSSS Vendor For response spectra analysis, the total unidi rectional seismic response is obtained by combining the individual modal responses ut ilizing the square root of the sum of the squares method. For systems having modes with closely spaced frequencies, this method is modified to include the possible effect of these modes. The groups of closely spaced modes are chosen such that the difference between the frequencies of the first mode and the last mode in the group does no t exceed 10% of the lower frequency.
Combined total response for systems which have such closely spaced modal frequencies is obtained by adding to the square root of the sum of the squares of all modes the product of the responses of the modes in each group of closely spaced modes and a coupling factor . This can be represent ed mathematically as:
02-01 3.7-35 Reformatted July 2014 KK N1K1NMK S1j 2 i N1i 2 TRR2RR j j j+=+==== (3.7-32) Where:
R T = total response.
R i = absolute value of response of mode i.
N = total number of modes considered.
S = number of groups of closely spaced modes.
M j = lowest modal number associated wi th group j of closely spaced modes.
N j = highest modal number associated with group j of closely spaced modes.
K = coupling factor with:
1 2KK K K 1+= (3.7-33) and:
2/12jjj])(1[= (3.7-34) djjj t 2+= (3.7-35)
RN 01-113RN 01-113 3.7-36 Reformatted July 2014 Where: j = frequency of closely spaced mode j.
j = fraction of critical damping in closely spaced mode j.
t d = duration of the earthquake.
An example of this equation applied to a system can be suppli ed with the following considerations. Assume t hat the predominant contribut ing modes have frequencies as given below:
Mode 1 2 3 4 5 6 7 8 Frequency 5.0 8.0 8.3 8.6 11.0 15.5 16.0 20
There are two groups of closely spaced modes, namely with modes {2, 3, 4} and {6, 7}.
Therefore:
S = 2 number of groups of closely spaced modes.
M 1 = 2 lowest modal number associated with group 1.
N 1 = 4 highest modal number associated with group 1.
M 2 = 6 lowest modal number associated with group 2.
N 2 = 7 highest modal number associated with group 2.
N = 8 total number of modes considered.
The total response for this system is, as derived from the expansion of Equation (3.7-32):
6776344324422332 2 8 2 3 2 2 2 1 2 TRR2RR2RRRR2]R....RRR[R++++++++= (3.7-36)
For time history analysis, each earthquake component direction is analyzed separately.
For each of these analysis, at each time step, the global response of interest is obtained by superposition of the i ndividual modal responses.
The preceding gives the criter ia and justification for meeti ng the intent of Regulatory Guide 1.92, Revision 1. RN 01-113 RN 01-11302-01 3.7-37 Reformatted July 2014 3.7.3.8 Analytical Procedures for Piping 3.7.3.8.1 Balance of Plant Scope The piping system geometry, cross sectional dimension and physical properties of each pipe segment and the restraint conditions ar e supplied as inputs to the PIPDYN II computer program during construction.
During operations, the approved programs are specified in the VCSNS piping design gui de. The mass of each piping segment is lumped at the element nodes by the computer. Additional concentrated masses are specified separately for valves, actuators, a nd other concentrated we ights at the centers of gravity for the individual assembly or subassemblies to represent both bending and torsional effects of the assembly.
The restraint conditions of supports are specified in three translational and three rotational directions in the model, in either global or local coordinates for each support point. The restraints may be free, rigid, or elastic with a specified spring constant for each translational or rotational direction.
When coupling effects between any two joint degrees of freedom are significant, a 6 by 6 stiffness matrix is used to describe an elastic foundation. Moment release at nodal points is used for pin connections or flexible joints whenever applicable.
The computer then formulates a discrete system of equations based upon the input data. The resulting homogeneous equations are solved as an eigenvalue problem. The floor response spectrum method is used in calculating the responses of each mode including nodal displacements, end forces, and moments and support loads. These modal responses are combined by the square root of the sum of the squares method for all modes with frequencies less than 30 Hz. In addition, the effect s of the modes not included are added to the square root of the sum of the squares response as one term, using the highest frequency from the square root of the s quares response under 30 Hz to obtain the total response. The definit ion and grouping method of combining close modes, described in Section 3.7.3.7 is appli ed in nodal deflection, element end forces and moments, acceleration, and support loads.
The responses from the two horizontal and the vertical component of an earthquake are
calculated separately as described above.
These responses are then combined using the square root of the sum of the squares method. T he resultant end moments are finally used in the applicable ASME Code, Sect ion III, equations fo r stress evaluation.
For certain 2 inch and smaller pipes or cold pipes of larger sizes, a simplified floor response spectrum dynamic analysis is perform ed. An iterative analytical procedure is
followed for each pipe size, schedule, and response spectrum input, while using span length as a variable. The dynamic system equations are solved for the various span
lengths and associated support reactions and pipe stresses are determined. Maximum allowable span lengths are t hen selected on the basis of a stress and/or natural frequency criteria for each of the various par ameters. Seismic su pports/restraints are then spaced to be within the allowable span limits established by this stress/frequency 02-01 3.7-38 Reformatted July 2014 criteria. The pipe frequency is determined using a multi-span, simple supported model with the maximum spans established as descri bed in Section 3.7.3.4. Justification for this approximate analysis method has been demons trated by comparison of the results of detailed dynamic analyses of various piping systems with the resu lts of the simplified method.
A scaling factor (gamma factor) is applied to the response spectrum envelopes for the vertical direction. This is used to account for amplification due to flexure of floor slabs and, therefore, is a function of the location of the components on a particular floor.
Typical scaling factors used for seismic analysis are presented in Table 3.7-8 and Figures 3.7-44 through 3.7-46.
Where relative structural movement exists within the same structure or where piping spans between adjacent structures, the effect of differential piping support movements are evaluated. The relative movements of pipe supports are considered separately in each of the spatial directions. The results are then combined by t he square root of the sum of the squares method.
3.7.3.8.2 Components and Equipment Provided by the NSSS Vendor The Class 1 piping systems are analyzed to the rules of the ASME Code, Section III, NB 3650. When response spectrum methods are used to evaluate piping systems supported at different elevatio ns, the following procedures are used. The effect of differential seismic movement of piping supports is included in the piping analysis according to the rules of the ASME Code, Se ction III, Paragraph NB 3653. According to ASME definitions, these displacements cause secondary stresses in the piping system. The response quantity of interest induced by differential seismic motion of the support is computed statically by considering the building response on a mode-by- mode basis.
In the response spectrum dynamic analysis for evaluation of piping systems supported at different elevations, the most severe floor response s pectrum corresponding to the support locations is used.
The selection, location, and use of snubbers for these systems is based upon the stress requirements of the ASME Code, Secti on III, Articles NB/
NC/ND-3600 or other controlling design criteria.
3.7.3.9 Multiply Suppor ted Equipment Components with Distinct Inputs 3.7.3.9.1 Balance of Plant Scope Any equipment supported at different locations (elevations and/or floors) is analyzed by imposing a single conservative response spectrum at each location. This response spectrum is constructed in such a way that it conservatively envelopes the pertinent response spectra of the different locations.
3.7-39 Reformatted July 2014 3.7.3.9.2 Components and Equipment Provided by the NSSS Vendor When response spectrum methods are used to evaluate Reactor Coolant System primary components interconnected between fl oors, the procedures of the following paragraphs are used. There are no components in the We stinghouse scope of analysis which are connected between buildings.
The primary components of the Reactor Coolant System are supported at no more than two floor elevations.
A response spectrum analysis is first made assuming no relative displacement between support points. The response spectra used in this analysis are the most severe floor response spectra.
Secondly, the effect of differential seis mic movement of components interconnected between floors is considered statically in the integrated syst em analysis and in the detailed component analysis. The results of the building analysis are reviewed on a mode-by-mode basis to determine the differ ential motion in each mode. Per ASME Code rules, the stress caused by differential seismic motion is clearly secondary for piping (NB 3650) and component supports (NF 3231). For components, the differential motion is evaluated as a free end displacemen t, since, per NB 3213.19, examples of a free end displacement are motions "that would occur because of relative thermal
expansion of piping, equipment, and equipm ent supports, or because of rotations imposed upon the equipment by sources other than the piping".
The effect of the differential motion is to impose a rotation on the component by t he building. This motion, then, being a free end displacement and being similar to thermal expansion loads, causes stresses which are evaluat ed with ASME Code methods including the rules of NB 3227.5 used for stresses originat ing from restrained free end displacements.
The results of these two st eps, the dynamic inertia analysi s and the static differential motion analysis, are combined absolutely with due consideration for the ASME classification of the stresses.
3.7.3.10 Use of Constant Vertical Static Factors 3.7.3.10.1 Balance of Plant Scope The response spectrum method is used for the vertical seismic subsystem dynamic analysis. However, for the cases where the equipment's lowest frequency in the vertical direction is more than 30 Hz, the maximum floor acceleration is used for equipment design.
3.7-40 Reformatted July 2014 3.7.3.10.2 Components and Equipment Provided by the NSSS Vendor Constant vertical load factor s are not used as the vertical floor response load for the seismic design of safety-related components and equipment within Westinghouse's scope of responsibility.
3.7.3.11 Torsional Effect s of Eccentric Masses If the torsional effect of a valve operator or other eccentric mass is likely to have a significant effect upon the results of the analysis described in Section 3.7.3.1 for Seismic Category I systems, t he eccentric mass and its mom ent arm are included in the mathematical model described in Sections 3.7.3.3 and 3.7.3.8.
3.7.3.12 Buried Seismic Ca tegory I Piping Systems Seismic analysis of buried safety class piping is performed in three phases, as follows:
- 1. Calculation of maximum soil strain and curvature resulting from the propagation of seismic waves, considering an estimated relative contribution of each wave type (shear wave, compression wave, and Rayleigh wave).
- 2. Determination of the extent to which the pipe deforms elastically as a result of soil strain and curvature, or as a result of relative ground movement at building/soil interfaces, considering the influence of friction forces between soil and pipe and treatment of soil as a c ontinuous elastic support.
- 3. Calculation of the stresse s in the pipe which result from such elastic deformation and comparison with allowable stre ss as described by Table 3.9-2.
References [21] through [24]
were used as the basis for the analytical determination of seismic stresses in buried safety class pipi ng. Sources for various parameters used in the analysis are as follows:
- 1. Maximum acceleration for SSE, Section 2.5.2.10.
- 2. Fill surface elevations and subsoil conditions, Sections 2.
5.4.4 and 2.5.4.8.
- 3. Properties (max tGand,,Q,c,µ) of Zone I and Zone II fill soils, Sections 2.5.4.5.2 and 2.5.6.4.
- 4. Properties (Qand,c, t)of Zone III fill soil are assumed, based upon Section 2.5.4.5.2 and US Navy Design Manual , NAVDOCK DM-7, "Soil Mechanics, Foundations and Earth Structures" Bureaus of Yards and Docks, 1962.
3.7-41 Reformatted July 2014 3.7.3.13 Interaction of Other Pipi ng with Seismic Category I Piping A Seismic Category I piping system is analyzed by including the piping extending to at least the first restraint in each of the th ree mutually orthogonal directions beyond the defined Seismic Category I boundaries. Wh enever necessary, piping segments and restraints beyond the region described abov e are included to ensure that both the elastic reaction and the effects of masses of the non-Seismic Ca tegory I piping on the Seismic Category I piping are adequately represented. Those portions of piping which form interfaces between Seismic Categor y I and non-Seismic Cat egory I are designed to satisfy Seismic Category I requirements.
3.7.3.14 Seismic Analyses for Reactor Internals Fuel assembly component stresses induced by horizontal seismic disturbances are
analyzed through the use of finite element computer modeling.
The time history floor response based on a standard seismic time history normalized to SSE levels is used as the seismic input. The reactor internals and the fuel assemblies are modeled as spring and lumped mass systems or beam elements.
The component seismic response of the fuel assemblies is analyzed to determine design adequacy. A detailed discussion of the analyses performed for typical fuel assemblies is contained in References [4] and [17].
Fuel assembly lateral structural damping obtained experimentally is presented in Figure 3-4 of Reference [18]. The data indicate that no damping values less than 10% were obtained for fuel assembly displacements greater than 0.11 inches.
The distribution of fuel assembly amplitudes decreases as one appr oaches the center of the core. The average amplit ude for the minimum displacement fuel assembly is well above 0.11 inches for the SSE.
Fuel assembly displacement time history for the SSE seismic input is illustrated in Figure 2-3 of Reference [18].
The control rod drive mechanisms (CRDM) are seismically analyzed to confirm that system stresses under the combined loading conditions as described in Section 3.9.1 do not exceed allowable levels as defined by the ASME Code, Section III for Upset and Faulted conditions. The CRDM is mathematically modeled as a system of lumped and distributed masses. The model is analyz ed under appropriate seismic excitation and the resultant seismic bending moments along the length of the CRDM are calculated.
The corresponding stresses are then combi ned with the stresses from the other loadings required and the comb ination is shown to meet the ASME Code Section III requirements. RN 01-113 3.7-42 Reformatted July 2014 3.7.3.15 Analysis Procedure for Damping 3.7.3.15.1 Balance of Plant Scope The composite modal damping approach is used to account for element-associated damping. This is based upon the use of the mass as a weighting function in generating the composite modal damping.
The formulation leads to:
j T M={}[]{} Where:
j = Equivalent modal damping factor of the j(th) mode.
[]M = The modified mass matrix construct ed from element matrix formed by the product of the damping factor as prov ided in Table 3.7-9 for the element and its mass matrix.
{} = The j(th) normalized model vector.
In the cases of equipment which consists of elements of the same damping factor, the above equation is reduced to a single damping factor.
3.7.3.15.2 Components and Equipment Provided by the NSSS Vendor Analysis procedures for damping for subsystems in Westinghouse's scope of
responsibility are given in Section 3.7.2.15.2.
3.7.4 SEISMIC INSTRUMENTATION 3.7.4.1 Comparison with Regulatory Guide 1.12 The seismic instrumentation provided is in general conformance with Regulatory Guide 1.12 (see Appendix 3A) for a maximum foundation acceleration of less than 0.3g.
Since the soil-structure interaction is negligib le, the "Free Field" triaxial time-history accelerograph is omitted as permitted by ANSI N18.5
[19].
3.7-43 Reformatted July 2014 3.7.4.2 Location and Descrip tion of Instrumentation The types and locations of seismic instrumentation are as follows:
- 1. Triaxial Time-His tory Accelerometer
One triaxial accelerometer is loca ted at each of the following locations
- a. Reactor Building foundation mat outside of the Reactor Building (see Figure 3.7-47).
- b. Reactor Building ring girder outside the Reactor Building (see Figure 3.7-48).
The output of both triaxial sensor uni ts (accelerometers) is recorded by a solid-state recording system in the Contro l Building. The accelerometer located on the Reactor Building foundation mat also fu nctions as a seismic trigger for the actuation of the solid-state recordi ng system. An alarm is sounded when the recording system is activated.
- 2. Triaxial Response Spectrum Recorder
One triaxial response spectrum recorder, capable of permanently recording peak response as a function of frequency for both horizontal motions and vertical
motion, is provided at each of the following locations:
- a. Reactor Building foundation mat ou tside the Reactor Building (see Figure 3.7-47).
- b. Steam generator support (see Figure 3.7-51).
- c. On the Intermediate Bu ilding roof at elevation 463'-0" (see Figure 3.7-52).
- d. On the foundation of the Auxiliary Building at el evation 374'-0" (see Figure 3.7-53).
Each triaxial response spectrum recorder will record 12 frequencies, 1/3 of an octave apart, beginning with 2 Hz and ending with 25.4 Hz.
- 3. Triaxial Seismic Switch
A triaxial seismic switch, located at the Reactor Building foundation mat, is provided to actuate an alarm in the control room (see Figure 3.7-47).
The seismic monitoring instrumentat ion shown in Table 3.7-11 should be maintained with a high level of availability. Each of these instruments shall be demonstrated operable by the performanc e of the channel check, channel 3.7-44 Reformatted July 2014 calibration, and analog channel operational test as established in the appropriate test procedures.
Each of the above seismic monitoring in struments actuated during a seismic event greater than or equal to 0.01g shall be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a channel calibration performed within 5 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission within 10 days describin g the magnitude, frequency spectrum, and resultant effect upon facility f eatures important to safety.
The availability of the seismic instrument ation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the
response of those features important to safe ty. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required.
Criteria for the selection of types and locations of seis mic instrumentation are in accordance with Regulatory Guide 1.12 (see Appendix 3A). Where multiple
locations for a particular instrument are possible, the location selected is based upon analytical results which show that an amplified response is expected at the selected location. If an earthquake o ccurs, the recorded responses of the previously discussed seismic instrumentation, except for the triaxial seismic switch, are compared to calculated responses as discussed in Section 3.7.4.4.
For instruments using a plate as a m ounting adapter, Table 3.
7-10 presents the calculated lowest natural frequency (of three dimensions) of each mechanical system monitored (adapter plate, plus inst rument). For instruments not using a plate adapter, the instruments are rigidly bolted to the concrete surface as a mounting accordance with the instru ment manufacturer's instructions.
Instrument assemblies are specified and designed to be free of spurious resonances within the frequen cy range of the instrument.
3.7.4.3 Control Room Operator Notification Control room signals available to the operator are as follows:
- 1. Indication and an audible alar m are actuated when the triaxial seismic switch at the Reactor Building foundation signals that the OBE peak ground acceleration has been exceeded in either of the horizontal directions or in the vertical direction.
- 2. Indication and a common alarm are act uated when any of the 12 elements of each triaxial section of the triaxial response spectrum recorder at the Reactor Building foundation mat exceeds the frequency setpoi nt. Two setpoints are provided for each element. Exceeding the first setpoint illuminates a yellow light at an 3.7-45 Reformatted July 2014 acceleration equivalent to 2/3 OBE desi gn response spectra. A red light is illuminated if acceleration exceeds the OBE design resp onse spectra.
- 3. Indication and an audible alarm are actuated when the accelerometer at the Reactor Building foundation mat detects a cceleration greater than 0.01g in either horizontal direction or great er than 0.0067g in the vertical direction signifying that the triaxial time-history accelerometer recording system has started. The recording system is located in the relay room below the control room.
