ML18117A094
| ML18117A094 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 04/26/2018 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18117A085 | List:
|
| References | |
| L-2018-103 | |
| Download: ML18117A094 (197) | |
Text
TABLE OF CONTENTS
Section Title Page
6.0 ENGINEERED
SAFETY FEATURES 6.1-1
6.1 Criteria
6.1-2
6.1.1 Engineered
Safety Features Basis for Design 6.1-2 Reliability and Testability of Engineered Safety Features 6.1-4 Missile Protection 6.1-4 Engineered Safety Features Performance Capability 6.1-5 Engineered Safety Features Components Capability 6.1-6 Accident Aggravation Prevention 6.1-7 Sharing of Systems 6.1-7 6.1.2 Related Criteria 6.1-8
6.2 Safety
Injection System 6.2-1 6.2.1 Design Basis 6.2-1 Emergency Core Cooling System Capability 6.2-1 Inspection of Emergency Core Cooling System 6.2-2 Testing of Emergency Core Cooling System Components 6.2-2 Testing of Emergency Core Cooling System 6.2-3 Testing of Operational Sequence of Emergency Core Cooling System 6.2-4 Codes and Classifications 6.2-4 Service Life 6.2-4 6.2.2 System Design and Operation 6.2-5 System Description 6.2-5 Injection Phase 6.2-6 Recirculation Phase 6.2-7 pH Control 6.2-10 Recirculation Phase - Hot Leg 6.2-10 Recirculation Phase - Cold Leg 6.2-10 Alternate Flow Path to Low Head-Safety Injection 6.2-11 Cooling Water 6.2-11 Component Cooling System 6.2-11 Intake Cooling Water System 6.2-11 Change-Over from Injection Phase to Recirculation Phase 6.2-12 Location of Major Components Required for Recirculation 6.2-13 Components 6.2-13 Accumulators 6.2-14 Boron Injection Tank 6.2-14 Refueling Water Storage Tank 6.2-15 Safety Injection Pumps 6.2-16 Heat Exchangers 6.2-17 Valves 6.2-18 Motor Operated Valves 6.2-20 Manual Valves 6.2-21 Accumulator Check Valves 6.2-22 Relief Valves 6.2-23 Leakage Limitations 6.2-24 Piping 6.2-24 Pump and Valve Motors 6.2-27 Motors Outside the Containment 6.2-27 Motors Inside the Containment 6.2-27 Electrical Supply 6.2-28 Protection Against Dynamic Effects 6.2-29 6-i Revised 04/17/2013 TABLE OF CONTENTS (Continued)
Section Title Page
6.2.3 Design
Evaluation 6.2-30 Range of Core Protection 6.2-30 System Response 6.2-32 Single Failure Analysis 6.2-33 Reliance on Interconnected Systems 6.2-34 Component Function Evaluation 6.2-35 Passive Systems 6.2-35 Emergency Flow to the Core 6.2-36 Recirculation Loop Leakage 6.2-36 Pump NPSH Requirements 6.2-37 Residual Heat Removal Pumps 6.2-37 Safety Injection Pumps 6.2-37 6.2.4 Inspections and Tests 6.2-38 Inspection Capability 6.2-38 Pre-Operational Testing 6.2-38 Component Testing 6.2-38 System Testing 6.2-39 Post-Operational Testing 6.2-40 Component Testing 6.2-40 System Testing 6.2-40
6.2.5 References
6.2.42
6.3 Emergency
Containment Cooling 6.3-1 6.3.1 Design Basis 6.3-1 Containment Heat Removal Systems 6.3-1 Performance Objectives 6.3-1 6.3.2 System Design 6.3-2 Cooling System Characteristics 6.3-2 Filtering System Characteristics 6.3-4 Redundancy Provisions 6.3-4 Actuation Provisions 6.3-5 Environmental Protection 6.3-5 6.3.3 Design Evaluation 6.3-6 Flow Tests and Analysis 6.3-6 Reliance on Interconnected Systems 6.3-6 6.3.4 Minimum Operating Conditions 6.3-6 6.3.5 Tests and Inspections 6.3-6 Inspections of Containment Pressure-Reducing Systems 6.3-6 Testing of Containment Pressure-Reducing Systems Components 6.3-7 Testing of Operational Sequence of Containment Pressure-Reducing Systems 6.3-7 6.3.6 Motors for Emergency Containment Fans 6.3-8 General 6.3-8 Insulation 6.3-8 Bearings 6.3-8 6.3.7 References 6.3-8
6-ii Revised 04/17/2013
TABLE OF CONTENTS (Continued)
Section Title Page
6.4 Containment
Spray System 6.4-1
6.4.1 Design
Bases 6.4-1 Containment Heat Removal Systems 6.4-1 Inspection of Containment Pressure Reducing Systems 6.4-2 Testing of Containment Pressure - Reducing Systems Components 6.4-2 Testing of Containment Spray Systems 6.4-3 Testing of Operational Sequence of Containment Pressure-Reducing Systems 6.4-4 Performance Objectives 6.4-4 Inspection of Air Cleanup Systems 6.4-4 Engineered Safety Features Components Capability 6.4-4 Accident Aggravation Prevention 6.4-5 Service Life 6.4-6 Codes and Classifications 6.4-6 6.4.2 System Design and Operation 6.4-6 System Description 6.4-6 Recirculation Phase 6.4-7 Cooling Water 6.4-8 Change-Over 6.4-8 Components 6.4-8 Pumps 6.4-8 Spray Nozzles 6.4-9 Heat Exchangers 6.4-10 Valves 6.4-10 Piping 6.4-10 NaTB Baskets 6.4-10 Motors for Pumps and Valves 6.4-11 Electrical Supply 6.4-12 Environmental Protection 6.4-12 Material Compatibility 6.4-12 6.4.3 Design Evaluation 6.4-12 Range of Containment Protection 6.4-12 System Response 6.4-13 Single Failure Analysis 6.4-13 Reliance on Interconnected Systems 6.4-14 6.4.4 Inspections and Tests 6.4-14 Inspection Capability 6.4-14 Component Testing 6.4-14 System Testing 6.4-15 Operational Sequence Testing 6.4-15 6.4.5 References 6.4-16
6-iii Revised 04/17/2013
TABLE OF CONTENTS (Continued)
Section Title Page
6.5 Leakage
Detection and Provisions for the Reactor and Auxiliary Coolant Loop 6.5-1
6.5.1 Leakage
Detection Systems 6.5-1 Design Bases 6.5-1 Monitoring Reactor Coolant Leakage 6.5-1 Monitoring Radioactivity Releases 6.5-1 Principles of Design 6.5-2 Systems Design and Operation 6.5-2 Reactor Coolant System 6.5-3 Containment Air Particulate Monitor 6.5-3 Containment Radioactive Gas Monitor 6.5-5 Component Cooling Liquid Monitors 6.5-7 Reactor Vessel Head Leakage Detection System 6.5-7 Condenser Air Ejector Gas Monitor R-*-15, RAD-*-6417 6.5-8 Steam Generator Liquid Sample Monitor 6.5-8 Pump Activity 6.5-9 Liquid Inventory 6.5-10 Residual Heat Removal Loop 6.5-10 Component Cooling Loop 6.5-11 6.5.2 Leakage Provisions 6.5-12 Design Basis 6.5-12a Design and Operation 6.5-12a Reactor Coolant System 6.5-12a Residual Heat Removal Loop 6.5-13 Component Cooling Loop 6.5-14 6.5.3 References 6.5-15
6.6 Containment
Isolation 6.6-1 6.6.1 Design Basis 6.6-1 6.6.2 System Design 6.6-2 6.6.2.1 Barriers 6.6-2 6.6.2.2 Containment Isolation Valves 6.6-4 6.6.2.3 Isolation Actuation 6.6-7 6.6.3 Testing and Inspection 6.6-7
6.7 Environmental
Qualification Testing of Safeguard Components 6.7-1 6.7.1 General 6.7-1
6-iv Revised 04/17/2013
LIST OF TABLES
Table Title
6.2-1 Safety Injection System - Code Requirements
6.2-2 Instrumentation Readouts on the Control Board for Operator Monitoring During Recirculation
6.2-3 Quality Standards of Safety Injection Systems Components
6.2-4 Accumulator Design Parameters
6.2-5 Boron Injection Tank
6.2-6 Refueling Water Storage Tank Design Parameters
6.2-7 Pump Parameters
6.2-8 Residual Heat Exchangers Design Parameters
6.2-9(a) Single Failure Analysis - Safety Injection System
6.2-9(b) Loss of Recirculation Flow Path
6.2-10 Component Functions Evaluation
6.2-11 DELETED
6.2-12 Maximum Potential Recirculation Loop Leakage
6.2-13 Residual Heat Removal System Design, Operation and Test Conditions
6.3-1 Emergency Containment Cooling Unit Characteristics 6.3-2 Emergency Containment Cooler Fan and Motor Data 6.4-1 Containment Spray System - Code Requirements
6.4-2 Containment Spray Pump Design Parameters
6.4-3 Single Failure Analysis - Containment Spray System
6.4-4 Sodium Tetraborate Decahyrate (NaTB) Basket Design Parameters 6.6-1 Containment Piping Penetrations and Isolation Barriers
6.6-2 Single Failure Analysis - Containment Purge Valves
6.6-3 Automatic Containment Isolation Valve Closure Times
6-v Revised 04/17/2013
LIST OF FIGURES
Figure Title
6.2-1 Residual Heat Removal System (Unit 3)
6.2-1a DELETED
6.2-1b DELETED
6.2-2 Containment Post LOCA Recirculation
6.2-3 Safety Injection Pump Performance Characteristics
6.2-4 Range of Protection by Safety Injection System
6.2-5 Residual Heat Removal System (Unit 4)
6.2-6 Safety Injection System (Unit 3)
6.2-7 Safety Injection System (Unit 3)
6.2-8 Safety Injection System (Unit 4)
6.2-9 Safety Injection System (Unit 4)
6.2-10 Safety Injection Accumulator System Inside Containment (Unit 3)
6.2-11 Safety Injection Accumulator System Inside Containment (Unit 4)
6.2-12 Containment Sump Strainer Piping Lay-out (Unit 4)
6.2-13 Containment Sump Strainer Piping Lay-out (Unit 3)
6.3-1 Emergency Containment Cooler
6.3-2 DELETED
6.3-3 Filter Spray System
6.3-4 Plan of Containment Laydown and Storage Area, El 58' -0"
6.3-5 DELETED
6.4-1 Containment Spray Pump Performance Characteristics
6.4-2 Containment Spray System (Unit 3)
6.4-3 Containment Spray System (Unit 4)
6.6-1 Valve Arrangements - Sheets 1 thru 20
6.6-2 Legend and General Notes
6-vi Revised 04/17/2013 6 ENGINEERED SAFETY FEATURES The safety objective in reactor design and operation is control of reactor
fission products. The methods used to assure this objective are:
a) Core design to preclude release of fission products from the fuel clad (Section 3) boundary.
b) Retention of fission products by the reactor coolant system boundary for whatever leakage occurs (Sections 4, and 6).
c) Retention of fission products by the containment boundary for operational and accidental releases beyond the reactor coolant boundary (Sections 5, and 6).
d) Limit fission product release to minimize population exposure. (Sections 2, and 11).
The Engineered Safety Features are the provisions which implement methods b, c
and d above to prevent the occurrence or to minimize the effects of serious
accidents.
The Engineered Safety Features are the Containment System, detailed in Section 5;
the core Safety Injection System, detailed in Section 6.2; the Emergency
Containment Cooling and Filtering System, detailed in Section 6.3; and the
Containment Spray System, detailed in Section 6.4. Leak detection is discussed
in Section 6.5.
Evaluation of techniques and equipment used to accomplish the safety objective
including accident cases are detailed in Section 5, 6 and 14.
6.1-1 Revised 04/17/2013
6.1 CRITERIA
Criteria applying in common to all engineered safety features are given in
Section 6.1.1. Thereafter, criteria which are related to engineered safety
features but are more specific to other features or systems are listed and cross
referenced in this section.
Those criteria which are specific to one of the engineered safety features are
discussed in the description of that system.
6.1.1 ENGINEERED
SAFETY FEATURES CRITERIA
Engineered Safety Features Basis for Design
Criterion: Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant
pressure boundary, and their protection systems. Such engineered
safety features shall be designed to cope with any size reactor
coolant piping break up to and including the equivalent of a
circumferential rupture of any pipe in that boundary, assuming
unobstructed discharge from both ends. (1967 Proposed GDC 37)
The design, fabrication, testing and inspection of the core, reactor coolant
system pressure boundary and their protection systems give assurance of safe and
reliable operation under all anticipated normal, transient, and accident
conditions. However, engineered safety features are provided in the facility to
back up the safety provided by these components. These engineered safety
features have been designed to cope with any size reactor coolant pipe break up
to and including the circumferential rupture of any pipe assuming unobstructed
discharge from both ends.
The release of fission products from the reactor fuel is limited by the Safety
Injection System which, by cooling the core, keeps the fuel in place and
substantially intact and limits the metal-water reaction to an insignificant
amount.
6.1-2 Revised 04/17/2013 The Safety Injection System consists of high and low head centrifugal pumps
driven by electric motors, and passive accumulator tanks which are self energized
and which act independently of any actuation signal or power source. The release
of fission products from the containment is limited in three ways:
- 1. Blocking the potential leakage paths from the containment. This is accomplished by:
- a. A steel-lined, concrete containment with testable penetrations.
- b. Isolation of process lines by the Containment Isolation System which imposes double barriers in each line which penetrates the containment.
- 2. Reducing the fission product concentration in the containment atmosphere by containment spray.
- 3. Reducing the containment pressure and thereby limiting the driving potential for fission product leakage by cooling the containment atmosphere
using the following independent systems:
- a. Containment Spray System
- b. Emergency Containment Cooling System
6.1-3 Revised 04/17/2013 Reliability and Testability of Engineered Safety Features Criterion: All engineered safety features shall be designed to provide such functional reliability and ready testability as is necessary to avoid
undue risk to the health and safety of the public. (1967 Proposed
A comprehensive program of testing is formulated for all equipment systems and system control vital to the functioning of engineered safety features. The
program consists of performance tests of individual pieces of equipment in the
manufacturer's shop, integrated tests of the system as a whole, and periodic
tests of the actuation circuitry and mechanical components to assure reliable
performance.
The initial tests of individual components and the integrated test of the system
as a whole complement each other to assure performance of the system as designed
and to prove proper operation of the actuation circuitry.
The engineered safety features components are designed to provide for routine
periodic testing.
Missile Protection
Criterion: Adequate protection for those engineered safety features, the failure of which could cause an undue risk to the health and safety of the
public, shall be provided against dynamic effects and missiles that
might result from plant equipment failures. (1967 Proposed GDC 40)
Equipment failure might result in dynamic effects or missiles. For engineered
safety features which are required to ensure safety in the event of such
equipment failure, protection is provided primarily by the provisions which are
taken in the design to prevent the generation of missiles. In addition, protection is also provided by the layout of equipment or by missile barriers in
certain cases. Reference is made to Appendix 5E for a discussion of missile
protection.
6.1-4 Revised 04/17/2013 Layout and structural design specifically protect injection paths leading to
unbroken reactor coolant loops against damage as a result of the maximum reactor
coolant pipe rupture. (However, dynamic effects of postulated primary loop pipe
ruptures have been eliminated from the Turkey Point design basis based on the
resolution of Generic Letter 84-04, "Asymmetric LOCA Loads," in NRC letter dated
November 28, 1988.) Injection lines penetrate the main missile barrier, and the
injection headers are located in the missile- protected area between the missile
barrier and the containment outside wall. Individual injection lines, connected
to the injection header, pass through the barrier and then connect to the loops.
Separation of the individual injection lines is provided to the maximum extent
practicable. Movement of the injection line, associated with rupture of a
reactor coolant loop is accommodated by line flexibility and by the design of the
pipe supports such that no damage outside the missile barrier is possible.
All hangers, stops and anchors are designed in accordance with USAS B31.1 Code
for Pressure Piping and ACI 318 Building Code Requirements for Reinforced
Concrete which provide minimum requirements on material, design and fabrication
with ample safety margin for both dead and dynamic loads.
Engineered Safety Features Performance Capability
Criterion: Engineered safety features, such as the emergency core cooling system and the containment heat removal system, shall provide sufficient
performance capability to accommodate the failure of any single
active component without resulting in undue risk to the health and
safety of the public. (1967 Proposed GDC 41)
Each engineered safety feature provides sufficient performance capability to accommodate any single failure of an active component and still function in a
manner to avoid undue risk to the health and safety of the public.
6.1-5 Revised 04/17/2013 The extreme upper limits of a once-in-a-lifetime public exposure are taken as the levels and time periods presently outlined in 10 CFR 50.67 at the plant exclusion radius and at the low population zone distance. The accident condition considered is the hypothetical case of a release of fission products per Alternative Source Term (AST) Methodology. Also, the total loss of all outside power is assumed concurrently with this accident.
Under the above accident condition, the Emergency Containment Cooling System and
the Containment Spray System are designed and sized to supply the necessary post-
accident cooling capacity to reduce the containment pressure following blowdown
and cooling of the core by safety injection.
Engineered Safety Features Components Capability
Criterion: Engineered safety features shall be designed so that the capability of these features to perform their required function is not impaired
by the effects of a loss-of-coolant accident to the extent of causing
undue risk to the health and safety of the public. (1967 Proposed
All active components of the Safety Injection System (with the exception of hot leg and low head injection line isolation valves) and the Containment Spray
System are located outside the containment and not subject to containment
accident conditions. The accumulators, emergency coolers, their related active
valves and containment spray headers are located in a missile shielded area
within containment.
Instrumentation, motors, cables and penetrations located inside the containment
are selected to meet the most adverse accident conditions to which they may be
subjected. These items are either protected from containment accident conditions
or are designed to withstand, without failure, exposure to the worst combination
of temperature, pressure, and humidity expected during the required operational
period.
The Safety Injection System and other engineered safeguard system pipes are
restrained such that the rupture of any one pipe will not impair the capability
of the safeguards to prevent undue risk to health and safety of the public.
6.1-6 Revised 04/17/2013 Accident Aggravation Prevention
Criterion: Protection against any action of the engineered safety features which would accentuate significantly the adverse after-effects of a loss of
normal cooling shall be provided. (1967 Proposed GDC 43)
The reactor is maintained subcritical following a reactor coolant system pipe rupture accident. Introduction of borated cooling water into the core results
in a net negative reactivity addition. The control rods insert and remain
inserted.
The delivery of cold safety injection water to the reactor vessel following
accidental expulsion of reactor coolant does not cause further loss of integrity
of the Reactor Coolant System boundary.
Sharing of Systems
Criterion: Reactor facilities may share systems or components if it can be shown that such sharing will not result in undue risk to the health and
safety of the public. (1967 Proposed GDC 4)
Certain components of the Auxiliary, Emergency, and Waste Disposal Systems are shared by the units 3 and 4. Appendix A presents a functional evaluation of the
components of the system which are shared by the two units. In addition, any
emergency and/or shutdown functions of each system are indicated, together with
the ability of the system to meet the emergency condition with either a failure
of an active component or during maintenance outage of a single item of
equipment.
Only certain components of shared equipment may be called upon to fulfill either
an emergency or emergency and shutdown function. As previously stated it is not
considered a credible event that both units can simultaneously develop accident
conditions where each accident is independent of, and not related in any way, to
the other. Thus the criterion in two unit plant design is to have the capability
to deal with the affected unit, while maintaining safe control of the second
unit. For a two unit plant, the worst situation which is credible is when an
accident condition on one unit causes tripping of that unit which in turn leads
to the tripping of the second unit. Further, in the event that the loss of the
output of the two units leads to the loss of all outside AC supply to the
Station, the emergency diesel power supply is required to control the accident
situation on the one unit, and maintain the second unit in a safe condition.
6.1-7 Revised 04/17/2013 Safety injection pump discharge on all four pumps is normally aligned to both
units. Two pumps take suction from each refueling water storage tank.
Within each unit the residual heat removal pumps and heat exchangers serve dual
functions. Although the normal duty of the residual heat removal exchangers and
residual heat removal pumps is performed during periods of reactor shutdown, during all reactor operating periods this equipment is aligned to perform the low
head safety injection function. In addition, during the recirculation phase of a
loss-of-coolant accident, the capability of this system may be divided between
the core cooling and the containment spray cooling functions. Demonstration
testing of the system, performed during each refueling period before startup, in
addition to the appropriate administrative controls, provides assurance of
correct system alignment for the safety function of components.
During the injection phase, the flow path for the safety injection pumps does not
depend on any portion of other systems with the exception of the suction line
from the refueling water storage tank. During the recirculation phase, if either
safety injection or containment spray is required, suction is provided by the
residual heat removal pumps.
6.1.2 RELATED
CRITERIA
The following are criteria which, although related to all engineered safety
features, are more specific to other plant features or systems and, therefore
are discussed in other sections as listed.
Name Discussion
Quality Standards (1967 Proposed GDC 1) Section 1.3 Records Requirements (1967 Proposed GDC 5) Section 4.1 Performance Standards (1967 Proposed GDC 2) Section 4.1
Instrumentation and Control Systems (1967 Proposed GDC 12) Section 7.1
Engineered Safety Features Actuation System (1967 Proposed GDC 15) Section 7.5 Emergency Power (1967 Proposed GDC 24) Section 8.2
6.1-8 Revised 04/17/2013
6.2 SAFETY
INJECTION SYSTEM
6.2.1 DESIGN
BASIS
Emergency Core Cooling System Capability Criterion: An Emergency Core Cooling System with the capability for accomplishing adequate emergency core cooling shall be provided.
This core cooling system and the core shall be designed to prevent
fuel and clad damage that would interfere with the emergency core
cooling function and to limit the clad metal water reaction to
acceptable amounts for all sizes of breaks in the reactor coolant
piping up to the equivalent of a double-ended rupture of the
largest pipe. The performance of such emergency core cooling
system shall be evaluated conservatively in each area of
uncertainty. (GDC 44)
Adequate emergency core cooling is provided by the Safety Injection System (which constitutes the Emergency Core Cooling System) whose components operate
in three modes. These modes are delineated as passive accumulator injection, active safety injection and recirculation.
The primary purpose of the Safety Injection System is to automatically deliver
cooling water to the reactor core in the event of a loss-of-coolant accident.
This limits the fuel clad temperature and thereby ensures that the core will
remain intact and in place, with its heat transfer geometry preserved. This
protection is afforded for:
a) All pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant loop, assuming unobstructed
discharge from both ends.
b) A loss of coolant associated with the rod ejection accident.
c) A steam generator tube rupture.
The basic design criteria for loss of coolant accident evaluations are:
- 1. The cladding temperature is to be less than:
- a. The melting temperature of the fuel cladding. b. The temperature at which gross core geometry distortion, including clad fragmentation may be expected.
- 2. The total core metal-water reaction will be limited to less than 1 percent.
6.2-1 Revised 03/11/2016 Thus, the core geometry is retained to such an extent that effective cooling
of the core is not impaired.
For any rupture of a steam pipe and the associated uncontrolled heat removal
from the core, the Safety Injection System adds shutdown reactivity so that
with a stuck rod, no off-site power and minimum engineered safety features, there is no consequential damage to the Reactor Coolant System and the core
remains in place and intact.
Redundancy and segregation of instrumentation and components is incorporated
to assure that postulated malfunctions will not impair the ability of the
system to meet the design objectives. The system is effective in the event of
loss of normal plant auxiliary power coincident with the loss of coolant, and
can accommodate the failure of any single component or instrument channel to
respond actively in the system.
The ability of the Safety Injection System to meet its capability objectives
is presented in Section 6.2.3. The analysis of the accidents is presented in
Section 14.
Inspection of Emergency Core Cooling System
Criterion: Design provisions shall, where practical, be made to facilitate physical inspection of all parts of the Emergency Core Cooling
System, including reactor vessel internals and water injection
nozzles (GDC 45).
Design provisions are available to facilitate access to the critical parts of
the reactor vessel internals, injection nozzles, pipes, valves and safety
injection pumps for visual or boroscopic inspection for erosion, corrosion and
vibration wear evidence, and for non-destructive inspection where such
techniques are desirable and appropriate.
Testing of Emergency Core Cooling System Components
Criterion: Design provisions shall be made so that components of the Emergency Core Cooling System can be tested periodically for
operability and functional performance. (1967 Proposed GDC 46).
The design provides for periodic testing of active components of the Safety
Injection System for operability and functional performance.
Power sources are arranged to permit individual actuation of each active component of the Safety Injection System.
