L-2016-198, Redacted Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 4, Reactor Coolant System
Text
TABLE OF CONTENTS
Section Title Page 4.0 REACTOR COOLANT SYSTEM
4.1 Design Bases 4.1-1 4.1.1 Performance Objectives 4.1-1 4.1.2 General Design Criteria 4.1-2 Quality Standards 4.1-2 Performance Standards 4.1-3 Records Requirements 4.1-4 Missile Protection 4.1-4 4.1.3 Principal Design Criteria 4.1-5 Reactor Coolant Pressure Boundary 4.1-5 Monitoring Reactor Coolant Leakage 4.1-6 Reactor Coolant Pressure Boundary Capability 4.1-6 Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention 4.1-8 Reactor Coolant Pressure Boundary Surveillance 4.1-9
4.1.4 Design Characteristics 4.1-10 Design Pressure 4.1-10 Design Temperature 4.1-11 Seismic Loads 4.1-11 4.1.5 Cyclic Loads 4.1-12 4.1.6 Service Life 4.1-13 4.1.7 Codes and Classifications 4.1-14 4.1.8 References 4.1-15
4.2 System Design and Operation 4.2-1 4.2.1 General Description 4.2-1 4.2.2 Components 4.2-2 Reactor Vessel 4.2-2 Reactor Vessel Support Structure 4.2-3 Pressurizer 4.2-3 Steam Generators 4.2-5 Steam Generator Support Structure 4.2-6 Reactor Coolant Pumps 4.2-6 Pump Support Structure 4.2-10 Pressurizer Relief Tank 4.2-10 Piping 4.2-11 Valves 4.2-13 4.2.3 Pressure-Relieving Devices 4.2-13 4.2.4 Protection Against Proliferation of Dynamic Effects 4.2-14 4.2.5 Materials of Construction 4.2-14a 4.2.6 Maximum Heating and Cooling Rates 4.2-18 4.2.7 Leakage 4.2-18 Leakage Prevention 4.2-19 Locating Leaks 4.2-19a
4-i Rev. 17 TABLE OF CONTENTS
Section Title Page 4.2.8 Water Chemistry 4.2.20 4.2.9 Reactor Coolant Flow Measurements 4.2-20 4.2.10 Reactor Coolant Subcooled Margin Monitor 4.2.21 4.2.11 Reactor Coolant Vent System 4.2-22 4.2.12 Reactor Vessel Drain Down Level Indication System 4.2-22 4.2.13 References 4.2-23
4.3 System Design Evaluation 4.3-1 4.3.1 Safety Factors 4.3-1 Reactor Vessel 4.3-1 Steam Generators 4.3-4 Reactor Coolant Pumps 4.3-7 4.3.2 Reliance on Interconnected Systems 4.3-8 4.3.3 System Integrity 4.3-8 4.3.4 Overpressure Protection 4.3-9 4.3.5 System Accident Potential 4.3-10 4.3.6 Redundancy 4.3-10 References 4.3-11
4.4 Tests and Inspections 4.4-1 4.4.1 Reactor Coolant System Inspection 4.4-1 Non-Destructive Inspection of Materials and Components 4.4-1 In-Service Inspection Capability 4.4-5
APPENDICES
Appendix 4A Determination of Reactor Pressure Vessel Reference Nil-Ductility Transition Temperature (RT NDT)
Appendix 4B Procedure for Plugging a Tube in a Steam Generator
Appendix 4C Replacement Steam Generator Design
4-ii Rev. 16 10/99
LIST OF TABLES
Table Title 4.1-1 Reactor Coolant System Design Parameters and Pressure Settings
4.1-2 Reactor Vessel Design Data
4.1-2a Chemical Analyses in Weight Percent Reactor Vessel Surveillance Material
4.1-3 Pressurizer and Pressurizer Relief Tank Design Data
4.1-4 Steam Generator Design Data
4.1-5 Reactor Coolant Pumps Design Data
4.1-6 Reactor Coolant Piping Design Data
4.1-7 Reactor Coolant System Design Pressure Drop
4.1-8 Design Thermal and Loading Cycles - 60 yrs
4.1-9 Reactor Coolant System-Code Requirements
4.1-10 Component Cyclic or Transient Limits
4.2-1 Materials of Construction of the Reactor Coolant System Components
4.2-2 Reactor Coolant Water Chemistry Specification
4.2-3 Steam Generator Water (Steam Side) Chemistry Specification
4.2-4 Furnace Sensitized RCS Stainless Steel Components
4.3-1 Summary of Primary Plus Secondary Stress Intensity for Components of the Reactor Vessel 4.3-2 Summary of Cumulative Fatigue Usage Factors for Components of the Reactor Vessel 4.3-2a Summary of Estimated Cumulative Fatigue Usage Factors for Pressure Bearing Components of the Reactor Coolant Pumps 4.3-3 Stresses Due to Maximum Steam Generator Tube Sheet Pressure Differential (2485 psi) 4.3-4 Ratio of Allowable Stresses to Computed Stresses for a Steam Generator Tube Sheet Pressure Differential of 2485 psi 4.3-5 Summary of Results of Charpy V-Notch and Drop Weight Tests for Reactor Vessel Plates and Forgings and Beltline Welds
4.3-6 [DELETED]
4.4-1 Reactor Coolant System Quality Assurance Program
4.4-2 Surveillance Capsule Withdrawal Schedule - Turkey Point Units 3 & 4
4-iii Revised 07/25/2012
C26 LIST OF FIGURES
Figure Title
4.2-1 Reactor Coolant System (Unit 3)
4.2-1a Deleted
4.2-1b Deleted
4.2-2 Arrangement of Reactor Vessel Longitudinal Section-Part 1
Arrangement of Reactor Vessel Cross Section-Part 2
Specification Drawing For Replacement Reactor Vessel Closure Head-Turkey Point Unit 3 & 4- Part 3
Turkey Point Unit 3 & 4 Replacement Reactor Vessel Closure Head- As-Built Dimensions- Part 4
4.2-3 Pressurizer
4.2-4 Steam Generator
4.2-5 Reactor Coolant Controlled Leakage Pump
4.2-6 Reactor Coolant Pump Estimated Performance
Characteristic
4.2-7 Radiation Induced Increase in Transition Temperature
for A 302-B Steel
4.2-8 Reactor Coolant Pump Motor Lube Oil Fire Protection
4.2-9 Reactor Coolant System (Unit 3)
4.2-10 Reactor Coolant System Reactor Coolant Pumps (Unit 3)
4.2-11 Reactor Coolant System PORV Control (Unit 3)
4.2-12 Reactor Coolant System (Unit 4)
4.2-13 Reactor Coolant System (Unit 4)
4.2-14 Reactor Coolant System Reactor Coolant Pumps (Unit 4)
4.2-15 Reactor Coolant System PORV Control (Unit 4)
4.3-1 Reactor Vessel Stress Analysis: Areas Examined
4.3-2 Reactor Vessel Stress Analysis: Details - Upper
4.3-3 Reactor Vessel Stress Analysis: Details - Lower
4-iv Revised 07/14/2005 4.0 REACTOR COOLANT SYSTEM The Reactor Coolant System, shown in Flow Diagrams, Figures 4.2-1, and 4.2-9
through 4.2-13 consists of three similar heat transfer loops connected in
parallel to a reactor vessel. Each loop contains a circulating pump and a
steam generator. The system also includes a pressurizer, pressurizer relief
tank, connecting piping, and instrumentation necessary for operational
control and protection.