3.7.4.4 Comparison of M easured and Predicted Responses In the event of an earthquake, the control room supervision determines whether or not
there are initial indications t hat the OBE acceleration level has been exceeded. This is accomplished by inspection of the indications and alarms described in Section 3.7.4.3.
In approximately 15 minutes after the occu rrence of the earthquak e, data from the triaxial time history accelerometer located on the Reactor Building foundation mat will be analyzed and retrieved from the relay room recording/analysis system for use by the control room supervision in assessing whether the OBE design of the plant has been exceeded. This is accomplished as follows:
- 1. The triaxial time history accelerometer on the Reactor Building foundation mat records acceleration as a function of time. The data is recorded in solid-state and then processed via computer analysis to pr oduce:(a) comparison s of the recorded earthquake to the foundation OBE design for each directional component, (b) a calculation of the Cumulative Absolute Velocity as compared to a threshold criterion for each directional component, and (c) a summary determination of exceedance of the plant OBE design basis. The summary position on exceedance of the plant OBE design can then be used by the control room supervision to determine if shut do wn of the plant is warranted.
- 2. Plant operator walkdowns will also commence after the occurrence of an earthquake to evaluate any unusual plant conditions which might exist, such as, pump or valve leakages, excess equipment vi brations or deforma tions, structural cracking, fallen objects, etc. Reports on t he extent of damage, or lack thereof, will be used by the control room supervision in conjunction with the seismic
instrumentation results and other control ro om indications to establish a position on whether the plant OBE design has been exceeded and if shut down is required.
This decision should be made within approximately eight (8) hours of the earthquake occurrence
- 3. If the decision to shut down the plant is made, it should be conducted in a controlled process using ex isting plant procedures.
- 4. As part of the evaluation of the effect s of the earthquake, the measured responses from the other sensors loca ted throughout the plant will be compared to the design response for their respective locations. These comparisons will be used to 3.7-46 Reformatted July 2014 evaluate the overall impact of the earthquake as it relates to the design of the plant and components.
3.7.5 SEISMIC DESIGN CONTROL 3.7.5.1 Balance of Plant Scope Safety class components and equipment are designed using floor response spectra as input with the exception of certain instruments, generally locally mounted devices, which are procured using fixed values of 1.5g (wall mounted devices) and 3.0g (pipe mounted devices) as conservative values for multiple locations. Vendors are responsible for the design and qualification by analysis or test of components and equipment within their scope. This responsibility includes review of the design, analyses, or tests by one or more qualified engineers ot her than the engineer(s) who originated the design, performed the analyses, or developed the tests.
Safety class piping systems are also design ed using floor response spectra as input. Gilbert is responsible for design and qualification of piping systems by analysis during construction (see Section 3.7.3). This resp onsibility includes review of the design and analyses by one or more qualified engineers ot her than the engineer(s) who originated the design or performed the analyses. During operations, VCSNS staff is responsible for design and qualification of BOP pipi ng sysgtems by analyses which may be performed by various contractors per VCSNS piping design guides or their predecessors.
The snubber vendor maintains acceptance and qualification test reports on file for each type of snubber which has been used as a pi ping restraint for the Virgil C. Summer Nuclear Station.
The loading conditions and transients analyz ed are described for each system in the piping system design specific ation. The analytical procedure described in Section 3.7.3.8 is followed and the re sults are compared to the applicable design Section of the ASME Code or other controlling design requirements.
For certain small diameter piping (2 in ch and smaller), the seismic design is implemented by the application of simplified seismic support criteria which are subjected
to independent review.
3.7.5.2 Components and Equipment Provided by the NSSS Vendor The following procedure is implemented for Westinghouse supplied safety-related mechanical equipment that falls within one of the many categories which have been analyzed as described in Sections 3.7.2 and 3.7.3 and has been shown to be rigid with all natural frequencies greater than or equal to 33 Hz.
02-01 3.7-47 Reformatted July 2014 1. Equivalent static acceleration factors fo r the horizontal and vert ical directions are included in the equipment s pecification. The vendor must certify the adequacy of the equipment to meet the seismic require ments as described in Section 3.7.3.
- 2. When the floor response spectra are developed the c ognizant engineer responsible for the particu lar component checks to ensure that the acceleration factors are less than those gi ven in the equipment specification. If accelerations exceed those in the equipment specifications , the designs are rechecked to verify equipment adequacy.
All other Westinghouse supplied safety-related equipment is analyzed or tested as
described in Sections 3.
7.2, 3.7.3, and 3.10.
3.
7.6 REFERENCES
- 1. Newmark, N. M., "Earthquake Response Analysis of Reactor Structures," First International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, September, 1971.
- 2. Roesset, J. M., Whitman, R.
V. and Dobry, R., "Modal Analysis for Structures with Foundation Interaction," Journal of the Structural Division, Proceedings of the
American Society of Civil Engineers, March, 1973.
- 3. "Damping Values of Nuclear Plant Components," WCAP-7921-AR, May, 1974.
- 4. Gesinski, T. L. and Chiang, D., "Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP- 8236 (Proprietary), December, 1973 and WCAP-8288 (Non-Proprietary), January, 1974.
- 5. ICES STRUDL, "The Structural Design Language," User's Manual, Volumes I, II, and III; Massachusetts Institute of Technol ogy, Department of Civil Engineering.
- 6. Nigam, N. C. and Jennings, P. C., "Calculation of Response Spectra from Strong Motion Earthquake Records," Bulletin of Seis mological Society of America, Vol. 59, No. 2, April, 1969.
- 7. ICES DYNAL, "User's Manual," Mc Donnell Douglas Automation Company, September 1, 1971.
- 8. Lysmer, J., Seed, H. B., Udaka, T., Hwang, R. N. and Tsai, C.
F., "Efficient Finite Element Analysis of Seismic Structure -
Soil- Structure Inte raction," Earthquake Engineering Research Center, University of California, Berkeley, California, Report No. EERC 75-34, November, 1975.
- 9. Wiegal, R. L., ed., Earthquake Engineeri ng, Chapter 14, "Soil-Pile Foundation Interaction," by Penzien, J., Prentice-Hall, 1970.
3.7-48 Reformatted July 2014 10. Lin, C. W., "How to Lump the Masses - A Guide to the Piping Seismic Analysis," ASME paper 74-NE-7 present ed at the Pressure Vessels and Piping Conference, Miami, Florida, June, 1974.
- 11. Biggs, J. M., Introduction to Structural Dynamics, McGraw-Hill, New York, 1964.
- 12. Thomas, T. H., et al., "Nuclear Reac tors and Earthquakes," TI D-7024, U.S. Atomic Energy Commission, Washington, D.C., August, 1963.
- 13. Biggs, J. M., et al., Structural Design for Dynamic Loads, Chapter 8, McGraw-Hill, New York, 1950.
- 14. Tsai, N. C., Neihoff, D., Swatta, M. and Hadjian, A. H., "The Use of Frequency-Independent Soil-Stru cture Interaction Pa rameters," Nuclear Engineering and Design, Vo
- l. 31, pp. 168-183, 1974.
- 15. Rosset, J. M., Dobry, R.
and Whitman, R. V., "Modal A nalysis for Structures with Foundation Interaction," American Society of Civil Engineers National Structural
Engineering Meeting, Cleveland, Ohio, April 24-28, 1972; Meeting Preprint 1688.
- 16. Biggs, J. M. and Whitman, R. V., "So il-Structure Interaction in Nuclear Power Plants," Proceedings of the Third Japanes e Earthquake Engineering Symposium, Tokyo, Japan, 1970.
- 17. Gesinski, T. L., "Fuel Assembly Safe ty Analysis for Combined Seismic and Loss of Coolant Accident," WCAP-7950, July, 1972.
- 18. "Safety Analysis of the 8-Grid 17 x 17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8326, Addendum 1 (Proprietary), March, 1974 and WCAP-8288, Addendum 1 (Non-Proprietary), April, 1974.
- 19. American National Standards Institute, "Earthquake Instrumentation Criteria for Nuclear Power Plants," ANSI N18.5-1974.
- 20. Newmark, N. M., Blume, J. A., and K apur, K. K., "Design Response Spectra for Nuclear Power Plants," Presented at t he American Society of Civil Engineers National Structural Engineering Meeting, San Francisco , California, April 9-13, 1973.
- 21. Yeh, C. K., "Seismic Analysis of Slender Buried Beams," Bulletin of the Seismological Society of America, Vol. 64, No. 5, October, 1974.
- 22. Shah, H. H., and Chu, S. L., "Seismic Analysis of Underground Structural Elements," Journal of the Power Division, Proceedings of the American Society of Civil Engineers, Vol. 100, No. P01, July, 1974.
3.7-49 Reformatted July 2014 23. Iqbal, M. A., and Goodling, E. C., "Seismic Design of Buried Piping," 2nd American Society of Civil Engineers Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, LA, December, 1975.
- 24. American Society of Mechanical Engi neers, Boiler and Pressure Vessel Code, Section III, SubSection NC 3650.
- 25. Crandall, S. H., and Mark , W. D., "Random Vibration in Mechanical Systems," Academic Press, New York, 1963.
- 26. Seismic Confirmatory Program, Equipment Margin Study, V. C. Summer Nuclear Station Unit 1, OL No. NPF-12 November 1983.
3.7-50 Reformatted July 2014 TABLE 3.7-0 FREQUENCY INTERVALS USED FOR CALCULATION OF RESPONSE SPECTRA Frequency Range (Hz) Increments (Hz) 0.2 - 3.0 0.0622 3.0 - 3.6 0.1425 3.6 - 5.0 0.1857 5.0 - 8.0 0.20 8.0 - 15.0 0.3532 15.0 - 18.0 0.5414 18.0 - 22.0 0.80 22.0 - 34.0 1.450
3.7-51Reformatted PerAmendment 02-01TABLE 3.7-1DAMPING FACTORSPERCENT OF CRITICAL DAMPING (1)Regulatory Guide 1.61 ValueComponent or StructureValue Used OBE & SSE OBE SSEReactor Building 2.0 2.0 5.0Concrete Support Structures Inside theReactor Building 2.0 4.0 7.0Assemblies & StructuresBolted & Riveted 2.5 4.0 7.0Welded 2.0 2.0 4.0Vital Piping SystemsLarger than 12 Inch Diameter2.0/3.0 (2)2.0 3.012 Inch Diameter and Smaller1.0/2.0 (2)1.0 2.0NOTE:(1)See Reference [1](2)Code Class N-411, "Alternative Damping Values for Seismic Analysis of Class 1, 2 and 3 Piping Sections, Section III,Division 1," is acceptable for piping analyses for systems supported on all building structures per Regulatory Guide1.84, Rev. 25.
02-01 3.7-52Reformatted PerAmendment 02-01TABLE 3.7-2DAMPING FACTORSPERCENT OF CRITICAL DAMPINGFOR STRUCTURES WITH SOIL INTERACTIONRange of ShearWave VelocityDamping Factor On Rock C 6000 fps2-5On Firm Soil C 2000 fps5-7On Soft Soil C 2000 fps7-10NOTE: See reference [1].
3.7-53AMENDMENT 97-01AUGUST 1997TABLE 3.7-3DAMPING VALUES USED FOR SEISMIC SYSTEMSANALYSIS FOR WESTINGHOUSE SUPPLIED EQUIPMENTDamping (Percent of Critical)
Item UpsetConditions(OBE)FaultedCondition(SSE, DBA)Primary Coolant Loop System - Components and Large Piping (1)2 (2)4 (2)Small Piping 1 (2)2 (2)Welded Steel; Structures 2 4Bolted and/or Riveted Steel Structur es 4 7(1)Generally applicable to 12 inch or larger diameter piping.(2)Code Case N-411, "Alternative Damping Values for Seismic Analysis of classes 1, 2and 3 Piping Sections, Section III, Division 1," is acceptable for piping analyses forsystems supported on all building structures per Regulatory Guide 1.84, Rev. 25.
3.7-54AMENDMENT 97-01AUGUST 1997TABLE 3.7-4FOUNDATION SEISMIC DESIGN PARAMETERSFoundationMaterialCompressionalWave Velocity(ft/sec)In-SituDensity(lbs/cu ft)Poisson'sRatio ()Shear Modulus (G) orModulus of Rigidity(lbs/sq in)Modulus of SubgradeReaction (1) (lbs/cu ft)Saprolite1000-3000110-135 0.351 x 10 4 to 3.5 x 10 45 x 10 6 BLWeathered andjointed rock(12,000-13,000)
(2)(140-160)(2)0.305 x 10 52 x 10 8 BLSound rock 15,000 165 0.202 x 10 68 x 10 8 BL(1)"BL" is contact area of square foundation. For other contact area shapes, subgrade modulus must be modified (see Barkan, 1962).(2)Numbers in parentheses are estimated values.NOTE: Young's Modulus, E = 2(1 + )G.
3.7-55AMENDMENT 97-01AUGUST 1997TABLE 3.7-5FOUNDATION ELEVATION AND FOUNDATION TYPE SEISMIC CATEGORY I STRUCTURES StructureFoundationElevation (ft)FoundationTypeReactor BuildingNorth 341 (1) to 396Fill Concrete East 348 (1) to 396Fill Concrete South 362 (1) to 396Fill ConcreteWest 367 (1) to 396Fill ConcreteEntire Structure 396 to 408MatControl BuildingNorth 366 (1) to 407Fill Concrete East 371 (1) to 407Fill Concrete South 366 (1) to 407Fill ConcreteWest 371 (1) to 407Fill ConcreteEntire Structure 407 to 411MatDiesel Generator Building Entire StructureTable 3.7-6CaissonsService Water Pumphouse 386MatService Water Intake Structure 367MatService Water Discharge Structure 408MatAuxiliary BuildingNorth 354 (1) to 384Fill Concrete South 368 (1) to 370Fill ConcreteNorth 384 to 388Mat South 370 to 374MatFuel Handling BuildingEntire StructureTable 3.7-6CaissonsIntermediate BuildingEntire StructureTable 3.7-6Caissons(1) Bottom elevation determined in field to suit actual condition.