6.2-2 Revised 04/17/2013 The safety injection pumps can be tested periodically during operation using
the full flow recirculation lines provided. The residual heat removal pumps
are used every time the residual heat removal loop is put into operation. All
remote operated valves can be exercised and actuation circuits can be tested
as required by the Technical Specifications.
Testing of Emergency Core Cooling System
Criterion: Capability shall be provided to test periodically the operability of the Emergency Core Cooling System up to a location as close to
the core as is practical. (1967 Proposed GDC 47).
An integrated safeguards test can be performed during refueling outages. This
test would not introduce flow into the Reactor Coolant System but would
demonstrate the operation of the valves, pump circuit breakers, and automatic
circuitry upon initiation of safety injection. A test is performed during
refueling outages with the safety injection pumps to demonstrate the ability
to introduce flow into the reactor coolant system.
The accumulator tank pressure and level are continuously monitored during
reactor operation.
The safety injection piping up to the final isolation valve is maintained full
of borated water at refueling water concentration while the reactor is in
operation. Any makeup that may be required to maintain the piping full of
water is obtained from a borated water source in accordance with plant
operating procedures.
For the Safety Injection System to be considered operable, it must be filled
with water to ensure that it can reliably perform its intended function. To
address this, Generic Letter 2008-01 was issued to discuss the consequences of
gas entrained in systems such as the Safety Injection System that could
compromise their operability. In response to this, a Gas Accumulation
Management Program (GAMP) was established to provide long term void
management. Locations in the Safety Injection System piping where gases could
potentially accumulate are periodically monitored using ultrasonic testing
and/or vented to verify the system is filled. From the results of this
monitoring, the GAMP ensures that gas accumulated within the Safety Injection
System is identified, evaluated, trended, and effectively controlled to
prevent unacceptable degradation of performance of any structures, systems or
components, ultimately to ensure system operability.
6.2-3 Revised 04/17/2013 Filling and venting operations and periodic system operational and leakage
tests are required to ensure that the Safety Injection System piping and
components are not damaged from water-hammer loads that may result from pump
flows into voided discharge lines. The system must be periodically verified
full by venting the accessible discharge piping high points. The inaccessible
discharge piping high points that may be susceptible to gas accumulation are
deemed appropriate to proactively provide the capability to vent each of these
locations and to allow for future monitoring and trending, if it becomes
necessary.
Flow in each of the hot leg high head injection lines, in the hot and cold leg
injection headers and in the main flow line for the residual heat removal
pumps is monitored by a flow indicator. Pressure instrumentation is also
provided for the main flow paths of the high head and residual heat removal
pumps. Level and pressure instrumentation are provided for each accumulator
tank.
Testing of Operational Sequence of Emergency Core Cooling System
Criterion: TCapability shall be provided to test initially, under conditions as close as practical to design, the full operational sequence
that would bring the Emergency Core Cooling System into action, including the transfer to alternate power sources. (1967 Proposed
GDC 48). The design provides for capability to test initially, to the extent practical, the full operational sequence up to the design conditions for the Safety
Injection System to demonstrate the state of readiness and capability of the
system. Details of the operational sequence testing are presented in Section
6.2.4. Codes
and Classifications
Table 6.2-1 tabulates the codes and standards to which the safety injection
system components are designed.
Service Life
All portions of the system located within the containment are designed to
operate without benefit of maintenance and without loss of functional
performance for the duration of time the component is required during the
accident.
6.2-4 Revised 04/17/2013
6.2.2 SYSTEM
DESIGN AND OPERATION
System Description
Adequate emergency core cooling following a loss-of-coolant accident is
provided by the Safety Injection System shown in Figures 6.2-l and 6.2-5
through 6.2-11. The system components operate in the following possible
modes:
a) Injection of borated water by the passive accumulators.
b) Injection of borated water from the refueling water storage tank by the safety injection pumps.
c) Injection by the residual heat removal pumps also drawing borated water from the refueling water storage tank.
d) Recirculation of spilled coolant injected water and Containment Spray System drainage back to the reactor from the containment recirculation
sump by the safety injection pumps aligned to take suction from the
residual heat removal pumps.
The initiation signal for core cooling by the safety injection pumps and the
residual heat removal pumps is the Safety Injection Signal which is actuated
by any of the following:
a) Low pressurizer pressure (2/3).
b) High containment pressure (2/3, Hi level approximately 10% of containment design pressure).
c) Steam line differential pressure two out of three per loop (between steam line and main steam header).
d) High steam flow in one out of two per steam line (2/3 lines) with low T avg. (2/3) or low steam line pressure (2/3).
e) Manual Actuation.
6.2-5 Revised 04/17/2013 Injection Phase
The principal components of the Safety Injection System which provide
emergency core cooling immediately following a loss of coolant are the
accumulators (one for each loop), the four safety injection (high head) pumps
and the two residual heat removal (low head) pumps.
The accumulators, which are passive components, discharge into the cold legs
of the reactor coolant piping when pressure decreases to approximately 660
psig, thus rapidly assuring core cooling for large breaks. They are located
inside the containment, and are protected against possible missiles.
The safety injection signal, opens the Safety Injection System isolation
valves and starts the safety injection pumps and the residual heat removal
pumps. The items on Figures 6.2-1 through 6.2-11 marked with an "S" receive
the safety injection signal (refer also to Figure 7.2-2). The high head
safety injection pumps take suction from the refueling water storage tank.
The residual heat removal pumps deliver to all three cold legs through the
piping between the accumulators and the cold legs. The high head safety
injection pumps deliver into two separate headers one going to the cold legs
and one to the hot legs. Motor operated valves 864A and 864B at the outlet of
the refueling water storage tank and valves 862A and 862B for residual heat
removal pump suction from the tank are normally open, and are only closed
momentarily for testing or during refueling for maintenance, and are
maintained open by keeping the motor circuit breakers locked open in the motor
control centers, Motor operated valves 866A and 866B, which control the two
flow paths to the hot legs, are maintained inoperative by keeping the motor
circuit breakers locked open at the motor control centers. The header to the
cold legs divides into three injection lines connecting to the pipes from the
accumulators close to the reactor coolant system cold leg piping. The ability
is provided to isolate the pumps on separate headers.
For large breaks, the Reactor Coolant System would be depressurized and voided
of coolant rapidly (about 10 seconds for the largest break) and a high flow
rate is required to quickly recover the exposed fuel rods and limit possible
core damage. To achieve this objective, one residual heat removal pump (high
flow, low head) is required to deliver borated water to the cold legs of the
reactor coolant loops. Two pumps are available in order to provide for an
active component failure. Delivery from these pumps supplements the
accumulator discharge.
6.2-6 Revised 04/17/2013 Because the injection phase of the accident is terminated before the refueling
water storage tank is completely emptied, all pipes are kept filled with water
before recirculation is initiated. Water level indication and alarms on the
refueling water storage tank give the operator ample warning to terminate
injection phase. Additional level alarms are provided in the containment
sumps which also gives back-up indication when injection can be terminated and
recirculation initiated.
For small breaks the depressurization of the Reactor Coolant System by the
Safety Injection System can be augmented by steam dump and auxiliary feed
water addition. As is stated in 14.3.2, use of the steam dump is not required
to meet the core cooling objectives. For small breaks (4" and smaller) steam
dump will be employed to facilitate the recovery from the accident, and to
reduce the reactor coolant pressure to the cut-in pressure of the residual
heat removal pumps.
The decision to initiate steam dump will be based on the rate of decrease of
Reactor Coolant System pressure as indicated by the pressurizer pressure
compared with steam generator pressure. For large breaks (6" and larger) the
reactor pressure drops below the steam side pressure quite rapidly. Before
any gap activity could be released due to clad bursting, the Reactor Coolant
System pressure becomes less than the steam generator pressure. As discussed
in 14.3.2 the expected clad temperatures for break sizes 4" and smaller are
limited to a value below which clad bursting is expected. If a small tube
leak existed prior to the accident, the only activity that could be released
during a steam dump would be the activity initially in the coolant. The
activity released in this manner would be a fraction of that released for a
full tube rupture. The consequences of a steam generator tube rupture are
discussed in 14.2.4.
Recirculation Phase
After the injection operation, coolant spilled from the break and water
collected from the containment spray is cooled and returned to the Reactor
Coolant System by the recirculation system.
Those portions of the Safety Injection System located outside of the
containment which are designed to circulate, under post-accident conditions, radioactively contaminated water collected in the containment, meet the
following requirements:
6.2-7 Revised 04/17/2013 a) Shielding to maintain radiation levels within the guidelines set forth in 10CFR50.67. (Section 11.2)
b) Collection of discharges from pressure relieving devices into closed systems.
c) Means to limit radioactivity leakage to the environs, within guidelines set forth in 10CFR50.67.
When the break is large, depressurization occurs due to the large rate of mass
and energy loss through the break to containment. The system is arranged so
that the residual heat removal pumps take suction from the sump in the
containment floor and delivers flow through the residual heat exchangers to
the suction side of the safety injection pump. The safety injection pumps
inject the spilled reactor coolant and borated refueling water back to the
core. The system is arranged to allow either of the residual heat removal
pumps to provide flow to the safety injection pumps. Only one residual heat
removal pump should be operated during sump recirculation after the MHA.
There are two sump return lines which lead from the containment to the
residual heat removal pumps. The arrangement of recirculation equipment is
shown on Figures 6.2-1 and 6.2-5.
Filtration of the water entering the residual heat removal pump suction piping
is accomplished by a series of stainless steel modular sump strainers serving
both containment sump suction inlets. The strainer modules provide sufficient
strainer surface area considering the mechanistically determined post LOCA
debris load under a MHA such that sufficient available NPSH for the pumps for
the required recirculation flow can still be maintained.
Unit 3 The individual strainer modules are interconnected to each other and to the
north and south containment sump suction inlets by an interconnecting pipe
run. The strainer module plenums are connected in series by piping between the
plenum nozzles. The entire strainer system is then connected in parallel to
the containment sump suction inlets via a tee to the north sump inlet before
terminating at the south sump inlets. Two containment sump suction inlets, each with a 14" diameter outlet, are provided at the 14'-0" elevation. The
openings in the interconnected strainer modules are 3/32", providing
protection against clogging the containment spray nozzle 3/8" openings.
6.2-8 Revised 04/17/2013 Unit 4 Three strainer assemblies are installed on elevation 14'-0". Each strainer assembly consists of five horizontally mounted modules. Each module has thirteen disks. All disks have a 48" width, 30" height and a nominal 1/2" in thickness. Each disk is separated by a screened 1" gap resulting in twelve gaps for each module. Each strainer assembly has a total of 1204.6 ft 2 of strainer surface area. The A, B, and C strainer assemblies have a total
strainer area of 3614 ft
- 2.
Strainer assembly A connects directly through piping to the plenum located
over the south sump pit. Strainer assemblies B and C connect to the plenum
through an 18" diameter pipe.
An extension pipe is installed from the plenum to the ECCS suction line in the south sump pit with a slip connection. A cross connection pipe between the plenum and the north sump pit connects to the existing ECCS suction pipe with a slip connection. The remaining volumes of the north and south sump pits are
filled with concrete.
The maximum allowable debris size that can pass through the strainers is controlled by limiting the perforated plate hole size to a nominal diameter of 0.0938". Particle retention is 100% of particles larger than 0.103",
which prevents clogging the spray nozzle 3/8" openings.
On Unit 4 only, debris interceptors are installed at the entrances to the
biological shield walls on the 14-foot elevation of the containment to limit
the quantity of debris reaching the sump screens and strainer modules during
recirculation.
Recirculation may start with a minimum water depth of 3.03 feet on the containment floor at elevation 14'-0". This is equivalent to a volume of 247,000 gallons. The maximum velocity of approach to the Screens is less than 1/2 ft/sec.
6.2-9 Revised 04/17/2013 pH Control The sodium Tetraborate Decahydrate (NaTB), stored in porous baskets at the lower elevation of containment ensures that the post-LOCA sump pH achieves a minimum value of 7.0 at the onset of spray recirculation and maintains the post-LOCA sump pH between 7 and 8.048 during long-term recirculation. This will minimize the evolution of iodine and prevent stress corrosion cracking of ECCS systems and will ensure that environmental qualification pH limits in Doc Pac 1001, Section 6.4 are not exceeded. Sump water chemistry is discussed in Section 6.4.2.
Recirculation loop leakage is discussed in Section 6.2.3.
Recirculation Phase - Hot Leg
The hot leg recirculation flow path via the high head safety injection pumps
is provided for continuation of the recirculation phase. A residual heat
removal train supplies flow to the SI pumps suctions. The SI pumps discharge
to the hot leg injection header. Hot leg recirculation is required to prevent
boric acid plate-out on the fuel cladding from reducing core cooling following
a cold leg break. See Table 6.2-9(a) for the evaluation of the ability to
achieve hot leg recirculation.
Recirculation Phase - Cold Leg
For the recirculation phase of the accident the reactor coolant water which
eventually drains to the containment recirculation sump is recirculated
through the sump line from the containment to the suction of the residual heat
removal pump through two independent and redundant recirculation lines. Each
line has two motor operated valves. The first valve is located as close as
possible to the containment such that the line outside the containment can be
isolated in the event of a passive failure. During recirculation one
recirculation train will be in service which includes either of the two
residual heat removal pumps and either one or two residual heat exchangers.
The flow will go from the discharge of the residual heat removal pump through
the residual heat exchanger and then into the high head injection path via the
safety injection pumps.
6.2-10 Revised 04/17/2013 Alternate Flow Path to Low Head-Safety Injection An alternate path to the normal low head safety injection path is provided by
MOV 872 using the RHR pumps (See Figures 6.2-1, 6.2-10 and 6.2-11). This
alternate flow path is provided for use in the long term post-LOCA operating
mode after switchover to cold leg recirculation mode. This flow path is
provided in the event a beyond-design-basis passive failure occurs in the
normal low head flowpath. It is not included in the Turkey Point LOCA design
basis analyses or emergency operating procedures. Use of both RHR pumps, at
the same time, during recirculation is not allowed.
Alternate Flow Path for Simultaneous Hot Leg and Cold Leg recirculation In the event that a beyond-design-basis passive failure occurs in the normal hot leg recirculation flowpath, the system is capable of simultaneous cold and hot leg recirculation using the alternate hot leg injection flowpath. In this lineup, hot leg recirculation is established by directing cooled sump water from the RHR heat exchanger, through the shutdown cooling return line, into the 'C' hot leg. Concurrent cold leg recirculation would then be provided by either (1) the normal high-head cold leg safety injection flowpath, or (2) by direct low-head cold leg injection from the RHR pumps. These lines provide defense-in-depth protection for a beyond-design-basis passive failure and are not included in the Turkey Point LOCA design basis analyses or emergency operating procedures.
Cooling Water Component Cooling System
During the recirculation mode, the Component Cooling System is used to cool
the recirculation fluid as it passes through the residual heat exchanger.
One of the three component cooling pumps and two of the three component
cooling heat exchangers provide the core and containment cooling function
during recirculation.
Intake Cooling Water System
The intake cooling water system is provided with normally cross-connected, redundant headers such that the three CCW heat exchangers normally receive
flow from both headers. One of the three intake cooling water pumps is
required to operate during the recirculation phase.
6.2-11 Revised 04/17/2013 Change-Over from Injection Phase to Recirculation Phase
The sequence, from the time of the safety injection signal, for the changeover
from the injection to the recirculation is as follows:
a) First, sufficient water is delivered to provide the required net positive suction head (NPSH) of the residual heat removal pumps to
change to recirculation.
b) Second, the first low level alarm on the refueling water storage tank sounds. The operator, at this point, takes appropriate action to assure
that sufficient NPSH exists for the operating pumps to run or secure
pumps if necessary. This alarm also serves to alert the operator to
begin operation to switchover to the recirculation mode.
c) The second low level alarm on the refueling water storage tank sounds.
At this time, the operator performs the remaining switchover
realignments.
The changeover from injection to recirculation is effected by the operator in
the control room via a series of manual switching operations after unlocking
and closing L.O. breakers to valves 750, 751, 862A and B, 863A and B, 864A and
B and 866A and B outside of the control room.
Remote operated valves for the injection phase of the Safety Injection System
which are under manual control, (these are valves which normally are in their
ready position and do not receive a safety injection signal) have their
positions indicated on a common portion of the control board. At any time
during operation when one of these valves is not in the ready position for
injection, it is shown visually on the board. Reference is made to Table
6.2-2 which is a listing of the instrumentation readouts on the control board
which the operator can monitor during recirculation.
6.2-12 Revised 04/17/2013 Location of the Major Components Required for Recirculation
The residual heat removal pumps are located in separate residual heat removal
rooms at -4'-0" elevation in the auxiliary building. The residual heat
exchangers are located in an adjacent room. The two heat exchangers and each
of the two pumps are separated from each other by a concrete shield wall.
The high head safety injection pumps are located in a common room with partial
shielding between each pair. The component cooling pumps and component
cooling heat exchangers for each unit are located in separate roams in the
Auxiliary Building, the component cooling pumps and heat exchangers being in
the same room.
The intake cooling water pumps are located in the intake structure and
redundant piping to the component cooling heat exchangers is run underground.
Components
All associated components, piping, structures, and power supplies, of the
Safety Injection System are designed to Class I seismic criteria.
All components inside the containment are capable of withstanding or are
protected from differential pressure which may occur during the rapid pressure
rise.
Motors which operate only during or after the postulated accident are designed
as if used in continuous service. Periodic operation of the motors and the
tests of the insulation will ensure that the motors remain in a reliable
operating condition.
All motors, instruments, transmitters, and their associated cables located
inside the containment which are required to operate following the accident
are designed to function under the post-accident temperature, pressure and
humidity conditions.
The quality standards of all safety injection system components is tabulated
in summary form in Table 6.2-3.
6.2-13 Revised 04/17/2013 Accumulators
The accumulators are pressure vessels filled with borated water and
pressurized with nitrogen gas. During normal operation each accumulator is
isolated from the Reactor Coolant System by two check valves in series.
Should the Reactor Coolant System pressure fall below the accumulator
pressure, the check valves open and borated water is forced into the Reactor
Coolant System. Mechanical operation of the swing-disc check valves is the
only action required to open the injection path from the accumulators to the
core via the cold leg.
The accumulators are passive engineered safety features because the gas forces
injection; no external source of power or signal transmission is needed to
obtain a fast-acting, high-flow capability when the need arises. One
accumulator is attached to each of the cold legs of the Reactor Coolant
System.
The design capacity of the accumulators is based on the assumption that flow
from one accumulator spills onto the containment floor through the ruptured
loop, and the flow from the remaining accumulators provides sufficient water
to fill the volume outside of the core barrel below the nozzles, the bottom
plenum, and penetrate the core. Connections for remotely draining or filling
the fluid space, during normal operation, are provided.
The level of borated water in each accumulator tank is adjusted remotely as
required during normal operation. Refueling water is added using a safety
injection pump. Water level is reduced by draining to the reactor coolant
drain tank. Samples of the solution in the tanks are taken at the sampling
station to confirm boron concentration.
Redundant level and pressure indicators are provided with read outs on the
control board. Each indicator is equipped with high and low level alarms.
The accumulator design parameters are given in Table 6.2-4.
Boron Injection Tank
The Boron Injection Tank (BIT) on each unit has been bypassed and abandoned in
place. It is no longer part of the safety injection system flow path on
either unit. The design parameters of the tank are given in Table 6.2-5.
6.2-14 Revised 04/17/2013 Refueling Water Storage Tank (RWST)
In addition to its usual duty to supply borated water to the refueling canal
for refueling operations, this tank provides borated water to the safety
injection pumps, the residual heat removal pumps and the containment spray
pumps for the loss-of-coolant accident. During operation it is aligned to the
suction of the pumps. It is constructed of epoxy coated carbon steel.
The capacity of the RWST is based on the requirement for filling the refueling
canal, given in Table 9.5-1. This capacity provides an amount of borated
water to assure:
a) A volume sufficient to refill the reactor vessel above the nozzles.
b) The volume of borated refueling water needed to increase the concentration of initially spilled water to a point that assures no
return to criticality with the reactor at cold shutdown.
c) A sufficient volume of water on the floor to permit the initiation of recirculation.
The water in the RWST is borated to a minimum concentration of 2400 ppm boron (approximately 1.4 weight percent boric acid). This ensures the reactor
remains sub-critical during the most limiting Post-LOCA case analyzed for EPU conditions. At 32
°F the solubility limit of boric acid is 2.7 weight percent.
Therefore the concentration of boric acid in the RWST is well below the solubility limit at 32
°F. A technical specification minimum level alarm and a high level alarm are
provided. Nominal RWST level indication with high and low level alarms are
also provided. The low level alarm setpoint has been determined to provide
sufficient NPSH for the containment spray pumps
A dynamic response analysis similar to that performed for the Containment
Structure has been performed to determine the horizontal loads applied to this
tank for a 5% ground acceleration based on yield stresses and a 15% ground
acceleration based on maximum deflection. Waves generated in the tank have
been taken into account as per "Nuclear Reactors and Earthquake", TID 7024. A
membrane stress analysis of the vertical cylindrical tank was performed
considering the discontinuities at the base and top.
The design parameters are given in Table 6.2-6.
6.2-15 Revised 04/17/2013 Safety Injection Pumps
The four high-head safety injection pumps for supplying borated water to the
Reactor Coolant System are horizontal centrifugal pumps driven by electrical
motors. Parts of the pump in contact with borated water are stainless steel
or equivalent corrosion resistant material. A minimum flow bypass line is
provided on each pump discharge to recirculate flow to the RWST in the event
the pumps are started with the normal flow paths blocked. The design
parameters are presented in Table 6.2-7 and Figure 6.2-3 gives the performance
characteristic of these pumps.
The two residual heat removal (low head) pumps of the Auxiliary Coolant System
are used to inject borated water at low pressure to the Reactor Coolant
System. They are also used to recirculate fluid from the containment sump and
send it to the suction of the spray pumps or to suction of the high head
safety injection pumps. These pumps are of the in-line centrifugal type, driven by electric motors. Parts of the pumps which contact the borated water
are stainless steel or equivalent corrosion resistant material. A minimum
flow bypass line is provided on the discharge of each RHR pump to recirculate
fluid to the suction of the residual heat removal pumps should these pumps be
started with their normal flow paths blocked. The duration the pumps can
operate on these mini recirculation lines has been determined to be about 44
minutes when aligned to take suction from the RWST. The duration the pumps
can operate on mini recirculation is reduced to 10 minutes when the pumps are
aligned to take suction form the sump due to the elevated temperature of the
sump fluid The original recirculation line on the RHR heat exchanger
discharge is currently closed via administrative controls. The design
parameters are presented in Table 6.2-7.
The pressure containing parts of the pumps are castings conforming to ASTM
A-351 Grade F8 or F8M. Stainless steel forgings are procured per ASTM A-182
Grade F304 or F316 or ASTM A-336, Grade F8 or F8M, and stainless plate is
constructed to ASTM A-240, Type 304 or 316. All bolting material conforms to
ASTM A-193. Materials such as weld-deposited Stellite or Colmonoy are used at
points of lose running clearances in the pumps to prevent galling and to
assure continued performance ability in high velocity areas subject to
erosion.
All pressure containing parts of the pumps are chemically and physically
analyzed and the results are checked to ensure conformance with the applicable
ASTM specification. In addition, all pressure containing parts of the pump
are liquid penetrant inspected in accordance with Appendix VIII of Section
VIII of the ASME Boiler and Pressure Vessel Code.
6.2-16 Revised 04/17/2013 The acceptance standard for the liquid penetrant test is USAS B31.1, Code for
Pressure Piping, Case N-10. The pump design is reviewed with special
attention to the reliability and maintenance aspects of the working
components.
Specific areas include evaluation of the shaft seal and bearing design to
determine that adequate allowances have been made for shaft deflection and
clearances between stationary parts.
Where welding of pressure-containing parts is necessary, a welding procedure
including joint detail is submitted for review and approval by Westinghouse.
The procedure includes evidence of qualification necessary for compliance with
Section IX of the ASME Boiler and Pressure Vessel Code Welding Qualifications.
This requirement also applies to any repair welding performed on pressure-
containing parts.
The pressure-containing parts of the pump are assembled and hydrostatically
tested to 1.5 times the design pressure for 30 minutes.
Each pump is given a complete shop performance test in accordance with
Hydraulic Institute Standards. The pumps are run at design flow and head, shut off head and three additional points to verify performance
characteristics. Where NPSH is critical, this value is established at design
flow by means of adjusting suction pressure.