4.1 DESIGN BASES
4.1.1 PERFORMANCE OBJECTIVES
The Reactor Coolant System transfers the heat generated in the core to the
steam generators where steam is generated to drive the turbine generator.
Borated demineralized light water is circulated at the flow rate and
temperature consistent with achieving the reactor core thermal-hydraulic
performance presented in Section 3. The water also acts as a neutron
moderator and reflector, and as a solvent for the neutron absorber used in
chemical shim control.
The Reactor Coolant System provides a boundary for containing the coolant
under operating temperature and pressure conditions. It serves to confine
radioactive material and limits to acceptable values its release to the
secondary system and to other parts of the unit under conditions of either
normal or abnormal reactor operation. During transient operation the
system's heat capacity attenuates thermal transients generated by the core or
extracted by the steam generators. The Reactor Coolant System accommodates
coolant volume changes within the protection system criteria.
The thermal hydraulic effects consequent on loss of power to the reactor
coolant pumps are reduced to acceptable levels by appropriate selection of
the inertia of the reactor coolant pumps. The layout of the system ensures
natural circulation capability following a loss of flow to permit cooldown
without overheating the core. Part of the system's piping is used by the
Emergency Core Cooling System to deliver cooling water to the core during a
loss-of-coolant accident.
4.1-1 Rev. 13 10/96 4.1.2 1967 NRC PROPOSED GENERAL DESIGN CRITERIA (GDC)
The following discussion refers to Turkey Point Plant commitments to the 1967 Proposed General Design Criteria as documented by the U.S. Atomic Energy Commission in Reference 1. Due to the vintage of the Turkey Point Plant, there is no correlation between the 1967 Proposed GDC and those criteria currently contained in 10 CFR 50, Appendix A.
General design criteria which apply to the Reactor Coolant System are given
below.
Quality Standards
Criterion: Those systems and components of reactor facilities which are essential to the prevention, or the mitigation of the consequences, of nuclear accidents which could cause undue risk
to the health and safety of the public shall be identified and
then designed, fabricated, and erected to quality standards that
reflect the importance of the safety function to be performed.
Where generally recognized codes and standards pertaining to
design, materials, fabrication, and inspection are used, they
shall be identified. Where adherence to such codes or standards
does not suffice to assure a quality product in keeping with the
safety function, they shall be supplemented or modified as
necessary. Quality assurance programs, test procedures, and
inspection acceptance criteria to be used shall be identified.
An indication of the applicability of codes, standards, quality
assurance programs, test procedures, and inspection acceptance
criteria used is required. Where such items are not covered by
applicable codes and standards, a showing of adequacy is
required. (1967 Proposed GDC 1)
The Reactor Coolant System is of primary importance with respect to its
safety function in protecting the health and safety of the public.
Quality standards of material selection, design, fabrication and inspection
conform to the applicable provisions of recognized codes and good nuclear
practice (Section 4.1.7). Details of the quality assurance programs, test
procedures and inspection acceptance levels are given in Section 4.3.1 and
4.4. Particular emphasis is placed on the assurance of quality of the
reactor vessel to obtain material whose properties are uniformly within
tolerances appropriate to the application of the design methods of the code.
4.1-2 Rev. 13 10/96 Performance Standards Criterion: Those systems and components of reactor facilities which are essential to the prevention or to the mitigation of the consequences of nuclear accidents which could cause undue risk to
the health and safety of the public shall be designed, fabricated, and erected to performance standards that will enable
such systems and components to withstand, without undue risk to
the health and safety of the public, the forces that might
reasonably be imposed by the occurrence of an extraordinary
natural phenomenon such as earthquake, tornado, flooding
condition, high wind or heavy ace. The design bases so
established shall reflect: (a) appropriate consideration of the
most severe of these natural phenomena that have been officially
recorded for the site and the surrounding area and (b) an
appropriate margin for withstanding forces greater than those
recorded to reflect uncertainties about the historical data and
their suitability as a basis for design. (1967 Proposed GDC 2)
All piping, components and supporting structures of the Reactor Coolant System are designed as Class I equipment; i.e., they are capable of
withstanding:
(a) The design seismic ground acceleration within code allowable working stresses.
(b) The maximum potential seismic ground acceleration acting in the horizontal and vertical direction simultaneously with no loss of capability to perform their safety function.
Details are given in Section 4.1.4.
The Reactor Coolant System is located in the containment building whose
design, in addition to being a Class I structure, also considers accidents or
other applicable natural phenomena. Details of the containment design are
given in Section 5.
Records Requirements
Criterion: The reactor licensee shall be responsible for assuring the maintenance throughout the life of the reactor of records of the design, fabrication, and construction of major components of the
plant essential to avoid undue risk to the health and safety of
the public. (1967 Proposed GDC 5)
4.1-3 Rev. 13 10/96 Records of the design, of the major Reactor Coolant System components and the related engineered safety features components are maintained in the offices
of Florida Power and Light Company and will be retained there throughout the
life of the unit.
Records of fabrication are maintained in the manufacturers' plants as
required by the appropriate Code, or other requirements pending submittal to
Westing-house or Florida Power and Light Company. They are available at any
time to Florida Power and Light throughout the life of the unit.
Construction records are available at the construction site and in the
offices of Florida Power and Light Company where they will be retained for
the life of the unit.
Missile Protection Criterion: Adequate protection for those engineered safety features, the failures 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)
The dynamic effects during blowdown following a loss-of-coolant accident are
evaluated in the detailed layout and design of the high pressure equipment
and barriers which afford missile protection. Fluid and mechanical driving
forces are calculated, and consideration is given to possible damage due to
fluid jets and secondary missiles which might be produced.
The steam generators are supported, guided and restrained in a manner which
prevents rupture of the steam side of a generator, the steam lines and the
feedwater piping as a result of forces created by a Reactor Coolant System
pipe rupture. These supports, guides and restraints also prevent rupture of
the primary side of a steam generator as a result of forces created by a
steam or feedwater line rupture.
The mechanical consequences of a pipe rupture are restricted by design such
that the functional capability of the engineered safety features is not
impaired.
A discussion on missile protection is given in Section 6.2.
4.1-4 Rev. 13 10/96 Environmental and Dynamic Effects The following general design criteria is contained in 10 CFR 50, Appendix A.
Criterion: The reactor coolant system shall be designed to accommodate the effects of and to be compatible with the environmental conditions
associated with normal operation, maintenance, testing and
postulated accidents (10 CFR 50, Appendix A - GDC 4).