3.7-56AMENDMENT 97-01AUGUST 1997TABLE 3.7-6FOUNDATION DATA FOR SAFETY CLASS STRUCTURES SUPPORTED ON CAISSONS StructureElevation of Cap (ft)Underlying SoilElevation in No. 3Rock for Seismic InputDiesel Gen. Building 394 to 421 Zone I, II, III fill and silty fine sand, medium
dense to dense 369 to 363Fuel HandlingBuildin g 409 to 430 Zone I, II, III fill and silty fine sand, medium
dense to dense 345Intermedi ateBuildin g 394 to 409 Zone I, II, III fill and silty fine sand, medium
dense to dense 375 3.7-57Reformatted PerAmendment 02-01TABLE 3.7-7DIMENSIONS OF SEISMIC CATEGORY I STRUCTURES StructurePlanDimensions (ft)Height (ft)EmbedmentDepth (ft)TopElevation (ft)Reactor Building134, OD 206 39 602Control Building84 x 141 98 28 505Diesel GeneratorBuildin g66 x 6783 and 5641 and 14 477Service Water Pumphouse 70 x 79 73 49 459Service Water Intake Structure166 x 1821 and 30--388 and 396.5Service Water Discharg e Structure35 x 336 to 15 15 425Auxiliary Building120 x 190------North--127 51 511South--141 65 511Fuel Handling Buildin g75 x 12395 and 10226 and 5 511Intermediate Buildin g85 x 19876 and 7839 and 26 485 3.7-58Reformatted PerAmendment 02-01TABLE 3.7-7aRESPONSE LOADS FOR SEISMIC CATEGORY I STRUCTURESAcceleration (g)Displacement (in)Elevation (Ft)Mass(Kip-Sec 2/Ft)X YVert.X YVert.Service Water Pumphouse390'184.9380.2000.1600.1610.7650.6120.063425'211.7360.2350.1880.1730.8990.7190.068436'107.6930.2620.2090.1751.0020.7990.069441'55.4160.2760.2200.1861.0550.8410.073459'82.3280.3270.2610.1901.2500.9980.074Control Building412'257.930.1440.1680.0780.0150.0270.010425'141.150.1830.2120.0830.0280.0470.010 436120.460.2080.2600.1100.0350.0610.017448'95.910.2630.3090.1240.0550.0740.019463'150.390.3280.3650.1300.0780.0870.019482'191.860.4290.4330.1370.1000.1030.020505'196.970.5010.4900.1420.1170.1140.020Auxiliary Building374'490.550.1170.1170.0820.0110.0090.014388'679.640.1340.1290.0980.0190.0160.012397'227.160.1500.2030.0890.0280.0380.015412'1086.120.2320.2680.1060.0550.0570.015436'900.960.3380.3630.1090.0850.0830.015463'658.050.5010.4440.1160.1370.1040.016485'390.810.5540.4850.1180.1540.1140.016511'50.150.6500.4980.1200.1750.1200.016Intermediate Building412'398.450.1800.3080.2090.0310.0720.025436'432.540.2740.4610.2220.0430.1110.026463'368.400.3570.5740.2350.0650.1360.027485'106.240.3720.5410.2420.0650.1400.027Reactor Building Shell435'203.540.1610.1570.1570.0400.0400.005462'180.060.2020.2000.1950.0790.0790.008481'148.760.2430.2420.2200.1090.1090.009500'173.730.2960.2950.2420.1390.1390.010 3.7-59Reformatted PerAmendment 02-01TABLE 3.7-7a (Continued)RESPONSE LOADS FOR SEISMIC CATEGORY I STRUCTURESAcceleration (g)Displacement (in)Elevation (Ft)Mass(Kip-Sec 2/Ft)X YVert.X YVert.Reactor Building Shell (Cont)523'180.060.3500.3500.2670.1760.1760.012546'153.110.3970.3990.2880.2110.2120.013582'517.950.4760.4750.3010.2570.2580.014Interior Concrete407'468.510.1180.1140.1120.0050.0050.002427'144.290.1680.1620.1240.0120.0110.003431'59.240.1800.1700.1260.0140.0130.003435'136.400.1920.1830.1290.0160.0160.003439'47.110.2010.1950.1300.0180.0170.003445'109.500.2190.2210.1310.0210.0200.003462'330.750.2980.3040.1340.0280.0270.004475'64.350.3420.3470.1340.0320.0300.004Diesel Generator Building427'127.70.4660.4700.1070.2280.2160.006436'117.20.4740.4780.1110.2640.2400.006463'84.10.5200.5140.1110.3360.3480.006476'67.70.8411.4200.0900.3840.3840.006Fuel Handling Building436'882.820.3030.3200.2300.1010.1300.023463'6.780.3040.3280.2300.2090.2790.036495'8.850.8460.9740.2250.3370.4560.051512'6.180.8140.8840.2250.4050.5500.059 3.7-60Reformatted PerAmendment 02-01TABLE 3.7-7bRESULTS OF TRANSIENT ANALYSIS - TYPICAL TORSIONAL MODELTransient Analysis Sum X-EQSystem Accelerations (Total)TranslationBending ResponseTorsional ResponseBending ResponseJoint X Y Z X Y Z 7010.174155E 020.968478E 01
-0.216468E 000.958479E 010.19445 1E 010.240496E 01-0.339760E-020.232497E 010.317496E-010.216497E 01-0.128979E-010.950479E 0170102 0.151535E 020.968478E 01
-0.297819E 000.822482E 01-0.166849E 010.682485E 01-0.339760E-020.232497E 010.317496E-010.216497E 01-0.128979E-010.950479E 0170103 0.151535E 020.968478E 01
-0.135537E 010.950479E 01-0.414053E 010.216497E 01-0.339760E-020.232497 E 01 0.317496E-010.216497E 01-0.128979E-010.950479E 0170104 0.192796E 020.968478E 01 0.66 432 9E 000.822482E 01-0.386272E 010.216497E 01-0.339760E-020.232497E 010.317496E-010.216497E 01-0.128979E-010.950479E 0170105 0.192796E 020.968478E 01
-0.795434E 000.992478E 01-0.166849E 010.682485E 01-0.339760E-020.232497E 010.317496E-010.216497E 01-0.128979E-010.950479E 01 7020.14316 1E 020.984478E 01
-0.200542E 000.950479E 01-0.223558E 010.232497E 01-0.339760E-020.232497E 010.270121E-010.216497E 01-0.128978E-010.950479E 0170202 0.125049E 020.984478E 01
-0.299360E 000.822482E 01-0.143435E 010.232497E 01-0.339960E-020.232497E 010.270121E-010.216497E 01-0.128978E-010.950479E 0170203 0.125049E 020.984478E 01
-0.135695E 010.950479E 010.37761 5E 010.224497E 01-0.339760E-020.232497E 010.270121E-010.216497E 01-0.128978E-010.950479E 0170204 0.162128E 020.984478E 01 0.66 278 4E 000.822482E 010.35726 8E 010.224497E 01-0.339760E-020.232497E 010.270121E-010.216497E 01-0.128978E-010.950479E 0170205 0.162128E 020.984478E 01
-0.792828E 000.992478E 01-0.143435E 010.232497E 01-0.339760E-020.232497E 010.270121E-010.216497E 01-0.128978E-010.950479E 01 7030.13084 9E 020.984478E 01
-0.137671E 000.950479E 01-0.197403E 010.232497E 01-0.339755E-020.232497E 010.221927E-010.216497E 01-0.128976E-010.950479E 0170302 0.115773E 020.984478E 01
-0.303186E 000.822482E 01-0.131435E 010.232497E 01-0.339755E-020.232497E 010.221927E-010.215497E 01-0.128976E-010.950479E 0170303 0.115773E 020.984478E 01
-0.135311E 010.950479E 01-0.320794E 010.216497E 01-0.339755E-020.232497E 010.221927E-010.216497E 01-0.128976E-010.950479E 0170304 0.146807E 020.984478E 01 0.65 893 4E 000.822482E 010.30373 0E 010.224497E 01-0.339755E-020.232497E 010.221927E-010.216497E 01-0.128976E-010.950479E 0170305 0.146807E 020.984478E 01
-0.795415E 000.992478E 01-0.131435E 010.232497E 01-0.339755E-020.232497E 010.221927E-010.216497E 01-0.128976E-010.950479E 01 02-01 3.7-61AMENDMENT 97-01AUGUST 1997TABLE 3.7-7cREACTOR BUILDING MODAL PARTICIPATION FACTORSModeParticipation FactorFrequency (cps)*1 (dominant mode) 26.48 4.35*2 22.68 11.47*3 12.18 15.68*4 22.25 29.15*5-8.03 32.96*6 13.69 40.64*7-6.36 45.87 8-0.97 56.10 9-1.26 56.77 10-0.11 66.38 11 0.06 76.14 12-1.01 79.20 13 0.37 111.01 14-1.79 147.12 15-0.66 157.47 16 0.69 208.84 17 0.37 213.11 18 0.01 239.89 19 0.35 312.96 20-0.12 385.35* Significant modes 3.7-62AMENDMENT 97-01AUGUST 1997TABLE 3.7-8GAMMA SCALING FACTORSGamma Scaling FactorReactor Building Shell and Interior Concrete 1.0Control BuildingSee Figure 3.7-44Diesel Generator Building 1.0Fuel Handling Building 1.0Auxiliary BuildingSee Figure 3.7-45Intermediate BuildingSee Figure 3.7-46Service Water Pumphouse 1.0 3.7-63Reformatted PerAmendment 02-01TABLE 3.7-9DAMPING FACTORSPERCENT OF CRITICAL DAMPING (1)Component or StructureWorking Stress, No MoreThan AboutOne-Half Yield PointAt, or Just BelowYield PointBeyond Yield Point withPermanent Strain GreaterThan Yield Point Limit StrainReactor Building 2.0 5.0Not applicableConcrete Support StructuresInside the Reactor Building 2.0 5.0Assemblies L Structuresa.Bolted & Riveted 2.5 5.0b.Welded 2.0 5.0Vital Piping Systems (2)Other Concrete Structures above Ground 2.0 5.0 (1)
Reference:
"Seismic Design Criteria for Nuclear Reactor Facilities" by Nathan M. Newmark and William J. Hall,Proceedings, Fourth World Conference on Earthquake Engineering, January 13 - January 18, 1969, Santiago, Chile.(2)Refer to Table 3.7-1 for damping factors.
3.7-64Reformatted PerAmendment 02-01TABLE 3.7-10SEISMIC INSTRUMENTATION SENSING ELEMENTS (1)Ident. No.Sensing ElementDescriptionLocation FSARFigureReferenceMounting TypeLowest NaturalFrequency ofMounting (2)IYM-1780AccelerometerReactor building,foundation mat
,outside reactor building3.7-47Bolted tofoundation (3)IYM-1782Seismic switchReactor building, foundation mat
,outside reactor building3.7-47Bolted tofoundation (3)IYM-1783Response spectrum recorderwith switchesReactor building, foundation mat
,outside reactor building3.7-47Bolted tofoundation (3)IYM-1784AccelerometerReactor building, top of ring girder,outside reactor building3.7-48Bolted to ring girder concrete (3) 3.7-65Reformatted PerAmendment 02-01TABLE 3.7-10 (Continued)SEISMIC INSTRUMENTATION SENSING ELEMENTS (1)Ident. No.Sensing ElementDescriptionLocation FSARFigureReferenceMounting TypeLowest NaturalFrequency ofMounting (2)IYM-1785Responsespectrum recorderSteam generator C, upper lateral support,inside Reactor Building3.7-51Bolted to adapterplate30.9 HzIYM-1786Response spectrum recorderIntermediate Building roof, elevation 4 63'3.7-52Bolted toconcrete floor (3)IYM-1787Response spectrum recorderAuxiliary Building foundation elevation 374'3.7-53Bolted tofoundation NOTES:1.All listed sensing elements are triaxial sensors.2.Lowest natural frequency, in three dimensions, of mounting plate.
3.This instrument does not use a plate as a mounting adapter. The instrument is bolted rigidly to the concrete surfacethat it monitors. The instrument is free from spurious resonances within its frequency range.
02-01 3.7-66Reformatted PerAmendment 02-01TABLE 3.7-11SEISMIC MONITORING INSTRUMENTATIONINSTRUMENTS AND SENSORLOCATIONSMEASUREMENTRANGEMINIMUMINSTRUMENTSOPERABLE 1.Triaxial Time-History AccelerographsSystem, including the followingcomponents:a.IYM-1780 Reactor Building FoundationMat Accelerometer/Trigger0.1 to 40 Hz0.01 to 1.0g 1b.IYM-1784 Reactor Building Ring GirderAccelerometer0.1 to 40 Hz0.01 to 1.0g 1 2.Triaxial Seismic Switcha.IYM-1782 Reactor Building FoundationMat0.1 to 30 Hz0.01 to 0.25 g 13.Triaxial Response-Spectrum Recorders a.IYM-1783 - Reactor Building FoundationMat(1)1b.IYM-1785 - Steam Generator Support(1)1c.IYM-1786 - Intermediate Bldg.,Elev. 463'(1)1 d.IYM-1787 - Auxiliary Bldg. Foundation(1)1 With control room indication and/or alarm.(1)Range varies for the multiple elements of the instrument, i.e., 1.6g at 2 Hz, 10g at5 Hz, 34g at 10 Hz, 12g at 16 Hz.
Figure 3.7-49 Delet ed by Amendment 96-03
Amendment 96-03
September 1996
Figure 3.7-50 Delet ed by Amendment 96-03
Amendment 96-03
September 1996
""""""""""""""
3.8-141Reformatted PerAmendment 02-01TABLE 3.8-5TOLERANCES FOR LOCAL BULGES, FLATSPOTS OR DISCONTINUITIES INCYLINDRICAL PORTION OF REACTOR BUILDING LINERChord LengthCenterlineTheoreticalOffset (in)Tolerance Measuredin One Plate MoreThan 12 Inches fromWeld (in)Tolerance Measuredacross Weld (in)10' - 0"2-3/8 7/16 5/815' - 0"5-1/2 1 1-1/220' - 0" 9-5/8 1-3/4 2-5/8 3.8-142Reformatted PerAmendment 02-01TABLE 3.8-6DISPLACEMENT MEASUREMENT LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTCylinder Base and Cylinder Wall Radial Displacement - Direct Current DisplacementTransducer (DCDT) LocationsInstrumentNo. & GageElevationAzimuthNotesDCDT 1, 2, 3420'-0" 59, 101, 162-30'Radial displacementDCDT 4, 5, 6420'-0" 243-20', 308, 34 7Radial displacementDCDT 7, 8, 9483'-0" 243-20', 308, 34 7Radial displacementCylinder Wall and Dome Radial Displacement - Scale and Jig Transit LocationsScale 10, 11483'-0" 59, 162-30'Radial displacementScale 12, 13, 14557'-0" 59, 101, 162-30'Radial displacementScale 15, 16, 17557'-0" 243-20', 308, 34 7Radial displacementScale 18, 19, 20576'-5" 59, 101, 162-30'Radial displacementScale 21, 22, 23576'-5" 243-20', 308, 34 7Radial displacementCylinder Ledge Vertical Displacement - Invar Tape and DCDT LocationsDCDT 30, 31, 32576'-5" 59, 101, 162-30'Vertical displacementDCDT 33, 34, 35576'-5" 243-20', 308, 34 7Vertical displacementDome Apex Vertical Displacement Invar Tape and DCDT LocationsDCDT 36599'-0"-Vertical displacementDCDT 37, 40, 43, 46, 49, 52472'-6"-Vertical displacementDCDT 38, 41, 44, 47, 50, 53472'-6"-Radial displacementDCDT 39, 42, 45, 48, 51, 54472'-6"-Tangential displacement 02-01 02-01 3.8-143Reformatted PerAmendment 02-01TABLE 3.8-6 (Continued)DISPLACEMENT MEASUREMENT LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTEquipment Hatch Displacement - (DCDT) LocationsInstrumentNo. & GageElevationAzimuthNotesDCDT 55492'-6" 101Vertical displacementDCDT 58486'-6" 101Vertical displacementDCDT 61480'-6" 101Vertical displacementDCDT 64464'-6" 101Vertical displacementDCDT 67458'-6" 101Vertical displacementDCDT 70452'-6" 101Vertical displacementDCDT 56492'-6" 101Radial displacementDCDT 59486'-6" 101Radial displacementDCDT 62480'-6" 101Radial displacementDCDT 65464'-6" 101Radial displacementDCDT 68458'-6" 101Radial displacementDCDT 71452'-6" 101Radial displacementDCDT 57492'-6" 101Tangential displacementDCDT 60486'-6" 101Tangential displacementDCDT 63480'-6" 101Tangential displacementDCDT 66464'-6" 101Tangential displacementDCDT 69458'-6" 101Tangential displacementDCDT 72452'-6" 101Tangential displacement 02-01 3.8-144Reformatted PerAmendment 02-01TABLE 3.8-7STRAIN GAGE LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTCylinder Wall and Base Junction - Strain Gage and Rebar LocationsInstrumentNo. & GageElevationAzimuthNotesPrimary Strain Gage
73, 75, 77410'-6" 8Meridional strainRedund. Strain Gage 1073, 1075, 1 077410'-6" 8Meridional strainPrimary Strain Gage
74, 76, 78410'-6" 8Hoop strainRedund. Strain Gage 1074, 1076, 1 078410'-6" 8Hoop strainPrimary Strain Gage
79, 81, 83412'-6" 8Meridional strainRedund. Strain Gage 1079, 1081, 1 083412'-6" 8Meridional strainPrimary Strain Gage
80, 82, 84412'-6" 8Hoop strainRedund. Strain Gage 1080, 1082, 1 084412'-6" 8Hoop strainCylinder Wall and Dome Junction - Strain Gage and Rebar LocationsPrimary Strain Gage
85, 87, 89553'-6" 8Meridional strainRedund. Strain Gage 1085, 1087, 1 089553'-6" 8Meridional strainPrimary Strain Gage
86, 88, 90553'-6" 8Hoop strainRedund. Strain Gage 1086, 1088, 1 090553'-6" 8Hoop strainPrimary Strain Gage
91, 93, 95557'-0" 8Meridional strain 02-01 3.8-145Reformatted PerAmendment 02-01TABLE 3.8-7 (Continued)STRAIN GAGE LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTCylinder Wall and Base Junction - Strain Gage and Rebar LocationsInstrumentNo. & GageElevationAzimuthNotesRedund. Strain Gage 1091, 1093, 1095557'-0" 8Meridional strainPrimary Strain Gage
92, 94, 96557'-0" 8Hoop strainRedund. Strain Gage 1092, 1094, 1 096557'-0" 8Hoop strainEquipment Access Opening - Strain Gage and Rebar LocationsPrimary Strain Gage 97, 99, 101, 103, 105, 107, 109, 111, 113, 115, 117, 119472'-6"-Meridional strainRedund. Strain Gage 1097, 1099, 1101, 1103,1105, 1107, 1109, 1111, 1113, 1115, 1117, 1119472'-6"-Meridional strainPrimary Strain Gage 98, 100, 102, 104
,106, 108, 110, 112, 114, 116, 118, 120472'-6"-Hoop strainRedund. Strain Gage 1098, 1100, 1102, 1104,1106, 1108, 1110, 1112, 1114, 1116, 1118, 1120472'-6"-Hoop strainPrimary Strain Gage 121, 123, 125 488'-6" 101Meridional strainRedund. Strain Gage 1121, 1123, 1 125488'-6" 101Meridional strain 3.8-146Reformatted PerAmendment 02-01TABLE 3.8-7 (Continued)STRAIN GAGE LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTInstrumentNo. & GageElevationAzimuthNotesMain Strain Gage 122, 124, 126488'-6" 101Hoop strainRedund. Strain Gage, 1122, 1124, 1 126488'-6" 101Hoop strainMain Strain Gage, 127, 129, 131 484'-10" 101Meridional strainRedund. Strain Gage 1127, 1129, 1 131484'-10" 101Meridional strainMain Strain Gage 128, 130, 132 484'-10" 101Hoop strainRedund. Strain Gage 1128, 1130, 1 132484'-10" 101Hoop strainMain Strain Gage 133, 135, 137 460'-9" 101Meridional strainRedund. Strain Gage 1133, 1135, 1 137460'-9" 101Meridional strainMain Strain Gage 134, 136, 138 460'-9" 101Hoop strainRedund. Strain Gage 1134, 1136, 1 138460'-9" 101Hoop strainMain Strain Gage 139, 141, 143 454'-9" 101Meridional strainRedund. Strain Gage 1139, 1141, 1 143454'-9" 101Meridional strainMain Strain Gage 140, 142, 144 454'-9" 101Hoop strainRedund. Strain Gage 1140, 1142, 1 144454'-9" 101Hoop strain 3.8-147Reformatted PerAmendment 02-01TABLE 3.8-7 (Continued)STRAIN GAGE LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTCylinder Wall - Strain Gage and Rebar LocationsInstrumentNo. & GageElevationAzimuthNotesMain Strain Gage 165, 167, 169420'-6" 8Meridional strainRedund. Strain Gage 1165, 1167, 1 169420'-6" 8Meridional strainMain Strain Gage 166, 168, 170 420'-6" 8Hoop strainRedund. Strain Gage 1166, 1168, 1 170420'-6" 8Hoop strainMain Strain Gage 171, 173, 175 450'-6" 8Meridional strainRedund. Strain Gage 1171, 1173, 1 175450'-6" 8Meridional strainMain Strain Gage 172, 174, 176 450'-6" 8Hoop strainRedund. Strain Gage 1172, 1174, 1 176450'-6" 8Hoop strainMain Strain Gage 177, 179, 181 483'-0" 8Meridional strainRedund. Strain Gage 1177, 1179, 1 181483'-0" 8Meridional strainMain Strain Gage 178, 180, 182 483'-0" 8Hoop strainRedund. Strain Gage 1178, 1180, 1 182483'-0" 8Hoop strainMain Strain Gage 183, 185, 187 520'-6" 8Meridional strain 3.8-148Reformatted PerAmendment 02-01TABLE 3.8-7 (Continued)STRAIN GAGE LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTInstrumentNo. & GageElevationAzimuthNotesRedund. Strain Gage 1183, 1185, 1187520'-6" 8Meridional strainMain Strain Gage 184, 186, 188 520'-6" 8Hoop strainRedund. Strain Gage 1184, 1186, 1 188520'-6" 8Hoop strainUnder Vertical Tendon Anchorage at 7 32' 20"Primary Strain Gage 189Refer to Figure 3.8-33VerticalRedund. Strain Gage 1189Refer to Figure 3.8-33VerticalPrimary Strain Gage 190Refer to Figure 3.8-33HorizontalRedund. Strain Gage 1190Refer to Figure 3.8-33HorizontalPrimary Strain Gage 191Refer to Figure 3.8-33 45Redund. Strain Gage 1191Refer to Figure 3.8-33 45Adjustment in location of instruments and gages may be made in the field to clear interferen ces. 02-01 3.8-149Reformatted PerAmendment 02-01TABLE 3.8-8CRACK PATTERN AREA LOCATIONS FOR REACTOR BUILDINGSTRUCTURAL ACCEPTANCE TESTArea No.Center LineElevationCenter Line AzimuthDimensions 145~563'-9-1/2" 2667'-0" x ~41'-0" 146488'-6" 2667'-0" x 7'-0" 147415'-6" 2667'-0" x 7'-0" 148483'-0"302-27'7'-0" x ~12'-2" 149485'-0"111-42'25'-10" x 27'-2"NOTE:Adjustment in location of whitewash areas may be made in the field to clearinterferen ces.