Details of the component cooling and intake cooling water pumps which serve
the Safety Injection System are presented in Section 9.
Heat Exchangers
The two residual heat exchangers of the Auxiliary Coolant System cool the
recirculated sump water. These heat exchangers are sized for the cooldown of
the Reactor Coolant System. Table 6.2-8 gives the design parameters of the
heat exchangers.
The ASME Boiler and Pressure Vessel Code has strict rules regarding the wall
thicknesses of all pressure containing parts, material quality assurance
provisions, weld joint design, radiographic and liquid penetrant examination
of materials and joints, and hydrostatic testing of the unit as well as
requiring final inspection and stamping of the vessel by an ASME Code
inspector.
6.2-17 Revised 04/17/2013 The designs of the heat exchangers also conform to the requirements of TEMA (Tubular Exchanger Manufacturers Association) for Class R heat exchangers.
Class R is the most rugged class of TEMA heat exchangers and is intended for
units where safety and durability are required under severe service
conditions. Items such as: tube spacing, flange design, nozzle location, baffle thickness and spacing, and impingement plate requirements are set forth
by TEMA Standards.
In addition to the above, additional design and inspection requirements were
imposed to ensure rugged, high quality heat exchangers such as: confined-type
gaskets, main flange studs with two nuts on each end to ensure permanent leak
tightness, general construction and mounting brackets suitable for the plant
seismic design requirements, tubes and tube sheet capable of withstanding full
shell side pressure and temperature with atmospheric pressure on the tube
side, ultrasonic inspection in accordance with Paragraph N-324.3 of Section
III of the ASME Code of all tubes before bending, penetrant inspection in
accordance with Paragraph N-627 of Section III of the ASME Code of all welds
and all hot or cold formed parts, a hydrostatic test duration of not less than
thirty minutes, the witnessing of hydro and penetrant tests by a qualified
inspector, a thorough final inspection of the unit for good workmanship and
the absence of any gouge marks or other scars that could act as stress
concentration points, a review of the radiographs and of the certified
chemical and physical test reports for all materials used in the unit.
The Residual Heat Exchangers are conventional vertical shell and U-tube type
units. The tubes are seal welded to the tube sheet. The shell connections
are flanged to facilitate shell removal for inspection and cleaning of the
tube handle. Each unit has a SA-285 Grade C carbon steel shell, a SA-234
carbon steel shell end cap, SA-213 TP-304 stainless steel tubes, SA-240 Type
304 stainless steel channel, SA-240 Type 304 stainless steel channel cover and
SA-240 Type 304 stainless steel tube sheet.
Valves All parts of valves used in the Safety Injection System in contact with borated water are austenitic stainless steel or equivalent corrosion resistant
material. The motor operators on the injection line isolation valves are
capable of rapid operation. All valves required for initiation of safety
injection or isolation of the system have remote position indication in the
control room.
6.2-18 Revised 04/17/2013 Valving is specified for exceptional tightness. The majority of valves, except those which perform a control function, are provided with backseats
which are capable of limiting leakage to less than 1.0 cc per hour per inch of
stem diameter, assuming no credit taken for valve packing. Manual and motor
operated gate and globe valves that are normally operated and in containment
are not backseated. Other normally opened valves are typically backseated.
Normally closed globe valves are installed with recirculation flow under the
seat to prevent leakage of recirculated water through the valve stem packing.
Relief valves are totally enclosed. Control and motor-operated valves, 2 1/2" and above, are typically provided with double-packed stuffing boxes and stem
leakoff connections which are piped to the Waste Disposal System except where
the need for bonnet relief has been deemed necessary inside the containment
for pressure locking considerations, i.e., MOV-*-744A, MOV-*-744B and MOV-*-
751.
The check valves which isolate the Safety Injection System from the Reactor
Coolant System are installed adjacent to the reactor coolant piping to reduce
the probability of an injection line rupture causing a loss-of-coolant
accident.
Three relief valves are provided to protect lines required for SI/ECCS
functions against overpressurization due to RCS back-leakage. One is located
outside the containment upstream of MOV-*-843A&B to prevent overpressure in
the header and relieves to the RWST. The high head safety injection piping
leading to the hot legs is protected by the second, that is the relief valve
which is inside the containment in the piping associated with containment
Penetration No. 18. This valve relieves to the pressurizer relief tank. The
third valve is located on the residual heat removal loop common header leading
to the accumulator pipes. This valve is inside containment and relieves to
the pressurizer relief tank.
The relieving capacity of these valves is based on a flow several times
greater than the expected leakage rate through the check and isolation valves.
They will also prevent overpressurization due to thermal expansion. The
relief valves on the accumulator protect them from pressures in excess of the
design values.
6.2-19 Revised 04/17/2013 Motor Operated Valves
The pressure containing parts (body, bonnet and discs) of the valves employed
in the Safety Injection System are designed per criteria established by the
USAS B16.5 or MSS SP66 specifications. The materials of construction for
these parts are procured per ASTM A182, F316 or A351, GR F8M, or F8. All
material in contact with the primary fluid, except the packing, is austenitic
stainless steel or equivalent corrosion resisting material. The pressure
containing cast components are radiographically inspected as outlined in ASTM
E-71 Class 1 or Class 2. The body, bonnet and discs are liquid penetrant
inspected in accordance with ASME Boiler and Pressure Vessel Code Section
VIII, Appendix VIII. The liquid penetrant acceptable standard is as outlined
in USAS B31.1 Case N-10. When a gasket is employed the body-to-bonnet joint
is designed per ASME Boiler and Pressure Vessel Code Section VIII or USAS
B16.5 with a fully trapped, controlled compression, spiral wound gasket with
provisions for seal welding, or of the pressure seal design with provisions
for seal welding. The body-to-bonnet bolting and nut materials are procured
per ASTM A193 and A194, respectively.
The entire assembled unit is hydrotested as outlined in MSS SP-61 with the
exception that the test is maintained for a minimum period of 30 minutes per
inch of wall thickness. Any leakage is cause for rejection. The seating
design is of the Darling parallel disc design, the Crane flexible wedge
design, or the equivalent. These designs have the feature of releasing the
mechanical holding force during the first increment of travel. Thus, the
motor operator has to work only against the frictional component of the
hydraulic unbalance on the disc and the packing box friction. The discs are
guided throughout the full disc travel to prevent chattering and provide ease
of gate movement. The seating surfaces are hard faced (Stellite No. 6 or
equivalent) to prevent galling and reduce wear.
The stem material is ASTM A276 Type 316 condition B or precipitation hardened 17-4 PH stainless procured and heat treated to Westinghouse Specifications.
These materials are selected because of their corrosion resistance, high
tensile properties, and their resistance to surface scoring by the packing.
The valve stuffing box is designed with a lantern ring leak-off connection
with a minimum of a full set of packing below the lantern ring and a maximum
of one-half of a set of packing above the lantern ring; a full set of packing
is defined as a depth of packing equal to 1-1/2 times the stem diameter. The
experience with this stuffing box design and the selection of packing and stem
materials has been very favorable in both conventional and nuclear power
plants.
6.2-20 Revised 04/17/2013 The RHR suction isolation valve MOV-*-751 and the RHR low head safety injection isolation valves (MOV-*-744A, MOV-*-744B) are provided with a single
set of packing and no stem leakoff.
The motor operator is extremely rugged and is noted throughout the power
industry for its reliability. The unit incorporates a "hammer blow" feature
that allows the motor to impact the discs away from the fore or backseat upon
opening or closing. This "hammer blow" feature not only impacts the disc but
allows the motor to attain its operational speed.
The valve is assembled, hydrostatically tested, seat-leakage tested (fore and
back), operationally tested, cleaned and packaged per specifications. All
manufacturing procedures employed by the valve supplier such as hard facing, welding, repair welding and testing are submitted to Westinghouse for
approval.
For those valves which must function on the safety injection signal, stroke-
time certified operators are provided. For all other valves in the system, the valve operator completes its cycle from one position to the other within
180 seconds. Valve stroke times may be increased as a result of evaluations
performed in accordance with NRC Generic Letter 89-10.
Valves are required to function under the maximum differential pressures
determined in accordance with the requirements of NRC Generic Letter 89-10.
Safety related power operated gate valves were evaluated for their
susceptibility to pressure locking and thermal binding as required by NRC
Generic Letters 89-10 and 95-07. The following motor operated valves each
have a design feature (either a hole drilled in one of the discs or a bonnet
vent to provide relief from the inter-disc space) which preclude the potential
for pressure locking as described in the two generic letters:
MOV-*-350, MOV-*-750/751, MOV-*-843A/B, MOV-*-860A/B, MOV-*-861A/B, MOV-*-866A/B, MOV-*-869, MOV-*-872, and MOV-*-744A/B.
Manual Valves
The stainless steel manual globe, gate and check valves are designed and built
in accordance with the requirements outlined in the motor operated valve
description above.
6.2-21 Revised 04/17/2013 The carbon steel valves are built to conform with USAS B16.5. The materials
of construction of the body, bonnet and disc conform to the requirements of
ASTM A105 Grade II, A181 Grade II or A216 Grade WCB or WCC. The carbon steel
valves pass only non-radioactive fluids and are subjected to hydrostatic test
as outlined in MSS SP-61, except that the test pressure is maintained for at
least 30 minutes per inch of wall thickness. Since the fluid controlled by
the carbon steel valves is not radioactive, the double packing and seal weld
provisions are not provided.
The pressure containing parts of this valve assembly are designed in
accordance with MSS SP-66. All parts in contact with the operating fluid are
of austenitic stainless steel or of equivalent corrosion resistant materials
procured to applicable ASTM or WAPD specifications. The cast pressure-
containing parts are radiographed in accordance with ASTM E-94 and the
acceptance standard as outlined in ASTM E-71. The cast pressure-containing
parts, machined surfaces, finished hard facings, and gasket bearing surfaces
are liquid penetrant inspected per ASME B&PV Code,Section VIII and the
acceptance standard is as outlined in USAS B31.1 Code Case N-10. The final
valve is hydrotested per MSS SP-66, except that the test pressure is
maintained for at least 30 minutes. The seat leakage is conducted in
accordance with the manner prescribed in MSS SP-61, except that the acceptable
leakage is 2cc/hr/in, nominal pipe diameter.
The valve is designed with a low pressure drop configuration with all
operating parts contained within the body, which eliminates those problems
associated with packing glands exposed to boric acid. The clapper arm shaft
is manufactured from 17-4 PH stainless steel heat treated to Westinghouse
Specifications, The clapper arm shaft bushings are manufactured from Stellite
No. 6 material, The various working parts are selected for their corrosion
resistant, tensile, and bearing properties.
The disc and seat rings are manufactured from a forging. The mating surfaces
are hard faced with Stellite No. 6 to improve the valve seating life. The
disc is permitted to rotate, providing a new seating surface after each valve
opening.
The valves are intended to be operated in the closed position with a normal
differential pressure across the disc of approximately 1550 psi. The valves
shall remain in this position except for testing and safety injection and are
not subjected to impact loads caused by sudden flow reversal.
6.2-22 Revised 04/17/2013 When the valve is required to function a differential pressure of less than 25
psig will shear any particles that may attempt to prevent the valve from
functioning. Although the working parts are exposed to the boric acid
solution contained within the reactor coolant loop, a boric acid "freeze up" is not expected with this low a concentration.
The experience derived from the check valves employed in the Emergency
Injection System of the Carolina - Virginia Tube Reactor in a similar system
indicates that the system is reliable and workable.
The CVTR Emergency Injection System, normally maintained at containment
ambient conditions was separated from the main coolant piping by a single six
inch check valve. Leak detection was provided at a proper elevation to
accumulate any leakage coming back through the check valve. A level alarm
provided a signal on excessive leakage. The pressure differential was 1500
psi and the system was stagnant. The valve was located 2 to 3 feet from
the main coolant piping which resulted in some heatup and cooldown cycling.
The CVTR went critical late in 1963 and operated until 1967. During that
time, the level sensor in the detection never alarmed due to check valve
leakage.
Relief Valves The accumulator relief valves are sized to pass nitrogen gas at a rate in excess of the accumulator gas fill line delivery rate. The relief valves will
also pass water in excess of the expected leak rate, but this is not necessary
because the time required to fill the gas space gives the operator ample
opportunity to correct the situation. For an inleakage rate 15 times the
manufacturing test rate, there will be about 1000 days before water will
reach the relief valves. Prior to this, level and pressure alarms would have
been actuated.
The safety injection test line relief valve at Penetration No. P-17 is provided to relieve any pressure, above design, that might build up in the
piping. The valve will pass a nominal 50 gpm which is far in excess of the
manufacturing design leak rate of 24 cc/hr.
The relief valve in the high head safety injection line leading to the hot leg at penetration P-18 will pass a nominal flow of 20 gpm, which is approximately
twice the maximum technical specification permitted leak rate through valves
- -874A/B and MOV-*-866A A/B. Valves *-874 A/B and MOV-*-866 A/B isolate this
portion of the SI system from the reactor coolant system.
6.2-23 Revised 04/17/2013 Leakage Limitations Valving is specified for exceptional tightness. The majority of valves, except those which perform a control function, are provided with backseats
which are capable of limiting leakage to less than one cubic centimeter per
hour per inch of stem diameter assuming no credit for packing in the valve.
Normally closed globe valves are installed with recirculation flow under the
seat to prevent stem leakage from the more radioactive fluid side of the seat.
Manual and motor operated gate and globe valves that are normally open and
inside containment are not backseated. Other normally open valves are
backseated.
Motor operated valves, which are exposed to recirculation flow, are provided
with double-packed stuffing boxes and stem leakoff connections which are piped
to the pressurizer relief tank (PRT) enroute to the waste disposal system, except the RHR suction isolation valve MOV-*-751, the HHSI hot leg injection
valves MOV-*-866A/B and the RHR low head safety injection isolation valves (MOV-*-744A, MOV-*-744B) which have one set of packing and no stem leak-off
connections.
The specified leakage across the valve disc required to meet the equipment
specification and hydrotest requirements is as follows:
a) Conventional globe - 3 cc/hr/in. of nominal pipe size.
b) Gate valves - 3 cc/hr/in. of nominal pipe size; 10 cc/hr/in for 300 and 150 pound USA Standard. c) Motor-operated gate valves - 3 cc/hr/in. of nominal pipe size; 10 cc/hr/in for 300 and 150 pound USA Standard. d) Check valves - 3 cc/hr/in. of nominal pipe size; 10 cc/hr/in for 300 and 150 pound USA Standard e) Accumulator check valves -2 cc/hr/in. of nominal pipe size
Relief valves are totally enclosed. Leakage from components of the
recirculation loop including valves is tabulated in Table 6.2-12.
Piping All Safety Injection System piping in contact with borated water is austenitic
stainless steel. Piping joints are welded except for the flanged connections
at the safety injection, containment spray pumps and accumulator fill line
The piping beyond the accumulator stop valves is designed for Reactor Coolant System conditions (2485 psig, 650
°F). All other piping connected to the accumulator tanks is designed for 700 psig and 400
°F. 6.2-24 Revised 04/17/2013 The safety injection pump suction piping (210 psig at 300
°F) from the refueling water storage is designed for low pressure losses to meet NPSH (net positive suction head) requirements of the pumps.
The safety injection high pressure branch lines (1500 psig at 300
°F) to the hot legs are designed for high pressure losses to limit the flow rate out of a potential rupture of a branch line at the connection to the reactor coolant
loop.
The piping is designed to meet the minimum requirements set forth in (1) the
USAS B31.1 Code for the Pressure Piping, (2) Nuclear Code Case N-7, (3) USAS
Standards B36.10 and B36.19 and (4) ASTM Standards, and (5) supplementary
standards plus additional quality control measures.
Minimum wall thicknesses are determined by the USAS Code formula in the power
piping Section 1 of the USAS Code for Pressure Piping. This minimum thickness
is increased to account for the manufacturer's permissible tolerance of minus
12-1/2 per cent on the nominal wall. Purchased pipe and fittings have a
specified nominal wall thickness that is no less than the sum of that required
for pressure containment, mechanical strength, and manufacturing tolerance.
Thermal and seismic piping flexibility analyses are performed. Special
attention is directed to the piping configuration at the pumps with the
objective of minimizing pipe imposed loads at the suction and discharge
nozzles. Piping is supported to accommodate expansion due to temperature
changes during the accident.
Pipe and fitting materials are procured in conformance with all requirements
of the ASTM and USAS specifications. All materials are verified for confor-
mance to specification and documented by certification of compliance to ASTM
material requirements. Specifications impose additional quality control upon
the suppliers of pipes and fittings as listed below:
a) Quality assurance records will be maintained on both the purchased pipe and fittings.
b) Pipe branch lines between the reactor coolant pipes and the isolation stop valves conform to ASTM A376 and meet the supplementary requirement
S6 ultrasonic testing.
c) Fittings conform to the requirements of ASTM A403. Fittings 3 inch and above have requirements for UT inspection similar to S6 of A376.
6.2-25 Revised 04/17/2013 Shop fabrication of piping subassemblies is performed by reputable suppliers
in accordance with specifications which define and govern material
procurement, detailed design, shop fabrication, cleaning, inspection, identification, packaging and shipment.
Welds for pipes sized 2-1/2" and larger are butt welded. Reducing tees are
used where the branch size exceeds 1/2 of the header size. Branch connections
of sizes that are equal to or less than 1/2 of the header size are of a design
that conforms to the USAS rules for reinforcement set forth in the USAS B31.1
Code for Pressure Piping. Bosses for branch connections are attached to the
header by means of full penetration welds.
All welding is performed by welders and welding procedures qualified in
accordance with the ASME Boiler and Pressure Vessel Code Section IX, Welding
Qualifications. The Shop Fabricator is required to submit all welding
procedures and evidence of qualifications for review and approval prior to
release for fabrication. All welding materials used by the Shop Fabricator
must have prior approval.
All high pressure piping butt welds containing radioactive fluid, at greater than 600°F temperature and 600 psig pressure or equivalent, are radiographed.
The remaining piping butt welds are randomly radiographed. The technique and acceptance standards are those outlined in UW-51 of the ASME B&PV Code Section
VIII. In addition, butt welds are liquid penetrant examined in accordance
with the procedure of ASME B&PV Code,Section VIII, Appendix VIII and the
acceptance standard as defined in the USAS Nuclear Code Case N-10. Finished
branch welds are liquid penetrant examined on the outside and where size
permits, on the inside root surfaces.
A post-bending solution anneal heat treatment is performed on hot-formed
stainless steel pipe bends. Completed bends are then completely cleaned of
oxidation from all affected surfaces. The shop fabricator is required to
submit the bending, heat treatment and clean-up procedures for review and
approval prior to release for fabrication.
General cleaning of completed piping subassemblies (inside and outside
surfaces) is governed by basic ground rules set forth in the specifications.
For example, these specifications prohibit the use of hydrochloric acid and
limit the chloride content of service water and demineralized water.
6.2-26 Revised 04/17/2013 Packaging of the piping subassemblies for shipment is done so as to preclude
damage during transit and storage. Openings are closed and sealed with
tight-fitting covers to prevent entry of moisture and foreign material.
Flange facings and weld end preparations are protected from damage by means of
c-over plates securely fastened in position. The packing arrangement proposed
by the shop fabricator is subject to approval.
Pump and Valve Motors
Motors Outside the Containment
Motor electrical insulation systems are supplied in accordance with USAS, IEEE
and NEMA standards and are tested as required by such standards. Temperature
rise design selection is such that normal long life is achieved even under
accident loading conditions.
Although the motors which are provided only to drive engineered safety
features equipment are normally run only for test, the design loading and
temperature rise limits are based on accident conditions. Normal design
margins are specified for these motors to make sure the expected lifetime
includes allowance for the occurrence of accident conditions.
Motors Inside the Containment
The motor operators for the valves inside containment are designed to with-
stand containment environment conditions following the loss of coolant
accident so that the valves can perform the required function during the
recovery period.
Periodic operation of the motors and tests of the insulation ensure that the
motors remain in a reliable condition.
Tests to demonstrate the adequacy of valve motor operators of the type used in
the design to be functional after exposure to temperature, pressure and
radiation are being conducted in two groups.
6.2-27 Revised 04/17/2013 The first group is the exposure of valve motor operators to both temperature
and pressures similar to that predicted for the incident:
a) Operator is located inside a pressure vessel which is exposed to approximately 330°F at 90 psig.
b) Operator will be cycled approximately three times under simulated valve operating loads.
c) Pressures and temperatures will be reduced in step changes to 285°F at 60 psig, 219°F at 20 psig and 152°F at atmosphere or less.
d) Operator will be cycled approximately three times at each of the levels of change. Full recordings of pertinent data will be taken throughout
the tests.
e) Unit shall be examined after completion of test and operating data compared to data prior to exposure.
The second group test is the radiation test on a motor from the valve
operator.
a) Two production line motors are being used for this test. One is to be exposed to 1.5 X 10 8 rads of gamma for approximate period of one month.
The other motor will be used for the final comparative analysis.
b) Both units will be tested for coil resistance, insulation meggering both before and after motor vibration and reversing operations.
Electrical Supply
Details of the normal and emergency power sources for the Safety Injection
System are presented in Section 8.
6.2-28 Revised 04/17/2013 Protection Against Dynamic Effects
All three cold leg injection lines and the hot leg injection header penetrate
the containment via the auxiliary building, pipe and valve room.
The portion of the high head injection system within the containment is
connected to the accumulator injection lines attached to each loop cold leg
piping and to the hot leg piping of two loops. The portion of the low head
injection system within the containment is connected to the accumulator
injection lines attached to each loop cold leg piping.
For most of their routing, these lines are outside the primary and secondary
shielding, and hence they are protected from missiles originating within these
areas.
All hangers, stops and anchors are designed in accordance with USAS B31.1 Code
for Pressure Piping and ACI 318-63 Building Code Requirements for Reinforced
Concrete which provide minimum requirements on materials, design and
fabrication with ample safety margins for both dead and operational dynamic
loads over the life of the equipment. In addition to the normal load
conditions the requirements of Table 5A-1 for the loading combinations shown
are used in design of supports. Specifically, these standards require the
following:
(1) All materials used are in accordance with ASTM specifications which establish quality levels for the manufacturing process, minimum strength
properties, and for test requirements which ensure compliance with the
specifications.
(2) Qualification of welding processes and welders for each class of material welded and for types and positions of welds.
(3) Maximum allowable stress values are established which provide an ample safety margin on yield strength for normal loads and ultimate strength
for design basis accident or maximum hypothetical seismic loads.
6.2-29 Revised 04/17/2013
6.2.3 DESIGN
EVALUATION
Range of Core Protection
The measure of effectiveness of the Safety Injection System is the ability of
the pumps and accumulators to keep the core flooded or to reflood the core
rapidly where the core has been uncovered for postulated large area ruptures.
The result of the performance is to sufficiently limit any increase in clad
temperature below a value where emergency core cooling objectives are met (Section 6.2.1). The range of core protection as a function of break diameter
provided by the various components of the Safety Injection System is presented
in Figure 6.2-4.
Figure 6.2-4 was developed from the results of the loss of coolant accident
studies presented in Section 14.3. Simulations of a sufficient number of
break sizes were performed to demonstrate that the Safety Injection components
meet the emergency core cooling requirements. The injection from the follow-
ing combination of components was analyzed as discussed below.
Bar A Two Safety Injection Pumps Bar B One Residual Heat Removal Pump and Two Accumulators Bar C Two Safety Injection Pumps and Two Accumulators Bar D Two Safety Injection Pumps, One Residual Heat Removal Pump and Two Accumulators
- With one safety injection pump, associated with the shutdown non-accident
unit, inoperable and the failure of an emergency diesel generator, associated
with the accident unit, the emergency core cooling equipment available is
represented by Bar D (two out of four safety injection pumps, one out of two
residual heat removal pumps, and two out of three accumulators). With these systems, the calculated maximum fuel cladding temperature is limited to 2200
°F, which meets the emergency core cooling design objectives for all break sizes up to and including the double-ended severance of the reactor coolant pipe.
Section 14.3.2 discusses the analyzed design basis as represented by Bar D.
- No credit is taken for the accumulator which is attached to the ruptured loop, nor for spillage from broken lines.
6.2-30 Revised 04/17/2013 The limiting single active short-term failure assumed in the Best Estimate Large Break LOCA (BELOCA) analysis in Section 14.3.2 is the loss of a
safeguards train, and the pumped emergency core cooling flow rates are
conservatively calculated based on minimum (bounding) safeguards assumptions
designed to minimize pumped Emergency Core Cooling System flow to the core.