The NRC documents in their letter of November 28, 1988 (Reference 1) that the
leakage detection systems at Turkey Point Units 3 and 4 satisfy the
requirements of Generic Letter 84-04, and that the primary loop piping
complies with the criteria of GDC 4 from 10 CFR 50, Appendix A. GDC 4 allows
the use of plant-specific Leak-Before-Break analysis to eliminate the dynamic
effects of postulated pipe ruptures in high energy piping from the design
basis of a plant. Plants with an NRC-approved Leak-Before-Break analysis may
remove pipe whip restraints and jet impingement barriers. Turkey Point Units
3 and 4 received NRC approval (Reference 2) for elimination of the dynamic
effects of postulated pipe ruptures in reactor coolant piping from the design
basis of the plant. The Turkey Point analysis for the Leak-Before-Break
Methodology is documented in the Westinghouse report WCAP-14237 (Reference
3).
4.1.3 PRINCIPAL DESIGN CRITERIA
The criteria which apply solely to the Reactor Coolant System are given
below:
Reactor Coolant Pressure Boundary Criterion: The reactor coolant pressure boundary shall be designed, fabricated and constructed so as to have an exceedingly low
probability of gross rupture or significant uncontrolled leakage
throughout its design lifetime. (1967 Proposed GDC 9)
The Reactor Coolant System in conjunction with its control and protective provisions is designed to accommodate the system pressures and temperatures
attained under all expected modes of station operation or anticipated system
interactions, and maintain the stresses within applicable code stress limits.
Fabrication of the components which constitute the pressure boundary of the
Reactor Coolant System is carried out in strict accordance with the
applicable codes. In addition there are areas where equipment specifications
for Reactor
4.1-5 Rev. 13 10/96 Coolant System components go beyond the applicable codes. Details are given in Section 4.4.1.
The materials of construction of the pressure boundary of the Reactor Coolant
System are protected by control of coolant chemistry from corrosion phenomena
which might otherwise reduce the system structural integrity during its
service lifetime.
System conditions resulting from anticipated transients or malfunctions are
monitored, and appropriate action is automatically initiated to maintain the
required cooling capability and to limit system conditions to a safe level.
The system is protected from overpressure by means of pressure relieving
devices, as required by Section III of the ASME Boiler and Pressure Vessel
Code.
Isolatable sections of the system are provided with overpressure relieving
devices discharging to closed systems such that the system code allowable
relief pressure is not exceeded within the protected section.
Monitoring Reactor Coolant Leakage Criterion: Means shall be provided to detect significant uncontrolled leakage from the reactor coolant pressure boundary.
(1967 Proposed GDC 16)
The methods by which significant leakage from the Reactor Coolant System is
detected are discussed in Section 6.5.
Further details are supplied in Section 4.2.7.
Monitoring of reactor vessel flange leakage is discussed in Section 4.2.2.
Reactor Coolant Pressure Boundary Capability
Criterion: The reactor coolant pressure boundary shall be capable of accommodating without rupture the static and dynamic load imposed on any boundary component as a result of an inadvertent and
sudden release of energy to the coolant. As a design reference, this sudden release shall be taken as that which would result
from a sudden reactivity insertion such as rod ejection (unless
prevented by positive mechanical means), rod dropout, or cold
water addition. (1967 Proposed GDC 33)
4.1-6 Revised 04/17/2013 The reactor coolant boundary is shown to be capable of accommodating without rupture, the static and dynamic loads imposed as a result of a sudden
reactivity insertion such as a rod ejection. Details of this analysis are
provided in Section 14.
The operation of the reactor is such that the severity of an ejection
accident is inherently limited. Since RCC assemblies are used to control
load variations only and boron dilution is used to compensate for core
depletion, only the rod cluster control assemblies in the controlling groups
are inserted in the core at power, and at full power these rods are only
partially inserted. A rod insertion limit monitor is provided as an
administrative aid to the operator to ensure that this condition is met.
By using the flexibility in the selection of control rod groupings, radial
locations and position as a function of load, the design limits the maximum
fuel temperature for the highest worth ejected rod to a value which precludes
any resultant damage to the primary system pressure boundary from possible
excessive pressure surges.
The failure of a rod mechanism housing causing a rod cluster to be rapidly
ejected from the core is evaluated as a theoretical, though not a credible
accident. While limited fuel damage could result from this hypothetical
event, the fission products are confined to the Reactor Coolant System and
the reactor containment. The environmental consequences of rod ejection are
less severe than from the maximum hypothetical loss-of-coolant accident, for
which public health and safety is shown to be adequately protected.
Reference is made to Section 14.
Rod drop out is positively prevented by the mechanical design of the core and
rod cluster control assemblies.
4.1-7 Revised 04/17/2013 Since no means are provided to isolate individual loops, and since natural circulation occurs if the system is hot and the reactor coolant pumps are not
running, it is not possible to preferentially cool a large volume of water
that could be swept into the core. It is also not possible to rapidly add
cold unborated water to the system. Therefore reactivity insertion from cold
water addition does not pose any threat to the integrity of the Reactor
Coolant System.
Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention Criterion: The reactor coolant pressure boundary shall be designed and operated to reduce to an acceptable level the probability of
rapidly propagating type failure. Consideration is given (a) to
the provisions for control over service temperature and
irradiation effects which may require operational restrictions, (b) to the design and construction of the reactor pressure vessel
in accordance with applicable codes, including those which
establish requirements for absorption of energy within the
elastic strain energy range and for absorption of energy by
plastic deformation and (c) to the design and construction of
reactor coolant pressure boundary piping and equipment in
accordance with applicable codes. (1967 Proposed GDC 34)
The reactor coolant pressure boundary is designed to reduce to an acceptable
level the probability of a rapidly propagating type failure.
In the core region of the reactor vessel it is expected that the notch
toughness of the material will change as a result of exposure to fast
neutrons. This change is evidenced as a shift in the Nil Ductility
Transition Temperature (NDTT), which is factored into the operating
procedures in such a manner that full operating pressure is not reached until
the affected vessel material is above the Design Transition Temperature (DTT), and in the ductile material region. The DTT is a minimum of NDTT plus 60 F and dictates the procedures to be followed in the hydrostatic test and in unit operations to avoid excessive cold stress. The pressure during startup and shutdown, at temperatures below NDTT, is maintained below the
threshold of concern for safe operation.
4.1-8 Rev. 13 10/96 The value of the DTT is increased during the life of the unit as required by the expected shift in the NDTT temperature, and as confirmed by the
experimental data obtained from irradiated specimens of reactor vessel
materials. Further details are given in Section 4.1.6 and Appendix 4A.
All pressure-containing components of the reactor coolant system are
designed, fabricated, inspected and tested in conformance with the applicable
codes. Further details are given in Section 4.1.7.