3.8-150Reformatted PerAmendment 02-01TABLE 3.8-9THERMOCOUPLE LOCATIONS FOR REACTOR BUILDING STRUCTURALACCEPTANCE TESTThermocouple No.ElevationAzimuth150, 153412'-0" 59, 308151, 152, 154416'-0" 162-30', 243-20', 347155, 156, 157, 158, 159~483'-0" 59, 162-30', 243-20', 308, 347160, 161, 162, 163, 164576'-0" 59, 162-30', 243-20', 308, 347NOTE:Adjustment in the location of thermocouples may be made in the field to clearinterferen ces. 02-01 02-01 3.8-151AMENDMENT 97-01AUGUST 1997TABLE 3.8-10LOAD COMBINATIONS FOR INTERNAL CONCRETE STRUCTURES 1.Load Combinations for Service Load Conditions a.U = 1.4D + 1.7L b.U = 1.4D + 1.7L + 1.9E c.U = (0.75) (1.4D + 1.7L + 1.7T o + 1.7R o)d.U = (0.75) (1.4D + 1.7L + 1.9E + 1.7T o + 1.7R o)e.U = 1.2D + 1.9E f.U = 1.4D + 1.7L + 1.4F 2.Load Combinations for Factored Load Conditions a.U = D + L + T o + R o + E'b.U = D + L + T o + R o + 1.5P o c.U = D + L + T o + R o + 1.25P o + 1.0 (Y r + Y j + Y m) + 1.25E d.U = D + L + T o + R o + 1.0P a + 1.0 (Y r + Y j + Y m) + 1.0E'"U" is the section strength required to resist the loads based upon the strength designmethods of ACI 318-71.
3.8-152AMENDMENT 97-01AUGUST 1997TABLE 3.8-11LOAD COMBINATIONS FOR INTERNAL STRUCTURAL STEEL STRUCTURES 1.Load Combinations for Service Load Conditions a.S = D + L b.S = D + L + E c.1.5S = D + L + R o + T o + E 2.Load Combinations for Factored Load Conditions a.1.6S = D + L + T o + R o + E'b.1.6S = D + L + T a + R a + P a c.1.6S = D + L + T a + R a + P a + Y r + Y j + Y m + E d.1.7S = D + L + T a + R a + P a + Y r + Y j + Y m + E'"S" is the required section strength based upon the elastic stress design method andallowable stresses defined in AISC, "Specification for the Design, Fabrication andErection of Structural Steel for Buildings," 1969 Edition.
3.8-153Reformatted PerAmendment 02-01TABLE 3.8-11aPOLYVINYL CHLORIDE WATERSTOPSBuildingElevationExpected Normal OperationRadiation LevelAuxiliary Building (1)386'-0" 10 4 to 10 6 Rad, TID (2), 40 yr410'-0" 10 6 Rad, TID, 40 yr412'-0" 10 4 to 10 6 Rad, TID, 40 yrControl Building (1)411'-0"NegligibleIntermediateBuildin g (1)407'-0" 411'-0" 10 4 Rad, TID, 40 yr 10 4 Rad, TID, 40 yr412'-0" 10 6 Rad, TID, 40 yrFuel Handling Buildin g (1)412'-9" 416'-8" 10 4 to 10 6 Rad, TID, 40 yr 10 4 to 10 6 Rad, TID, 40 yrDiesel Generator Buildin g (1)400'-0" 421'-0"Negligible Negligib leReactor Building (3)387'-0" to 387'-8" 6.5 x 10 3 Rad, TID, 40 yr388'-0" 6.5 x 10 3 Rad, TID, 40 yr394'-10" 6.5 x 10 3 Rad, TID, 40 yr396'-0" 6.5 x 10 3 Rad, TID, 40 yrNOTES:1.Radiation levels are from Table 3.11-3. Shielding provided by concrete cover onwaterstops is neglected.2.TID - Total integrated dose.3.Shielding provided by concrete cover on waterstops is considered.
02-01 02-01 02-01 02-01 3.8-154Reformatted PerAmendment 02-01TABLE 3.8-12LOAD COMBINATIONS AND ACCEPTANCE CRITERIAFOR CONCRETE STRUCTURES 1.Load Combinations for Service Load Conditionsa.Load Combinations Considered when Working Stress Design Method is UsedAcceptance Criteria(1)D + L S (1)(2)D + L + E S(3)D + L + W SWhen thermal stresses due to T o and R o are present, the following loadcombinations are also considered:Acceptance Criteria(4)D + L + T o + R o 1.3S(5)D + L + T o + R o + E 1.3S(6)D + L + T o + R o + W 1.3SBoth cases of L having its full value or being completely absent are checked.b.Load Combinations Considered when Strength Design Method is UsedAcceptance Criteria(1)1.4D + 1.7L U (2)(2)1.4D + 1.7L + 1.9E U(3)1.4D + 1.7L + 1.7W U 02-01 02-01 3.8-155Reformatted PerAmendment 02-01TABLE 3.8-12 (Continued)LOAD COMBINATIONS AND ACCEPTANCE CRITERIAFOR CONCRETE STRUCTURESWhen thermal stresses due to T o and R o are present, the following loadcombinations are also considered:Acceptance Criteria(4)(0.75) (1.4D + 1.7L + 1.7T o + 1.7R o)U(5)(0.75) (1.4D + 1.7L + 1.9E +1.7T o + 1.7R o)U(6)(0.75) (1.4D + 1.7L + 1.7W +1.7T o + 1.7R o)UBoth cases of L having its full value or being completely absent area checked.In addition, the following load combinations are considered:Acceptance Criteria(7)1.2D + 1.9E U(8)1.2D + 1.7W U 2.Load Combinations for Factored Load ConditionsAcceptance Criteriaa.D + L + T o + R o + E'Ub.D + L + T o + R o + Wt Uc.D + L + T a + R a + 1.5P a Ud.D + L + T a + R a + 1.25P a + 1.0(Y j + Y r + Y m) + 1.25E Ue.D + L + T a + R a + 1.0P a + 1.0(Y j + Y r + Y m) + 1.0E'UBoth cases of L having its full value or being completely absent are checked.
02-01 02-01 3.8-156Reformatted PerAmendment 02-01TABLE 3.8-12 (Continued)LOAD COMBINATIONS AND ACCEPTANCE CRITERIAFOR CONCRETE STRUCTURESNOTES:(1)S - For concrete structures, S is the required section strength based upon theworking stress design method and the allowable stresses defined in Section 8.10of ACI 318-71. Increases in allowable stresses for concrete due to seismic or wind loadings are not used.(2)U - For concrete structures, U is the section strength required to resist designloads based upon the strength design methods described in ACI 318-71.
3.8-157Reformatted PerAmendment 02-01TABLE 3.8-13LOADS AND LOAD COMBINATIONS FOR THE SERVICE WATERINTAKE AND DISCHARGE STRUCTURES 1.Definitions and NomenclatureD = dead weight of concrete and weight of soil above the Intake Structure E = is the Operating Basis Earthquake E' = is the Safe Shutdown Earthquake H = is the lateral earth pressure F = hydrostatic pressure 2.Load Combinations U is defined as required strength to resist design loads.a.End of ConstructionU = 1.4D + 1.7HU = 0.9D + 1.7H, where D reduces the effect of Hb.Operating ConditionU = 1.4D + 1.7H U = 1.4D + 1.7H + 1.9E U = 0.9D + 1.7H, where D reduces the effect of H U = 1.4D + 1.4F + 1.7H U = 1.4D + 1.4F + 1.7H + 1.87E (Critical Loading Combination)
U = 0.9D + 1.87Ec.Extreme Environmental ConditionsU = D + H + E'
U = D + H U = D + H + 1.25E + F (1)U = 0.9D + H + E' + F (1), where D reduces effect of HNOTE:(1) F for discharge structure only.
02-01 02-01 02-01 3.8-158Reformatted PerAmendment 02-01TABLE 3.8-14LOAD COMBINATIONS AND ACCEPTANCE CRITERIAFOR STEEL STRUCTURES1.Load Combinations for Service Load Conditionsa.Load Combinations Considered when Elastic Working Stress Design Methodsare Used Acceptance Criteria(1)D + LS (1)(2)D + L + E S(3)D + L + W SWhen thermal stresses due to To and Ro are present, the following loadcombinations are also considered:Acceptance Criteria(4)D + L + T o + R o 1.5S(5)D + L + T o + R o + E1.5S(6)D + L + T o + R o + W1.5SBoth cases of L having its full value or being completely absent are checked.b.Load Combinations Considered when Plastic Design Methods are UsedAcceptance Criteria(1)1.7D + 1.7LY (2)(2)1.7D + 1.7L + 1.7E Y(3)1.7D + 1.7L + 1.7W YWhen thermal stresses due to To and Ro are present, the following load combinations are also considered:Acceptance Criteria(4)1.3 (D + L + T o + R o)Y(5)1.3 (D + L + E +T o + R o)Y(6)1.3 (D + L + W +T o + R o)YBoth cases of L having its full value or being completely absent are checked.
02-01 02-01 02-01 02-01 3.8-159Reformatted PerAmendment 02-01TABLE 3.8-14 (Continued)LOAD COMBINATIONS AND ACCEPTANCE CRITERIAFOR STEEL STRUCTURES2.Load Combinations for Factored Load Combinationsa.Load Combinations Considered when Elastic Working Stress DesignMethods are Used. (3)Acceptance Criteria(1)D + L + T o + R o + E'1.6S(2)D + L + T o + R o + W t 1.6S(3)D + L + Ta + Ra + Pa1.6S(4)D + L + T a + R a + P a + 1.0 (Y j + Y r + Y m) + E1.6S(5)D + L + T a + R a + P a + 1.0 (Y j + Y r + Y m) + E'1.7Sb.Load Combinations Considered when Plastic Design Methods are UsedAcceptance Criteria(1)D + L + T o + R o + E'0.9Y(2)D + L + T o + R o + W t 0.9Y(3)D + L + T a + R a + 1.5P a 0.9Y(4)D + L + T a + R a + 1.25P a + 1.0 (Y j + Y r + Y m) + 1.25E0.9Y(5)D + L + T a + R a + 1.0P a + 1.0 (Y j + Y r + Y m) + E'0.9Y 02-01 3.8-160Reformatted PerAmendment 02-01TABLE 3.8-14 (Continued)LOAD COMBINATIONS AND ACCEPTANCE CRITERIAFOR STEEL STRUCTURESNOTES:(1)S - For structural steel, S is the required section strength based upon the elasticdesign methods and the allowable stresses defined in Part I of the AISC,"Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," 1969 Edition. Increases in allowable stresses for steel due to seismic or wind loadings are not used.(2)Y - For structural steel, Y is the required section strength required to resist designloads and is based upon plastic design methods described in Part 2 of the AISC,"Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," 1969 Edition.(3)For these combinations, the plastic modulus of steel shapes is used in computingthe required section strength, S.
3.8-161AMENDMENT 97-01AUGUST 1997TABLE 3.8-15LOAD COMBINATIONS USED TO CHECK AGAINSTSLIDING, OVERTURNING AND FLOTATIONMinimum Factors of SafetyLoad CombinationOverturningSlidingFlotation 1.D + H + E 1.5 1.5-2.D + H + W 1.5 1.5-3.D + H + E'1.1 1.1-4.D + H + W t 1.1 1.1-5.D + F--1.5 3.8-162AMENDMENT 97-01AUGUST 1997TABLE 3.8-16CONCRETECOMPARISON OF ALLOWABLE COMPRESSIVE STRESSESServiceFactoredMembraneMembraneplus BendingMembraneMembraneplus BendingSection 3.8.1.5 0.4f c'(1)0.60f c'(1)0.765f c'(2)0.765f c'(2)ASME CodeTable CC-3431-1Table CC-3421-1NOTES:1.Section 3.8.1.5.1.2, Items 1 and 2.2.Section 3.8.1.5.1.1, Item 1.
3.8-163Reformatted PerAmendment 02-01TABLE 3.8-17REINFORCING STEELCOMPARISON OF ALLOWABLE STRESSES AND STRAINSServiceFactoredStressStrainStressStrainSection 3.8.1.5 24 ksi (1)-0.9f y (2)0.9y (2)ASME Code 30 ksi-0.9f y y (3)NOTES:1.Section 3.8.1.5.1.2, Item 7.2.Section 3.8.1.5.1.1, Item 1. Note that the FSAR strain limit of 1.5y results fromReference 3 of Section 3.8.1.2.1. However, due to use of a linear concrete stressdistribution in conjunction with 0.9f y, the reinforcing steel strain is actually limitedto 0.9y.3.ASME Code CC-3422.1.