Therefore, the Emergency Core Cooling System flow assumed for a large break
LOCA only includes one residual heat removal pump and two safety injection
pumps in operation, which matches the Safety Injection component design basis
represented by Bar D.
The remaining three combinations (Bars A, B, C,) represent degraded cases with
operation of less than the minimum design emergency core cooling equipment.
These cases are shown only to present the capability of individual portions of
the systems and to demonstrate the overall margins to loss of clad geometric
integrity and to clad burst of the system. The operation of two safety
injection pumps, (Bar A) provides core protection for break sizes up to an
equivalent break diameter of approximately 4 inches. The operation of two
safety injection pumps would allow flow spilling from a broken safety
injection line to go uncorrected by operator action. Isolation of the broken
line by operator action would increase the range of protection.
In Section 14.3.2 an analysis is presented that demonstrated the ability of
the two high head pumps to keep the core hot spot covered for all breaks up to
a 4" equivalent diameter hole. When the hot spot remains covered, no core
damage is expected.
The cases represented by Bars B and C are presented to demonstrate the
redundancy between the low and high head pumping systems to provide protection
for large area rupture. For large area ruptures analyzed (14.3.2) the clad
temperatures are turned around by the accumulator injection. The active
pumping components serve only to complete the refill started by the
accumulators. Either two safety injection pumps or one residual heat removal
pump provides sufficient addition of water to continue the reduction of clad
temperature initially caused by the accumulator.
Bars A, B, and C are presented solely to demonstrate the redundant sufficiency
of the system and do not represent the design basis of the system which is
depicted by Bar D. The four inch break represents the upper limit of
protection for the high head safety injection system. Two safety injection
pumps limit the peak clad temperatures to levels well below the temperature
where clad bursting is expected.
6.2-31 Revised 04/17/2013 For all breaks four inches and smaller the release of fission products will be
limited to the activity initially in the coolant. For larger ruptures, up to
and including the double ended rupture of the reactor coolant pipe, the
accumulators together with various combinations of high and low head pumps
limit the peak clad temperatures to levels well below the temperature where
any loss of clad geometric integrity is expected (Bars B and C).
System Response To provide protection for large area ruptures in the Reactor Coolant System
the Safety Injection System must respond to rapidly reflood the core following
the depressurization and core voiding that is characteristic of large area
ruptures. The accumulators act to begin the rapid reflooding function with no
dependence on the normal or emergency power sources, and also with no
dependence on the receipt of an actuation signal.
Operation of this system with two of the three available accumulators deliver-
ing their contents to the reactor vessel (one accumulator spilling through the
break) prevents fuel clad melting and limits metal-water reaction to an
insignificant amount (<1%).
The function of the safety injection or residual heat removal pumps is to
complete the refill of the vessel and ultimately return the core to a sub-
cooled state. As discussed earlier the flow from either two safety injection
pumps or one residual heat removal pump is sufficient to complete the refill
with no loss of level in the core. Moreover, there is sufficient excess water
delivered by the accumulators to tolerate a delay in starting the pumps.
Initial response of the injection systems is automatic, with appropriate
allowances for delays in actuation of circuitry and active components. The
active portions of the injection systems are automatically actuated by the
safety injection signal (Section 7). In addition, manual actuation of the
entire injection system and individual components can be accomplished from the
control room. In analysis of system performance, delays in reaching the
programmed trip points and in actuation of components are conservatively
established on the basis that only emergency on-site power is available.
6.2-32 Revised 04/17/2013 The starting sequence of the safety injection and residual heat removal pumps
and the related emergency power equipment is designed so that delivery of the
full rated flow is reached within 25 seconds after the process parameters
reach the setpoints for the injection signal. See Section 8.2.
The delay of 35 seconds is assumed in the analysis of the loss-of-coolant
accident as described in Section 14.
For the small break analysis, an additional delay time is allowed to account
for the receipt of safety injection signal, either from low pressurizer
pressure or from high containment pressure.
Single Failure Analysis
A single active failure analysis is presented in Table 6.2-9(a). All credible
active system failures are considered. The analysis of the loss-of-coolant
accident presented in Section 14 is consistent with the single failure
analysis.
It is based on the worst single failure (generally a safeguards train)
affecting both the safety injection and residual heat removal pumping systems.
The analysis shows that the failure of any single active component will not
prevent fulfilling the design function.
In addition, an evaluation of an alternative flow path to maintain core
cooling is given in Table 6.2-9(b) in the event a beyond-design-basis single
passive failure disrupts hot leg or cold leg recirculation.
Failure analyses of the component cooling and intake cooling water system
under loss-of-coolant accident conditions are described in Section 9.3 and
9.6, respectively.
6.2-33 Revised 04/17/2013 Reliance-on Interconnected Systems
During the injection phase, the flow path for the high head safety injection
pumps does not depend on any portion of other systems, with the exception of
the suction line from the RWST. During the recirculation phase of the
accident for small breaks, suction to the safety injection pumps is provided
by a residual heat removal pump.
The residual heat removal (low head) pumps are normally used during reactor
shutdown operations. Whenever the reactor is at power, the pumps are aligned
for emergency duty.
Operation of a single residual heat removal loop is permitted for decay heat
removal when fuel is in the reactor vessel and the refueling cavity is flooded
to greater than or equal to 23 feet above the reactor vessel flange. A loss of
this single residual heat removal loop has been evaluated to ensure that
adequate natural circulation cooling can be maintained for decay heat removal.
The analysis utilized GOTHIC thermal-hydraulic analysis software to evaluate
natural circulation cooling conditions with both the reactor vessel upper
internals assembly installed and with the upper internals assembly removed. In
each case, the results of the analysis demonstrate that no fuel design limits
will be exceeded for at least 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and this is more than enough time to
initiate emergency procedures to maintain core cooling. Furthermore, the
calculated increase in heat flux in the core volume is only a small fraction
of the CHF for the fuel, and is considered a minimal (essentially negligible)
increase. This negligible impact on fuel CHF ensures that no fuel design
limits are approached or challenged during the fuel cavity boil-off period.
Also, the possibility of inadvertent criticality is minimized since the
capability exists to inject boron into the reactor vessel to mitigate a
reduction in boron concentration during boil-off, and demonstrated natural
circulation flow paths exist in the core volume to minimize boron
stratification. The presence of the natural circulation flow paths provide
assurance that, in the event of a loss of the single residual heat removal
loop, the backup decay heat removal capability afforded by the 23 feet of
water above the vessel flange can be credited, regardless of whether the upper
internals assembly is installed or removed. For more detailed information
regarding the analysis, see Reference 1.
6.2-34 Revised 04/17/2013 Component Function Evaluation
Table 6.2-10 is an evaluation of the main components, which have been
previously discussed and a brief description of how each component functions
during normal operation and during an incident.
Passive Systems
The accumulators are a passive safety feature in that they perform their
design function in the total absence of an actuation signal or power source.
The only moving parts in the accumulator injection train are in the two check
valves.
The working parts of the check valves are exposed to fluid of relatively low
boric acid concentration contained within the reactor coolant loop. Even if
some unforeseen deposition accumulated, calculations have shown that a
differential pressure of about 25 psi will shear any particles in the bearing
that may attempt to prevent the valve from functioning.
The isolation valve at each accumulator is designed to withstand full reactor
coolant system pressure and is only closed momentarily for testing or when the
reactor is intentionally depressurized. The isolation valve (865) is normally
open and is maintained inoperative by keeping the motor circuit breaker locked
open in the motor control center.
The check valves operate in the closed position with a nominal differential
pressure across the disc of approximately 1550 psi. They remain in this
position except for testing or when called upon to function. Since the valves
operate normally in the closed position and are therefore not subject to the
abuse of flowing operation or impact loads caused by sudden flow reversal and
seating, they do not experience any wear of the moving parts, and function as
required.
When the Reactor Coolant system is being pressurized following refueling
outages, the accumulator check valves are tested for leakage. The testing
performed ensures that the associated check valve is in the closed position
and verifies seat leakage is acceptable while maintaining a minimum test
pressure differential across each valve. This testing is completed prior to
the reactor coolant system pressure exceeding 1000 psig, at which point the
discharge isolation valves are opened and de-energized, placing the
accumulators in their required standby configuration for normal operation.
6.2-35 Revised 04/17/2013 The accumulators are located inside the reactor containment and protected from
the reactor coolant system piping and components by a missile barrier.
Accidental release of the gas charge in the three accumulators would cause an
increase in the containment pressure of approximately 0.1 psi. This release
of gas has been included in the containment pressure analysis for the
loss-of-coolant accident, Section 14.3.
During normal operation, the flow rate through the reactor coolant piping is
approximately five times the maximum flow rate from the accumulator during
injection. Therefore fluid impingement on reactor vessel components during
operation of the accumulator is not restricting.
Emergency Flow to the Core
Special attention is given to factors that could adversely affect the
accumulator and safety injection flow to the core. These factors are:
a) Steam binding in the core, including flow blockage due to loop sealing
b) Carryover of accumulator water during blowdown
c) Short circuiting of the accumulator from the core to another part of the Reactor Coolant System
d) Loss of accumulator water through the break
All of the above are considered in the analysis, and are discussed
quantitatively in Section 14.
Recirculation Loop Leakage
Table 6.2-12 summarizes the maximum potential leakage from the leak sources of
the recirculation loop which goes through the residual heat removal pumps, a
residual heat exchanger and the high head safety injection pumps. In the
analysis, a maximum leakage is assumed from each leak source. For
conservatism, a leakage of 5 drops per minute was assumed from each flange
although each flange would be adjusted to essentially zero leakage. The total
maximum potential leakage resulting from all sources is listed in Table
6.2-12.
6.2-36 Revised 04/17/2013 During external recirculation, significant margin exists between the design
and operating conditions of the residual heat removal system components, as
shown in Table 6.2-13. In addition, during normal cooldown, operation of the
residual heat removal system is initiated when the reactor coolant pressure and temperature have been reduced to 450 psig and 350
°F respectively. The maximum accident operating conditions are far less rigorous. In view of the design margins, it is considered that the leakage rates in Table 6.2-12 are
conservative.
Leakage detection exterior to containment is achieved through use of sump
level detection. The Auxiliary Building sump pumps start automatically in the
event that liquid accumulates in the sump and an alarm in the control room
indicates that water has accumulated in the sump. Valving is provided to
permit the operator to isolate individually the residual heat removal pumps.
Pump NPSH Requirements
Residual Heat Removal Pumps
The NPSH of the residual heat removal pumps is evaluated for normal shutdown
operation, and both the injection and recirculation phase operation of the
design basis accident. Recirculation operation gives the limiting NPSH
requirement, and sufficient NPSH to initiate recirculation is determined by
the containment water level switches. During recirculation an adequate margin
exists between the available and required NPSH.
Safety Injection Pumps
The NPSH for the safety injection pumps is evaluated for both the injection
and recirculation phase operation of the design basis accident. The end of
injection phase operation gives the limiting NPSH requirement and the NPSH
available is determined from the elevation head and vapor pressure of the
water in the RWST, and the pressure drop in the suction piping from the tank
to the pumps.
6.2-37 Revised 04/17/2013
6.2.4 INSPECTIONS
AND TESTS
Inspection Capability
All components of the Safety Injection System can be inspected periodically to
demonstrate system readiness.
The pressure containing components are inspected for leaks from pump seals, valve packing, flanged joints and safety valves during system testing.
In addition, to the extent practical, the critical parts of the reactor vessel
internals, pipes, valves and pumps are inspected visually or by boroscopic
examination for erosion, corrosion, and vibration wear evidence, and for
non-destructive test inspection where such techniques are desirable and
appropriate.
Pre-Operational Testing
Component Testing
Pre-operational performance tests of the components are performed in the
manufacturer's shop. The pressure-containing parts of the pump are
hydrostatically tested in accordance with Paragraph UG-99 of Section VIII of
the ASME Code. Each pump is given a complete shop performance test in
accordance with Hydraulic Institute Standards. The pumps are run at design
flow and head, shut-off head and at additional points to verify performance
characteristics. NPSH is established at design flow by means of adjusting
suction pressure for a representative pump. This test is witnessed by
qualified Westinghouse personnel.
The remote operated valves in the safety injection systems will be
motor-operated. Shop tests for each valve include a hydrostatic pressure
test, leakage tests, a check of opening and closing time, and verification of
torque switch and limit switch settings. The ability of the motor operator to
move the valve with the design differential pressure across the gate is
demonstrated by opening the valve with an appropriate hydrostatic pressure on
one side of the valve.
The recirculation piping and accumulators are initially hydrostatically tested
at 150 percent of design pressure. The intake cooling water and component
cooling water pumps are tested prior to initial operation.
6.2-38 Revised 04/17/2013 System Testing
An initial functional test of the core cooling portion of the safety injection
systems is conducted during testing of the reactor coolant system before
initial startup. The purpose of the initial systems test is to demonstrate
the proper functioning of actuation circuits and to evaluate the dynamics of
placing the system in operation. This test is performed following the
flushing and hydrostatic testing of the system.
The functional test is performed with the Reactor Coolant System initially
cold and at low pressure. The safety injection system valving is set to
initially simulate the system alignment for power operation.
To initiate the test, the safety injection block switch is moved to the
unblock position to provide control power allowing the automatic actuation of
the safety injection relays from the low pressure signals from the pressurizer
instrumentation. Simultaneously, the breakers supplying outside power to the
4160 volt buses are tripped manually and operation of the emergency diesel
system automatically commences. The high head safety injection pumps and the
residual heat removal pumps are started automatically following the prescribed
diesel loading sequence. The valves are operated automatically to align the
flow path for injection into the reactor coolant system.
The rising water level in the pressurizer provides indication of systems
delivery. Flow into the reactor coolant system is terminated at a prescribed
pressurizer level by manually stopping safety injection pumps.
Tests are performed to provide information to confirm valve operating times, pump motor starting times, the proper automatic sequencing of, load addition
to the emergency diesels, and delivery rates of injection water to the reactor
coolant system.
Tests are performed for the various modes of operation needed to demonstrate
performance at partial effectiveness, i.e., to demonstrate the proper loading
sequence with loss of one of the diesel generator power sources. These latter
cases are performed without delivery of water to the reactor coolant system, but include starting of all pumping equipment involved in each test.
The systems are accepted only after demonstration of proper actuation and
after demonstration of flow delivery and shutoff head within design require-
ments.
6.2-39 Revised 04/17/2013 Flow is introduced into the Reactor Coolant loops through the accumulator
discharge line to demonstrate operability of the check valves and remotely
actuated stop valve and confirm L/D ratios of accumulator discharge lines used
in the calculation.
Post-Operational Testing
Component Testing
Routine periodic testing of the safety injection system components and all
necessary support systems at power is planned. No inflow to the Reactor
Coolant System will occur whenever the reactor coolant pressure is above 1500
psi. If such testing indicates a need for corrective maintenance, the
redundancy of equipment in these systems permits such maintenance to be
performed without shutting down or reducing load under conditions defined in
the Technical Specifications. These conditions include such matters as the
period within which the component should be restored to service and the
capability of the remaining equipment to meet safety limits within such a
period.
The accumulator discharge remote stop valves, MOV-*-865A, B, C, may be tested
for stroke time while the unit is in a hot standby to cold shutdown condition
for a refueling outage. Test capabilities are provided to periodically examine
the leakage back through the check valves and to ascertain that these valves
seat. It is expected that this test will be routinely performed when the
reactor is being returned to power after a refueling outage.
The residual heat removal and high head safety injection pumps will be run
periodically. Idle intake cooling water and component cooling water pumps
will be put in service periodically as part of the normal rotation of
machinery use. The content of the accumulators and the RWST are sampled
periodically to determine that the required boron concentration is present.
6.2-40 Revised 04/17/2013 System Testing System testing can be conducted during shutdown to demonstrate proper
automatic operation of the Safety Injection System. The system test
demonstrates the operation of the valves, pump circuit breakers, and automatic
circuitry. Isolation valves in the injection line will be blocked closed so
that flow is not introduced into the reactor coolant system. The test is
considered satisfactory if control board indication and visual observations
indicate all components have operated and sequenced properly.
During periodic RHR/HHSI pump testing, independent of the integrated
safeguards testing, it is verified that the pumps attain their required
discharge heads. The accumulators and the injection piping up to the final
isolation valve are maintained full of borated water while the unit is in
operation. The accumulators and the high head injection lines are refilled
with borated water as required by using the safety injection pumps.
Flow in each of the safety injection headers and in the main flow line for the
residual heat removal pumps is monitored by a flow indicator. Pressure
instrumentation is also provided for the main flow paths of the safety
injection and residual heat removal pumps.
The external recirculation flow paths are leak checked during periodic retests
at the normal operating pressures. This is accomplished by running each pump
which could be utilized during external recirculation (safety injection and
residual heat removal pumps) in turn at near shutoff head conditions and
checking the discharge and recirculation test lines. The suction lines are
tested by running the residual heat removal pumps and opening the flow path to
the safety injection pumps in the same manner as described above. Note that
an external pressure source may also be used.
During the above test, all system joints, valve packings, pump seals, leakoff
connections or other potential points of leakage are visually examined. Valve
gland packing, pump seals, and flanges are adjusted or replaced as required to
reduce the leakage to acceptable proportions. For power operated valves, final packing adjustments are made, and the valves are put through an
operating cycle before a final leakage examination is made.
The entire recirculation loop except the recirculation line to the residual
heat removal pumps is pressurized during periodic testing of the engineered
safety features components. Portions of the recirculation line to the
residual heat removal pump are capable of being leak checked during unit
shutdown. The unisolable suction line to the containment sump is leak tested
as part of the containment integrated leak rate test (ILRT).
6.2-41 Revised 04/17/2013 6.
2.5 REFERENCES
- 1. Engineering Evaluation PTN-ENG-SENS-07-032, "RHR System Operation with the Reactor Cavity Filled and the Vessel Upper Internals In Place," Revision 0. 2. Letter from R.C. Jones (USNRC) to N.J. Liparulo (Westinghouse), "Acceptance for Referencing of the Topical Report WCAP-12945(P), Westinghouse Code Qualification Document for Best-Estimate Loss of Coolant Analysis," June 28, 1996. 3. DELETED 4. Unit 3 PC/Ms that implemented the changes to address Generic Safety Issue 191, "Assessment of Debris Accumulation on PWR Sump Performance": PC/M 06-030, Containment Recirculation Sump Debris GSI-191 Resolution PC/M 06-075, Containment Recirculation Sump Cross-Connect Core Bore PC/M 07-006, Reactor Coolant Pump And Pressurizer Surge Line Insulation Replacement 5. Unit 4 PC/Ms that implemented the changes to address Generic Safety Issue 191, "Assessment of Debris Accumulation on PWR Sump Performance": PC/M 06-031, Containment Recirculation Sump Debris GSI-191 Resolution PC/M 06-049, Interim Containment Recirculation Sump Debris GSI-191 Resolution PC/M 06-055, Bioshield Wall Scupper Seals PC/M 06-071, Containment Recirculation Sump Cross-Connect Core Bore PC/M 07-007, Reactor Coolant Pump Insulation Replacement PC/M 07-081, Containment Spray Pump Seal Modification 6. Westinghouse Technical Report WCAP-17152-P Rev. 0, "Turkey Point Units 3 & 4 Extended Power Uprate Engineering Report," August 2012. 6.2-42 Revised 04/17/2013 TABLE 6.2-1
SAFETY INJECTION SYSTEM - CODE REQUIREMENTS
Component Code
Refueling Water Storage Tank AWWA D100-65
Residual Heat Exchanger
Tube Side ASME Section III*Class C
Shell Side ASME Section VIII
Accumulators ASME Section III Class C
Boron Injection Tank**** ASME Section III Class C
Valves USAS B16.5**
Piping USAS B31.1***
- ASME Section III - American Society of Mechanical Engineers, Boiler
and Pressure Vessel Code Section III, Nuclear Vessels - 1965
- Aloyco valve weld ends in accordance with Westinghouse Spec. No. G-676241, Dwg. No. 498B932, hydrostatically retested at system test pressure after
installation
- USAS B31.1 - Code for Pressure Piping - 1955
- The units 3/4 BIT have been bypassed and removed from the system flowpath.
Rev. 15 4/98 TABLE 6.2-2 Sheet 1 of 2
INSTRUMENTATION READOUTS ON THE CONTROL BOARD FOR OPERATOR
MONITORING DURING RECIRCULATION
Valves
System Valve Number
ACS HCV 3-758 ACS HCV 3-605 ACS MOV 3-744 A, B
ACS MOV 3-749 A, B
Instrumentation
System Channel Number
SIS FI 3-940
SIS FI 3-943
SIS FI 3-932
SIS FI 3-933
SIS LS3-1570
SIS LS3-1571
Note: Prefix 3 indicates that the component also is supplied for unit 4.
Rev. 16 10/99 TABLE 6.2-2 Sheet 2 of 2
Instrumentation
System Channel Number
ACS Fl 3-605
ACS LI 3-613 A ACS LI 3-614 A ACS TR 3-604 A, B
ACS TR 3-606
RCS LRCA 3-459
RCS LICA 3-460
RCS LICA 3-461
Pumps
SIS Safety Injection
ACS Component Cooling
ICS Intake Cooling
Rev. 16 10/99 TABLE 6.2-3 Sheet 1 of 3
QUALITY STANDARDS OF SAFETY INJECTION SYSTEM COMPONENTS
RESIDUAL HEAT EXCHANGER
A. Tests and Inspections
- 1. Hydrostatic Test
- 2. Radiograph of longitudinal and girth welds (tube side only)
- 3. UT of tubing or eddy current tests
- 4. Dye penetrant test of welds
- 5. Dye penetrant test of tube to tube sheet welds
- 6. Gas leak test of tube to tube sheet welds before hydro and
expanding tubes
B. Special Manufacturing Process Control
- 1. Tube to tube sheet weld qualifications procedure
- 2. Welding and NDT and procedure review
- 3. Surveillance of supplier quality control and product
COMPONENT COOLING HEAT EXCHANGER
A. Test and Inspections
- 1. Hydrostatic Test
- 2. Dye penetrant test of welds
B. Special Manufacturing Process Control
- 1. Welding and NDT and procedure review
- 2. Surveillance of supplier quality control and product
SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PUMPS
A. Tests and Inspections
- 1. Performance Test
- 2. Dye penetrant of pressure retaining parts
- 3. Hydrostatic test
TABLE 6.2-3 (cont'd) Sheet 2 of 3
B. Special Manufacturing Process Control
- 2. Surveillance of suppliers quality control system and product
A. Tests and Inspections
- 1. Hydrostatic test
- 2. Radiography of longitudinal and girth welds
- 3. Dye penetrant/magnetic particle of weld
B. Special Manufacturing Process Control
- 2. Surveillance of suppliers quality control and product
VALVES
A. Tests and Inspections
(a) 200 psi and 200 F or below (cast or bar stock)
- 1. Dye Penetrant Test
- 2. Hydrostatic Test
- 3. Seat Leakage Test
(b) Above 200 psi and 200 F
(i) Forged Valves
- 1. UT of billet prior to forging
- 2. Dye penetrant 100% of accessible areas after forging
- 3. Hydrostatic Test
- 4. Seat Leakage Test
(ii) Cast Valves
- 1. Radiographic 100%*
- 2. Dye Penetrant all accessible areas*
- 3. Hydrostatic Test
- 4. Seat Leakage Test
- For valves in radioactive services only
TABLE 6.2-3 Sheet 3 of 3
(c) Functional Tests Required for:
- 1. Motor Operated Valves
- 2. Auxiliary Relief Valves
B. Special Manufacturing Process Control
procedure review
- 2. Surveillance of suppliers quality control and product
- 3. Special Weld process procedure qualification (e.g. hard facing)
PIPING
A. Tests and Inspections
Butt Weld Inspection Schedule.
Pipe LIQUID PENETRANT RADIOGRAPHY Specification Finished Random All Welds
Class No. Root Pass Weld 10% 100%
2501 R N&R N R
2502 R N&R N R
2503 R N&R N R
1501 R N&R N R
1502 R N&R N R
901 R N&R N R
902 R N&R N R
903 R N&R N R
601 R N&R N R
602 R N&R N R
301 R N&R N R
302 R N&R N R
151 R N&R N R
152 R N&R N R
N -- for non-radioactive service
R -- for radioactive service
B. Special Manufacturing Process Control
Surveillance of suppliers quality control and product.
TABLE 6.2-4
ACCUMULATOR DESIGN PARAMETERS
Number 3
Type Stainless steel clad/
carbon steel
Design pressure, psig 700
Design temperature, oF 300 Operating temperature, oF 70 - 120
Normal pressure, psig 660
Minimum pressure, psig 600
Total volume, ft.