Reactor Coolant Pressure Boundary Surveillance
Criterion: Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance of critical areas by appropriate means to assess the structural and leaktight
integrity of the boundary components during their service
lifetime. For the reactor vessel, a material surveillance
program conforming with current applicable codes shall be
provided. (1967 Proposed GDC 36)
The design of the reactor vessel and its arrangement in the system permits access during the service life to the entire internal surfaces of the vessel
and to the following external zones of the vessel: the flange seal surface, the flange O.D. down to the cavity seal ring, the closure head except around
the drive mechanism adapters and the nozzle to reactor coolant piping welds.
The reactor arrangement within the containment provides sufficient space for
inspection of the external surfaces of the reactor coolant loop components
and piping, except for the area of pipe within the primary shielding
concrete.
Monitoring of the NDTT properties of the core region plates, forgings, weldments and associated heat treated zones are performed in accordance with
the version of ASTM E185,"Recommended Practice for Surveillance Tests on
Structural Materials in Nuclear Reactors," required by 10 CFR 50, Appendix H.
Samples of reactor vessel plate materials are retained and catalogued in case
future engineering development shows the need for further testing.
4.1-9 Rev. 13 10/96 The material properties surveillance program includes not only the conventional tensile and impact tests, but also fracture mechanics tests.
The fracture mechanics specimens are the Wedge Opening Loading (WOL) type
specimens. The observed shifts in NDTT of the core region materials with
irradiation will be used to confirm the calculated limits on heatup and
cooldown transients.
To define permissible operating conditions below DTT, a pressure range is
established which is bounded by a lower limit for pump operation and an upper
limit which satisfies reactor vessel stress criteria. To allow for thermal
stresses during heatup or cooldown of the reactor vessel, an equivalent
pressure limit is defined to compensate for thermal stress as a function of
rate of change of coolant temperature. Since the normal operating
temperature of the reactor vessel is well above the maximum expected DTT, brittle fracture during normal operation is not considered to be credible.
4.1.4 DESIGN CHARACTERISTICS
Design Pressure
The Reactor Coolant System design and operating pressure together with the
safety, power relief and pressurizer spray valves set points, and the
protection system set point pressures are listed in Table 4.1-1. The design
pressure allows for operating transient pressure changes. The selected
design margin considers core thermal lag, coolant transport times and
pressure drops, instrumentation and control response characteristics, and
system relief valve characteristics. The design pressures and data for the
respective system components are listed in Tables 4.1-2 through 4.1-6. Table
4.1-7 gives the design pressure drop of the system components.
4.1-10 Design Temperature The design temperature for each component is selected to be above the maximum
coolant temperature in that component under all normal and anticipated
transient load conditions. The design and operating temperatures of the
respective system components are listed in Tables 4.1-2 through 4.1-6.
Seismic Loads
The seismic loading conditions are established by the "design earthquake" and "maximum potential earthquake". The former is selected to be typical of the
largest probable ground motion based on the site seismic history. The latter
is selected to be the largest potential ground motion at the site based on
seismic and geological factors and their uncertainties.
For the "design earthquake" loading condition, the nuclear steam supply
system is designed to be capable of continued safe operation. Therefore, for
this loading condition, critical structures and equipment needed for this
purpose are required to operate within normal design limits. The seismic
design for the "maximum potential earthquake" is intended to provide a margin
in design that assures capability to shut down and maintain the nuclear
facility in a safe condition. In this case, it is only necessary to ensure
that the Reactor Coolant System components do not lose their capability to
perform their safety function. This has come to be referred to as the "no-loss-of-function" criteria and the loading condition as the "no-loss-of-function earthquake" loading condition.
The criteria adopted for allowable stresses and stress intensities in vessels
and piping subjected to normal loads plus seismic loads are defined in
Appendix 5A.
4.1-11 For the combination of normal plus design earthquake loadings, the stresses in the support structures are kept within the limits of the applicable codes.
For the combination of normal plus no-loss-of-function earthquake loadings, the stresses in the support structures are limited to values necessary to
ensure their integrity, and to keep the stresses in the Reactor Coolant
System components within the allowable limits as given in Appendix 5A.
4.1.5 CYCLIC LOADS
All components in the Reactor Coolant System are designed to withstand the
effects of cyclic loads due to reactor coolant system temperature and
pressure changes. These cyclic loads are introduced by normal power changes, reactor trip, and startup and shutdown operation. The number of thermal and
loading cycles used for design purposes and their bases are given in Table
4.1-8. During unit startup and shutdown, the rates of temperature and
pressure changes are limited as indicated in Section 4.4.1. The cycles are
estimated for equipment design purposes and are not intended to be an exact
representation of actual transients or actual operating experience. For example the number of cycles for unit heatup and cooldown at 100 F per hour was selected as a conservative estimate based on an evaluation of the expected requirements. The resulting number could be increased
significantly; however, it is the intent to represent a conservative
realistic number rather than the maximum allowed by the design.
Although loss of flow and loss of load transients are not included in Table
4.1-8 since the tabulation is only intended to represent normal design
transients, the effect of these transients have been analytically evaluated
and are included in the fatigue analysis for primary system components.
Table 4.1-10 provides the component cyclic or transient limits for the Reactor Coolant System and Secondary Coolant System as was relocated from Technical Specification Section 5.6 by License Amendments 251 and 247 (Reference 8) for Units 3 and 4, respectively.
4.1-12 Revised 07/25/2012 C26 Over the range from 15% full power up to but not exceeding 100% of full power, the Reactor Coolant System and its components are designed to
accommodate 10% of full power step changes in unit load and 5% of full power
per minute ramp changes without reactor trip. The turbine bypass and steam
dump system make it possible to accept a step load decrease of 50% of full
power without reactor trip.
4.1.6 SERVICE LIFE
The service life of Reactor Coolant System pressure components depends upon
the material irradiation, unit operational thermal cycles, quality
manufacturing standards, environmental protection, and adherence to
established operating procedures.
The reactor vessel is the only component of the Reactor Coolant System which
is exposed to a significant level of neutron irradiation and it is therefore
the only component which is subject to any appreciable material irradiation
effects. The NDTT shift of the vessel material and welds, due to radiation
damage effects, is monitored by a radiation damage surveillance program which
conforms with ASTM - E 185 standards.
Reactor vessel design is based on the transition temperature method of
evaluating the possibility of brittle fracture of the vessel material, as
result of operations such as leak testing and heatup and cooldown.
To establish the service life of the Reactor Coolant System components as
required by the ASME (part III), Boiler and Pressure Vessel Code for Class "A" Vessels, the unit operating conditions have been established for the
initial 40 year design life. These operating conditions include the cyclic
application of pressure loadings and thermal transients. The evaluation for
extended plant design life concludes that the 40-year design cycles envelope
the 60-year extended design life.