3.8-164Reformatted PerAmendment 02-01TABLE 3.8-18LOAD COMBINATION ALLOWABLE STRESSESFROM ASME CODE TABLES CC 3421-1 and CC 3431-1CODE ALLOWABLE CONCRETE COMPRESSION STRESSES (PSI)AT DISCONTINUITIES - BASE ANDSPRING LINE OF WALLAWAY FROM DISCONTINUITIES - WALLMID-HEIGHT AND DOME APEXLOAD COMBINATIONTYPEMEMBRANEMEMBRANE+ BENDINGMEMBRANEMEMBRANE+ BENDINGINITIAL PRESTRESSD + F I S0.35 fc' = 17500.60 fc' = 30000.35 fc' = 17500.45 fc' = 2250NORMAL WINTER OPERATIOND + F S0.30 fc' = 15000.60 fc' = 30000.30 fc' = 15000.45 fc' = 2250D + F + P v S0.30 fc' = 15000.60 fc' = 30000.30 fc' = 15000.45 fc' = 2250D + F + T o S0.45 fc' = 22500.60 fc' = 30000.45 fc' = 22500.60 fc' = 3000D + F + T o + P v S0.45 fc' = 22500.60 fc' = 30000.45 fc' = 22500.60 fc' = 3000SEVERE ENVIRONMENTALD + F + E o S0.40 fc' = 20000.60 fc' = 30000.40 fc' = 20000.45 fc' = 2250D + F + E o + P v S0.40 fc' = 20000.60 fc' = 30000.40 fc' = 20000.45 fc' = 2250D + F + E o + T o S0.45 fc' = 22500.60 fc' = 30000.45 fc' = 22500.60 fc' = 3000D + F + E o + T o + P v S0.45 fc' = 22500.60 fc' = 30000.45 fc' = 22500.60 fc' = 3000EXTREME ENVIRONMENTALD + F + E ss U0.60 fc' = 30000.85 fc' = 42500.60 fc' = 30000.75 fc' = 3750D + F + E ss + P v U0.60 fc' = 30000.85 fc' = 42500.60 fc' = 30000.75 fc' = 3750D + F + E ss + T o U0.75 fc' = 37500.85 fc' = 42500.75 fc' = 37500.85 fc' = 4250D + F + E ss + T o + P v U0.75 fc' = 37500.85 fc' = 42500.75 fc' = 37500.85 fc' = 4250ABNORMAL/EXTREMEENVIRONM ENTALD + F + E o + 0.5E ss + W U0.60 fc' = 30000.85 fc' = 42500.60 fc' = 30000.75 fc' = 3750D + F + E o + 0.5E ss + W + T o0.75 fc' = 37500.85 fc' = 42500.75 fc' = 37500.85 fc' = 4250 F I IS INITIAL PRESTRESS; F IS PRESTRESS AT STARTUP 02-01 02-01 02-01 02-01 02-01 02-01 3.8-165Reformatted PerAmendment 02-01TABLE 3.8-19COMPARISON OF PREDICTED STRESSES AND STRAINS WITH ASME CODE ALLOWABLES AT BASECONCRETE STRESSES (PSI)MEMBRANE LINERLINER STRAINS(MICRO IN/IN)CODEALLOWABLEPREDICTEDSTRESSES(PSI)CODE PREDICTEDALLOW. MEMBRANEMEM-MEMB.MERIDIONAL HOOPMERID-MEM-MERID-LOAD COMBINATIONTYPEBRANE+BENDI.F.O.F.MEMB.I.F.O.F.MEMB.IONAL HOOPBRANEIONAL HOOPINITIAL PRESTRESS D+F S-1750-3000-1477 391-543-252 67-93-17982-4855-2000-572 12NORMAL WINTEROPERATIO N D+F S-1500-3000-1368 330-519-234 56-94-16658-4499-2000-529 11 D+F+P v S-1500-3000-1430 360-539-245 61-92-17124-4624-2000-544 12D+F+T o S-2250-3000-2153 1177-488-1492 944-274-25558-16516-2000-716-314D+F+T o+P v S-2250-3000-2215 1207-504-1503 949-277-26024-16641-2000-731-314SEVEREENVIRO NMENTALD+F+E o S-2000-3000-1344-78-711-230-13-122-16475-4449-2000-524 11D+F+E o+P v S-2000-3000-1406-48-727-241-8-125-16941-4574-2000-538 12D+F+E o+T o S-2250-3000-2129 739-695-1388 875-257-25375-16466-2000-710-314D+F+E o+T o+P v S-2250-3000-2191 799-696-1499 880-310-25841-16591-2000-725-314EXTREME ENVIRO NMENTALD+F+E ss U-3000-4250-1332-282-807-228-48-138-16485-4424-5000-524 12D+F+E ss+P v U-3000-4250-1394-252-823-239-48-141-16849-4549-5000-536 12D+F+ss+T o U-3750-4250-2117 565-776-1486 840-323-25283-16441-5000-707-314D+F+E ss+T o+P v U-3750-4250-2179 595-792-1497 845-326-25749-16566-5000-713-314ABNORMAL/EXTREMEENVIRO NMENTALD+F+E o+ 0.5E ss+W U-3000-4250-1326-386-856-227-66-147-16336-4411-5000-519 11D+F+E o+ 0.5E ss+W+T o U-3750-4250-2111-461-825-1485 822-332-25236-16428-5000-706-334 02-01 02-01 02-01 02-01 02-01 3.8-166Reformatted PerAmendment 02-01TABLE 3.8-20COMPARISON OF PREDICTED STRESSES AND STRAINS WITH ASME CODE ALLOWABLES AT WALL MID-HEIGHTCONCRETE STRESSES (PSI)MEMBRANE LINERLINER STRAINS(MIC RO IN/IN)CODEALLOWABLEPREDICTEDSTRESSES(PSI)CODE PREDICTEDALLOW. MEMBRANEMEM-MEMB.MERIDIONAL HOOPMERID-MEM-MERID-LOAD COMBINATIONTYPEBRANE+BENDI.F.O.F.MEMB.I.F.O.F.MEMB.IONAL HOOPBRANEIONAL HOOPINITIAL PRESTRESS D+F S-1750-2250-791-757-774-1445-1439-1442-11244-18237-2000-205-516NORMAL WINTEROPERATIO N D+F S-1500-2250-751-720-736-1335-1329-1332-10625-16871-2000-198-476 D+F+P v S-1500-2250-777-744-761-1387-1370-1379-10853-17270-2000-202-487D+F+T o S-2250-3000-1821 462-680-2407-148-1278-21417-27675-2000-462-740D+F+T o+P v S-2250-3000-1847 438-705-2459-199-1329-21648-28074-2000-466-752SEVERE ENVIRO NMENTALD+F+E o S-2000-2250-890-865-878-1319-1311-1315-11637-16940-2000-232-468D+F+E o+P v S-2000-2250-916-889-903-1371-1352-1362-11865-17339-2000-235-479D+F+E o+T o S-2250-3000-1960 317-822-2391-130-1261-22429-27744-2000-496-732D+F+E o+T o+P v S-2250-3000-1962 293-834-2443-181-1312-22657-28143-2000-500-744EXTREMEENVIRO NMENTALD+F+E ss U-3000-3750-960-938-949-1311-1302-1309-12143-16975-5000-249-464D+F+E ss+P v U-3000-3750-986-962-974-1363-1353-1358-12371-17374-5000-253-475D+F+ss+T o U-3750-4250-2030 244-893-2383-121-1252-22995-27779-5000-515-728D+F+E ss+T o+P v U-3750-4250-2056 220-918-2435-172-1304-23163-28178-5000-517-740ABNORMAL/EXTREMEENVIRO NMENTALD+F+E o+ 0.5E ss+W U-3000-3750-996-975-986-1307-1297-1302-12402-16993-5000-258-462D+F+E o+ 0.5E ss+W+T o U-3750-4250-2066 207-930-2379-116-1248-23193-27797-5000-522-726 02-01 02-01 02-01 02-01 02-01 3.8-167Reformatted PerAmendment 02-01TABLE 3.8-21COMPARISON OF PREDICTED STRESSES AND STRAINS WITH ASME CODE ALLOWABLES AT WALL SPRING LINECONCRETE STRESSES (PSI)MEMBRANE LINERLINER STRAINS(MICRO IN/IN)CODEALLOWABLEPREDICTEDSTRESSES(PSI)CODE PREDICTEDALLOW. MEMBRANEMEM-MEMB.MERIDIONAL HOOPMERID-MEM-MERID-LOAD COMBINATIONTYPEBRANE+BENDI.F.O.F.MEMB.I.F.O.F.MEMB.IONAL HOOPBRANEIONAL HOOPINITIAL PRESTRESS D+F S-1750-3000-1101-264-683-670-606-638-14023-9373-2000-390-183NORMAL WINTEROPERATIO N D+F S-1500-3000-1034-255-645-617-557-587-13149-8656-2000-367-167 D+F+P v S-1500-3000-1071-268-670-639-575-607-13456-8846-2000-376-170D+F+T o S-2250-3000-2107 932-588-1702 612-545-24043-19625-2000-633-436D+F+T o+P v S-2250-3000-2144 919-613-1724 594-565-24330-19815-2000-641-440SEVEREENVIRO NMENTALD+F+E o S-2000-3000-1073-271-672-586-521-554-13410-8459-2000-378-158D+F+E o+P v S-2000-3000-1110-284-697-608-539-574-13637-8649-2000-384-162D+F+E o+T o S-2250-3000-2146 916-615-1671 648-512-24304-19428-2000-644-427D+F+E o+T o+P v S-2250-3000-2177 903-637-1693 630-532-24591-19618-2000-652-431EXTREME ENVIRO NMENTALD+F+E ss U-3000-4250-1093-279-686-570-503-537-13541-8360-5000-383-153D+F+E ss+P v U-3000-4250-1130-292-711-592-521-557-13828-8380-5000-391-158D+F+ss+T o U-3750-4250-2166 908-629-1656 666-495-24435-19329-5000-649-422D+F+E ss+T o+P v U-3750-4250-2203 895-634-1677 648-515-24712-19519-5000-657-426ABNORMAL/EXTREMEENVIRO NMENTALD+F+E o+ 0.5E ss+W U-3000-4250-1103-283-693-562-494-528-13608-8310-5000-386-150D+F+E o+ 0.5E ss+W+T o U-3750-4250-2176 904-636-1647 675-486-24502-19279-5000-652-420 02-01 02-01 02-01 02-01 02-01 02-01 3.8-168Reformatted PerAmendment 02-01TABLE 3.8-22COMPARISON OF PREDICTED STRESSES AND STRAINS WITH ASME CODE ALLOWABLES AT DOME APEXCONCRETE STRESSES (PSI)MEMBRANE LINERLINER STRAINS(MIC RO IN/IN)CODEALLOWABLEPREDICTEDSTRESSES(PSI)CODE PREDICTEDALLOW. MEMBRANEMEM-MEMB.MERIDIONAL HOOPMERID-MEM-MERID-LOAD COMBINATIONTYPEBRANE+BENDI.F.O.F.MEMB.I.F.O.F.MEMB.IONAL HOOPBRANEIONAL HOOPINITIAL PRESTRESS D+F S-1750-2250-1380-1508-1444-1470-1494-1482-18295-19272-2000-438-482NORMAL WINTEROPERATIO N D+F S-1500-2250-1293-1412-1353-1377-1399-1388-17132-18046-2000-410-451 D+F+P v S-1500-2250-1339-1452-1396-1422-1443-1433 17503-18410-2000-419-459D+F+T o S-2250-3000-2345-212-1279-2432-198-1315-27781-28713-2000-671-712D+F+T o+P v S-2250-3000-2391-252-1322-2447-242-1344-28152-29077-2000-680-721SEVERE ENVIRO NMENTALD+F+E o S-2000-2250-1293-1413-1353-1356-1372-1364-17119-17895-2000-411-446D+F+E o+P v S-2000-2250-1339-1453-1396-1401-1416-1409-17496-18259-2000-421-455D+F+E o+T o S-2250-3000-2345-213-1279-2401-171-1286-27768-28562-2000-672-707D+F+E o+T o+P v S-2250-3000-2391-253-1322-2456-215-1336-28079-28926-2000-679-717EXTREMEENVIRO NMENTALD+F+E ss U-3000-3750-1293-1414-1354-1345-1358-1352-17112-17819-5000-412-443D+F+E ss+P v U-3000-3750-1339-1454-1397-1390-1402-1396-17423-18123-5000-419-453D+F+ss+T o U-3750-4250-2345-214-1280-2400-159-1279-27761-28486-5000-672-704D+F+E ss+T o+P v U-3750-4250-2391-254-1323-2445-201-1323-28132-28850-5000-682-713ABNORMAL/EXTREMEENVIRO NMENTALD+F+E o+ 0.5E ss+W U-3000-3750-1293-1414-1354-1340-1351-1346-17109-17780-5000-412-442D+F+E o+ 0.5E ss+W+T o U-3750-4250-2345-214-1280-2395-150-1273-27758-28447-5000-672-703 02-01 02-01 02-01 02-01 02-01 02-01
()' I '<t _j w ' I ... . .. . ... .. .... .. RN 13-022 SOUTH CAROLINA ELECTRIC & GAS CO. VIRGIL C. SUMMER NUCLEAR STATION RB Recirculation Sumps for Residual Heat Removal and RB Spray Pumps Figure 3.8-17
RN 11-027
3.9-43 Reformatted Per RN 07-024 TABLE 3.9-1 BALANCE OF PLANT COMPONENT AND PLANT OPERATING CONDITIONS Operating Condition or Initiating Event Plant Condition System and Component Condition Startup Normal Normal Standby Normal Normal Part Load
Normal Normal Full Load
Normal Normal Shutdown Normal Normal Uncontrolled RCC Assembly Withdrawal at Power
Upset Upset Uncontrolled RCC Assembly Withdrawal from a Subcritical Condition
Upset Upset RCC Assembly Misalignment
Upset Upset Chemical and Volume Control System Malfunction
Upset Upset Practical Loss of Forced Reactor Coolant Flow
Upset Upset Startup of an Inactive Reactor Coolant Loop Upset Upset Loss of External Electrical Load Upset Upset 3.9-44 Reformatted Per RN 07-024 TABLE 3.9-1 (Continued)
BALANCE OF PLANT COMPONENT AND PLANT OPERATING CONDITIONS Operating Condition or Initiating Event Plant Condition System and Component Condition Turbine Trip Upset Upset Loss of Normal Feedwater
Upset Upset Station Blackout
Upset Upset Excessive Heat Removal D ue to Feedwater System
Malfunction
Upset Upset Excessive Load Increase
Upset Upset Slow Loss of Reactor C oolant which Actuates Emergency Core Flooding
Emergency Upset Minor Secondary System Pipe Break
Emergency Upset Loading Fuel Assembly into Improper Position Emergency Upset Complete Loss of Forced Reactor Coolant Flow Emergency Upset Waste Gas Decay Tank Rupture
Emergency Upset Loss of Coolant Accident
Faulted Emergency or Faulted (1) Rupture of a Steam Pipe Faul ted Emergency or Faulted (1) 3.9-45 Reformatted Per RN 07-024 TABLE 3.9-1 (Continued)
BALANCE OF PLANT COMPONENT AND PLANT OPERATING CONDITIONS Operating Condition or Initiating Event Plant Condition System and Component Condition Steam Generator Tube Rupture Faulted Emergency Single Reactor Coolant Pump Locked Rotor
Faulted Emergency Fuel Handling Accident
Faulted Emergency Rupture of a Control Rod Drive Me chanism Housing Faulted Emergency
(1) Faulted component conditions have been considered for components in a run of pipe subject to pipe rupture between pipe anchors and/or pipe rupture restraints.
3.9-46 Reformatted Per RN 07-024 TABLE 3.9-2 BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings Nonactiv Class 2 and 3 e Upset 1.a al pressure (upset) e Emergency Note 1.b pressure (normal) Active ASME Code Class 2 and 3 Normal Pr not exceeded, per ANSI, l pressure (normal) e Upset 2.a al pressure (upset) lve 02-01 e ASME Code Valves Normal Pr not exceeded, per ANSI B16.5, 1968 (see Note 1 for definition of Pr)
Internal pressure (normal) Deadweight Normal piping loads at valv ends Note InternDeadweight Upset piping loads at valvends Operating Basis Earthquake (OBE) InternalDeadweight Emergency piping loads at valve ends Safe Shutdown Earthquake (SSE) Interna Valves B16.5, 1968 Deadweight Normal piping loads at valvends Note InternDeadweight Upset I piping loads at vaends OBE 3.9-47 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings Active ASMss 2 and 3 2.b Internlve Emergency Note 2.c al pressure (normal) FaultedE Code ClaValves (Cont'd)
Upset Note al pressure (upset) Deadweight Upset II piping loads at va ends InternDeadweight Emergency piping loads at valve ends
SSE (i) Note 2.c as emergency, above, s ASME Code Class 2 and 3 Vessels, Normal In accordance with the ASME Upset Notes 3.a and d essure (upset) d 02-01 Sameexcept that faulted piping loadare applied at valve ends Internal pressure (normal) Designed in Accordance with Division1 of Section VIII of the ASME Code Code, Section VIII, Division 1 Nozzle loads from attached piping (normal) Concentrated loads from supports Internal prNozzle loads from attache piping (upset) Concentrated loads from supports OBE 3.9-48 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings ASME Codend 3 Vessels, Interna d Emergency Notes 3.c and d essure (normal)
ASME Code Class 2 Vessels, ivision Normal In accordance with the ASME al pressure (normal)
Upset essure (upset) d 02-01 Class 2 aDesigned in Accordance with Division1 of Section VIII of the ASME Code (Cont'd) Upset Notes 3.b and d l pressure (Upset) Nozzle loads from attache piping (upset)
Concentrated loads from supports Internal prNozzle loads from attachedpiping (emergency)
Concentrated loads from supports SSE InternDesigned in accordance with D2 of Section VIII of the ASME Code Code, Section VIII, Division 2 Nozzle loads from attached piping (normal) Concentrated loads from supports Note 4.a Internal prNozzle loads from attache piping (upset) Concentrated loads from supports
OBE 3.9-49 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings ASME Code Class 2 Vssels, ivision 4.b Intern d Emergency Note 4.c essure (Normal)
ASME Code Class 2 and 3 Piping Normal In accordance with the ASME weight pressure stress ly) Upset Note 5.a pressure stress icle loads 02-01 eDesigned in Accordance with D2 of Section VIII of the ASME Code (Cont'd) Upset Note al pressure (Upset) Nozzle loads from attache piping (upset)
Concentrated loads from supports Internal prNozzle loads from attachedpiping (emergency)
Concentrated loads from supports SSE Dead Code, Section III, for Normal design conditions Longitudinal (normal) Thermal expansion stress Soil loads (buried piping onDeadweight Longitudinal (upset) Thermal expansion stress OBE (ii) Flow transients (ii) Soil loads and veh (buried piping only) Safety valve reactions, as applicable (ii) 3.9-50 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings ASME Codend 3 Piping Deadpressure stress ehicle loads Emergency 5.c pressure stress piping only) Faulted Class 2 a(Cont'd) Upset Notes 5.b and d weight Longitudinal (upset)
Flow transients Soil loads and v(buried piping only) Safety valve reactions, as applicable (ii) Jet loadings Note Deadweight Longitudinal (normal) Flow transients (ii) Soil loads (buried Jet loadings SSE (ii) (iii) Note 5.c weight pressure stress piping only) Nonactive ASME Code Class 2 and 3 Normal In accordance with the pressure (normal) 02-01 DeadLongitudinal (normal) Flow transients (ii) Soil loads (buried Pipe rupture loads SSE (ii) Internal Pumps ASME Code, Section IIINozzle loads from attachedpiping (normal)
Concentrated loads from supports 02-01 3.9-51 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings Nonactiv Class 2 and 3 6.a Intern d Upset 6.b al pressure (upset) d Emergency Note 6.c essure (normal)
Active ASME Code Class 2 and 3 Normal In accordance with the ASME al pressure (normal)
Upset Notes 7.a and d re (upset) d 02-01 e ASME Code Pumps (Cont'd)
Upset Note al pressure (upset) Nozzle loads from attachepiping (upset)
Concentrated loads from supports OBE Note InternNozzle loads from attache piping (upset) Concentrated loads from supports Internal prNozzle loads from attached piping (emergency) Concentrated loads from supports SSE Intern Pumps Code, Section III Nozzle loads from attachedpiping (normal) Support loads Internal pressuNozzle loads from attache piping (upset) Support loads OBE 02-01 3.9-52 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings Active ASMss 2 and 3 Interna d Emergency Notes 7.c and d re (normal)
Code Class 2 and 3 Pipe Supports Pipe support ons for the In accordance with the ASME weight d loads and ow Code Class 2 and 3 Supports for Same as for In accordance with the ASME d loads and ow 02-01 E Code Cla Pumps (Cont'd) Upset Notes 7.b and d l pressure (upset) Nozzle loads from attachepiping (upset)
Support loads Internal pressuNozzle loads from attachedpiping (emergency) Support loads
SSE Deaddesign conditiare consistent withthe design conditions establishedpiping Code, Section III, Subsection NF, Winter 1973 Addenda Superimpose reactions Dynamic loads (seismic, fl transient) Thermal expansion Anchor and support movements Environmental loads Deadweight Pumps and Vessels supported component Code, Section III, Subsection NF, as applicable Superimpose reactions
Dynamic loads (seismic, fltransient) Thermal expansion Anchor and support movements Environmental loads 02-01 02-01 02-01 02-01 02-01 3.9-53 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
BALANCE OF PLANT COMPONENT LOADING CONDITIONS Compnent o Component Condition Design Limits and Loading Combinations Types of Loadings Code Classers Snub the In accor OTES 2 and 3 Snubbber design conditions are consistent with design conditions established for the piping dance with the ASME Code, Section III, Subsection NF, Winter 1976 Addenda Dynamic loads (seismic, flow transient)
N: d component condition is specified for active valves in a run of pipe subject to a postulated pipe rupture. Operabil ity i) on a root mean square basis: alve reactions.
ii) ated for portions of piping systems subject to postulated pipe rupture.
i) Faulte (has been demonstrated through test and analysis.