3 1200 Minimum water volume at
operating conditions, ft.
3 875 (2)
Boron concentration, ppm 2300 - 2600
Relief valve set point (1), psig 700
NOTES :
- 1. The relief valves have soft seats and are designed and tested to ensure zero leakage at normal operating pressure
- 2. A minimum Technical Specification volume of 872 ft.
3 for the tank above has been approved by the NRC as Operating Licensee Amendment Nos. 143/138, dated May 29, 1991. This approved minimum tank volume takes credit for
the accumulator water stored in the piping run between the accumulator
tank and the first check valve which, when added to the tank volume, equals 875 ft.
3
Revised 04/17/2013 TABLE 6.2-5
BORON INJECTION TANK
Number 1
Type Vertical
Total Volume (1) 900 gal Design Pressure 1750 psig
Design Temperature 300 o F Operating Pressure 0 - 1750 psig
Operating Temperature Ambient
Fluid (1) 1950 +400, -0 ppm Boric Acid Solution
Material Austenitic Stainless Steel
Code Class C,Section III ASME
NOTES :
- 1. The Units 3/4 BIT are empty, bypassed, and abandoned in place.
Rev. 15 4/98 TABLE 6.2-6
REFUELING WATER STORAGE TANK DESIGN PARAMETERS
Number 1
Material Carbon Steel with Epoxy Lining
Total volume, gal. 332,000
Normal pressure, psig atmospheric
Operating temperature, F ambient
Design pressure, psig atmospheric
Design temperature, F 200
Nominal boron concentration, ppm 2500 Boron concentration, ppm 2400 - 2600
Revised 04/17/2013 TABLE 6.2-7
PUMP PARAMETERS
Safety Injection Pump Design Parameters Number 4 (shared - Reference Appendix A)
Design pressure, discharge, psig 1,750
Design temperature, F 300
Design flow rate, gpm 300
Max. flow rate, gpm 625
Design head, ft 2,500 Shutoff head, ft 3,450 Material 11 - 13 Chrome
Motor H.P. 350
Type Horizontal centrifugal
Residual Heat Removal Pump Design Parameters
Number of pumps 2 (per unit)
Type Inline centrifugal
Design pressure, discharge, psig 600
Design temperature F 400
Design flow, gpm 3750
Design head, ft. 240
Material Austenitic Stainless steel
Maximum flow rate, gpm 3750 with NPSH of 14 ft.
Shutoff head, ft. 325
Motor H.P. 300
Rev. 16 10/99 TABLE 6.2-8
RESIDUAL HEAT EXCHANGERS DESIGN PARAMETERS
Number 2
Design heat duty, Btu/hr (Normal) 29.4 x 10 6
Design UA, Btu/hr/
oF 1.41 x 10 6
Design Cycles (50 o F - 400 oF) 200 Type Vertical Shell and U-tube
Tube-Side Shell-side
Design pressure psig 600 150
Design flow, lb/hr 1.875 x 10 6 5.12 x 10 6
Inlet temperature, oF 140 107.4
Outlet temperature, oF 124.4 113.2
NOTE :
- 1. Tube-side and shell-side parameters will vary from the design values for the Extended Power Uprate Project. Refer to Westinghouse Uprating Engineering Report WCAP-17152-P, Rev.0, August 2012 (Reference 6).
Revised 04/17/2013 TABLE 6.2-9 (a) Sheet 1 of 4 SINGLE FAILURE ANALYSIS - SAFETY INJECTION SYSTEM
COMPONENT MALFUNCTIONCOMMENTS A. Accumulator (injection phase) Delivery to broken loop Totally passive system with one accumulator per loop. Evaluation based on two accumulators delivering to the core and one spilling from ruptured loop B. Pump: (injection phase) 1) High head safety injection Fails to startFour provided for two units. Evaluation based on operation of two** 2) Residual heat removal Fails to startTwo provided. Evaluation based on operation of one 3) Component cooling* Fails to startThree provided. One required for recirculation
- 4) Intake cooling water Fails to start Three provided. Evaluation based on operation of one. (See also Section 6.2-3)
C. Automatically Operated Valves: Open on SIS -
Injection phase
- 1) High head cold leg isolation injection header valve Fails to open Two parallel valves. One required to open. 2) Residual heat removal pump isolation valve at injection line Fails to openTwo parallel valves. One required to open.* Recirculation phase
- Evaluation credits two HHSI pumps under a single train failure, one from the accident Unit and one from the opposite Unit.
Revised 09/20/2017 TABLE 6.2-9 (a) Sheet 2 of 4 COMPONENT MALFUNCTION COMMENTS D. Valves Operated from Control Room for Recirculation: (recirculation phase cold leg) 1) Containment Sump Fails to openTwo lines parallel with two valves in recirculation isolation series in each line, one pair of valves in either line is required to open.
- 2) Safety injection pump suction, suction valve at residual heat exchanger discharge Fails to open Two parallel valves. One required to open. 3) Isolation valve on the test line returning to the re-fueling water storage tank Fails as is Two valves in series. One required to close. 4) Isolation valve at suction header from refueling water storage tank Fails tocloseTwo valves in series. One required to close. 5) Isolation valves at residual heat removal pump suction line from refueling water storage tank Fails to close Two valves in series. One required to close. The status of all active components of the Safety Injection System is indicated on the control board. Reference is made to Table 6.2-2. Revised 09/20/2017 TABLE 6.2-9 (a) Sheet 3 of 4 COMPONENT MALFUNCTIONCOMMENTS D. 6) High head injection cold leg injection valve (outside containment) Fails to close(for hot leg recirculation) Two parallel valves. Remote isolation capability is provided by an upstream manual valve equipped with a reach rod.
- 7) Residual heat removal pump injection valve at injection line (inside containment) Fails to closeTwo parallel valves,both required to close for piggy-back recirculation. Remote isolation capability of direct cold leg injection downstream of each RHR heat exchanger is provided by manual isolation valves equipped with a reach rod.
E. Valves Operated from Control Room for Recirculation:
(recirculation phase - hot leg)
- 1) High head safety injection hot leg injection valve (outside containment) Fails to openThere is a manual bypass valve installed in parallel with this valve that can be opened utilizing a reach rod.
- 2) High head safety injection hot leg valves (inside containment) Fails to openTwo parallel valves. One required to open.
Revised 04/06/2018 TABLE 6.2-9 (a) Sheet 4 of 4 COMPONENT MALFUNCTIONCOMMENTS F. Valves Operated from Control Room for Recirculation: (recirculation phase - return to cold leg) 1) High head safety injection hot leg injection valve (outside containment) Fails to closeThe two high head safety injection valves inside containment can be closed to isolate this flow path. 2) High head safety injection hot leg injection valves (inside containment) Fails to closeThe high head safety injection valves outside containment can be closed to isolate this flow path.
Revised 09/20/2017 TABLE 6.2-9 (b)
LOSS OF RECIRCULATION FLOW PATH
Flow Path Indication of Loss of Flow Path Alternative Flow Path
From containment sump to low head
injection pump to residual heat
exchanger to high head injection
pumps to hot or cold legs No flow on high head cold leg
injection header or high head hot
leg injection header (independent
flow monitor on each injection
path). Cold Leg Recirculation: From
containment sump to low head
injection pump to residual heat
exchanger to cold legs via the
normal or alternate direct
injection line.
Hot Leg Recirculation: From
containment sump to low head
injection pump to residual heat
exchanger to alternate hot leg
injection path through shutdown
cooling supply line to the RCS 'C'
hot leg. Use of this flowpath
requires concurrent cold leg
recirculation.
NOTE: As shown on Figure 6.2-1, there are valves at all locations where alternative flow paths are provided. The alternate flowpaths are defense-in-depth measures for a beyond-design-basis passive failure that could occur in the normal hot leg or cold leg recirculation lineups. Te alternate flowpaths are not
analyzed in the design basis LOCA safety analyses and are not credited lineups in the emergency
operating procedures.
Revised 04/17/2013 TABLE 6.2-10 Sheet 1 of 2
COMPONENT FUNCTIONS EVALUATIONS
Normal Operating Normal Operating Accident Accident Component Function Arrangement Function Arrangement Boron Injection Tank None N/A N/A N/A (Unit 3/4 BIT are bypassed and abandoned in place)
Refueling Water Storage tank for Line up to suction Source of Lined up to Storage Tanks refueling operations of safety injection borated water suction of (1 per unit, can residual heat removal for core and safety injection be shared) and spray pumps, one spray nozzles residual heat to each unit removal, and spray one to each unit
Accumulators None Line up to cold legs of Supply borated Line up to cold (3 per unit) reactor coolant piping water to core legs of reactor coolant piping
Safety Injection None Line up to hot and cold Supply borated Line up to hot Pumps (4 shared) legs of reactor coolant water to core and cold legs piping, two to each unit of reactor coolant piping, two to each unit
Residual Heat Supply water to Lined up to take suction Supply borated Lined up to take Removal Pumps core to remove from refueling water water to core suction from (2 per unit) residual heat storage tank refueling water during shutdowns storage tank
Intake Cooling Water Supply cooling canal Two pumps in service Supply cooling One pump in Pumps (3 per unit) water to component canal water to service (See cooling heat component cool- Section 6.2-3) exchangers ing heat exchangers
Rev. 15 4/98 TABLE 6.2-10 Sheet 2 of 2
Normal Operating Normal Operating Accident Accident Component Function Arrangement Function Arrangement Component Cooling Supply cooling water One pump in service Supply cooling One pump in Pumps to station nuclear water to resi- service (3 per unit) components dual heat exchangers and S.I. pump bearings
Residual Heat Remove residual Line up for Cool water from Lined up for Exchangers heat from core recirculation containment recirculation (2 per unit) during shutdown sump for core cooling and containment spray
Component Cooling Remove heat from Three heat exchangers Cool water for Two heat ex-Heat Exchangers component cooling in service residual heat changers in (3 per unit) water exchangers service
Rev 8 7/90 TABLE 6.2-11 ACCUMULATOR INLEAKAGE
[DELETED]
Rev. 16 10/99 TABLE 6.2-12 MAXIMUM POTENTIAL RECIRCULATION LOOP LEAKAGE
Items No. of Type of Leakage Control and Unit Leakage to Leakage to Units Leakage Rate Used in the Analysis Atmosphere, Waste Holdup cc/hr Tank, cc/hr
_____________________________________________________________________________________________________________
- 1. Residual Heat Removal 2 Mechanical seal with auxiliary 0 0 Pumps backup packing with leakoff in between.
- 2. Spray Pumps 2 Mechanical Seal 0 0
- 3. Safety Injection Pumps 4 Mechanical Seal 0 0
- 4. Flanges: Gasket - adjusted to zero leakage following any test - 5 drops/min/ flange used in analysis
- a. Pumps 14 210 0 b. Valves bonnet to body 78 1170 0 (larger than 2")
- c. Control Valves 5 75 0
- 5. Valves - Stem Leakoffs 10 Backseated, double packing with 0 10 leakoff - 1 cc/hr/in stem diameter
- 6. Valves - Bonnet Equalizing 2 Backseated, single packing with 0 0 zero leakage
- 7. Misc. Small Valves 90 Flanged body packed stems - 1 drop/mn 270 0 used
- 8. Containment Spray 2 Gasket - 30 cc/hr for each of the 60 0 orifices 1785 10
Revised 04/17/2013 TABLE 6.2-13
RESIDUAL HEAT REMOVAL SYSTEM DESIGN, OPERATION AND TEST CONDITIONS
Pipes
Heat and
Pumps Exchangers Valves Fittings
Design Conditions Pressure, psig 600 600 665 700 Temperature, oF 400 400 400 400
Operating Conditions (Max) (NOTE 1) Pressure, psig 160 160 160 160 Temperature, o F (NOTE 2) 200 200 200 200
Test Pressure, psig 1200 900 1100 900
Allowable Pressure at Operating Temp. psig >600 >600 >690 >850
NOTES :
- 1. During post loss-of-coolant recirculation.
- 2. Maximum calculated RHR heat exchanger outlet temperature was calculated to be 198.8 oF for Extended Power Uprate. Refer to Westinghouse Uprating Engineering Report WCAP-17152-P, Rev. 0, August 2012 (Reference 6).
Revised 04/17/2013
6.3 EMERGENCY
CONTAINMENT COOLING
6.3.1 DESIGN
BASIS
Containment Heat Removal Systems
Criterion: Where an active heat removal system is needed under accident conditions to prevent exceeding containment design pressure, this
system shall perform its required function, assuming failure of any
single active component. (GDC 52)
Adequate heat removal capability for the Containment is provided by two
separate, engineered safety features systems. These are the Containment Spray
System, whose components are described in Section 6.4 and the Emergency
Containment Cooling System whose components operate as described in Section
6.3.2. Although
these systems operate simultaneously during a MHA to keep the
containment pressure from exceeding design pressure, each subsystem is provided
with redundant equipment such that they can accomplish their safety function
under any single active failure condition.
Performance Objectives
The Emergency Containment Cooling system is designed to provide the following
engineered safeguard functions:
(a) Cooling System: Remove sufficient heat from the reactor containment following a MHA, coincident with the operation of the Containment Spray
System (See Section 6.4), to keep the containment pressure from exceeding
design pressure. The emergency fan cooling units continue to remove heat
after the MHA and reduce the pressure to atmospheric. (See Fig. 6.3-1).
Refer to Sect. 14.3.4 for heat removal capacity.
(b) Filtering System: The Emergency containment filtering system is no longer credited to reduce the iodine concentration in the containment atmosphere due to the implementation of the Alternative Source Term (AST) methodology for Turkey Point (Reference 1). The AST methodology implementation credits the use of sodium tetraborate decahydrate (NaTB) inside containment for post-LOCA containment sump fluid pH control and for post-LOCA iodine removal/retention in conjunction with containment spray system operation. Therefore, the emergency containment filtering system has been disconnected and abandoned in place.
6.3-1 Revised 04/17/2013 Details of the site boundary and control room dose calculations are given in Section 14.3.5 for the MHA.
The housing of the emergency containment filtering systems have been abandoned in place after removal of the demisters, HEPA and charcoal filters. These housings contribute to the heat sink inventory in Unit 3 and Unit 4 containments, credited in the accident analysis with mitigating peak containment pressure and temperature, as described in UFSAR Section 14.3.4.
Portions of other systems which share functions and become part of this
emergency containment cooling system when required, are designed to meet the
criterion of this section. Any single active component failure in such systems
will not degrade the heat removal capability of containment cooling.
Where supporting systems are located outside containment, the following features
are incorporated in the design for operation under post accident conditions:
(a) Means for isolation of any section under malfunction or failure conditions (expected fault conditions).
(b) Means to detect and control radioactivity leakage into the environs, to the limits consistent with guidelines set forth in 10 CFR 50.67.
6.3.2 SYSTEM
DESIGN
The systems are described in the flow diagrams, Section 9. The Emergency
Containment Cooling System components and their supports meet the requirements
for Class I structures.
Cooling System Characteristics
The Emergency Containment Cooling System, in each containment consists of three
fan cooling units; each consisting of a motor, fan, bare tube cooling coil, instrumentation and controls. The units are located above the refueling floor, around the inside of each containment between the containment wall and the
secondary compartment shield walls. The location of the three fan cooling units
provides individual isolation and prevents recirculation between units.
6.3-2 Revised 04/17/2013 Each fan is designed to provide 25,000 CFM at both the design condition of 283 o F and 59 psig and the in-place testing condition at atmospheric pressure. The
emergency cooling unit design pressure and temperature (59 psig/283 o F) conservatively envelope the licensed containment design basis conditions of 55
psig and 283 o F. The fans are of the directly driven vane-axial type (See Table 6.3-1).
The temperature of the component cooling water from each fan cooling unit is
recorded in the control room. The total return water flow from all fan cooling
units is indicated in the control room. Following a loss-of-coolant accident, the safety injection signal will automatically energize motor circuits to start
the dedicated Train A and B emergency cooling fans.
The third swing emergency containment cooling fan automatically starts in the event of a valve or fan
failure in either of the other two emergency containment cooling units. This
ensures that two emergency containment cooling units are available post
accident. The third swing emergency containment cooling fan can also be
manually started. These units are not normally in service during reactor
operation.
The original design basis allowed all three emergency containment fans to start on a safety injection signal. With various postulated single active failures, two or three fans could operate. For the Thermal and Extended Power Uprate
Projects, revised containment and component cooling water system thermal
analyses were performed. The revised containment analyses have demonstrated that
two emergency containment fan coolers and one train of Containment Spray are
required for short-term accident mitigation. To ensure long-term containment
post-accident conditions can be reduced to within applicable electrical
equipment qualification bases, two emergency containment fan coolers and one
train of Containment Spray are required at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The revised analyses have
demonstrated that significant operating margin can be gained if only two
emergency containment fans operate. Based on these analyses, the Thermal Uprate
Project modified the plant configuration such that only two of the three
emergency containment fans can start on a safety injection signal. To maintain
adequate long-term containment cooling capability, the third emergency
containment fan is required to be available and will automatically start if
there is a failure of one of the other units.
During the emergency operation the air-steam mixture is forced upward through the coils and discharged into the upper regions of the containment. The
condensate will drain downward and out the bottom of the unit from whence it
will flow by gravity through the floor drain system to the containment sump.
6.3-3 Revised 04/17/2013 Filtering System Characteristics - DELETED Redundancy Provisions
The cooling water requirements for three emergency fan cooling units can be
supplied by any one of the component cooling water pumps. Component cooling
water pumps will be automatically started as part of the emergency loading
sequence on loss of normal A.C. power, coincident with a requirement for
engineered safety features operation (safety injection signal).
The emergency containment cooling units are supplied by individual lines from
the component cooling water loop. Each inlet line is provided with a power
operated shut-off valve and a drain valve. Similarly, each discharge line from
a cooler is provided with a power operated shut-off valve and bypass valve.
This allows each cooler to be isolated individually for maintenance.
The cooling coils and water lines are completely closed inside the containment.
No radioactive leakage is expected into these coils because there is no
interface with the RCS, and the coolers operate with the component cooling water
pressure greater than the containment pressure.
During normal operation, a minimal cooling water flow through the emergency fan
cooling units is maintained. A bypass is provided around the full flow valve on
the discharge line from each unit. The full flow valve opens automatically in
the event that its associated fan starts. On loss of instrument air in Unit 3
an accumulator maintains air supply to the full flow valve for a short time, then it fails open. On a loss of instrument air the bypass valve fails closed.
On Unit 4 the full flow valve will fail open upon a loss of instrument air and
the bypass valve will fail closed. The full flow valve fails closed on a loss
of power to its associated fan and the bypass valve fails open allowing only
minimal bypass flow to be diverted to an inactive Emergency Containment Cooler.
6.3-4 Revised 04/17/2013 Actuation Provisions Receipt of safety injection signal will energize motor control circuits to start the dedicated Train A and B emergency fan cooling units. The swing emergency cooling fan shall start automatically following failure of a dedicated Train.
Overload protection for the fan motors is provided at the Motor Control Center
by overcurrent trip devices in the motor starters. The fan motors can be
operated from the control room, with one exception. The fan motors, power
cables and control circuits for the Unit 3 and 4 emergency containment filter
units 3V3A, 3V3B, 3V3C, 4V3A, 4V3B, 4V3C were disconnected and abandoned in
place, and the control switches were removed from the control room.
Environmental Protection
The fan motor enclosures, electrical insulation and bearings are designed for operation during accident conditions, with one exception. The Unit 3 and 4 emergency filtering system is no longer required to operate during accident conditions. ECF units 3V3A, 3V3B, 3V3C, 4V3A, 4V3B, 4V3C and related SSCs and design features were disconnected, removed and/or abandoned in place. These ECs removed the demisters, HEPA and charcoal filters from the Unit 3 and 4 ECF units and isolated the ECF dousing lines from the Unit 3 and 4 containment spray system. The Unit 3 and 4 ECF units are no longer functional or credited with post LOCA iodine removal and retention.
6.3-5 Revised 04/17/2013
6.3.3 DESIGN
EVALUATION
Flow Tests and Analysis
Provisions have been included in the emergency cooling system to balance and adjust water flow to satisfy post-accident conditions.
Reliance on Interconnected Systems The Operation of the Emergency Containment Cooling System is dependent on the following systems:
- 1. Component cooling water 2. Electrical System 3. Intake Cooling Water System The redundancy provisions of these systems are described in Sections 6, 8 and 9.
6.3.4 MINIMUM
OPERATING CONDITIONS Technical specifications establish limiting conditions regarding the operability of the Emergency Containment Cooling units when the reactor is critical.
6.3.5 TESTS
AND INSPECTIONS
Inspection of Containment Pressure-Reducing Systems
Criterion: Design provisions shall be made to extent practical to facilitate the periodic physical inspection of all important components of the
containment pressure-reducing systems, such as pumps, valves, spray
nozzles, tanks and sumps. (GDC 58)
Design provisions are made to the extent practical to facilitate access for
visual inspection of all important components of the Emergency Containment
Cooling System.
The fan motors and the fan rotating assemblies are statically and dynamically
tested for proper balance.
6.3-6 Revised 04/17/2013 Testing of Containment Pressure-Reducing Systems Components Criterion: The containment pressure-reducing systems shall be designed to the extent practical so that components, such as pumps and valves, can
be tested periodically for operability and required functional
performance. (GDC 59)
The Emergency Containment Cooling System is designed to the extent practical so
that the components can be tested, after any component maintenance for
operability and functional performance. Provisions for hydro-testing of the
coils during shutdown are provided.
Testing of Operational Sequence of Containment Pressure-Reducing Systems
Criterion: A capability shall be provided to test initially under conditions as close to practical to the design and the full operational sequence
that would bring the containment pressure-reducing systems into
action including the transfer to alternate power sources. (GDC 61)
Means are provided to test initially to the extent practical the full
operational sequence of the Emergency Containment Cooling System including
transfer to alternate power sources.
6.3-7 Revised 04/17/2013
6.3.6 MOTORS
FOR EMERGENCY CONTAINMENT FANS
General These totally enclosed fan cooled motors will have a useful life of sixty (60)
years under the normal containment service conditions as demonstrated by the
appropriate EQ documentation package (See Appendix 8A). Internal heaters will
dispel moisture condensation when motor is idle, with one exception. The fan
motors, power cables and control circuits for the Unit 3 and 4 emergency
containment filter units 3V3A, 3V3A, 3V3B, 3V3C, 4V3A, 4V3b, 4V3c were
disconnected and abandoned in place, and the control switches were removed from
the control room.
Insulation
The insulation will be a special Class B suitable for MHA conditions. The
insulation system is described in Table 6.3-2.
Bearings The bearings will be specially selected, conservatively rated ball bearings with
clearances, held to closer tolerances than standard bearings.
The lubricant will be a high temperature, radiation resistant grease listed in
Table 6.3-2.
The bearing housing will be so constructed that a sudden pressure wave will in
no way impair the lubrication or serviceability of the bearings.
Qualification of materials is stated in Section 6.7.
6.
3.7 REFERENCES
- 1. Turkey Point Units 3 and 4 - Issuance of Amendments Regarding Alternative Source Term (Tac Nos. ME 1624 and ME 1625), Issued June 23, 2011, ML 110800666.
6.3-8 Revised 04/17/2013 TABLE 6.3-1
EMERGENCY CONTAINMENT COOLING UNIT CHARACTERISTICS
(Each Unit) (Each Unit)
Component Design Characteristics Flow Accident Arrangement
Emergency Containment Remove heat at rate of 25,000 cfm 2 operating, one on standby and automatically start if a failure occurs in another unit. Cooling Units (3) 60 x lO 6 BTU/hr Emergency Containment Remove 60 x 10 6 BTU/hr 2,000 GPM (1) Type Description Cooling Coil one/ Bare staggered tube, air/water air/water cooling Unit (3) admiralty metal; inlet temp. outlet temp.
150 psig design.
283 o F/125 o F 278 o F/185 o F NOTE:
- 1. 2,000 gpm - Continuous Operation 3,200 gpm - 1 Month (Post-LOCA Recirculation Limit) 3,600 gpm - 1 week 5,000 gpm - 24 Hours (Initial Safety Injection) 5,500 gpm -1 Hour
Revised 04/17/2013
6.4 CONTAINMENT
SPRAY SYSTEM
6.4.1 DESIGN
BASES
Containment Heat Removal Systems
Criterion: Where an active heat removal system is needed under accident conditions to prevent exceeding containment design pressure, this
system shall perform its required function, assuming failure of
any single active component. (GDC 52)
Adequate containment heat removal capability for the Containment is provided by two separate, engineered safety feature systems. The Containment Spray
System, whose components operate in the sequential modes described in 6.4.2, and the Emergency Containment Cooling and Filtering System which is discussed
in Section 6.3.