The number of thermal and loading cycles used for design purposes are listed
in Table 4.1-8. Component Cyclic or Transient Limits are listed in Table
4.1-10
Metal fatigue considerations, including Reactor vessel underclad cracking, have also been evaluated for the extended plant life as discussed in UFSAR
Chapter 16. The analysis associated with reactor vessel underclad crack
growth has been projected to the end of the period of extended operation, in
accordance with 10 CFR 54.21(c)(1)(ii) as indicated in UFSAR Chapter 16.
Underclad cracking has been evaluated by Westinghouse in WCAP-15338, "A
Review of Cracking Associated with Weld Deposited Cladding in Operating PWR
Plants."
4.1-13 Revised 07/25/2012 C26 4.1.7 CODES AND CLASSIFICATIONS
All pressure-containing components of the Reactor Coolant System are
designed, fabricated, inspected and tested in conformance with the applicable
codes listed in Table 4.1-9. The Reactor Coolant System is classified as
Class I for seismic design, requiring that there will be no loss of function
of such equipment in the event of the assumed maximum hypothetical ground
accelerations acting in the horizontal and vertical directions
simultaneously, when combined with the primary steady state stresses.
Reactor coolant system valves, fittings and piping were designed, fabricated, inspected and tested in conformance with the Code requirements listed in
Table 4.1-9. Hydrostatic testing of piping and fittings is done after
installation at the pressure given in Table 4.1-6, which is the reactor
coolant system test pressure also. This is 1 1/4 times design pressure and
is a necessary deviation from Code Case N-10.
Reactor Coolant System inservice Inspection of ASME Code Class 1, 2, and 3
components shall be performed in accordance with Section XI of the ASME
Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR
50, Section 50.55a(g), except where specific written relief has been granted by
the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Prior approval is not required for ASME BP&V Code Cases listed in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME XI, Division
- 1. All provisions of the Code Case must be met along with limitations issued in Regulatory Guide 1.147, if any. Code Cases not listed in Regulatory Guide 1.147 must receive specific permission for use (i.e. Relief Request) from the USNRC prior to their use, pursuant to 10 CFR 50.55a(g)(6)(i).
Inservice inspection of the steam generator tubes ensure that the structural
integrity of this portion of the RCS will be maintained. The program for
inservice inspection of steam generator tubes is based on a modification of
Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator
tubing is essential in order to maintain surveillance of the conditions of
the tubes in the event that there is evidence of mechanical damage or
progressive degradation due to design, manufacturing errors, or inservice
conditions that lead to corrosion. Inservice inspection of steam generator
tubing also provides a means of characterizing the nature and cause of any
tube degradation so that corrective measures can be taken.
4.1-14 Rev. 15 4/98 4.
1.8 REFERENCES
- 1. NRC Letter, from G.E. Edison (NRC) to W.F. Conway (FPL), "Turkey Point Units 3 and 4 - Generic Letter 84-04, Asymmetric LOCA Loads," dated November 28, 1988.
- 2. NRC Letter, from R. P. Croteau (NRC) to J. H. Goldberg (FPL), "Turkey Point Units 3 and 4 - Approval to Utilize Leak-Before-Break Methodology
for Reactor Coolant System Piping (TAC Nos. M91494 and M91495)," dated
June 23, 1995.
- 3. Westinghouse WCAP-14237, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Turkey
Point Units 3 and 4 Nuclear Power Plants," dated December 1994.
- 4. WCAP-15092 Revision 3, Turkey Point Units 3 and 4 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal
Operation, Westinghouse, May 2000.
- 5. Low Temperature Overpressure Protection System Setpoints, Florida Power
& Light Company, 32 EFPY and 48 EFPY for Turkey Point, Units 3 & 4.
Westinghouse Electric Company LLC, June 2000.
- 6. PC/M 03-057, Rev. 01, "Reactor Vessel Closure Head Replacement"-Unit 3.
- 7. PC/M 03-058, Rev.0 "Reactor Vessel Closure Head Replacement"-Unit 4.
- 8. NRC (Paige) to FPL (Nazar), "Turkey Point Units 3 and 4. Issuance of Amendments Regarding Section 5.0 Design Features (TAC Nos. ME6334 and ME6335)," dated June 21, 2012.
4.1-15 Revised 07/25/2012 C26 TABLE 4.1-1 REACTOR COOLANT SYSTEM DESIGN PARAMETERS AND PRESSURE SETTINGS
DESCRIPTION PARAMETER Total Primary Heat Output, MWt 2652 Total Primary Heat Output, Btu/hr 9049 x 10 6 Number of Loops 3 Coolant Volume, including pressurizer volume, ft 3 9343 Total Reactor Coolant Flow, gpm (MMF) 270,000 Design Pressure 2485 psig Operating Pressure (at pressurizer) 2235 psig Safety Valves 2465 (+2%,-3%) psig Power Relief Valves : i) Normal Operation 2335 psig ii) OMS Actuation During Heatup and Cooldown a) RCS 285 F 440 psig (Note 2) b) Above 285F, the OMS setpoints are established by the Plant Curve Book,Section V, Figure 3D. The most restrictive cool-down rate curve (typically 100°F/hr) and the 60°F subcooling curve bound the setpoint
curve.(1) Pressurizer Spray Valves (Open) 2250 psig High Pressure Trip 2385 psig High Pressure Alarm 2300 psig Low Pressure Trip 1835 psig Low Pressure Alarm 2185 psig Hydrostatic Test Pressure 3107 psig
NOTES: 1. OMS is not normally in-service at RCS temperatures greater than 300 F. 2. Technical Specification LCO 3.4.9.3a indicates a PORV lift setting of <
448 psig, however, the field device for OMS setpoint actuation will be at 440 psig to provide buffer from the Technical Specifications value.
Revised 04/17/2013 C26C26C26C26C26C26C26 TABLE 4.1-2 REACTOR VESSEL DESIGN DATA
Design/Operating Pressure, psig 2485/2235 Hydrostatic Test Pressure, psig 3107 Design Temperature, F 650 Overall Height of Vessel and Closure Head, ft-in. (Bottom Head O.D. to top of Control Rod Mechanism Housing) 42-7
Water Volume, (with core and internals in place), ft 3 3667 Thickness of Insulation, min., in.