The following loads, if applicable, may be combined (i
- 1. Flow transients.
2 . Dynamic portion of safety v 3 (iFaulted piping conditions are postul 02-01 02-01 3.9-54 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
NOTES TO TABLE 3.9-2
- 1. Nonactive ASME Code Class 2 and 3 Valves The design of Code Class 2 and 3 valves encompasses the use of pressure temperature ratings. The design limits are in terms of Pr which is the primary pressure rating corresponding to the maximum transient temperature for each plant condition as specified in Articles NC-3511 and ND-3511 of the ASME Code, Section III, for Code Class 2 and 3 valves, respectively. To assure pressure retaining integrity the limits for Pr are set as follows:
- a. The primary pressure rating, Pr, is not exceeded by more than 10 percent when the component is subject to concurrent loadings associated with either:
(1) The normal plant condition or the upset plant condition and the vibratory motion of 50 percent of the SSE.
(2) Loading conditions associated with the emergency plant condition.
- b. Pr is not exceeded by more than 20 percent when the component is subject to concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
- 2. Active ASME Code Class 2 and 3 Valves
To provide pressure retaining integrity and assurance of operability for active valves of Code Class 2 and 3, Pr is not exceeded for the combinations of loading delineated.
- a. The primary pressure rating, Pr, is not exceeded when the component is subjected to concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of 50 percent of the SSE.
- b. The primary pressure rating, Pr, is not exceeded when the component is subject to loadings associated with the emergency plant condition.
- c. The primary pressure rating, Pr, is not exceeded when the component is subject to concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
3.9-55 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
- 3. ASME Code Class 2 and 3 Vessels (designed in accordance with Division 1 of Section VIII of the ASME Code) (See Item d, below).
To provide assurance of pressure retaining integrity for Code Class 2 and 3 vessels (designed in accordance with Division 1 of Section VIII of the ASME Code) the allowable stress value, S, (as specified in Appendix 1 of Section III of the ASME Code) is not exceeded by more than 10 percent when the component is subjected to the loading combinations identified by items a and b, below, and is not exceeded by more than 50 percent when the component is subjected to the loading combinations identified by item c, below.
- a. Concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of 50 percent of the SSE.
- b. Loadings associated with the emergency plant condition.
- c. The allowable stress value, S is not exceeded by more than 50 percent when the component is subjected to concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
- d. When a more detailed analysis is performed, Division 1 vessels satisfy, as a minimum, equations (1) and (2), below. Equation (1) is applicable to items a and b, above. Equation (2) is applicable to item c, above.
M 1.1 S 5.1bM (1) M 1.5 S 5.1bM (2) 02-01 Where: M = Primary membrane stress.
b = Primary bending stress.
S = Allowable stress value as specified in Appendix 1 of Section III of the ASME Code.
3.9-56 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
- 4. ASME Code Class 2 Vessels (designed in accordance with Division 2 of Section VIII of the ASME Code.)
To provide assurance of pressure retaining integrity for Code Class 2 Vessels, (designed in accordance with Division 2 of Section VIII of the ASME Code) the upset, emergency and faulted operating condition category design limits of Article NB-3200 of Section III of the ASME Code are not exceeded when the component is subjected to the following combinations:
- a. The design limits specified in Article NB-3223 of the ASME Code are not exceeded when the component is subjected to concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of 50 percent of the SSE.
- b. The design limits specified in Article NB-3224 of the ASME Code are not exceeded when the component is subjected to loadings associated with the emergency plant condition.
- c. The design limits specified in Article NB-3225 of the ASME Code are not exceeded when the component is subjected to concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
- 5. ASME Code Class 2 and 3 Piping
To provide assurance of pressure retaining integrity for Code Class 2 and 3 piping, the design limits specified in NC-3611.1(b)(4)(c)(b)(1) of the Winter 1972 Addenda to Section III of the ASME Code (i.e., 1.2Sh) are not exceeded when the piping is subjected to the loading combinations identified in items a and b, below. However, for short sections of piping exposed to jet impingement from postulated cracks or breaks in adjacent piping, a stress limit of 1.5Sh may be used.
The design limits specified in NC-3611.1(b)(4)(c)(b)(2) of the Winter 1972 Addenda to Section III of the ASME Code (i.e. 1.8Sh) are not exceeded when the piping is subjected to the loading combinations identified in item c below. However, for short sections of piping exposed to jet impingement from postulated cracks or breaks in adjacent piping, a stress limit of 2.4Sh may be used.
3.9-57 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
Whenever the 2.4Sh allowable stress limit is employed for Class 2 and 3 piping, an evaluation of the allowable collapse load will be performed in accordance with Appendix F to the ASME Code , Section III, Winter, 1972, A ddenda. It is recognized that neither the "collapse load" nor the "plastic instability load," as defined by the ASME Code, refer to geometrical instability. However, test results are available which do provide a basis for evaluating geometrical stability of fittings (elbows and tees), as well as straight pipe, at stress levels associated with the ASME Code "collapse" or "plastic instability" load limit. Additionally , test results reveal that, for thickness to radius ratios (t/r) less than 0.08, the mode of collapse is not in the form of geometrical instability at the fitting or discontinuity. The piping systems evaluated generally have t/r ratios less than 0.08.
- a. Concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of 50 percent of the SSE.
- b. Loadings associated with the emergency plant conditions (see Note 1).
- c. Concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
Thermal expansion effects of piping are not evaluated for loadings associated with emergency or faulted plant conditions. Therefore, only Equation 9 of Article NC-3651 of Section III of the ASME Co de is applied for the loading combinatio ns identified in items b and c, above. Thermal expansion effects are evaluated for the loading combinations identified in item a, above.
- 6. Nonactive ASME Code Class 2 and 3 pumps
In order to assure pressure retaining integrity for nonactive Code Class 2 and 3 pumps, the primary membrane stress is not exceeded by more than 10 percent of S (as specified in Appendix 1 of Section III of the ASME Code) and the sum of the primary membrane plus primary bending stresses is not exceeded by more than 65 percent of S when the component is subjected to the following load combinations:
- a. Concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of 50 percent of the SSE.
- b. Loadings associated with the emergency plant condition.
- c. In addition, the primary membrane stress is not exceeded by more than 20 percent of S, and the sum of the primary membrane and primary bending stresses is not exceeded by more than 80 percent of S when the component is subjected to concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
3.9-58 Reformatted Per RN 07-024 TABLE 3.9-2 (Continued)
- 7. Active ASME Code Class 2 and 3 Pumps To provide increased assurance that unacceptable deformations affecting operability of active Code Class 2 and 3 pumps do not result, the primary membrane stress does not exceed S (as specified in Appendix 1 of Section III of the ASME Code) and the sum of the primary membrane plus primary bending stresses is not exceeded by more than 50 percent of S when the component is subjected to the following loading combinations: (See item d, below).
- a. Concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of 50 percent of SSE.
- b. Loadings associated with the emergency plant condition.
- c. Concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
- d. The design limits given below are not exceeded for the applicable loading combinations. Analysis and/or testing confirms that operability is not impaired when the component is designed to these limits.
The primary membrane stress is not exceeded by more than 10 percent of S and the sum of the primary membrane plus primary bending stresses is not exceeded by more than 65 percent of S when the component is subjected to the following combinations:
(1) Concurrent loadings associated with either the normal plant condition or the upset plant condition and the vibratory motion of 50 percent of the SSE.
(2) Loadings associated with the emergency plant condition.
The primary membrane stress is not exceeded by more than 20 percent of S and the sum of the primary membrane and primary bending stresses is not exceeded by more than 80 percent of S which the component is subjected to concurrent loadings associated with the normal plant condition, the vibratory motion of the SSE and the dynamic system loadings associated with the faulted plant condition.
TABLE 3.9-3 DESIGN LOADING COMBINATIONS FOR NSSS SUPPLIED ASME CODE CLASS 2 AND 3 COMPONENTS AND SUPPORTS Conditions Classification Loading Combination Design and Normal Design pr essure Design temperature, (1) Dead weight, nozzle loads (2) Upset Upset condition pr essure, Upset condition metal temperature, (1) deadweight, OBE, nozzle loads (2) Faulted Faulted condition pressure, faulted condition metal temperature, (1) deadweight, SSE, nozzle loads (2) (1) Temperature is used to det ermine allowable stress only.
(2) Nozzle loads are those loads associated with the particular plant operating conditions for the component under consideration.
3.9-59 Reformatted Per RN 07-024 3.9-60 Reformatted Per RN 07-024 TABLE 3.9-4 STRESS CRITERIA FOR NSSS SUPPLIED SAFETY-RELATED ASME CLASS 2 AND CLASS 3 TANKS Condition Stress Limits Design and Normal The vessel shall conform to the requirements of ASME Section VIII, Division 1.
Upset m 1.1 S ( m or L) + b 1.65 S Faulted m 2.0 S ( m or L) + b 2.4 S
3.11-1 Reformatted July 2014 3.11 ENVIRONMENTAL QUALIFICATION OF MECHANICAL AND ELECTRICAL EQUIPMENT This section describes the program for environmental qualificat ion of electrical equipment. The Virgil C. Summer Nuclear Station does not have a licensing commitment to environmentally qualify safety-related mechanical equipment to the level of detail required to qualify electrical equipment. However, mechanical equipment has been specified and procured to satisfy requirem ents which assure that it can withstand the normal, abnormal, accident, and post-acci dent conditions to which it may be subjected. The environment al qualification program for the V. C. Summer Nuclear Station identifies the electrical equipment to be qualified, defin es the environmental conditions under normal, abnormal, accident, and post-accident operating conditions.
The program also documents the qualific ation tests and analyses employed to demonstrate the equipment's capability to per form design safety functions including post-accident monitoring when expos ed to normal, abnormal, accident, and post-accident environments as applicable. Seismic qualification is addressed in Section 3.10 for mechanical and electrical equipment.
3.11.1 ENVIRONMENTAL CONDITIONS AND EQUIPMENT IDENTIFICATION This section identifies: 1) the environmental design basis for electrical equipment, including the definition of the normal, abnormal, accident, and post-accident environments, and 2) the systems and electrical equipment that are required to perform a design safety function, including Regulatory Guide 1.97
[24] monitoring.
3.11.1.1 Environm ental Conditions Electrical equipment location for environmental qualification purposes is defined by environmental zones. Various environmental zones are encountered within the plant's building(s) (i.e. Auxiliary Bldg., Containment Bldg., Intermediate Bldg., etc.). The environmental zone boundaries are shown on plant layout drawings, called "Environmental Zone Maps", as document ed by drawings SS-021-001 through SS-021-017
[3]. The zone boundaries shown on drawings SS-021-001 through SS-021-017 were determined based on conti guous areas with similar environmental conditions.
An equipment qualification dat abase with environmental zone information as zones documented by drawing S-021-018
[4] provides a list of the environmental zones and conditions, including the normal, abnormal, and accident (including post-accident) environmental conditions. Environmental da ta is provided for the temperature, pressure, relative humidity, and radiation parameters for each env ironmental condition postulated to occur within each zone. Defi nitions used in determining the environmental conditions are as follows:
- a. Normal Conditions - planned, purposeful, unr estricted reactor operating modes that include startup, power range and hot standby (condenser available), shutdown, and refueling modes.
3.11-2 Reformatted July 2014 b. Abnormal Conditions - any deviation fr om normal conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operation impairment; planned testi ng including preoperat ional tests are also considered abnormal conditions(loss of non-safety related HVAC is an example of an abnormal condition).
- c. Accident Conditions - a single event not reasonably expected during the course of plant operations that has been hypothesiz ed for analysis purposes or postulated from unlikely but possible situations or t hat has the potential to cause a release of radioactive material (a reactor coolant pressure boundary rupture may qualify as an accident; a fuel cladding defect does not
). Postulated a ccident conditions include those resulting from: a Main St eam Line Break (MSLB) inside or outside of containment; a Loss of Coolant A ccident (LOCA); a High Energy Line Break/Superheated Blowdown Outside Cont ainment (HELB/SBOC)
- or other Line Break Accidents. Accident conditions ar e calculated for a post-accident period sufficient to ensure that steady state conditions have been reached.
The environmental parameters listed in dr awing S-021-018 are based on verified design calculations and do not include margins required in qualification testing or analyses as described in Section 3.11.2. The design basis used for preventing t he loss of ventilation for some zones is discussed in Section 3.11
.4. The basis for the estimated chemical and radiation environmental conditions is discussed in Section 3.11.5. Environmental
conditions listed in drawing S-021-018, and environmental zone boundaries shown on Environmental Zone Maps, drawings SS-021-001 through SS-021-017, will be revised when a modification in equipment or com ponents or the result s of temperature monitoring affect environmental qualification conditions.
To aid in the application of the qualificat ion program acceptance criteria that is discussed in Section 3.11.2, the environmental zones list ed in S-021-018 drawings have been classified as either a harsh or mild environment based on the following definitions:
Harsh Environments - Those zones or ar eas where the environmental conditions significantly exceed the normal or abnormal range as a result of a DBE. A harsh
environment area is an area or zone of the plant where one or mo re of the following environmental service conditions exist:
- Radiation:
TID (normal plus accident) > 1.0 E05 Rads (1.0 E04 for electronics) and the accident dose is greater than or equal to normal.
- Humidity Relative Humidity = 100%
3.11-3 Reformatted July 2014
- Pressure:
0.1 psig above atmospheric
- Temperature:
The accident temperature is a 20% change from the normal maximum temperature.
Mild Environments - Those zones or areas where the environmental conditions do not significantly exceed the normal or abnormal range as a result of a DBE. A mild environment area is an area or zone of the plant where the following environmental service conditions exist:
- Radiation:
TID (normal plus accident) 1.0 E05 Rads; or, if greater than 1.0 E05, the accident dose is less than TID normal. The 1.0 E05 limit is reduced to 1.0 E04 for electronics.
- Humidity:
Relative Humidity < 100%
- Pressure:
< 0.1 psig above atmospheric
- Temperature:
The accident temperature is < a 20% change from the normal maximum temperature.
3.11.1.2 Equipment Identification
- 1. General
Electrical equipment to be qualified includes equipment associated with systems that are essential to:
- a. Emergency reactor shutdown.
- b. Containment isolation.
3.11-4 Reformatted July 2014 c. Reactor core cooling.
- d. Containment heat removal.
- e. Reactor heat removal.
- f. Preventing significant release of radioactive material to the environment, or
- g. Provide Regulatory Guide 1.97 Category 1 and 2 indicating and post-accident monitoring. Non 1E, Regulatory Gu ide 1.97 Category 2 equipment installed in a geographic mild environment is ex cluded from the mild EQ Program This is equipment that:
- a. Performs the previous functions automatically,
- b. Is used by the operator to per form these functions manually, or
- c. The failure of which can prevent the satisfactory accomplishment of one or more of the previous safety functions.
"Safety-related" equipment is categorized in three groups by design safety function:
- a. Safety-related electrical equipment - designated as "Class 1E" per IEEE Standard 308.
[1]
- b. Safety-related "Active" Mechanical Equipment - that equipment which must move or change position to perform its design safety function (examples are pumps, motor operated valves, and safety relief valves).
- c. Safety-related "Passive" Mechanical E quipment - that equi pment which must only maintain its pressure integrity to perform its design safety function (examples are tanks, heat ex changers, and manual valves).
The design safety functions for spec ific equipment items are discussed on a system basis in Chapters 3 (Sections 3.4, 3.
5, and 3.6), 5, 6, 7, 8, 9, 12, and 15.
- 2. List of Equipment
The Equipment Qualification Database (EQDB)
[2] identifies electrical equipment which requires environmental qualification. The Database provides specific informative data such as: Tag number, manufacturer, model number, purchase order number, equipment qualification documentation package number, etc. It is a computerized Database loaded on the VCSNS VAX System. Refer to Section 3.11.3.1 for discussion.
3.11-5 Reformatted July 2014 An overview of major safety-related equipment and components other than Class 1E equipment that are requir ed to function during and/or subsequent to the design
basis accidents are listed in Tables 3.11-1 for equipment and components located inside the Reactor Building, and 3.11-2 for equipment and components located outside the Reactor Building. The remainde r of the information required to comply with General Design Criteri on (GDC) 4 is discussed in Sections 3.11.1.1 through 3.11.3.3.
- 3. Equipment Categorization
- a. For environmental qualific ation, "Environmentally Qualified" electrical equipment is grouped into one or more categories or designations based on the equipment's functional design requirements. A breakdown of environmental categories and designatio ns including their respective definitions are as follows:
- 1. Category A1:
Equipment that could experience t he environmental conditions of design basis (LOCA) accidents for which it must function to mitigate said accident. It must be qualified to demonstrate operability in the accident environment for the time required fo r accident mitigation with safety margin to failure.
- 2. Category A2:
Equipment that could experience t he environmental conditions of design basis line break accidents, includi ng Main Steam Line Break (MSLB),
High Energy Line Brea k/Superheated Blowdown Outside Containment (HELB/SBOC), and/or other line breaks for which it must function to mitigate said accidents. It must be qualified to demonstrate operability in its specifically applicable accident environment for the time required for accident mitigation with safety margin to failure.
- 3. Category A1 *:
Equipment that could experience increased radiation exposure due to Post-LOCA recirculation for which it must function to mitigate said accident. It must be qualified to demonstrate operability in the radiation environment for the time required fo r accident mitigation with safety margin to failure.
3.11-6 Reformatted July 2014 4. Category B1:
Equipment that could experience environmental conditions of design basis (LOCA) accidents through which it need not function for mitigation of said accident, but through which it must not fail in a manner
detrimental to plant safety or accident mitigation. It must be qualified to demonstrate the capability to withstand any LOCA accident environment
for the time during which it must not fail with safety margin to failure.
- 5. Category B2:
Equipment that could experience t he environmental conditions of design basis line break accidents, includi ng Main Steam Line Break (MSLB),
High Energy Line Brea k/Superheated Blowdown Outside Containment (HELB/SBOC), and/or other line breaks for which it need not function for
mitigation of said accidents, but through which it must not fail in a manner detrimental to plant safety or accident mitigation.