The primary purpose of the Containment Spray System is to spray cool water
into the containment atmosphere when appropriate in the event of a
loss-of-coolant accident (LOCA) or main steam line break (MSLB). Operation
of the Containment Spray System and the Emergency Containment Cooling System
will ensure that containment pressure does not exceed its design value which is 55 psig at 283
°F (100% R.H.). This protection is afforded for all pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant pipe. Pressure and temperature
transients for loss of coolant accident are presented in Section 14.
Although the water in the core after a loss-of-coolant accident is quickly
subcooled by the Safety Injection System, the Containment Spray System design
is based on the conservative assumption that the core residual heat is
released to the containment as steam.
The original design basis for containment heat removal considered
simultaneous operation of one spray pump and 2 of 3 emergency containment
coolers. The design basis relative to the emergency containment coolers was
changed to provide significant component cooling water system operating
margin at uprated conditions. For short-term accident mitigation, only 1 of
3 emergency containment coolers is required and long-term mitigation would
require 2 of 3 emergency containment coolers. This is the basis for the
containment pressure transient calculations in Section 14.
6.4-1 Revised 04/17/2013 Inspection of Containment Pressure Reducing Systems
Criterion: Design provisions shall be made to the extent practical to facilitate the periodic physical inspection of all important
components of the containment pressure reducing systems, such as
pumps, valves, spray nozzles and sumps. (GDC 58)
Where practicable, all active components and passive components of the Containment Spray Systems are inspected periodically to assure system
readiness. The pressure containing systems are inspected for leaks from pump
seals, valve packing, flanged joints and safety valves. During operational
testing of the containment spray pumps, the portions of the systems subjected
to pump pressure are inspected for leaks. Design provisions for inspection
of the Safety Injection System, which also function as part of the
Containment Spray System, are described in Section 6.2.5.
Testing of Containment Pressure - Reducing Systems Components
Criterion: The containment pressure reducing systems shall be designed, to the extent practical so that active components, such as pumps and
valves, can be tested periodically for operability and required
functional performance (GDC 59).
All active components in the Containment Spray Systems are adequately tested both in pre-operational performance tests in the manufacturer's shop and
in-place testing after installation. Thereafter, periodic tests are also
performed after any component maintenance. Testing of the components of the
Safety Injection System used for containment spray purposes are described in
Section 6.2.5.
The component cooling water pumps and the intake cooling water pumps which
supply the cooling water to the residual heat exchangers are in operation on
a relatively continuous schedule during plant operation. Those pumps not
running during normal operation may be tested by changing the operating
pump(s).
6.4-2 Revised 04/17/2013 Testing of Containment Spray Systems Criterion: A capability shall be provided to the extent practical to test periodically the delivery capability of the containment spray system at a position as close to the spray nozzles as is
practical. (1967 Proposed GDC 60)
Test lines (2-inch permanent for mini-flow, a permanent 6-inch for full flow for Unit 3 and Unit 4) for all the containment spray loops are located so
that all components up to the containment isolation valves may be tested.
The manual isolation valves are checked for leakage during local leak rate
testing (LLRT).
The spray nozzles in containment are periodically verified to be unobstructed
by verification of air flow by the use of thermography or other appropriate
means.
The Containment Spray System piping up to the final isolation valve is
maintained full of borated water at refueling water concentration while the
reactor is in operation to ensure that the Containment Spray System can
reliably perform its intended function. To address this, Generic Letter
2008-01 was issued to discuss the consequences of gas entrained in systems
such as the Containment Spray System that could compromise their operability.
In response to this, a Gas Accumulation Management Program (GAMP) was
established at the Turkey Point Nuclear Units to provide long-term void
management. Locations in the Containment Spray System in the pump discharge
and suction piping up to the first closed discharge line isolation valve
where gases could potentially accumulate are periodically monitored using
ultrasonic testing and/or vented to verify the system is filled. From the
results of this monitoring, the GAMP ensures that gas accumulated within the
Containment Spray System is identified, evaluated, trended, and effectively
controlled to prevent unacceptable degradation of performance of any
structures, systems or components, ultimately to ensure system operability.
Filling and venting operations and periodic system operational and leakage
tests are required to ensure that the Containment Spray System is not damaged
from water-hammer loads that may result from pump flows into voided discharge
lines. The system must be periodically verified full by venting the
accessible discharge piping high points. The inaccessible discharge piping
high points that may be susceptible to gas accumulation are deemed
appropriate to proactively provide the capability to vent each of these
locations and to allow for future monitoring and trending, if it becomes
necessary.
6.4-3 Revised 04/17/2013 Testing of Operational Sequence of Containment Pressure-Reducing Systems Criterion: A capability shall be provided to test initially under conditions as close as practical to the design and the full operational
sequence that would bring the containment pressure-reducing
systems into action, including the transfer to alternate power
sources. (1967 Proposed GDC 61)
Capability is provided to test initially to the extent practical the operational startup sequence beginning with transfer to alternate power
sources and ending with near design conditions for the Containment Spray
System, including the transfer to the alternate emergency diesel-generator
power source.
Performance Objectives
The Containment Spray System was originally designed to spray at least 2900
gpm of borated water into the containment building whenever the coincidence
of two out of three high and two out of three (high-high) containment
pressure signals occurs or a manual signal is given. Either of two
subsystems containing a pump and associated valving and spray headers are
independently capable of delivering one-half of this flow or 1450 gpm. The
current containment analysis models the spray flow as a function of
containment pressure. For example, for the limiting case of the containment
analysis, spray flow capability at 50 psig is 1293 gpm during injection phase
and 1575 gpm during cold leg recirculation phase.
Inspection of Air Cleanup Systems Criterion: Design provisions shall be made to the extent practical to facilitate physical inspection of all critical parts of containment air cleanup systems, such as ducts, filters, fans and dampers, (GDC 62) Access is available for visual inspection of Emergency Containment Filtering System components.
The NaTB Baskets shall be inspected to ensure there is no damage and that they can readily perform their safety related function of dispersing the NaTB for pH control.
Engineered Safety Features Components Capability Criterion: Engineered safety features shall be designed so that the capability of these features to perform their required function is not impaired by the effects of a loss-of -coolant accident to the extent of causing undo risk to the health and safety of the public (1967 proposed GDC 42)
The NaTB Baskets are designed to disperse the NaTB readily once in contact with flood water. A loss-of-coolant accident would not affect the engineered safety features of the NaTB Baskets.
6.4-4 Revised 04/17/2013 Accident Aggravation Prevention Criterion: Protection against any action of the engineered safety features which would accentuate significantly the adverse after-effects of normal cooling shall be provided. (1967 proposed GDC 43).
The NaTB Baskets disperse NaTB in the flood water raising the pH to the range of 7.0 to 8.048. This results in prevention of iodine generation, stress corrosion cracking in stainless steels and fuel scaling and debris generation due to aluminum.
The design basis is to provide sufficient heat removal capability, in
conjunction with the Emergency Containment Cooling System, to maintain the
post-accident containment pressure below the design pressure assuming that
the core residual heat is released to the containment as steam. A further
design basis is to remove sufficient containment heat to reduce the
containment temperature within the time limits of the Environmental
Qualification (EQ) curve. This requires a heat removal capacity of the
subsystem, with either pump operating, at least equivalent to two fan-coolers
heat removal capability at the containment design conditions.
The spray system is designed to operate over an extended time period, following a reactor coolant system failure to help restore and maintain
containment conditions at near atmospheric pressure. It can accomplish its
containment pressure reducing function (and consequent containment leakage
reduction) under the most limiting single active failure condition.
Portions of other systems which share functions and become part of the
containment cooling system when required are designed to meet the criteria of
this section. Any single failure of active components in such systems does
not degrade the heat removal capability of containment cooling.
Those portions of the spray systems located outside of the containment which
are designed to circulate, under post-accident conditions, radioactively
contaminated water collected in the containment meet the following
requirements:
a) Adequate shielding to maintain radiation levels within the guidelines of 10 CFR 100 (Section 11.2).
b) Collection of discharges from pressure relieving devices into closed systems.
c) Means to limit radioactivity leakage to the environs, consistent with guidelines set forth in 10 CFR 100.
6.4-5 Revised 04/17/2013 System active components are redundant. System piping located within the containment is redundant and separable in arrangement and is fully protected
from damage which may follow any reactor coolant system failure.
System isolation valves relied upon to operate for containment cooling are
redundant, with automatic actuation or manual actuation.
Service Life
All portions of the system located within containment are designed to
withstand, without loss of functional performance, the post-accident
containment environment and operate without benefit of maintenance for the
duration of time to restore and maintain containment conditions at near
atmospheric pressure.
The NaTB Baskets located in the containment sump will passively achieve a minimum pH of 7.0 at the onset of containment spray recirculation and will ensure that environmental qualification and chemical effects pH limits are not exceeded. The NaTB Baskets are located below the minimum post-LOCA flood height ensuring the NaTB is quickly and completely dissolved in the sump fluid.
Codes and Classifications
Table 6.4-1 tabulates the codes and standards to which the containment spray
system components are designed.
6.4.2 SYSTEM
DESIGN AND OPERATION
System Description
Post-accident containment heat removal is provided, in part, by the
Containment Spray System shown in Figures 6.4-2 and 6.4-3 whose components
operate in sequential modes. These modes are:
a) Spray a portion of the contents of the refueling water storage tank into the entire containment atmosphere using the containment spray
pumps.
b) Recirculation of water from the containment sump is provided by the diversion of a portion of the recirculation flow from the discharge of
the residual heat removal heat exchangers to the suction of the spray
pumps after injection from the refueling water storage tank has been
terminated.
6.4-6 Revised 04/17/2013 The bases for the selection of the various conditions requiring system actuation is presented in Section 14.
The principal components of the Containment Spray System which provides
containment cooling following a loss of coolant accident consists of two
pumps, spray ring headers and nozzles, and the necessary piping and valves.
The containment spray pumps are located in the Auxiliary Building and the
spray pumps take suction directly from the refueling water storage tank.
The Containment Spray System also utilizes the two residual heat removal
pumps, two residual heat exchangers and associated valves and piping of the
Safety Injection System for the long term recirculation phase of containment
cooling.
The spray system will be actuated by the coincidence of two out of three high
and two out of three (high-high) containment pressure signals. This starting
signal will start the pumps and open the discharge valves to the spray
header. If required, the operator can manually actuate the system from the
control room, and periodically, the operator will actuate system components
to demonstrate operability.
The system design conditions were selected to be compatible with the design
conditions for the low pressure injection system since both of these systems
share the same suction line.
The NaTB Baskets located in the containment sump will passively achieve a minimum pH of 7.0 at the onset of containment spray recirculation and will ensure that environmental qualification and chemical effects pH limits are not exceeded. The NaTB Baskets are located below the minimum post-LOCA flood height ensuring the NaTB is quickly and completely dissolved in the sump fluid.
Recirculation Phase
Revised limiting containment integrity analyses assume one spray pump in operation for the duration of the event (injection and recirculation phases).
The heat removal capacity of two of the three fan coolers operating in conjunction with one containment spray pump is sufficient to remove the corresponding energy addition to the vapor space resulting from steam boil o from the core assuming flow into the core from one residual heat removal pump at the beginning of recirculation without exceeding containment design pressure.
6.4-7 Revised 04/17/2013
Cooling Water The cooling water for the residual heat exchangers has been described in
Section 9.3.
Change-Over
The procedure for the changeover from injection to recirculation has been
described in Section 6.2.
Remote operated valves of the Containment Spray System which are under manual
control (that is, valves which normally are in their ready position and do
not receive a containment spray signal) have their positions indicated on a
common portion of the control board. At any time during operation when one
of these valves is not in the ready position for injection, it is shown
visually on the board.
Components
All associated components, piping, structures, and power supplies of the
Containment Spray System are designed to Seismic Class I criteria.
The Containment Spray System shares the refueling water storage tank liquid
capacity with the Safety Injection System. Refer to Section 6.2.2 for a
detailed description of this tank.
Pumps The two containment spray pumps are of the horizontal centrifugal type driven
by electric motors. These motors can be powered from both normal and
emergency power sources.
The design head of the pumps is sufficient to continue at rated capacity with
a minimum level in the refueling water storage tank against a head equivalent
to the sum of the design pressure of the containment, the head to the
uppermost nozzles, and the line and the nozzle pressure losses. Pump motors
are direct-coupled and large enough for the maximum power requirements of the
pumps. The materials of construction are stainless steel or equivalent
corrosion resistant material. Design parameters are presented in Table
6.4-2.
6.4-8 Revised 04/17/2013 The original containment spray pumps are designed in accordance with the
specifications discussed for the pumps in the Safety Injection System, Section 6.2.2. Spare containment spray pump pullout assemblies (a pullout
assembly is the pump less the casing, coupling, and motor) are used
interchangeably with the original pumps to enhance reliability and
availability of the pumps. The spare assemblies are designed, fabricated and
inspected to ASME Section III, Class 2 (less N stamp), 1980 Edition through
Winter 1982 Addenda. All parts in contact with a boric acid solution are
austenitic stainless steel type 304L or 316L. All pressure containment parts
of the pullout assemblies are chemically and mechanically analyzed and the
results are checked to ensure conformance with the applicable ASME
specification. In addition, all pressure retaining part materials are
nondestructively examined in accordance with the ASME Boiler & Pressure
Vessel Code, 1980 Edition through Winter 1982 Addenda.
The pump motors are direct-coupled and non-overloading to the end of the pump
curve.
Details of the component cooling pumps and intake cooling water pumps, which
serve the Safety Injection System, are presented in Section 9.
Spray Nozzles
The spray nozzles, of the ramp bottom design, are not subject to clogging by
particles less than 3/8 inch in maximum dimension, and are capable of
producing a mean drop size of approximately 700 microns in diameter at a flow
rate of 15.2 gpm @ 40 psid per nozzle.
During containment spray recirculation operation, the water is filtered
through a series of strainer modules with 3/32 inch openings. The spray
nozzles are bronze and have a 3/8 inch diameter orifice. The nozzles are
connected to five straight parallel horizontal fingers on each header. There
are 86 Spraco Model 1713 nozzles on each of two main headers located in Unit
4 containment and 86 nozzles in Unit 3's North header and 85 nozzles in the
South header.
The nozzles and headers are so oriented as to ensure adequate coverage of the
containment volume.
6.4-9 Revised 04/17/2013 Heat Exchangers
The two residual heat exchangers which are used during the recirculation
phase are described in Section 6.2.2.
Valves The valves for the Containment Spray System are designed in accordance to the
specifications discussed for the valves in the Safety Injection System.
Valving descriptions and valve details are shown in Section 6.2.
Piping The piping for the Containment Spray System is designed in accordance to the
specifications discussed for the piping in the Safety Injection System (Section 6.2.2).
The pump suction piping is designed for 200 psig at 300 o F and discharge piping is designed for 300 psig at 300 o F.
NaTB Baskets The NaTB Baskets provide (1) a means of containing the NaTB, (2) a means of measuring the quality of NaTB, and (3) a means for the NaTB to dissolve and mix with the sump fluid. The pH buffering function provided by NaTB helps minimize the evolution of iodine and prevent stress corrosion cracking of stainless steels and will ensure that environmental qualification pH limits in Doc Pac 1001, Section 6.4 are not exceeded. Limiting the PH to the range of 7.0 to 8.048 is also critical in preventing fuel scaling and the generation of debris due to chemical reaction of the Carbon Steel (CS) and Aluminum and other containment materials such as insulation.
In order to bound any potential acidic material addition to the containment such as boric acid leaks, the chemical effects analysis for determining sump strainer debris loading and fuel scaling was done assuming an additional amount of NaTB was added to containment equal to a small basket. With the additional basket added, a pH value of 8.079 was determined for analysis purposes. Utilizing a pH of 8.079 bounds the maximum pH of 8.048 calculated using 10 baskets.
6.4-10 Revised 04/17/2013 To support Alternate Source Term (AST), the location and storage geometry of the NaTB support the safety related function to provide pH buffering.
Therefore, the basket screens and the basket frame which support these functions are Safety Related. The NaTB Baskets are seismically designed. In response to NRC RAI question 2 (L-2009-195) the baskets are designed to withstand Class III loads which will preclude interaction with Class I SSC's.
The Locations for the NaTB Baskets were chosen based on the following considerations:
- Distribution in containment to ensure contact with flooding.
- Avoid seismic interaction with safety related SSC's.
- Outside the area of influence for High Energy Line Breaks (HELB).
- Does not interfere with plant operations.
- At least 30 feet unobstructed distance from existing sump strainers.
- Not in containment Sump High Flow areas.
The following is a list of NaTB Basket critical attributes necessary to accomplish the component safety functions:
- All 4 Leveling feet must be on the containment floor. No casters should support basket.
- Basket bottom screens must be no greater than 4.5" off the floor for small baskets and 3.5" off the floor for large baskets.
- Baskets must be in their specified locations.
- Baskets must be filled with NaTB chemical to a level between the minimum fill bar and the top of the basket.
- Basket covers must be installed to prevent dissolving any of the NaTB caused by any potential spray from leaking system Motors for Pumps and Valves
The motors for the Containment Spray System are designed in accordance to the
specifications discussed for motors in the Safety Injection System. (Section
6.2.2)
6.4-11 Revised 04/17/2013 Electrical Supply
Details of the normal and emergency power sources are presented in the
discussion of the Electrical System, Section 8.
Environmental Protection
The spray headers are located outside and above the primary and secondary
concrete shield. During operation a movable shield also provides missile
protection for the area immediately above the reactor vessel. The spray
headers are therefore protected from missiles originating within the shield.
All of the active components of the containment spray system are located
outside the containment, and hence are not required to operate in the
steam-air environment produced by the accident.
Material Compatibility
Parts of the system in contact with borated water, are stainless steel or an
equivalent corrosion resistant material, except for the spray headers (and
piping risers) which are carbon steel.
6.4.3 DESIGN
EVALUATION
Range of Containment Protection
During the injection phase following the maximum loss-of-coolant accidents (i.e., during the time that the containment spray pumps take their suction
from the refueling water storage tank) this system, in conjunction with two
emergency containment fan coolers, provides the design heat removal capacity
for the containment. After the injection phase, each train of the
recirculation system provides sufficient cooled recirculated water to keep
the core flooded as well as providing, if required, sufficient flow to the
suction of the containment spray pumps to satisfy the Containment Spray
System heat removal requirements. This applies for all reactor coolant pipe
sizes up to and including the hypothetical instantaneous circumferential
rupture of a reactor coolant pipe. Only one pumping train and one heat
exchanger are required to operate for this capability at the earliest time
recirculation is initiated.
6.4-12 Revised 04/17/2013 For the MHA concurrent with limiting cooling system heat removal capability, continued operation of recirculation spray may be required during long-term
recirculation to restore post-accident containment temperatures to pre-event
conditions.
During the injection and recirculation phases the spray water is raised to
the temperature of the containment in falling through the steam-air mixture.
The minimum fall path of the droplets is approximately 70 ft. from the spray
headers to the operating deck. The actual fall path is longer due to the
trajectory of the droplets sprayed out from the header. Heat transfer
calculations, based upon 700 micron droplets, show that thermal equilibrium
is reached in a distance of approximately five feet. Thus, the spray water
reaches essentially the saturation temperature. The model for spray droplet
heat removal is discussed in Section 14.3.
In addition to heat removal, the spray system is effective in scrubbing
fission products from the containment atmosphere. However, no credit is
taken for absorption of iodine in the analysis of the hypothetical accident (Section 14.3).
System Response
The starting sequence of the containment spray pumps and their related
emergency power equipment is designed so that delivery of the minimum
required flow is reached within 60 seconds which is the delay assumed for the
starting of emergency containment cooling (Section 14.3).
Single Failure Analysis
A failure analysis has been made on all active components of the system to
show that the failure of any single active component will not prevent
fulfilling the design function. This analysis is summarized in Table 6.4-3.
The analysis of the loss-of-coolant accident presented in Section 14 reflects
the single failure analysis.
6.4-13 Revised 04/17/2013 Reliance on Interconnected Systems
The Containment Spray System initially operates independently of other
engineered safety features following a loss-of-coolant accident. For
extended operation in the recirculation mode, water is supplied through the
residual heat removal pumps. Spray pump cooling is supplied from the
component cooling loop.
During the recirculation phase, the flow leaving the residual heat exchangers
may be directed to the suction of the containment spray pumps and the high
head safety injection pumps. The systems are designed, and administrative
controls provided to assure adequate safety injection and containment spray
flow in the recirculation phase is maintained without the need for manual
flow adjustments.
The Containment Spray System operates in conjunction with the Emergency Containment Cooling System during both the injection and recirculation phases
of a loss-of-coolant accident to provide the required heat removal
capability.
Normal and emergency power supply requirements are discussed in Section 8.
6.4.4 INSPECTIONS
AND TESTS
Inspection Capability
All components of the Containment Spray System can be inspected periodically
to demonstrate system readiness.
On an 18 month interval, during operational testing of the containment spray
pumps, the portions of the system subjected to pump pressure are inspected
for leaks. This includes pump seals, valve packings, flanged joints and
safety valves.
Component Testing
All active components in the Containment Spray System are tested both in
preoperational performance tests in the manufacturer's shop and in-place
testing after installation.
6.4-14 Revised 04/17/2013 The containment spray pumps can be tested singly by opening the valve in the
containment spray recirculation test line. Each pump, in turn, can be
started by operator action and checked for flow establishment. The spray
valves can be tested with the pumps shut down.
During these tests the equipment will be visually inspected for leaks.
Leaking seals, packing, or flanges may be tightened to eliminate leakage or
evaluated for continued service. Valves and pumps will be operated and
inspected as appropriate after any maintenance to ensure proper operation.
The containment spray nozzle availability is tested by blowing air through
the nozzles and observing the flow through the various nozzles in the
containment by use of thermography or other appropriate means.
System Testing
Permanent test lines for all containment spray loops are located so that the
system, up to the motor operated containment isolation valves, can be tested.
Permanent test line (Unit 3 and Unit 4) provisions are provided for the
containment spray pumps so that the system, up to the manual isolation valves
(*891A/B), can be tested to full containment spray pump flow. The manual
isolation valves and check valves can be checked separately for leakage
during local leak rate testing (LLRT).
The spray nozzles in containment are periodically verified to be unobstructed
by verification of air flow by the use of thermography or other appropriate
means.
Operational Sequence Testing
The functional test of the Safety Injection System described in Section 6.2
demonstrates proper transfer to the emergency diesel generator power source
in the event of loss of power. A test signal simulating the containment
spray signal will be used to demonstrate operation of the spray system up to
the manual isolation valves on the pump discharge line.
6.4-15 Revised 04/17/2013 6.
4.5 REFERENCES
- 1. NRC letter to FPL, dated September 20, 1991, Safety Evaluation of the Inservice Testing (IST) Program for Pumps and Valves - Turkey Point
Plant Units 3 and 4.
- 2. Westinghouse Letter, 94-JB-UP-5285, dated September 1, 1994.
(Attachment 4, Case 1)
- 3. Westinghouse Technical Report WCAP-17152, "Turkey Point Units 3 & 4 Extended Power Uprate Engineering Report," August 2012.
6.4-16 Revised 04/17/2013 TABLE 6.4-1
CONTAINMENT SPRAY SYSTEM-CODE REQUIREMENTS
Component Code
Valves USAS B16-5
Piping (including headers and
spray nozzles) USAS B31.1
TABLE 6.4-2
CONTAINMENT SPRAY PUMP DESIGN PARAMETERS
Quantity 2
Design pressure, discharge, psig 300
Design temperature, F 300
Design flow rate, gpm 1450
Design head, ft. 470
Shutoff head, ft. 617
Motor HP 250
Type Horizontal-Centrifugal TABLE 6.4-3 Sheet 1 of 2 SINGLE FAILURE ANALYSIS - CONTAINMENT SPRAY SYSTEM Component Malfunction Comments and Consequences A. Spray Nozzles Clogged Large number of nozzles renders clogging of a significant number of nozzles as incredible.
B. Pumps 1) Containment Spray Pump Fails to start Two provided. Containment peak pressure evaluation based on one pump operation for the duration of the event. In addition, two fan coolers are operating for the remainder of the transient.
- 2) Residual Heat Removal Pump Fails to start Two provided. Evaluation based on operation of one pump and two out of three containment cooling fans operating during long-term recirculation phase.
- 3) Intake Cooling water Pump Fails to start Three provided. operation of one pump during recirculation required.