(Unit 3 & 4 Vessels) 3
Thickness of Insulation, min., in (Unit 3 & 4 RVCH) 5
Number of Reactor Closure Head Studs 58
Diameter of Reactor Closure Head Studs, in. 6
Flange, ID, in. 149.6
ID at Shell, in. 155.5
OD across inlet/outlet nozzles, in. 230-5/16 / 240
Inlet Nozzle ID, in. Tapered 27-15/32 to 33-13/16
Outlet Nozzle ID, in. 28 31/32
Clad Thickness, min., in. 0.156
Lower Head Thickness, min., in. 4-3/4 plus
cladding Vessel Belt-Line Thickness, min., in. 7-3/4 plus
cladding Closure Head Thickness, in. 6-3/16 plus
cladding Reactor Coolant Inlet Temperature, F 535.5-549.2 Reactor Coolant Outlet Temperature, F 604.5-616.8 Reactor Coolant Flow, lb/hr 98.1 x 10 6 - 99.9 x 10 6
Revised 04/17/2013 C26 TABLE 4.1-2a CHEMICAL ANALYSES IN WEIGHT PERCENT REACTOR VESSEL SURVEILLANCE MATERIAL
Intermediate Lower Element Shell Shell Unit 3 Unit 4 Unit 3 Unit 4 C 0.20 0.22 0.20 0.21
Mn 0.64 0.67 0.61 0.67
P 0.010 0.010 0.010 0.011
S 0.010 0.009 0.008 0.009
Si 0.26 0.20 0.20 0.23
Ni 0.70 0.68 0.67 0.74 Cr 0.40 0.33 0.38 0.31
V 0.02 0.002 0.02 0.001
Mo 0.62 0.56 0.58 0.56
Co 0.011 0.017 0.015 0.015
Cu 0.058 0.054 0.079 0.056
Zr *0.001 0.005 *0.001 0.004
Sn 0.010 0.008 0.008 0.008
Ti *0.001 *0.001 *0.001 *0.001
Sb *0.001 *0.001
Zn 0.001 *0.001 0.001 *0.001
As *0.005 0.004 *0.005 0.005
B *0.003 *0.003 *0.003 *0.003
Al 0.005 0.008 0.005 0.008
N 2 0.003 0.001 0.003 0.002 Nb 0.002 0.001
W *0.001 *0.001
Pb *0.001 0.001
Ta 0.003 0.002
- Not detected. The number indicates the minimum limit of detection.
Revised 03/01/2011 C25 TABLE 4.1-3 Sheet 1 of 2 PRESSURIZER AND PRESSURIZER RELIEF TANK DESIGN DATA Pressurizer Design/Operating Pressure, psig 2485/2235 Hydrostatic Test Pressure (cold), psig 3107 Design/Operating Temperature o F 680/653 Water Volume, Full Power, ft 3
- 766 Steam Volume, Full Power, ft 3 534 Surge Line Nozzle Diameter, in./Pipe Schedule 14/Sch 140 Shell ID, in./Minimum Shell Thickness, in.
84/4.1 Minimum Clad Thickness, in.
0.188 Electric Heaters Capacity, kw (total)
- 1300(Design)
Heatup rate of Pressurizer using Heaters only, oF/hr 55 (approximately with Design heater capacity)
Power Relief Valves: #455C & 456
Number 2 Set Pressure (open), psig
i) Normal operation 2335 ii) OMS Actuation during Heatup or Cooldown
a) RCS 285 o F 440*** b) RCS > 285 o F Setpoint increases to 2335 psig at 554 o F as a segmented curve. Above 554 o F and up to 750 o F, the setpoint is maintained at a constant 2335 psig.
Capacity, lb/hr saturated steam/valve 179,000 Safety Valves
Number 3 Set Pressure, psig 2465 +1% (as left)
+2%/-3% (as found)
Capacity, lb/hr saturated steam/valve 313,826 Pressurizer Relief Tank Design pressure, psig 100 Rupture disc release pressure, psig 100 Design temperature, o F 340 Normal water temperature, o F 120 Total volume, ft 3 1300 Rupture disc relief capacity, lb/hr 931,964 Each
Revised 04/17/2013 C26C26C26C26C26C26 TABLE 4.1-3 Sheet 2 of 2 PRESSURIZER AND PRESSURIZER RELIEF TANK DESIGN DATA
- This volume corresponds to the maximum pressurizer level of 60% of span at full power conditions. Surge line volume is not included.
- Original as-built design. The extended power uprate analysis supports a minimum pressurizer heater capacity of 1000 kw.
448 psig, however, the field device for OMS setpoint actuation will be set at 440 psig to provide buffer from the Technical Specifications value.
Revised 04/17/2013 C26C26 TABLE 4.1-4 STEAM GENERATOR DESIGN DATA
Number of Steam Generators 3
Design Pressure, Reactor Coolant/Steam, psig 2485/1085
Reactor Coolant Hydrostatic Test pressure (tube side-cold), psig 3107
Design Temperature, Reactor Coolant/Steam, F 650/556
Reactor Coolant Flow, lb/hr, each 33.83 x 10 6 Total Heat Transfer Surface Area, ft.
2 , each 43,467 Steam Conditions at Full Load, Outlet Nozzle:
Steam Flow, lb/hr, each 3.196 x 10 6 Steam Temperature, F 516
Steam Pressure, psig 770
Feedwater Temperature, F 436.5
Overall Height, ft.-in. 63-1.6
Shell 0D, upper/lower, in. 166/127.5
Shell Thickness, upper/lower, in. 3.5/2.63
Number of U-tubes 3214
U-tube Diameter, in. 0.875
Tube Wall Thickness, (average), in. 0.050
Number of Manways/ID in. 3/16
Number of handholes/ID, in. 6/6
2200 MWt Zero Power Primary Side Fluid Volume, ft.
3 935 935 Primary Side Fluid Heat Content, BTU 24.31 x 10 6 23.7 x 10 6 Secondary Side Fluid Volume, ft.
3 4596 4596 Secondary Side Fluid Mass, lbs. 80,300 134,000
- The above Steam Generator design data has not been revised as part of the Steam Generator Repair Project or Thermal Uprate Project and should be regarded as historical reference only. Refer to FSAR Table 4C-1 for updated design data resulting from the Thermal Uprate Project.
Rev. 16 10/99 TABLE 4.1-5 REACTOR COOLANT PUMPS DESIGN DATA
Number of Pumps 3 Design Pressure/Operating Pressure, psig 2485/2235 Hydrostatic Test Pressure (cold), psig 3107 Design Temperature (casing), o F 650 RPM at Nameplate Rating 1188 Suction Temperature, o F 548.9 Net Positive Suction Head, ft. (required) 168 Developed Head, ft.
266 Capacity, gpm 88,500 Seal Water Injection, gpm 7.5 Seal Water Return, normal, gpm 2.5 Pump Discharge Nozzle, ID, in.
27-1/2 Pump Suction Nozzle ID, in.