It must be qualified to demonstrate the capability to withstand any such specifically applicable accident environm ent for the time which it must not fail with safety margin to failure.
- 6. Category B1 *:
Equipment that could experience increased radiation exposure due to Post-LOCA recirculation through which it need not function for mitigation of said accident, but through which it must not fail in a manner
detrimental to plant safety or accident mitigation. It must be qualified to demonstrate the capability to withstand any radiation environment for the time during which it must not fail with safety margin to failure.
- 7. Category C1:
Equipment that could experience environmental conditions of design basis (LOCA) accidents through which it need not function for mitigation of said accident, and whose failure is deemed not detrimental to plant safety or accident mitigation. It deemed not detrimental to plant safety or
accident mitigation. It need not be qualified for any LOCA accident environment, but must be qualified for the non-accident environment.
3.11-7 Reformatted July 2014 8. Category C2:
Equipment that could experience t he environmental conditions of design basis line break accidents, includi ng Main Steam Line Break (MSLB),
High Energy Line Brea k/Superheated Blowdown Outside Containment (HELB/SBOC), and/other line breaks th rough which it need not function for mitigation of said accidents and whose failure is deemed not detrimental to plant safety or accident mitigation. It need not be qualified for these accident environments, but must be qualified for the non-accident environment.
- 9. Category C1 *:
Equipment that could experience increased radiation exposure due to Post-LOCA recirculation through which it need not function for mitigation of said accident and whose failure is deemed not detrimental to plant
safety or accident mitigation. It need not be qualified for any Post-LOCA harsh accident environment, but must be qualified for the non-accident environment. Also equipment that c ould experience increased radiation exposure due to Post-LOCA recirculation but has completed its functional requirement prior to the environment becoming harsh and whose failure (after completing an y required function) is deemed not detrimental to plant safety or acci dent mitigation need not be qualified for any Post-LOCA harsh accident envir onment, but must be qualified for
the non-accident and Post-LOCA m ild accident environments.
- 10. Category D:
Equipment that would not experience environmental conditions of design basis accidents and that must be qual ified to demonstrate operability in the normal and abnormal service envir onment. This equipment is located outside containment.
- 11. Designation QR-H:
Equipment that is quality related and requires harsh environment qualification per design documents.
- 12. Designation QR-M:
Equipment that is quality related and requires mild environment qualification per design documents.
3.11-8 Reformatted July 2014 b. Environmental categories for Regulator y Guide 1.97 equipment are similar to those previously identified for other safe ty-related electrical equipment except that a prefix letter E is used.
A breakdown of Regulatory Gui de 1.97 environmental categories and respective definitions are as follows:
- 1. Category EA1:
Equipment that could experience t he environmental conditions of design basis (LOCA) accidents for which it must function to provide Regulatory Guide 1.97 monitoring information to the operator. It mu st be qualified to demonstrate operability in the accident environment for the time required with safety margin to failure.
- 2. Category EA2:
Equipment that could experience t he environmental conditions of design basis line break accidents, includi ng Main Steam Line Break (MSLB),
High Energy Line Brea k/Superheated Blowdown Outside Containment (HELB/SBOC), and/or other line breaks for which it must function to
provide Regulatory Guide 1.97 monitori ng information to the operator. It must be qualified to dem onstrate operability in it s specific applicable accident environment for t he time required with safe ty margin to failure.
- 3. Category EA1:*
Equipment that could experience increased radiation exposure due to Post-LOCA recirculation for which it must function to provide RG 1.97
monitoring information to the operat or. It must be qualified to demonstrate operability in the radiation environment for the time required
with safety margin to failure.
- 4. Category EB1:
Equipment that could experience environmental conditions of Design Basis (LOCA) accidents through which it need not function to provide RG 1.97 monitoring information to the operator, but through which it must not fail in a manner detrimental to any related RG 1.97 monitoring function. It must be qualified to demonstrate the capability to withstand any such accident environment for the time during which it must not fail with safety margin to failure.
3.11-9 Reformatted July 2014 5. Category EB2:
Equipment that could experience t he environmental conditions of design basis line break accidents, includi ng Main Steam Line Break (MSLB),
High Energy Line Brea k/Superheated Blowdown Outside Containment (HELB/SBOC), and/or other line breaks for which it need not function to
provide RG 1.97 monitoring informa tion to the operator, but through which it must not fail in a manner detrimental to any related RG 1.97 monitoring function. It must be qualif ied to demonstrate the capability to
withstand any such specifically applicable accident environment for the time during which it must not fail with safety margin to failure
- 6. Category EB1:*
Equipment that could experience increased radiation exposure due to Post-LOCA recirculation through whic h it need not function to provide RG 1.97 monitoring information to the operator, but through which it must not fail in a manner detrimental to any related RG 1.97 monitoring function. It must be qualified to demonstrate the capability to withstand any such accident environment for the time during which it must not fail with safety margin to failure.
- 7. Category EC1:
Equipment that could experience t he environmental conditions of design basis (LOCA) accidents through wh ich it need not function for
Regulatory Guide 1.97 monitoring purposes and whose failure is not deemed detrimental to the Regulatory Gu ide 1.97 monitoring function. It need not be qualified for a LOCA acci dent environment, but must be qualified for the non-accident environment.
- 8. Category EC2:
Equipment that could experience t he environmental conditions of design basis line break accidents, includi ng Main Steam Line Break (MSLB),
High Energy Line Brea k/Superheated Blowdown Outside Containment (HELB/SBOC), and other line breaks th rough which it need not function for Regulatory Guide 1.97 monitori ng purposes and whose failure is deemed not detrimental to the Regulatory Guide 1.97 monitoring
function. It need not be qualified for these accident environments, but must be qualified for the non-accident environment.
3.11-10 Reformatted July 2014 9. Category EC1:*
Equipment that could experience increased radiation exposure due to Post-LOCA recirculation through which it need not function to provide RG 1.97 monitoring information to the operator and whose failure is
deemed not detrimental to the RG 1.97 monitoring function. It need not be qualified for any Post-LOCA harsh accident environment, but must be qualified for the non-accident environment.
Also equipment that could experience increased radiation exposure due to Post-LOCA recirculation but has completed its RG 1.97 monitoring requirement prior to the environment becoming harsh, and whose failure (after completing any required function) is deemed not detrimental to the RG 1.97 monitoring function. It n eed not be qualified for any Post-LOCA harsh accident environment, but must be qualified for the non-accident and the Post-LOCA mild accident environment.
- 10. Category ED:
Equipment that would not experience environmental conditions of design basis accidents and that must be qual ified to demonstrate operability for its Regulatory Guide 1.97 monitoring function in its normal and abnormal
service environment. This equipment is located outside containment.
3.11.2 ENVIRONMENTAL QUALIFICATION PROGRAM ACCEPTANCE CRITERIA This section describes the environmental qualificat ion program acceptan ce criteria that were employed to meet the fo llowing general requirements:
- a. The equipment was designe d to have the capability of performing its design safety functions under postulated normal, abnormal, accident, and post-accident environments for the length of time for wh ich its function is required plus margin.
- b. The equipment environmental capability was demonstrated by appropriate testing, analyses, and/or oper ating experience.
- c. A quality assurance program meeting the requirements of 10CFR50, Appendix B, was established and implemented to provide assurance that all requirements have been satisfactorily accomplished.
The Virgil C. Summer Nuclear Station is committed to qualificat ion of electrical equipment requiring harsh qualification in accordance with NUREG-0588
[5], Cat. II, which relates to IEEE 323-1971 [6] for the original plant design. However, some equipment was qualif ied to IEEE 323-1974 [7] (NUREG 0588, Cat. 1) requirements.
New and replacement electrical equipment requiring harsh qualification is governed by 3.11-11 Reformatted July 2014 the current regulations of 10CFR 50.49. The status of elec trical equipment qualification in accordance with the applicable IEEE Standards and Regulatory requirements is documented in Virgil C. Summer Nucl ear Station Equipment Qualification Documentation Packages (EQDP's) for equipm ent which requires hard qualification and Equipment Qualification Files (EQF's) for equ ipment which requires mild qualification.
Refer to Section 3.11.3 for further discussion.
3.11.2.1 Conformance with Regulatory Requirements Conformance with General Design Criteria 1, 4, 23, and 50 is discussed in Section 3.1.
3.11.2.1.1 10CFR50 Appen dix A Criterion 4
[8] - Environmental and Missile Design Bases The scope of electrical equipment and the environmental requirements for GDC 4 are addressed in Section 3.11.1.
Refer to Sections 3.5 and 3.
6 for discussions related to missile protection.
Electrical equipment that is required to perform a design safety function is designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with normal operation, maintenan ce, testing, and postulated accidents, including loss-of-coolant accident s in accordance with GDC 4.
3.11.2.1.2 10CFR 50 Appendix B
[9] The Electrical Equipment Environmental Qualification Program is in compliance with the Virgil C. Summer Nuclear Station Quality Assurance Program. T he Quality Assurance Program meets the requirements of 10CFR50, Appendix B.
3.11.2.1.3 Regulatory Guide 1.89
[10] Compliance with Regulatory Guide 1.89, which pertains to the qualification of Class 1E electrical equipment, is also discussed in Appendix 3A. Class 1E electrical equipment is qualified in accordance with IEEE 323-1974, as endorsed by Regulatory Guide 1.89 with the following exceptions:
- a. In cases where the qualification has been demonstrated in accordance with IEEE 323-1971, documentation of that qualification is maintai ned in the form of auditable file packages. Refer to Se ction 3.11.3.2 for discussion.
- b. Specific criteria for assessing the acceptability of the orig inal scope of the environmental qualification pr ogram for safety relat ed electrical equipment requiring harsh qualification is provided by NUREG-0588 Category II, as related to IEEE 323-1971, or Category I as related to IEEE 323-1974.
3.11-12 Reformatted July 2014 Specific criteria for assessing the acceptability of the envir onmental qualification program for electrical equipment added or replaced in harsh environments are provided by 10CFR50.49
[11], dated January 17, 1983.
- c. Mild environment equipm ent qualification by test and/or analysis is not a requirement of 10CFR50.49. Mild environment equipment qualification files are established to provide design, procurement, and maintenance information in a readily accessible form.
3.11.2.1.4 Regulatory Guides 1.30
[12], 1.40 [13], 1.63 [14], and 1.73
[15] The detailed criteria contai ned in these documents as they relate to environmental qualification should be used in conjunction wit h the more comprehensive criteria of NUREG-0588 for evaluating the respective equipment environmental qualification.
Compliance with these Regulatory Gui des is discussed in Appendix 3A.
3.11.2.1.5 Regulatory Guide 1.97
[24] Regulatory Guide 1.97 imposes environmental qualification requirements on electrical components used for monitoring certain plant parameters after an accident. Design and qualification criteria for Regul atory Guide 1.97 instrumentati on, as described in Table 1 of the Regulatory Guide, st ates that the instrumentat ion should be environmentally qualified in accordance with Regulatory Gu ide 1.89 and the methodology as described
in NUREG-0588.
Regulatory Guide 1.97 describes three equi pment categories for design and equipment qualification requirements. V. C. Summer Nuclear Station Regulatory Guide 1.97 category 1 and 2 equipment conforms to Regulatory Guide 1.89 environmental
qualification requirements.
Regulatory Guide 1.97 Ca tegory 3 equipment has no special regulatory requirements but the regulatory position is that the equipment should be of high quality commercial grade and should be selected to withstand the specified
service environment.
3.11.2.2 Qualification Methodologies for Safety-Related Electrical Equipment Safety-related electrical equipment including Reg. Guide 1.97 monitoring equipment requiring harsh environmental qualification supplied by Westinghouse under the NSSS contract is qualified as out lined in Section 3.11.2.2.1.
All other safety-related electrical equipm ent including Reg. Guide 1.97 monitoring equipment requiring harsh environmental qu alification is qualified using the methodologies of Section 3.11.2.2.2. Safety-related electrical equipment requiring mild environmental qualification is qualified as discussed in Exception c to Regulatory Guide 1.89 (refer to Sect ion 3.11.2.1.3).
3.11-13 Reformatted July 2014 The documentation of the application of the methodologies for the specific equipment identified in Section 3.11.1.2, to demonstrate qualif ication to the environmental conditions defined in Section 3.11.1.1, is presented in Section 3.11.3.
3.11.2.2.1 Qualification Tests and Analyses Applicable to the NSSS Electrical Equipment Westinghouse has qualified its NSSS safety-related electrical equipment in accordance with IEEE-323-1971. The Westinghouse supplemental qualification program (Reference 16) is an NRC approved seismic and environmental qualification program, as stated in the Staff letter, D. B. Vassall o to C. Eicheldinger dated November 19, 1975. Mechanical equipment design basis considerati ons are described in Chapters 5, 6, 9, and 10. Mechanical and electrical components have been identified and classified
relative to their safety cla ssification in Section 3.2.
Comprehensive testing and/or analysis is co nducted for those electrical equipment and components which are required to function during and subsequent to any of the design basis accidents and that experience hostile environments. The program consists of performance tests of individual pieces of equipment in the manufacturer's shop, integrated tests of t he system as a whole in the field, and periodic inspection and tests of the activation circuitry and mechanical components to assure reliable performance, upon demand, throughout t he plant lifetime.
The initial qualificat ion tests of individual components and the integrat ed tests of the systems as a whole complement each other to assure performance of the system as designed and to prove proper operation of the actuation circuitry. For engineered safeguard features (ESF) equipment located inside the Reactor Building, qualification testing and/or analysis is performed under the e ffects of the conservative post accident
temperature, pressure, humidit y, radiation, and chemical environment, when applicable.
Routine periodic inspection and testing of ESF equipment is performed as outlined in Technical Specifications.
Chapter 6.0 describes the containment tem perature and pressure response to various sizes of in-containment main steam line rupt ures. Qualified equi pment located inside the Reactor Building is required to provide protection in the unlikely event of one of these breaks.
For the larger steam line breaks evaluated, the steam line pressure instrumentation, which is located outside containment, will init iate safety injection on low steam line pressure. RN 01-113 3.11-14 Reformatted July 2014 3.11.2.2.2 Qualification Tests and Analyses Applicab le to the BOP Electrical Equipment Balance of plant electrical equipment, incl uding cabling, is desi gned to accommodate the effects of, and to be compatible with, the environmental condi tions associated with the location of the equipment. The environmental conditions considered include those expected during normal operation, maintenance, testing, and, if applicable, post accident periods. The ESF mechanical and electrical equipm ent and instrumentat ion associated with balance of plant systems inside the Reactor Building are designed to perform required functions under the conservative accident and post-accident temperature, pressure, humidity, radiation, and chemical conditions.
Where design of balance of plant equipment to withstand dy namic effects of missiles, pipe whip, and jet forces was impractical, barriers were designed to protect such equipment (refer to Se ctions 3.5 and 3.6).
Safety-related equipment located outdoors is ei ther qualified for ex pected environmental conditions or is protect ed from such conditions.
Qualification of BOP electric al equipment is accomplished by type testing, analysis, and/or documented operating experience. Electrical equipment requiring harsh environmental qualification is qualified in accordance with IEEE 323 and ancillary daughter standards (e.g., IEEE Std. 317
[17], 334 [18], 382 [19], 383 [20]). Although type testing is the preferred met hod of qualification, equipment qualification usually involves some combination of the three methods. The qualification methods used depend on a number of factors, including:
- 1. Material used in construction of the equipment.
- 2. Applicable normal, abnormal, a ccident, and post-accident environmental conditions.
- 3. Operational requirements (during and after accidents).
- 4. Nature of safety function(s).
- 5. Size of equipment.
- 6. Dynamic characteristics of expected fa ilure modes (structural or functional).
In general, analysis is used to supplement test data, although equip ment requiring mild environmental qualification and simple components may lend themselves to analysis in lieu of full scale testing. The role of operat ing experience is generally limited to aiding in determining realistic performance goals.
3.11-15 Reformatted July 2014 Equipment samples selected for qualificat ion are of the same basic design and materials as the equipment to be installed at the Virgil C. Summer Nuclear Station. The sample is manufactured using similar tec hniques and processes as those used for the installed equipment. Any significant vari ations or deviations are noted in the qualification results with justification provided as necessary.
The list of electrical equipment subject to environmental qualificat ion is documented by the Equipment Qualification Database (E QDB). The EQDB lists the electrical equipment as identified in Section 3.11.1.2. NUREG-0588 se ts forth NRC positions in implementation of IEEE 323-1971 and 1974 versions of the "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations" and was used as the basis for assessing the acceptability of the original scope of the environmental
qualification program. Electr ical equipment for replacements or modifications to the plant is procured under the guidance of 10CFR50.49 and is qual ified under the requirements of IEEE 323-1974 and Regulatory Guide 1.89, if the equipment requires harsh environmental qualification.
3.11.2.2.2.1 Main Steam Line Break Inside Containment Equipment Qualification
Main steam line breaks (MSLB) inside containment have previously been discussed in Section 6.2.
The composite temperature profiles for environmental zones subject to MSLB
conditions are presented in dr awing S-021-018. Electrical equipment located in these environmental zones and required to mitigate the MSLB accident has been qualified to these temperature profiles as doc umented in the applicable EQDP's.
3.11.2.2.2.2 Main Steam Line Break Outsi de Containment Equipment Qualification
Main steam line breaks (MSLB) outside containment have previously been discussed in Section 3.6. However, as a resu lt of Information Notice No. 84-90
[21], South Carolina Electric and Gas Company has elected to address the HELB/SBOC (High Energy Line Break/Superheated Blowdown Outside Containm ent) accident scenario as a Virgil C.
Summer Plant requirement and as a modifica tion to the previously postulated MSLB accident scenario. A complete analysis of the HELB/SBOC environmental conditions
and equipment qualification is presented in Refe rence 22. An increase in temperature in certain areas of the East and West Penetration Access areas and the Intermediate Building is the only significant environmental change resulti ng from postulated HELB/SBOC ruptures relative to that of the MSLB acci dent environment previously evaluated and identified in Sect ion 3.6. All other environmental conditions remain unchanged.