- 4) Component Cooling Fails to start Three provided. Operation of one pump during recirculation required. C. Automatically operated Valves: (open on coincidence of 2/3
[Hi-Hi] containment pressure signals)
- 1) Containment spray pump discharge isolation valve Fails to open Two provided. operation of one required.
Revised 04/17/2013 TABLE 6.4-3 sheet 2 of 2 SINGLE FAILURE ANALYSIS - CONTAINMENT SPRAY SYSTEM Component Malfunction Comments and Consequences D. Valves Operated From Control Room
- 1) Containment sump recirculation isolation Fails to open Two lines in parallel, each with two valves in series. One line required.
- 2) Containment spray header isolation valve from heat exchangers residual Fails to open Two valves provided. Operation of one required.
- 3) Residual heat removal pump suction line Fails to open Two valves in series, one required to close.
Revised 04/17/2013 TABLE 6.4-4 Sheet 1 of 1 SODIUM TETRABORATE DECAHYDRATE (NaTB) BASKET DESIGN PARAMETERS
Quantity 2 Large 8 Small
Minimum total amount of NaTB 11061 lbm
Maximum total amount of NaTB 14264 lbm
Bulk density of NaTB 48.82 lbm/ft to 54.13 lbm/ft Small Size 3.0' x 3.0' x 2.5' Large Size 4.5' x 4.5' x 2.7'
Materials of Construction Stainless Steel Screen Mesh size (same for each size) Course - 4 Mesh - 0.047" wire cloth Small - 100 Mesh - 0.0045" wire cloth
Location Elevation 14 ft. in containment Maximum Height of Bottom Screen Large Basket 3.5 in. off floor Small Basket 4.5 in. off floor
Design Features:
- 1. Basket side and bottom panels are provided with an outer coarse stainless steel mesh screen providing support for a fine inner mesh screen minimizing any
leakage for material. Screen ensures that NaTB is quickly and completely
dissolved in the sump fluid.
- 2. Top of baskets are located below minimum Post-LOCA sump water level.
- 3. Baskets have adjustable legs providing bottom surface area for dissolution, protect NaTB from spills and provide adjustment for uneven floor.
- 4. Baskets are provided with stainless steel cover equipped with drip edge to prevent inadvertent dissolution of NaTB from spill and condensation.
- 5. The Baskets are freestanding structures and will not interact with walls or any SSCs.
Revised 01/25/2013
6.5 LEAKAGE
DETECTION AND PROVISIONS FOR THE REACTOR AND AUXILIARY COOLANT LOOP
6.5.1 LEAKAGE
DETECTION SYSTEMS
The leakage detection systems reveal the presence of significant leakage from
the reactor and auxiliary coolant loops.
DESIGN BASES
Monitoring Reactor Coolant Leakage
Criterion: Means shall be provided to detect significant uncontrolled leakage from the reactor coolant pressure boundary. (GDC 16)
Positive indications in the control room of leakage of coolant from the Reactor Coolant System to the containment are provided by equipment which permits
continuous monitoring of containment air activity. This equipment provides
indication of normal background which is indicative of a basic level of leakage
from primary systems and components. Any increase in the observed parameters is
an indication of change within the containment, and the equipment provided is
capable of monitoring this change. The basic design criterion is the detection of
deviations from normal containment environmental conditions including air
particulate activity, radiogas activity, and in addition, in the case of gross
leakage, the liquid inventory in the process systems and containment sump.
Monitoring Radioactivity Releases
Criterion: Means shall be provided for monitoring the containment atmosphere and the facility effluent discharge paths for radioactivity released from
normal operations, from anticipated transients, and from accident
conditions. An environmental monitoring program shall be maintained to
confirm that radioactivity releases to the environs of the plant have
not been excessive. (GDC 17)
6.5-1 The containment atmosphere, the ventilation exhausts from the Auxiliary Building, the component cooling loop liquid, the steam generator blowdown and the condenser
air ejector exhaust are monitored for radioactivity concentration.
Principles of Design
The principles for design of the leakage detection systems can be summarized as
follows:
- 1. Increased leakage could occur as the result of failure of pump seals, valve packing glands, flange gaskets or instrument connections. The maximum
leakage rate calculated for these types of failures is 50 gpm which would be
the anticipated flow rate of water through the pump seal if the seal failed
and the area between the shaft and housing were completely open.
- 2. The leakage detection systems shall not produce spurious annunciation from normal expected leakage rates but shall annunciate increasing leakage.
- 3. Increasing activity level shall be annunciated in the control room.
Operator action will be required to isolate the leak in the offending
system.
SYSTEMS DESIGN AND OPERATION
Various methods are used to detect leakage from either the reactor coolant or the
auxiliary system. Although described to some extent under each system
description, all methods are included here for completeness.
6.5-2
Reactor Coolant System During normal operation and anticipated reactor transients the following
methods are employed to detect leakage from the Reactor Coolant System. At
least one leakage detection system with a sensitivity capable of detecting a 1
gpm leak in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be operable (References 1 and 2). The sensitivities
of the containment air particulate monitors and containment radioactive gas
monitors have been demonstrated to meet the design and licensing bases
requirements applicable to Turkey Point for Reactor Coolant System Leakage
detection systems (Reference 3).
Containment Air Particulate Monitor
This channel takes continuous air samples from the containment atmosphere and
measures the air particulate radioactivity. The samples, drawn outside the
containment, are in a closed, sealed system and are monitored by a
scintillation counter-filter paper detector assembly. The filter paper
collects all particulate matter greater than 1 micron in size on its
constantly moving surface, which is viewed by a hermetically sealed
scintillation-photomultiplier combination. After passing through the gas
monitor, the samples are returned to the containment.
The filter paper has a 25-day minimum supply at normal speed. The filter
paper mechanism, an electromagnetic assembly which controls the filter paper
movement, is provided as an integral part of the detector unit.
The detector assembly is in a completely closed housing. The detector output
is amplified by a preamplifier and transmitted to the Radiation Monitoring
System cabinet. Lead shielding is provided to reduce the background radiation
level to where it does not interfere with the detector's sensitivity.
The activity is indicated on numerical displays and recorded by a multipoint
recorder. High-activity alarm indications are displayed on the Radiation
Monitoring cabinets. Local alarms provide operational status of supporting
equipment such as pumps, motors and flow and pressure controllers.
6.5-3 Revised 04/17/2013 The containment air particulate monitor is the most sensitive instrument of those available for detection of reactor coolant leakage into the containment.
This instrument must be capable of detecting particulate activity in
concentrations as low as 10
-9 uc/cc of (micro curies per cubic centimeter) containment air. The minimum required measuring range is 10
-9 to 10-6 uc/cc. The range of the instrument provided is approximately 10
-11 to 10-5 uc/cc (CO 60).
The sensitivity of the air particulate monitor to an increase in reactor coolant leak rate is dependent upon the magnitude of the normal baseline leakage into the containment. The sensitivity is greatest where base-line
leakage is low as has been demonstrated by the experience of Indian Point Unit No. 1, Yankee Rowe and Dresden Unit 1. Where containment air particulate activity is below the threshold of detectability, operation of the monitor with stationary filter paper would increase leak sensitivity to a few cubic centimeters per minute. Assuming a low background of containment air particulate radioactivity, a reactor coolant corrosion product radioactivity (Fe, Mn, Co, Cr) of 0.2 uc/cc (a value consistent with little or no fuel cladding leakage), and complete dispersion of the leaking radioactive solids into the containment air, the air particulate monitor is capable of detecting leaks as small as approximately 0.013 gpm (50 cc/minute) within twenty minutes after they occur. If only ten percent of the particulate activity is actually dispersed in the air, leakage rates of the order of 0.13 gpm (500 cc/minute)
are well within the detectable range.
For cases where base-line reactor coolant falls within the detectable limits of the air particulate monitor, the instrument can be adjusted to alarm on
leakage increases from two to five times the base-line value.
The containment air particulate monitor together with the other radiation
monitors mentioned in this Section are further described in Section 11.2.
6.5-4 Rev.13 10/96 Containment Radioactive Gas Monitor
This channel measures the gaseous beta radioactivity in the containment by taking the continuous air samples from the containment atmosphere, after they pass through the air particulate monitors, and drawing the samples through a
closed, sealed system to a gas monitor assembly.
Each sample is constantly mixed in the fixed, shielded volume, where it is viewed by a beta sensitive scintillation photomultiplier combination detector.
The sample is then returned to the containment.
The detector is in a completely enclosed housing containing a beta sensitive scintillating detector mounted in a constant gas volume container. Lead shielding is provided to reduce the background radiation level to a point where it does not interfere with the detector's sensitivity. A preamplifier
and impedance matching circuit are mounted at the detector.
The detector outputs are transmitted to the Radiation Monitoring System cabinets. The activity is indicated by numerical displays and recorded by a multipoint recorder. High-activity alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring System cabinets. Local alarms annunciate the supporting equipments' operational
status.
The containment radioactive gas monitor is inherently less sensitive (required
threshold at 10
-6 uc/cc) than the containment air particulate monitor, and would function in the event that significant reactor coolant gaseous activity exists from fuel cladding defects. The minimum required measuring range is
10-6 to 10-3 uc/cc. Assuming a reactor coolant activity of 0.3 uc/cc, the occurrence of a leak of two to four gpm would double the zero leakage background in less than an hour's time. In these circumstances this instrument is a useful backup to the air particulate monitor. The range of
the detector provided is approximately 10
-7 uc/cc to 10
-1 uc/cc (Xe 133).
6.5-5 Revised 11/16/2007 The containment air particulate and radioactive gas monitors have assemblies that are common to both channels. They are described as follows:
i) The flow assembly includes a pump unit and selector valves that provide
a representative sample (or a "clean" sample) to the detector.
ii) The pump unit consists of:
- 1. A pump to obtain the air sample.
- 2. A flowmeter to indicate the flow rate.
- 3. A flow control valve to provide flow adjustment.
- 4. A flow alarm assembly to provide low and high flow alarm signals.
iii) Selector valves are used to direct the desired sample to the detector
for monitoring and to allow flow when the channel is in maintenance or
"purging" condition.
iv) A temperature sensor and pressure sensor are used to protect the system from high sample stream temperatures and pressures. The output of either sensor is capable of isolating sample flow.
v) Purging is accomplished with a valve control arrangement whereby the
normal sample flow is blocked and the detector purged with a "clean"
sample. This facilitates detector calibration by establishing the
background level and aids in verifying sample activity level.
vi) The mass flow transmitter produces an analog signal based on sample stream flow. The microprocessor automatically regulates sample flow using the flow transmitter output signal as input to adjust the flow control valve. Visual indicators show sample flow rate, inlet vacuum, and sample pressure.
6.5-6 Rev.16 10/99 vii) A sample flow rate indicator is calibrated form 0 to 6.90 scfm.
Alarm lights are actuated by the following:
- 1. Flow alarm assembly (low and high flow)
- 2. The pressure/temperature sensor assembly (high pressure/temperature) 3. The filter paper sensor (paper drive malfunction)
- 4. Failure of any microprocessor controlled self test
Component Cooling Liquid Monitors
These channels continuously monitor the component cooling loops of the Auxiliary Coolant System for activity indicative of a leak of reactor coolant
from either the Reactor Coolant System, the recirculation loop, or the residual heat removal loop of the Auxiliary Coolant System. A scintillation counter is located in an in-line well in each component cooling pump suction header. Each detector assembly output is amplified by a preamplifier and transmitted to the Radiation Monitoring System cabinets in the control room.
The activity is indicated on a meter and recorded by a multipoint recorder.
High-activity alarm indications are displayed on the control board annunciator
in addition to the Radiation Monitoring System cabinets.
The measuring range of each monitor is 10.0 to 10 6 cpm.
Reactor Vessel Head Leakage Detection System
A sampling skid located in containment is capable of sampling and analyzing each CRDM Cooler Ventilation discharge and the containment atmosphere on an as-needed basis. The output from the analyzer is transmitted to the monitoring cabinet (3C367) located in the computer room. The monitoring cabinet is common to both units. An increasing trend in CRDM discharge radioactivity levels over containment background radioactivity levels indicates a Reactor Vessel Head Leak. The monitoring cabinet also contains
sample pump controls, sample valve status, and air filter unit control.
6.5-7 Rev. 16 10/99 Condenser Air Ejector Gas Monitors R-*-15, RAD-*-6417 R-*-15 monitors the discharge from the Steam Jet Air Ejector exhaust header for gaseous radiation which is indicative of a primary to secondary system leak. R-*-15 uses an inline beta scintillator to monitor a fixed volume sufficiently shielded to prevent background radiation from reducing maximum sensitivity. R-*-15 has a range of 1.0E-07 to 1.0E-01 uCi/cc for Kr-85. The effluent is routed to the atmosphere.
The R-*-15 detector output is transmitted to the Radiation Monitoring System cabinets in the control room. The activity is indicated by a meter and recorded by a multipoint recorder. High-activity alarm indications are displayed on the control board annunciator in addition to the Radiation Monitoring cabinets.
The Steam Jet Air Ejector effluent is also monitored by RAD-*-6417 SPINGs, which use a beta scintillation counter for low range noble gas and energy-compensated G-M detectors for medium and high range noble gas. SPINGs comply with Regulatory Guide 1.97 on high-range monitoring in response to post Three Mile Island initiatives. RAD-*-6417 has a range from 1.0E-07 to 1.0E+05 uCi/cc for Xe-133.
Steam Generator Liquid Sample Monitor
This channel monitors the liquid phase of the secondary side of the steam
generator for radiation, which would indicate a primary-to-secondary system
leak, providing backup information to that of the condenser air ejector gas
monitor. Samples from the bottom of each of the three steam generators are
mixed to a common header and the common sample is continuously monitored by a
scintillation counter and holdup tank assembly. Upon indication of a
high-radiation level (R-19), blowdown is automatically isolated. Each steam
generator is then manually sampled in order to determine the source of the
activity. This sampling sequence is achieved by manually obtaining steam
generator liquid samples at the primary sample sink for laboratory analysis
after allowing sufficient time for sample equilibrium to be established.
6.5-8 Revised 04/17/2013 A remote indicator panel, mounted at the detector location, indicates the radiation level and high-radiation alarm.
The measuring range of this monitor is 10.0 to 10 6 cpm.
A photomultiplier tube - scintillation crystal (NaI) combination, mounted in a hermetically sealed unit, is used to monitor liquid effluent activity. Lead shielding is provided to reduce the background level so it does not interfere with the detector's maximum sensitivity. The in-line, fixed-volume container
is an integral part of the detector unit.
During cold shutdown personnel can enter the containment and make a visual inspection for leaks. The location of any leak in the Reactor Coolant System would be determined by the presence of boric acid crystals near the leak. The leaking fluid transfers the boric acid crystals outside the Reactor Coolant
System and the process of evaporation leaves them behind.
If an accident involving gross leakage from the Reactor Coolant System
occurred it could be detected by the following methods.
Pump Activity
During normal operation only one charging pump is operating. If a gross loss of reactor coolant to another closed system occurred which was not detected by
the methods previously described, the speed of the charging pump would
indicate the leakage.
The leakage from the reactor coolant will cause a decrease in the pressurizer
liquid level that is within the sensitivity range of the pressurizer level indicator. The speed of the charging pump will automatically increase to try to maintain pressurizer programmed level. If the pump reaches a high speed
limit, an alarm is actuated.
6.5-9 Rev. 16 10/99 A break in the primary system would result in reactor coolant flowing into the containment sump. Gross leakage to this sump would be indicated by the frequency of operation of the containment sump pumps and by indication on the
containment sump level recorder.
Liquid Inventory
Gross leaks might be detected by unscheduled increases in the amount of reactor coolant makeup water which is required to maintain the normal level in the pressurizer. This is inherently a low precision measurement, since makeup
water is necessary as well for leaks from systems outside the containment.
A large tube side to shell side leak in the non-regenerative (letdown) heat exchanger would result in reactor coolant flowing into the component cooling water and a rise in the liquid level in the component cooling water head tank.
The operator would be alerted by a high water alarm for the head tank and high radiation and temperature alarms actuated by monitors at the component cooling
water pump suction header.
A high level alarm for the component cooling water head tank and high radiation and temperature alarms actuated by monitors at the component cooling pump suction header could also indicate a thermal barrier cooling coil rupture in a reactor coolant pump. However, in addition to these alarms, high
temperature and high flow on the component cooling outlet line from the pump would activate alarms. In the event of a thermal barrier rupture, high flow as sensed by FIC-626 will normally close MOV-*-626, which will terminate
leakage into the CCW system.
Gross leakage might also be indicated by a rise in the normal containment sump
level. High level in this sump will actuate an alarm.
The residual heat removal loop removes residual and sensible heat from the core and reduces the temperature of the Reactor Coolant System during the
second phase of shutdown.
6.5-10 Rev. 16 10/99 Leakage from the residual heat removal loop during normal operation would be detected by the component cooling loop radiation monitor (see analysis of
detection of leakage from the Reactor Coolant System in this section).
The physical layout of the two residual heat removal pumps will be within separate shielded and isolated rooms outside of the containment. This will permit the detection of a leaking residual heat removal pump by means of the radiation monitor located in the plant vent. Supplemental radiation monitoring will be provided by the plant vent gas monitoring system. Alarms
in the control room will alert the operator when the activity exceeds a preset level. Small leaks to the environment could be detected with these systems
within a short time after they occurred.
Should a large tube side to shell side leak develop in a residual heat exchanger or the seal water heat exchanger of a residual heat removal pump, the water level in the component cooling water head tank would change, rising or falling dependent upon specific operating conditions of the interfacing systems. The operator would be alerted by a high or low water level alarm.
The high water level alarm would also isolate RCV-*-609 to prevent uncontrolled overflow into the Waste Holdup Tank (WHT). Maximum CCW system pressure would then be controlled by RV-*-707, which would relieve to the WHT, as required. Radiation and temperature monitors at the component cooling
water pump suction header may also signal an alarm.
Leakage from the residual heat removal pumps is drained to separate sumps equipped with a sump pump. The failure of either sump pump or leakage greater than sump pump capacity will be indicated by annunciation in the control room as a means of detection of gross leakage, i.e., a seal failure, from a
residual heat removal pump.
Component Cooling Loop
Leakage from the component cooling loop inside the reactor containment will be
detected by level change in the containment sump or by decrease in component
cooling water head tank level via low level alarm.
Visual inspection inside the containment is possible during cold shutdown.
6.5-11 Rev. 16 10/99 Gross leakage from the component cooling loop would be indicated inside the containment by a rise in the liquid level of the containment sump. This sump
has a high level recorder and a high level alarm.
If the leakage is from a part of the component cooling loop outside the
containment, it would be detected by a drop in the head tank level below the
low level alarm setpoint and confirmed by the level in the sight glass
installed on the affected tank. If an external leak occurs in an area
provided with sump pumps which have control room indication for their
operation, the leak can be located by the indication of the associated sump
pump operation; otherwise the leak will be located by visual inspection of the
system. For an internal leak, e.g., from the CCW to any other system, the
leak may be located by sequential isolation of inspection of equipment in the
affected loop.
The Reactor Vessel Head Area Leakage Detection System supplements the
monitoring performed via primary side monitoring systems and containment area
radiation monitoring to provide detection of any leaks in the area of the
reactor head.
The system monitors samples drawn from the CRDM cooler discharge as well as
from the containment atmosphere for a reference. The system is designed to
detect small RCS leaks by monitoring particulate radiation levels in the
reactor vessel head area by means of these samples.
It includes a sampling skid located within the containment and a remote
control and display rack in the computer room. The sample pump delivers a
sample through a slow moving filter set-up. Particulates emitted are trapped
on the filter paper and are monitored by a beta scintillation detector. The
beta detector provides a digital signal to the Universal Digital Ratemeter
located inside in the remote control and display rack in the computer room.
6.5.2 LEAKAGE
PROVISIONS
Provisions are made for the isolation or containment of any leakage.
6.5-12 Rev. 16 10/99 DESIGN BASIS
The design provisions prevent uncontrolled leakage of reactor coolant or
auxiliary cooling water. This is accomplished by (1) isolation of the leak by
valves, (2) designing relief valves to accept the maximum flow rate of water
from the worst possible leak, (3) supplying redundant equipment which allows a
standby component to be placed in operation while the leaking component is
repaired and (4) routing the leakage to various sumps and holdup tanks.
DESIGN AND OPERATION
Various provisions for leakage avert uncontrolled leakage from the reactor and
auxiliary coolant loops.
When significant leakage from the Reactor Coolant System is detected, action
is taken to prevent the release of radioactivity to the atmosphere outside the
containment.
6.5-12a Rev 7 7/89 If either the containment air particulate beta activity or the radioactive gas activity exceed pre-set levels on the containment air particulate and
radioactive gas monitors, respectively, the containment purge supply and
exhaust duct valves and instrument air bleed valves are closed.
A high radiation alarm actuated by the steam generator liquid sample monitor
initiates closure of the isolation valves in the blowdown lines and sample
lines.
If the component cooling loop radiation monitor signals a high radiation
alarm, the valve in the component cooling water head tank vent line
automatically closes to prevent gaseous activity release.
If a leak from the Reactor Coolant System to the component cooling loop was a
gross leak or if the leak could not be isolated from the component cooling
loop before the inflow completely filled head tank, the relief valve on the
surge tank would lift. The discharge from this valve is routed to the waste
holdup tank in the Auxiliary Building.
A large leak in the Reactor Coolant System pressure boundary, which does not
flow into another closed loop, would result in reactor coolant flowing into
the containment sump.
High containment air particulate beta activity or high radioactive gas
activity will result in an alarm being activated by either the containment air
particulate or radioactive gas monitors, respectively. The containment purge
supply and exhaust duct valves and pressure relief line valves are closed.
This prevents the release of radioactivity to the atmosphere outside the
containment.
6.5-13 Revised 11/16/2007 If leakage from the residual heat removal loop into the component cooling loop occurs, the component cooling radiation monitor will actuate an alarm and the
valve in the component cooling water head tank vent line is automatically
closed to prevent gaseous radioactivity release. If the leaking component (i.e. , a residual heat exchanger) could not be isolated from the component
cooling loop before the inflow completely filled the head tank, the relief
valve on the surge tank would lift and the effluent would be discharged to the
auxiliary building waste holdup tank.
Gross leakage from the section of the residual heat removal loop inside the
containment, which does not flow into another closed loop, would result in
reactor coolant flowing into the containment sump.
Other leakage provisions for the residual heat removal loop are discussed is
Section 9.3.
Component Cooling Loop
Gross leakage from the section of the component cooling loop inside the
containment which does not flow into another closed loop will flow into the
containment sump. Outside the containment major leakage would be drained to
the auxiliary building sump. From here it is pumped to the waste holdup tank.
The location of the CCW head tank and associated piping prevents the auxiliary
building drain system from collecting any leakage from this portion of the CCW
system. Leakage from that tank would not be captured; however, the capacity
of the head tank is limited (300 gal) and would not constitute a major source
of leakage.
Other provisions made for leakage from the component cooling loop are
discussed in Section 9.3.
6.5-14 Revised 04/17/2013
6.5.3 References
- 1. FPL Letter L-88-481, "Application of Leak Before Break Technology to Primary Coolant System Piping", dated November 1, 1988.
- 2. NRC Letter dated November 28, 1988, "Turkey Point Units 3 and 4 Generic Letter 84-04, Asymmetric LOCA Loads".
- 3. Calculation PTN-BFJN-96-001, "Calculation Related to R-11/R-12 response Time to a 1 GPM Reactor coolant System Leak," Revision 0, 10/17/1996.
6.5-15 Revised 04/17/2013
6.6 CONTAINMENT
ISOLATION
6.6.1 DESIGN
BASIS System piping that penetrates the containment leakage limiting boundary is designed to maintain or establish isolation of the containment from the outside environment under the following postulated conditions:
(a) Any accident for which isolation is required and, (b) A coincident independent single failure or malfunction (expected fault condition) occurring in any active system component within the isolated bounds. The following General Design Criteria (GDC) must be satisfied by the containment penetrations:
Criterion: "Penetrations that require closure for the containment functions shall be protected by redundant valving and associated apparatus." (1967 Proposed GDC 53)
Following the events at Three Mile Island, this criterion was provided with the added clarification that the redundant barriers must not require operator action in order to effect isolation.
Piping penetrating the containment was designed for pressures at least equal to the containment design pressure. Isolation capability for fluid system lines penetrating the containment generally involves at least two redundant barriers against leakage of radioactive fluids to the outside environment.