31 Overall Unit Height, ft.
28.242 Water Volume, ft.
3 192 Pump-Motor Moment of Inertia, lb-ft 2 70,000 Motor Data:
Type AC Induction Single Speed, Air Cooled Voltage 4000 Insulation Class B Thermalastic Epoxy Phase 3 Frequency, Hz 60 Starting Current, maximum, amp 4800 Input (hot reactor coolant), kw 4360 Input (cold reactor coolant), kw 5674 Power, HP (nameplate) 6000
Revised 07/15/2016 C28 TABLE 4.1-6 REACTOR COOLANT PIPING DESIGN DATA
Design/Operating Pressure, psig 2485/2235 Hydrostatic Test Pressure, (cold) psig 3107 Design Temperature, o F 650 Design Temperature, (pressurizer surge line), o F 680 Reactor Inlet Piping, ID, in. 27-1/2 Reactor Inlet Piping, nominal thickness, in. 2.375 Reactor Outlet Piping, ID, in. 29 Reactor Outlet Piping, nominal thickness, in. 2.50 Coolant Pump Suction Piping, ID, in. 31 Coolant Pump Suction Piping, nominal thickness, in. 2.625 Pressurizer Surge Line Piping, ID, in./Pipe Schedule 12/Sch 140 Pressurizer Surge Line Piping, nominal thickness, in. 1.125
Water Volume, (all 3 loops) ft 3 783
- Surge line fitted with a 14"/12" adapter at the pressurizer
TABLE 4.1-7 REACTOR COOLANT SYSTEM DESIGN PRESSURE DROP
Pressure Drop, psi (estimated)
Across Pump Discharge Leg 1.3 / 1.2 Across Vessel, including nozzles 40.8 / 38.4 Across Hot Leg 1.2 / 1.1 Across Steam Generator 33.9 / 38.0 Across Pump Suction Leg 3.0 / 2.8 Total Pressure Drop 80.2 / 81.5
NOTE: The first value provided coincides with the maximum Best Estimate Flow (minimum steam generator tube plugging, minimum reactor vessel average temperature) and the second value coincides with the minimum Best Estimate Flow (maximum steam generator tube plugging, maximum reactor vessel average temperature).
Revised 04/17/2013 C26C26 TABLE 4.1-8 DESIGN THERMAL AND LOADING CYCLES - 60 YEARS
Transient Design Condition Design Cycles
- 1. Station heatup at 100 o F per hour 200
- 2. Station cooldown at 100 o F per hour 200
- 3. Station loading at 5% of full 14,500 (5) power/min
- 4. Station unloading at 5% of full 14,500 (5) power/min
- 5. Step load increase of 10% of 2000 full power (but not to exceed full power)
- 6. Step load decrease of 10% of full 2000
power
- 7. Step load decrease of 50% of full 200
power
- 8. Reactor trip 400
- 9. Hydrostatic test at 3107 psig 1 (3) pressure, 100 o F temperature
- 10. Hydrostatic test at 2485 psig 5 (4) pressure and 400 o F temperature
- 11. Steady state fluctuations 00 (1)
- 12. Feedwater Cycling at Hot Standby 2000 (2)
Notes: (1) Not counted, not significant contributor to fatigue usage factor.
(2) Not counted, Intermittent slug feeding at hot standby not performed.
(3) Limited by Seam Generator Analysis. Represents pre-operational hydrostatic test. (4) Limited by Reactor Coolant Pump Analysis.
(5) The loading and unloading design cycle limit for baffle-former bolts only is being lowered from 14,500 to 2,200 due to EPU RCS conditions.
Revised 04/17/2011 C26C26C26 TABLE 4.1-9 Sheet 1 of 2 REACTOR COOLANT SYSTEM - CODE REQUIREMENTS Component Codes Reactor Vessel(Note 1, Note 4, Note 5)
ASME III* Class A Control Rod Drive Mechanism Housings (Note 2)
ASME III* Class A Steam Generator Tube Side ASME III* Class A Shell Side ***
ASME III* Class C Reactor Coolant Pump Casing No Code (Design per ASME III-Article 4)
Pressurizer ASME III* Class A
Pressurizer Relief Tank ASME III* Class C
Pressurizer Safety Valves ASME III*
Reactor Coolant Piping ASA B31.1**
System valves, fittings, piping and tubing (Note 5, Note 6)
ASA B31.1**
Core Exit Thermocouple Seal Assemblies (Note 3) ASME III* Subsection (Head Port Adapters, Drive Sleeves NB, Class 1, 1986 Edition
______________
- ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
- ASA B31.1-1955 Code for Pressure Piping, plus Code Cases N-7 and N-10 where applicable.
- The shell side of the steam generator conforms to the requirements for Class A vessels and is so stamped as permitted under the rules of Section III.
Notes:
- 1. The Reactor Vessel Closure Head (RVCH) for Unit 3 and 4 has been replaced (Reference 6 and 7).The replacement RVCH design Code is ASME Boiler and
Pressure Vessel Code,Section III, Subsection NB, Class 1, 1989 Edition, no
Addenda.
- 2. The Control Rod Drive Mechanism (CRDM) pressure housings for Unit 3 and Unit 4
have been replaced (Reference 6 and 7). The replacement CRDM pressure housing
design Code is ASME Boiler and Pressure Vessel Code,Section III, Subsection
NB, Class 1, 1989 Edition, no Addenda.
- 3. The Core Exit Thermocouple Nozzle Adapter (CETNA) and Reactor Vessel Level Monitoring System (RVLMS) Nozzle Adapter for Unit 3 and Unit 4 have been
replaced (Reference 6 and 7). The replacement components' design Code is ASME
Boiler and Pressure Vessel Code,Section III, Subsection NB, Class 1, 1989
Edition, no Addenda.
- 4. Per Reference 6 and Reference 7 - The spare CRDM housing adapters are closed with a CRDM plug. The CRDM Plug design Code is ASME Boiler and Pressure
Vessel Code,Section III, Subsection NB, Class 1, 1989 Edition, no Addenda.
Revised 04/17/2013 C26C26C26 TABLE 4.1-9 Sheet 2 of 2 REACTOR COOLANT SYSTEM - CODE REQUIREMENTS Notes (cont'd)
- 5. The design Code for the Bottom Mounted Instrumentation (BMI) Tubing is the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB, 1989 edition, no addenda. The design Code for the BMI Supports is the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, 1998 edition, no addenda. 6. With the exception of the thermal stratification analysis of the pressurizer surge line which uses fatigue stress limits from the 1986 edition of the ASME B & PV Code,Section III, Subsection NB.
Revised 04/17/2013 C26 TABLE 4.1-10 COMPONENT CYCLIC OR TRANSIENT LIMITS COMPONENT CYCLIC OR TRANSIENT LIMIT DESIGN CYCLE OR TRANSIENT Reactor Coolant System 200 heatup cycles at < 100°F/h and 200 cooldown cycles at 100°F/h.
Heatup cycle - T avg from 200°F to 550°F. Cooldown cycle - T avg from 550°F to 200°F.
200 pressurizer cooldown cycles at 200°F/h from nominal pressure.
Pressurizer cooldown cycl e temperatures from
> 650°F to 200°F.
200 pressurizer cooldown cycles at 200°F/h from 400 psia.
Pressurizer cooldown cycle temperatures from > 650°F to 200°F.
80 loss of load cycles, without immediate Turbine or Reactor trip.
15% of RATED THERMAL POWER to 0% of RATED THERMAL POWER.
40 cycles of loss-of-offsite A.C electrical power.
Loss-of-offsite A.C electrical ESF Electrical System. 80 cycles of loss of flow in one reactor coolant loop.
Loss of only one reactor coolant pump. 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER.