3.11-16 Reformatted July 2014 Steamline breaks resulting in HELB/SBOC can only occur in the 436 ft. floor elevation of the East and West Penetration Access areas, 436 ft. floor elevation of the Intermediate Building, and in the Turbine Building. Br eaks postulated in the 4-inch steamlines supplying the turbine driven EFW Pump result in a break area too small to lead to a
superheated steam discharge per Westinghouse WCAP-10961
[23]; therefore, the licensing analysis for the 4-inch line remains unchanged. HELB/SBOC resulting from postulated breaks in large steam lines need not be addressed for the Turbine Building since equipment in this area is not required in order to mitigate the consequences of an HELB/SBOC.
For each environmental zone in which the ambient temperature profile for the HELB/SBOC exceeds that prev iously established for a MSLB, a composite temperature profile was generated. The composite temperature profiles for these environmental zones are presented in drawing S-021-018.
Electrical equipment located in these environmental zones and required to miti gate the HELB/SBOC accident has been
qualified to these te mperature profiles.
Qualification of the r equired equipment in superheat high energy line break environmental zones was accomplished by a se ries of detailed en gineering analyses. In some cases, these analyses are based on specific as-built hardware and location specific transient heat transfer analyses. Qu alification for each component is based on qualification test data, but in some ca ses this data has been extrapolated using acceptable analytical techniques. The evaluations specifically address the higher temperature effect of the HELB/SBOC accid ent and the capability of the equipment to withstand the accident conditions for the operat ing time required to perform its safety function. These evaluations have demons trated that the r equired equipment as installed, is environmentally qualified for the postulat ed environmental conditions.
3.11.2.2.3 QUALIFICATION MAINTENANCE Qualification is not a guarantee of performance for each component of a system. It is rather, assurance that the system can perform its safety function under all specified service conditions. Maintenance of the qualifi ed status of harsh en vironment electrical equipment requires scheduled maintenance to prevent components from exceeding their qualified life, and periodic testing to loca te components that may have failed, or be near failure.
Scheduled maintenance is also performed on mild environment electrical equipment to ensure proper operation of t he equipment throughout the established service life.
Scheduled maintenance activities required to maintain harsh and mild environment equipment qualification is specified for all safety-related electrical equipment. These activities are documented in the Equipment Qualific ation Database.
3.11-17 Reformatted July 2014 3.11.3 QUALIFICATION TEST RESULTS This section addresses the qualif ication test results applicable to Nuclear Steam Supply System and Balance of Plant Electrical equipment. This equipment is required to function under anticipated normal operating conditions, and/or is required to function to mitigate the consequences of design basis accidents, including LOCA and MSLB inside containment and MSLB outside containment.
In addition to the qualification results provided in this section, a comple te analysis of the High Energy Line Break/Superheated Blowdown Outside Containment accident environmental conditions, as discussed in Section 3.11.2.2.2.
2, is presented in Reference 22.
The results of the qualificat ion program for each type of electrical equipment are recorded in the applicable Equipment Qualif ication Documentation Packages (EQDP's) and/or Equipment Qualification Files (EQF's). The collection of various computer data files containing information relative to equipment qualification and qualified equipment are included in an Equipment Qualification Database (EQDB).
Electrical equipment and data relative to environmental qua lification are listed in the EQDB.
Westinghouse NSSS supplied safety-related equipment, which is required to function to mitigate the consequences of a postulated accident and which may be exposed to the elevated environmental condition s that may result from t he accident, is qualified by Westinghouse under the 1971 version of IEEE-323. The Westinghouse supplemental qualification program (Reference 16) has been accepted by the NRC staff as meeting the requirements of IEEE 323-19 71 in NRC letter D. B. Vassallo to C. Eicheldinger dated November 19, 1975.
A portion of the Westinghouse supplied electrical equipment has subsequently been upgraded to the qualification requirements of IEEE 323-1974, by WCAP 8587 and WCAP 8687. This equipment upgrade has been documented in their respective EQDP's and EQF's.
3.11.3.1 Qualified Equipment List The "Qualified Equipment/Components and Materials" list provides a listing of equipment, components, and materials for wh ich environmental qualification is
maintained. The list includes items subjec t to both mild and harsh environments and comprises two sections of the EQDB. One section is arranged alphanumerically by equipment number while the other section is arranged alphanumerically by system.
3.11.3.2 Auditable File The auditable files are arranged in equipment qualification documentation packages or equipment qualification files by Qualification File Number. The "Equipment Qualification File Index" (part of EQDB) provides a listing of the EQDP's and EQF's and specifies the components to which each qualif ication file applies.
3.11-18 Reformatted July 2014 3.11.3.3 Master Equipment List The "Master Equipment List" is contained wi thin the EQDB and provides a listing of equipment and materials to which the require ments of 10CFR50.49 for environmental qualification of electric al equipment subject to harsh environment applies.
3.11.4 LOSS OF VENTILATION Safety-related electrical equipment and component s, as identified in Section 3.11.1.2, are located in areas which are mechanically cooled or ventilated by safety-related or quality related HVAC systems.
Safety-related HVAC systems providing cooling or ventilation are desig ned to satisfy the following considerations:
- 1. Seismic Categor y I requirements.
- 2. Redundant active system component s are provided, as required.
- 3. Independent and redundant Class 1E power sources are provided.
- 4. Arrangement is such that single failure of an active or passive component does not result in loss of required cooling function.
In certain cases, such as rooms housing only one train of safety-related electrical equipment, each room is serviced by a si ngle air handling unit. In these cases the safety-related HVAC systems, such as those servicing the Residual Heat Removal/Reactor Building Spray Pump Rooms, the Charging/SI Pump Rooms, the ESF Switchgear and Speed Switch Rooms, and the Auxiliary Build ing Motor Control Center Rooms, are designed such that no single failure can cause loss of cooling to more than one room, and subsequently to more than one train of redundant safety-related electrical equipment.
Quality Related (QR) HVAC equi pment provides cooling or v entilation for some areas of the plant, outside the Reactor Building, which contain Class 1E electrical equipment. Class 1E equipment in these areas was evaluated on a case by case basis and is
designed to function in the unlikely event of abnormal environmental conditions caused
by loss of non-safety related HVAC. Loss of QR HVAC does not have an immediate effect on, or correlation to, the performance of the Class 1E equipment safety function. However, since credit is taken for normal c onditions of HVAC operation in determining qualified life of Class 1E equipment locat ed in harsh environmental areas and in determining service life of Class 1E equipment located in mild environmental areas, loss of QR HVAC systems could have a long term effect on equipment life expectancy.
Therefore, QR HVAC is procedurally cont rolled and operability of QR HVAC systems is monitored to ensure that air flow to t hese areas, and subsequently the environmental conditions upon which the Class 1E equipment qualification was based, is maintained.
3.11-19 Reformatted July 2014 Controls and electrical equipment necessa ry for operation of safety-related HVAC systems outside the Reactor Building, fo llowing a LOCA or high energy line break condition, are located such that they are not exposed to post-accident environmental conditions, or are designed to withstand t hese severe conditions. Controls and electrical equipment required by safety-related HVAC systems within the Reactor Building are capable of withs tanding the worst case environmental conditions resulting from a DBA.
Environmental test reports describing qualif ication of ventilation and cooling equipment located inside and outside the Reactor Build ing are referenced in the applicable equipment qualificat ion data packages, and equipment qualification files.
The preceding discussions result in the det ermination that loss of ventilation, although highly unlikely, will not prev ent performance of the Class 1E equipment safety function or affect the environmental qualif ication status of the safety-related electrical equipment.
3.11.5 ESTIMATED CHEMICAL AND RADIATION ENVIRONMENT This section presents the ju stification for the estimat ed chemical and radiation environments of Section 3.11.1.
3.11.5.1 Chemical Environment 3.11.5.1.1 Normal Operation Adverse chemical environmental conditions do not exist during normal plant operation.
3.11.5.1.2 Design Basis Accident The chemical spray environment for which electrical equipment inside containment must be qualified is based upon a maximum operating time for the spray system of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> based on IEEE 323 testing standards. During a design basis LOCA, the containment spray system will be operated for a minimum period of four hours and up to a maximum of 40 days as required to return containment pressure a nd temperature conditions to normal levels. Therefore, a period of up to 40 days has been used as a basis for
judging the adequacy of electrical equipment qualification.
The chemical spray environment for which electrical equipment inside containment must be qualified is based upon the following post-accident operating envelopes and spray
pH conditions:
Operating Period Spray pH Range 0-2 hours (minimum) 8.7-10.5 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> - 40 days (maximum) 8.0-8.5 RN 03-008 12-034 RN 03-008 3.11-20 Reformatted July 2014 These chemical spray environmental (pH) conditions are based upon analyzing the drawdown of the Refueling Water Stor age Tank (RWST) and Sodium Hydroxide Storage Tank (SHST) to develop a buffered borated water solution in the spray header. The analysis is performed for the range of boron sources (RWST, RGS, SIS), and sodium hydroxide concentrations (20-22 wt/%) required by the Technical Specifications in conjunction with the design, normal, and degraded operating modes discussed in Section 6.2.2.2.1.2.
Depending upon the spray system operating mode, the results of the analysis yield a spray pH range of 8.8 to 10.1 dur ing the drawdown of the RWST ( 23 to 65 minutes post LOCA). At the completion of the RWST drawdown, the spray system operation is maintained by recirculation of the Reactor Building Sump Water (pH:
7.5 - 8.5) and the injection of any remaining NaOH from the SH ST. Upon recirculation from the sump the spray system pH is maintained in the range of 8.7 to 10.5 until the SHST is emptied (approximately 10 to 40 minutes). Thereafter, the spray pH is equal to the sump water pH of 8.1 to 8.5. Refer to Section 6.2.2.3.1.4 for a detailed discussion of drawdown analysis.
3.11.5.2 Radiation Environment 3.11.5.2.1 Normal Operation The design basis radiation sources and dose rates for various plant systems and equipment during normal plant operation are discussed in Chapters 11 and 12. The neutron and gamma radiation source terms and energy spectra data for major equipment is summarized in Tables 12.1-3 through 12.1-17. Based upon these source terms and plant shielding, t he plant radiation exposure z ones for normal operation, shutdown, and refueling are presented in Fi gures 12.1-1 through 12.1-20. The total integrated (over 40 years) radiation doses re sulting for normal plant operation are given in EQDB drawing S-021-018.
3.11.5.2.2 Design Basis Accident The design basis post accident radiation so urces and doses for vital plant systems and equipment are discussed in Chapter 12A.
The radiation sour ces are based upon NUREG 0737, Section II.B.2. The post-acci dent total integrated (over 1 year) gamma and beta doses addressing the requirement s of NUREG-0588 are given in EQDB drawing S-021-018. RN 03-008 3.11-21 Reformatted July 2014 3.
11.6 REFERENCES
- 1. Institute of Electrical and Electronics Engineers (IEEE), "Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations", Standard 308, dated 1971.
- 2. Virgil C. Summer Nuclear Station Equipment Qualification Database.
- 3. Virgil C. Summer Nuclear Station Environmental Zone Maps SS-021-001 through SS-021-017.
- 4. Virgil C. Summer Nuclear Station Equi pment Qualification Da tabase Environmental Zone Information drawing S-021-018.
- 5. NUREG-0588, "Interim Staff Position on Environm ental Qualification of Safety-Related Electrical Equipment".
- 6. Institute of Electrical and Electronics Engineers (IEEE), "General Guide for Qualifying Class 1E Electrical Equipment for Nuclear Po wer Generating Stations", Standard 323, dated 1971.
- 7. Institute of Electrical and Electronics Engineers (IEEE), "Standard for Qualifying Class 1E Equipment Nuclear Power Generating Stations", Standard 323, dated 1974.
- 8. 10 CFR Part 50, Appendix A - General De sign Criterion 4, "Environmental and Missile Design Bases".
- 9. 10 CFR Part 50, Appendix B - Quality Assu rance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
- 10. U.S. Nuclear Regulatory Commission Re gulatory Guide 1.89, "Qualification of Class 1E Equipment for Nuclear Power Plants".
- 11. 10 CFR Part 50, Section 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants".
- 12. U.S. Nuclear Regulatory Commission Re gulatory Guide 1.30, "Q/A Requirements for the Installation, Inspection and Testi ng of Instrument and Electric Equipment".
- 13. U.S. Nuclear Regulatory Commission Regul atory Guide 1.40, "Q ualification Tests of Continuous Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants".
3.11-22 Reformatted July 2014 14. U.S. Nuclear Regulatory Commission Regul atory Guide 1.63, "Electric Penetration Assemblies in Containment Structures for Light Water-Cooled Nuclear Power Plants". 15. U.S. Nuclear Regulatory Commission Regul atory Guide 1.73, "Q ualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power
Plants".
- 16. Letter NS-CE-692, dated July 10, 1975, from C. Eicheldinger (Westinghouse) to D. B. Vassallo (NRC).
- 17. Institute of Electrical and Electronics Engineers (IEEE), "Electric Penetration Assemblies in Containment Structures for Nuclear Power G enerating Stations", Standard 317, dated 1972.
- 18. Institute of Electrical and Electronics Engineers (IEEE), "Standard for Type Tests of Continuous Duty Class 1E Motors fo r Nuclear Power Generating Stations", Standard 334, dated 1971.
- 19. Institute of Electrical and Electronics Engineers (IEEE), "Standard for Qualification of Safety-Related Valve Act uators", Standard 382, dated 1972.
- 20. Institute of Electrical and Electronics Engineers (IEEE), "Standard for Type Tests of Class 1E Electric Cables, Field Splices and Connections for Nuclear Power Generating Stations", Standard 383, dated 1974.
- 21. "Main Steam Line Break Effect on Environmental Qualificat ion of Equipment", Nuclear Regulatory Commission Information Notice No. 84-90, December 7, 1984.
- 22. "Evaluation to Address Environmental Qualification of Qualified Equipment Subjected to a High Energy (Steam) Li ne Break of Superheated Blowdown Outside Containment and a Main Steam Line Break (MSLB) Inside Containment", Gilbert Associates, Inc. Report No. 2616.
- 23. "Steamline Break Mass/Energy Releases for Equipment Environmental Qualification Outside Containment", Westinghouse Topical Report, WCAP-10961, October 1985.
- 24. U.S. Nuclear Regulatory Commission Regul atory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs
Conditions During and Following an Accident".
3.11-23AMENDMENT 97-01AUGUST 1997TABLE 3.11-0NUCLEAR STEAM SUPPLY SYSTEM CLASS 1E EQUIPMENTIN CONTAINMENTTABLE 3.11-0ABALANCE OF PLANTCLASS 1E EQUIPMENTThis information is contained in the computerized EquipmentQualification Database loaded on the VAX System 3.11-24AMENDMENT 97-01AUGUST 1997TABLE 3.11-1EQUIPMENT AND COMPONENTS, OTHER THAN CLASS 1E,INSIDE THE REACTOR BUILDING REQUIRED TO FUNCTIONDURING AND/OR AFTER AN ACCIDENT1.Containment Spray SystemPipingSpray Header and Nozzles2.Reactor Building Cooling Units3.Containment Isolation SystemIsolation ValvesMechanical Penetrations
Air Locks and Hatches4.Emergency Core Cooling SystemAccumulators Valves and Piping
_____________________
NOTES:1.Unless otherwise indicated, equipment is located outside thesecondary shield wall.2.Refer to drawing S-200-971, "Essential Equipment List", whichincludes a listing of equipment and components, other thanClass 1E, required for safe shutdown and/or design basis accident, condition IV, mitigation.
3.11-25AMENDMENT 97-01AUGUST 1997TABLE 3.11-2EQUIPMENT AND COMPONENTS, OTHER THAN CLASS 1E,OUTSIDE THE REACTOR BUILDING REQUIRED TO FUNCTIONDURING AND/OR AFTER AN ACCIDENT1.Service Water SystemPumpsValves and Piping Heat Exchangers2.Component Cooling Water SystemPumps Valves Heat Exchangers Surge Tank3.Containment Isolation SystemIsolation Valves Mechanical Penetrations Air Locks and Hatches4.Main Steam SystemValves and Piping to Turbine Driven Emergency Feedwater Pump5.HVAC Chilled Water SystemPumps Valves and Piping Chillers6.Building Ventilation and Cooling Systems for the Control Room, Relay Room; RHR/RB SprayPump Room s; Charging/SI Pump Rooms, Auxiliary Building Motor Control Center andSwitchgear Areas; ESF Switchgear and Speed Switch Rooms, Battery Rooms, Service Water Booster Pump Areas, Emergency Feedwater Pump Rooms; Diesel Generator Building; and Service Water Pumphouse.Air Handling UnitsFilters Fans Ducts, Dampers, Valves and Piping7.Emergency Core Cooling SystemRefueling Water Storage Tank Charging Pumps RHR Pumps RHR Heat Exchanger Valves and Piping (see Section 6.3) 3.11-26AMENDMENT 97-01AUGUST 1997TABLE 3.11-2 (Continued)8.Reactor Building Spray SystemReactor Building Spray PumpValves and Piping Refueling Water Storage Tank Sodium Hydroxide Storage Tank9.Spent Fuel Cooling SystemPumps Heat Exchangers Valves and Piping10.Emergency Feedwater SystemCondensate Storage Tank Motor Driven Emergency Feedwater Pumps Turbine Driven Emergency Feedwater Pump Valves and Piping
_____________________
NOTE:Refer to drawing S-200-971, "Essential Equipment List", which includes a listing of equipment and com ponents, other than Class 1E, required for safe shutdown and/ordesign basis accident, condition IV, mitigation.
3.11-27AMENDMENT 97-01AUGUST 1997TABLE 3.11-3POSTULATED ENVIRONMENTAL CONDITIONSTable 3.11-3 environmental data have been combined in a series of controlleddrawings.An equipment qualification database with environmental zone information, asdocumented by EQDB drawing S-021-018, provides a list of the environmental zonesand conditions, including the normal, abnormal, and accident (including post-accident) environmental conditions for each environmental zone.