Such releases might be due to rupture of a line within the containment concurrent with a LOCA, or due to rupture of a line within the containment which connects to a source of radioactive fluid within the containment.
These barriers, in the form of isolation valves or closed systems, are defined on an individual penetration basis. In addition to satisfying the following containment isolation criteria, valving is designed to facilitate normal operation and maintenance of the systems and to ensure reliable operation of
6.6-1 Rev. 12 5/95 other engineered safeguards systems. Table 6.6-1 identifies the barriers for each penetration. Normally open manual valves which may be in the penetration pressure boundary are not isolation barriers. Vents, drains and test connections are considered to be extensions of the penetration pressure boundary and are administratively controlled. Valves in these lines are not considered containment isolation valves, and they have not been included in the table.
6.6.2 SYSTEM
DESIGN NUREG 0578 established the penetration classifications of "essential" and "non-essential". Essential lines are those fluid system process lines that are required to mitigate or monitor an accident, or those which, if unavailable, could increase the magnitude of the accident. Essential lines may not be isolated in response to containment isolation signals; they may even be opened to ensure the required process function is met. Non-essential lines include all other fluid system penetrations. There are several penetrations for which neither classification is applicable; these include the hatches and airlocks as well as the spare penetrations.
The barriers for each penetration are identified in Table 6.6-1. Further discussion of the barrier design is provided below in Subsection 6.6.2.1 for all barriers, and Subsection 6.6.2.2 for valving.
6.6.2.1 BARRIERS The containment isolation features at Turkey Point may include valving, closed systems, caps, and blind flanges. Each of these devices is considered to be an effective barrier against leakage of the containment atmosphere. In some cases a single integral barrier has been credited as the means of isolation. In other cases, redundant barriers consisting of valves, closed systems or other closure devices are utilized. Subsection 6.6.2.2 discusses design requirements applicable to valving. This subsection discusses the design requirements for closed systems, integral barriers and other closure devices.
6.6-2 Rev. 12 5/95 A closed system, for containment isolation purposes, shall meet the following requirements:
(a) Withstand temperature and pressure equal to the containment design conditions; (b) Meet Seismic Category I design requirements; (c) Be protected against overpressure from thermal expansion of contained fluid when isolated; (d) Be protected against high energy line breaks and missiles unless it can be demonstrated that these hazards will not result in the need for containment isolation; and (e) Be capable of being leak tested, unless system integrity is being maintained during normal operation at a pressure equal to or greater than the containment design pressure (closed systems outside containment).
In addition, a closed system inside containment must not communicate with the primary coolant or the containment atmosphere; a closed system outside containment must not communicate with the outside atmosphere.
Branch lines located within the boundary of a closed system barrier must meet similar criteria out to the valve or barrier which provides isolation. These branch lines, which may be vents, drains, test connections, or intersystem connections are generally provided with valves which are subject to administrative controls to assure they support the containment function.
Main Steam, Steam Generator blowdown, feedwater and engineered safeguards lines penetrating the containment have been designed to assure that they are capable of withstanding the maximum hypothetical earthquake.
Penetrations which are provided with a pipe cap as a barrier are considered to be extensions of the containment liner. The cap, located either inside or outside containment, forms a single integral barrier against leakage. The
6.6-3 Rev. 12 5/95 caps may be welded or threaded; if threaded, the cap will be administratively controlled. Threaded caps will also be subject to local leak rate testing to provide continued assurance of their integrity.
A special example of the use of a single integral barrier involves the containment pressure monitoring instrumentation (Penetrations 46 (Unit 4) and 62 (Unit 3)). These penetrations provide a sensing line into containment for the pressure instrumentation used by the Engineered Safeguards Features Actuation System (ESFAS). This is an essential function which is required to ensure initiation of containment isolation and actuation of safety injection.
The leakage barrier includes the piping and tubing as well as the instrument pressure boundary. The instruments have been designed to monitor post-accident conditions. This single barrier is considered to be appropriate considering the required function provided by the monitors.
Flanges, when used as a penetration boundary at Turkey Point, rely on their gasketing as the barrier against leakage. Redundant barriers have been identified for such penetrations. These include, for example, double gaskets such as those used on the fuel transfer tube or two barriers where a single gasket flange is used. The containment personnel access and emergency escape airlocks, although provided with redundant seal rings at each door, also have an equalization line with a single valve through each side. For this reason, their redundant isolation barriers are identified as the door and valve on each side, rather than just the redundant seal rings on one door. Mechanical interlocks have been provided to assure that one door is closed at all times when containment integrity is required. The space between the seal rings may be tested to provide assurance that the door is properly sealed.
6.6.2.2 CONTAINMENT ISOLATION VALVES Containment isolation valves are provided with actuation and control equipment appropriate to the valve type. For example, air operated globe and diaphragm valves are generally provided with air diaphragm operators, with fail-safe operation ensured by redundant control devices in the instrument air supply to the valves. Motor operated gate valves are capable of being supplied from reliable on-site emergency power as well as their normal power source. Check
6.6-4 Rev. 12 5/95 valves and relief valves are considered to be automatic isolation valves but do not require actuation by control systems. Manual and remote manual valves which do not receive an automatic actuation signal are administratively controlled to preserve containment integrity.
Containment isolation "trip" valves are actuated to the closed position by the containment isolation signal, derived automatically from the signals as listed in Table 6.6-1. Non-automatic isolation valves, i.e., remote stop valves and manual valves, are used in lines which are normally closed.
Administrative controls are used to ensure they are in, or can be placed in, their required position in the event of an accident.
The objective of establishing valve closure times is to limit the release of radioactivity from containment. Fast closure times are typically required for valves in lines which provide direct communication between the containment and the outside atmosphere (i.e., containment purge and vent lines; see below). The stroke time of selected containment isolation valves have been evaluated to demonstrate they are capable of meeting their design objectives. The nominal stroke times of other automatic containment isolation valves, for which component specific analyses have not been performed, meet the guidance of the Standard Review Plan, Section 6.2.4. The capability of the valves to meet the required closure times is verified periodically during performance of the In-Service Testing (IST) Program.
Operability requirements for these automatic isolation valves are as provided in the Technical Specifications.
The large tight sealing butterfly valves used to isolate the containment purge ventilation ducts are equipped with air-cylinder operators, with spring returns. These valves fail to the closed position on loss of control signal or instrument air. The maximum closure time for these valves is 5 seconds from receipt of an automatic isolation signal. As discussed in Subsection 9.8.2, the opening angle of the valves has been limited. These valves can be closed either manually or automatically upon a signal of high radioactivity level in the containment or by the containment isolation signal. These valves are normally closed during reactor power operation. Refer to Table 6.6-2 for a single failure analysis of the containment purge valves. Debris screens are present upstream (with respect to Post LOCA flow) of the supply and exhaust
6.6-5 Rev. 12 5/95 purge valves. These debris screens will protect the containment isolation valves from debris that might follow as a result of a LOCA.
Isolation valves located inside containment are subject to the high pressure, high temperature, steam laden atmosphere resulting from an accident as well as a possible earthquake. Operability of these valves in the accident environment is ensured by proper design, construction and installation, as reflected by the following considerations:
(a) All components in the valve installation, including valve bodies, trim and moving parts, actuators, instrument air lines and control and power wiring, are constructed of materials sufficiently temperature resistant to be unaffected by the accident environment. Special attention is given to electrical insulation, air operator diaphragms and stem packing material.
(b) In addition to normal pressures, the valves are designed to withstand maximum pressure differentials in the reverse direction imposed by the accident conditions. This criterion is particularly applicable to the butterfly type isolation valves used in the containment purge lines.
(c) Valves are located in a manner to reduce the accelerations on the valves. Piping spans have been designed to adequately accommodate the loads to which the span would be subjected. Valves are mounted in the position recommended by the manufacturer.
(d) Earthquake forces on the operating parts of the valve are calculated to be small compared to the other forces present in the piping system.
(e) Control cables to the valve operators are designed and installed to assure that the flexure of the line does not endanger the control system. Appendages to the valve, such as position indicators and operators, are designed for structural adequacy.
Status monitoring of plant conditions, including containment integrity, during and following an accident, has been provided in accordance with the
6.6-6 Rev. 12 5/95 requirements of Regulatory Guide 1.97, Revision 3. This guide recommended that valve position indication be provided to permit the operators to ascertain the status of containment integrity and main steam safety valve positions. The valves which have been provided with position indication to comply with the Regulatory Guide 1.97 commitments are included in Table 6.6.1. 6.6.2.3 ISOLATION ACTUATION Phase A Isolation is initiated by a safety injection (SI) signal. Phase B Isolation is initiated by a high containment pressure coincident with high-high containment pressure. Containment Ventilation Isolation is initiated by a safety injection signal or high containment radioactivity (R-11 or R-12) signals. Additional information is provided in Table 7.2-1.
Table 6.6-1 provides a listing of isolation valves and their actuating signals. Automatic closure of the valves listed in Table 6.6-1 will achieve containment isolation in accordance with the requirements of the 1967 Proposed GDC 53 and the plant Technical Specifications. The listed valves are considered to be part of the isolation barrier capability even if they open in response to accident conditions. That is, selected valves in essential penetration flow paths may receive a signal to open (e.g., SI signal or Auxiliary Feedwater Actuation signal). Barriers that provide a containment integrity function, but which do not receive an isolation signal (e.g., manual valves), have also been included in Table 6.6-1.
6.6.3 TESTING
AND INSPECTION The Turkey Point Units 3 and 4 containment structure was designed such that the maximum allowable containment vessel leakage rate shall not exceed 0.25%
per day of containment free volume at the conditions of a Maximum Hypothetical Accident (MHA), that is, 49.9 psig and 276 o F. 10 CFR 50, Appendix J requires pre-operational and periodic operational verification by testing of the leak-tight integrity of the containment structure and lines penetrating containment. The testing consists of Type A, Type B, and Type C tests. Testing performed to meet 10 CFR 50, Appendix J provides assurance that leakage through the primary reactor containment and systems penetrating containment does not exceed this maximum allowable leakage rate. The
6.6-7 Rev. 12 5/95 combined leakage from all penetrations subject to Type B and C tests must be less than 0.6 times the maximum allowable leakage rate (La) prior to
increasing the Reactor Coolant System temperature above 200 o F.
Type A tests are the Integrated Leak Rate Tests (ILRT), which measure the
containment structure overall integrated leak rate. FSAR Section 5.1.7
describes the pre-operational ILRTs. Integrated leak rate tests are
performed periodically over the operating life of the plant.
Type B Local Leak Rate Tests (LLRT) detect local leaks across pressure-
containing or leakage limiting boundaries other than valves. Containment
penetrations that are required to be Type B tested include:
(a) Personnel and Emergency Air locks
(b) Equipment access hatch
(c) Fuel transfer tube flange
(d) Electrical penetrations
(e) Other containment penetrations whose design incorporates degradable mechanical parts such as resilient seals, gaskets or sealant compounds.
Type C Local Leak Rate tests measure containment isolation valve leakage
rates. Containment isolation valves in the following categories (from
10 CFR 50, Appendix J) shall be leak rate tested in accordance with the Type
C test requirements:
(a) Valves located in lines that provide a direct connection between the inside and outside atmospheres of the primary reactor containment under
normal operation, such as purge, ventilation, vacuum relief, and
instrumentation valves;
(b) Valves required to close automatically upon receipt of a containment isolation signal in response to controls intended to effect containment
isolation; and
6.6-8 Rev. 15 4/98 (c) Valves required to operate intermittently under post accident conditions.
In general, Type C tests are performed by pressurizing the valves from the
side facing the inside of containment. However, there are configurations in
which such testing is not practical. Alternative test configurations have
been reviewed to ensure that the reverse testing is equivalent to the
preferred test direction. In particular, the review must consider any
potential leak paths (such as valve packing or vents, drains, or test
connections) which would not be detected by the reverse flow test.
Periodic local leak rate testing and pre-operational leak rate testing
discussed above describe a reasonable approach to assuring that the
containment leakage is maintained below design limits during the life of the
plant. Periodic local leak testing provides an accurate method of monitoring
changes in the leakage characteristics of the containment. An integrated
leak rate test (ILRT) is performed less frequently. An ILRT is also
performed if major maintenance or modification to the containment was made.
Periodic testing is conducted on a performance basis as determined by the
Containment Leakage Rate Testing Program, which is in compliance with the
requirements of Appendix J of 10 CFR Part 50, Option B (as modified by
approved exemptions), and consistent with the guidance of Regulatory Guide
1.163, dated September 1995, in accordance with the plant Technical
Specifications. The testing program requirements, including test intervals, are based on both overall and individual component leakage rate performance.
Automatic and power-operated containment isolation valve closure times shall
be verified periodically by testing in accordance with the inservice test
requirements of ASME Section XI.
6.6-9 Rev. 15 4/98 TABLE 6.6-1 Sheet 1 of 18 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BARRIERS
TABLE LEGEND
SYSTEMS
AFWS Auxiliary Feedwater System BA Breathing Air System CSS Containment Spray System CVCS Chemical and Volume Control System CCWS Component Cooling Water System IAS Instrument Air System FW Feedwater System MS Main Steam PACVS Post Accident Containment Ventilation System PAHMS Post Accident Hydrogen Monitoring System RCS Reactor Coolant System RHRS Residual Heat Removal System SGBD Steam Generator Blowdown System SIS Safety Injection System SS Sample System Vent Containment Ventilation System WDS Waste Disposal System
ABBREVIATIONS Cntmt Containment Constr Construction CSIC Closed System Inside Containment CSOC Closed System Outside Containment Exch Exchanger ILRT Integrated Leak Rate Test LC Locked Closed LLRT Local Leak Rate Test NA Not Applicable PCV Pressure Control Valve PRT Pressurizer Relief Tank RCDT Reactor Coolant Drain Tank RCP Reactor Coolant Pump RM Remote Manual SG Steam Generator SI Safety Injection
FLOW DIRECTIONS
In Indicates that the predominate flow direction is into containment during normal operation.
Out Indicates that the predominate flow direction is out of containment during normal
operation.
DEFINITIONS Normal Position Expected position of valve during plant operation. Actual valve position may vary Depending on specific plant Conditions.
Rev. 16 10/99 TABLE 6.6-1 Sheet 2 of 18 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BARRIERS
TABLE LEGEND (Continued)
METHODS OF ACTUATION
Auto Self actuating valve (Check, PCV, Relief)
PWR Power operated valve (Air, MOV, Solenoid)
Manual Handwheel (or reach rod) operated valve
SIGNALS (Refer to Table 7.2-1 for additional information)
Phase A Containment Isolation Signal initiated by a Safety Injection Signal (Also known as a T-
signal) - closes non-essential automatic
containment valves.
Phase B Containment Isolation Signal initiated by containment pressure (Also known as a P-
signal) - closes selected automatic isolation
valves which are beneficial to remain open
following a Phase A signal. See Table 7.2-1
for cause of actuation.
AFAS Auxiliary Feedwater Actuation Signal - opens essential steam flow to the AFW pump turbine
and actuates auxiliary feedwater. See Table
7.2-1 for cause of actuation.
CVS Containment Ventilation System Isolation initiated by Safety Injection Signal or
Containment High Radiation Signal - isolates
non-essential containment ventilation lines.
FWIS Main Feedwater Isolation Signal - closes the main feedwater control and bypass valves and
the backup feedwater isolation valves. See
Table 7.2-1 for cause of actuation.
MSIS Main Steam Isolation Signal - isolates non-essential portions of the main steam lines.
See Table 7.2-1 for cause of actuation.
RM Remote Manual (Power operated)
SIS Safety Injection Signal - initiates Phase A isolation and actuates safety injection.
Revised 04/17/2013
TABLE 6.6-1 Sheet 17 of 18 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BARRIERS TABLE NOTES
- 1. Inside secondary shield. Open for shutdown cooling.
- 2. Normal residual heat removal loop flow path. Missile protected - between secondary shield and containment. Low head safety injection flow path, into 3 loops. Valve opens and remains open for normal safety injection.
- 3. With valve switch in auto during normal operation, valve auto opens on gas analyzer signal.
- 4. Charging line may provide supplemental safety injection flow. First valve inside containment is outside shield wall in a separate shielded area. 5. Redundant (Train A & Train B) solenoid operators are provided for POV-2604, 2605, 2606.
- 6. Indication is a demand signal.
- 7. Valves may be manually opened for post-accident hydrogen monitor operation.
- 8. Valves may be opened for hot leg recirculation.
- 9. Valve positions shown in this table reflect those expected during normal or post-accident operation without operator intervention. Position other than that shown may be acceptable if performed under appropriate control or authorization (e.g., see Notes 7, 8, or 13). Position for stop-check valves reflect disk position and not necessarily handwheel position.
- 10. Administratively controlled; breakers locked open.
- 11. CV-855 may be opened to pressurize SI Accumulators with fill N 2 and during flux mapping operations.
- 12. Pressure transmitter located outside of containment; boundary includes degradable materials so this penetration must be tested.
- 13. Valves may be opened for sump recirculation.
- 14. Valves open based on emergency containment fan start (which is on SI signal). 15. Valves CV-*-2816, / 2817 / 2818 / 2831 / 2832 / 2833 are normally closed. They automatically open when any of the AFW steam supply valves (MOV-*-1403, 1404, 1405) on corresponding unit open. These valves may remain open if the AFW pump is operating and the differential pressure across the valve is above approximately 900 psig.
Rev. 17 TABLE 6.6-1 Sheet 18 of 18 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BARRIERS TABLE NOTES (Continued)
- 16. Deleted
- 17. Valve located in main steam sample line. Post-accident position evaluated and found acceptable.
- 18. This valve is a containment isolation valve (i.e., closes on a Phase A, Phase B, or Containment Ventilation Isolation signal; is a normally closed manual isolation valve; or is a non-essential system check valve).
- 19. This is the primary barrier relied upon to support containment integrity.
- 20. This barrier is part of a containment airlock.
- 21. This barrier is associated with a containment purge supply or exhaust penetration.
- 22. This barrier is associated with an essential penetration. Its system safety function overrides the containment isolation safety function.
- 23. These valves are secondary system barriers. Where applicable, they are subject to secondary system/component requirements. Testing and leak tightness of these valves are not required for the LOCA events for which containment isolation is needed.
- 24. This penetration designation is actually a series of electrical penetrations. Each consists of active electrical canisters and spare canister locations, which are considered integral containment barriers.
- 25. Valves CV-*-2906, CV-*-2907, and CV-*-2908 fail open on loss of instrument air and fail closed on loss of electrical power. Valves CV-*-2810, CV-*-2812, and CV-*-2814 fail closed on loss of instrument air and fail open on loss of electrical power.
- 26. Valves CV-*-6275A, CV-*-6275B, and CV-*-6275C "fail-as-is" on loss of instrument air and "fail closed" on loss of electrical power.
Revised 7/28/2016 TABLE 6.6-2 SINGLE FAILURE ANALYSIS - CONTAINMENT PURGE VALVES
COMPONENT FAILURE MODE RESULTS
Valve Fails to close or fails 2nd purge air valve in Operator to seat or signal to close series will provide the not received. required isolation.
Instrument Air Failure or air. Purge air valves are closed Supply to the by spring, air is NOT Operator required for closure.
Solenoid Valve Fails to operate The 2nd purge air valve (i.e., does NOT in series will provide isolate the operator the required isolation.
cylinder from air supply and does NOT provide air bleed off for the cylinder).
Rev. 15 4/98 TABLE 6.6-3 Sheet 1 of 2 AUTOMATIC CONTAINMENT ISOLATION VALVE CLOSURE TIMES
A. PHASE "A" ISOLATION VALVES
VALVE NUMBER PENET SYSTEM/FUNCTION CONTROL ROOM CLOSURE TIME (Note 1) NUMBER LABEL (SECONDS)
CV-*-200A 14 CVCS Normal Letdown L/D Orifice Note 2 CV-*-200B 14 CVCS Normal Letdown L/D Orifice Note 2 CV-*-200C 14 CVCS Normal Letdown L/D Orifice Note 2 CV-*-204 14 CVCS Normal Letdown L/D Isol Note 2 MOV-*-381 25 CVCS Excess Letdown/ Seal Rtn Isol Note 2 RCP Seal Water Return CV-*-516 5 PRT Gas Analyzer Sample PRT to GA Note 2 CV-*-519A 7 PRT Makeup Primary Water PW to Cntmt Note 2 Supply CV-*-739 13 CCW Return from Excess Letdown Exc L/D Hx Notes 3,5 Heat Exchanger CV-*-855 42 Nitrogen Supply to Accumulators N2 to Accum Note 2 CV-*-956A 8 Pressurizer Steam Space Sample Przr Sample Note 2 CV-*-956B 9 Pressurizer Liquid Space Sample Przr Sample Note 2 CV-*-956D 55 Accumulator Sample Accum Sample Note 2 MOV-*-1417 21 CCW Supply to Normal Contain- NCC Inlet Notes 2,5 ment Coolers MOV-*-1418 22 CCW Return from Normal Contain- NCC Outlet Notes 2,5 ment Coolers CV-*-2821 23 Containment Sump Discharge Sump Isol Note 2 CV-*-2822 23 Containment Sump Discharge Sump Isol Note 2 SV-*-2911 33 Containment Air Sample PRMS Isol Note 2 SV-*-2912 32 Containment Air Sample Return PRMS Isol Note 2 SV-*-2913 33 Containment Air Sample PRMS Isol Note 2 CV-*-4658A 10 RC Drain Tank Vent RCDT Vent Note 2 CV-*-4658B 10 RC Drain Tank Vent RCDT Vent Note 2 CV-*-4659A 31 RC Drain Tank Gas Analyzer RCDT to GA Note 2 Sample CV-*-4659B 31 RC Drain Tank Gas Analyzer RCDT to GA Note 2 Sample CV-*-4668A 52 RC Drain Tank Pump Discharge RCDT Disch Note 2 CV-*-4668B 52 RC Drain Tank Pump Discharge RCDT Disch Note 2 SV-*-6385 5 PRT Gas Analyzer PRT to GA (7) Note 2 MOV-*-6386 25 CVCS Excess Letdown/RCP Seal Seal Rtn Isol Note 2 Water Return SV-*-6428 20 RCS Hot Leg Sample RCS Sample Note 2
Revised 07/28/2016
TABLE 6.6-3 Sheet 2 of 2 AUTOMATIC CONTAINMENT ISOLATION VALVE CLOSURE TIMES B. PHASE "B" ISOLATION VALVES VALVE NUMBER PENET SYSTEM/FUNCTION CONTROL ROOM CLOSURE TIME (Note 1) NUMBER LABEL (SECONDS)
MOV-*-626 43 CCW Return from RCP Thermal RCP Clg Wtr Notes 2,5 Barriers CLRS
MOV-*-716B 3 CCW Supply to RCPs RCP Clg Wtr Notes 2,5 MOV-*-730 4 CCW Return from RCP Oil Coolers RCP Clg Wtr Notes 2,5 C. CONTAINMENT VENTILATION ISOLATION VALVES (8) VALVE NUMBER PENET SYSTEM/FUNCTION CONTROL ROOM CLOSURE TIME (Note 1) NUMBER LABEL (SECONDS)
POV-*-2600 35 Containment Purge Supply Supply Iso 5 POV-*-2601 35 Containment Purge supply Supply Iso 5 POV-*-2602 36 Containment Purge Exhaust Exh Iso 5 POV-*-2603 36 Containment Purge Exhaust Exh Iso 5 CV-*-2819 63 Instrument Air Bleed Air Bleed Note 2 CV-*-2826 63 Instrument Air Bleed Air Bleed Note 2 Notes: 1. * = Unit number (3 or 4)
- 2. The isolation times of each automatic valve shall be within the limits established for testing in accordance with Section XI of ASME Boiler and Pressure Vessel code and applicable Addenda as required by 10 CFR 50.55a(g),
except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
- 3. Testing requirements are per letter L-89-358 from J. H. Goldberg to U. S. Nuclear Regulatory Commission dated October 3, 1989.
- 4. Deleted
- 5. Valve is not subject to Type C local leak rate testing of 10 CFR 50, Appendix J.
- 6. Deleted
- 7. The PRT to GA is labeled CV-*-6385
- 8. These valves are mounted in the Phase "A" Isolation Valves matrix.
Revised 07/28/2016
Revised 04/17/2013
6.7 ENVIRONMENTAL
QUALIFICATION TESTING OF SAFEGUARD COMPONENTS
6.7.1 GENERAL
The environmental safeguard components that are required for the mitigation
of a loss of coolant or a High Energy Line break (inside and outside
containment) accident and for the post-accident monitoring, are
described in Appendix 8A of the FSAR
6.7-1 Rev. 16 10/99