10cycles of inadvertent auxiliary spray. Spray water temperature differential to 560°F. 150 primary to secondary side leak tests. Pressurized to 2435 psig.
15 primary to secondary side leak tests. Pressurized to 2250 psig. 5 hydrostatic pressure tests. Pressurized to 2485 psig and 400°F.
Secondary Coolant System 50 hydrostatic pressure tests. Pressurized to 1085 psig.
10 hydrostatic pressure tests. Pressurized to 1356 psig. 15 secondary to primary side leak tests. Pressurized to 840 psig.
Revised 07/25/2012 C26 4.2 SYSTEM DESIGN AND OPERATION 4.2.1 GENERAL DESCRIPTION
The Reactor Coolant Systems of the two nuclear power units are essentially
identical and do not share any components. The following description applies
to either unit.* Each Reactor Coolant System consists of three similar heat
transfer loops connected in parallel to the reactor vessel. Each loop
contains a steam generator, a pump, loop piping, and instrumentation. The
pressurizer surge line is connected to one of the loops. Auxiliary system
piping connections into the reactor coolant piping are provided as necessary.
A flow diagram of the system is shown on Figures 4.2-1 and 4.2-9 through 4.2-
13 and of the reactor coolant pump on Figures 4.2-10 and 4.2-14.
The containment boundary shown on the flow diagram indicates those major
components which are to be located inside the containment. The intersection
of a process line with this boundary indicates a containment penetration.
Reactor Coolant System and components design data are listed in Table 4.1-1
through 4.1-6.
Pressure in the system is controlled by the pressurizer, where water and
steam pressure is maintained through the use of electrical heaters and
sprays. Steam can either be formed by the heaters, or condensed by a
pressurizer spray to minimize pressure variations due to contraction and
expansion of the coolant. Instrumentation used in the pressure control
system is described in Section 7. Spring-loaded code steam safety valves and
power-operated relief valves are connected to the pressurizer and discharge
to the pressurizer relief tank, where the discharged steam is condensed and
cooled by mixing with water.
- See Appendix 4C for description of replaced steam generator lower assemblies.
4.2-1 Rev. 11 11/93 4.2.2 COMPONENTS
Reactor Vessel
The reactor vessel is cylindrical in shape with a hemispherical bottom and a
flanged and gasketed removable upper head. Figure 4.2-2 is a schematic of
the reactor vessel. The materials of construction of the reactor vessel are
given in Table 4.2-1. The upper, intermediate and lower shell courses, and
the lower head ring, are cylindrical machined forgings.
The Reactor Vessel Closure Heads (RVCH) for Unit 3 and Unit 4 have been replaced with a heads manufactured from a mono-block forging instead of the forged flange and dished plate weldment head. This design eliminated the circumferential weld that welded the head dome plate to the forged flange section. Figure 4.2-2 Parts 3 and 4 is a schematic of the replacement head for the Unit 3 & 4 Reactor Vessel.
Coolant enters the reactor vessel through inlet nozzles in a plane just below
the vessel flange and above the core. The coolant flows downward through the
annular space between the vessel wall and the core barrel into a plenum at
the bottom of the vessel where it reverses direction. Approximately
ninety-five per cent of the total coolant flow is effective for heat removal
from the core. The remainder of the flow includes the flow through the RCC
guide thimbles, the leakage across the fuel assembly outlet nozzles, and the
flow deflected into the head of the vessel for cooling the upper flange. All
the coolant is united and mixed in the upper plenum, and the mixed coolant
stream then flows out of the vessel through exit nozzles located on the same
plane as the inlet nozzles.
A one-piece thermal shield, concentric with the reactor core, is located
between the core barrel and the reactor vessel. The shield is bolted and
welded to the top of the core barrel. The shield, which is cooled by the
coolant on its downward pass, protects the vessel by attenuating much of the
gamma radiation and some of the fast neutrons which escape from the core.
This shield minimizes thermal stresses in the vessel which result from heat
generated by the absorption of gamma energy. It is illustrated in Figure
3.2.3 and is further described in Section 3.2.3.
Fifty core instrumentation nozzles are located on the lower head.
4.2-2 Revised 07/14/2005 The reactor closure head and the reactor vessel flange are joined by 58 - 6 in. diameter studs. Two metallic O-rings seal the reactor vessel when the
reactor closure head is bolted in place. A leakoff connection is provided
between the two O-rings to monitor leakage across the inner O-ring. In
addition, a leak-off connection is also provided beyond the outer O-ring
seal.
The reactor vessel insulation is of the reflective type, supported from the
nozzles and consisting of inner and outer sheets of stainless steel spaced 3
inches apart and with multilayer aluminum foil. For Unit 3, The removable
vessel flange is of similar construction, while stainless steel foil filler
is used for the Unit 4 Reactor Vessel Flange insulation. The clearance to
the reactor vessel is 1/2 inch. The insulation provided for the reactor
closure flange is supported on the refueling seal ledge and vent shroud
support rings.
The Reactor Vessel Head Permanent insulation (i.e., within the IHA) for Unit 3 & 4 consists of self supporting panels, constructed of metallic reflective insulation, that are attached to one another with stainless steel buckles.
This configuration ensures clearance between the reactor vessel head and the bottom of the insulation. The vertical portions of this permanent insulation have removable portions to allow access to the reactor vessel head and CRDM nozzles.
The reactor vessel contains the core support assembly, upper plenum assembly, fuel assemblies, control rod cluster assemblies, surveillance specimens, and
in-core instrumentation. The reactor vessel internals are designed to direct
the coolant flow, support the reactor core, and guide the control rods in the
withdrawn position.
The reactor internals are described in detail in Section 3.2.3 and the
general arrangement of the reactor vessel and internals is shown in Figure
3.2.3-2.
Reactor vessel design data are listed in Table 4.1-2.
4.2-3 Revised 07/14/2005 Reactor Vessel Support Structure The reactor vessel support structure is described in Section 5 and shown in
Figure 5.1-20.
Pressurizer
The general arrangement of the pressurizer is shown in Figure 4.2-3, and the
design data are listed in Table 4.1-3.
The pressurizer maintains the required reactor coolant pressure during
steady-state operation, limits the pressure changes caused by coolant thermal
expansion and contraction during normal load transients, and prevents the
pressure in the Reactor Coolant System from exceeding the design pressure.
The pressurizer contains replaceable direct immersion heaters, multiple
safety and relief valves, a spray nozzle and interconnecting piping, valves
and instrumentation. The electric heaters located in the lower section of
the vessel maintain the pressure of the Reactor Coolant System by keeping the
water and steam in the pressurizer at system saturation temperature. The
heaters are capable of raising the temperature of the pressurizer and contents at approximately 55 F/hr during startup of the reactor.
The pressurizer is designed to accommodate positive and negative surges
caused by load transients. The surge line which is attached to the bottom of
the pressurizer connects the pressurizer to the hot leg of a reactor coolant
loop. During a positive surge, caused by a decrease in unit load, the spray
system, which is fed from the cold leg of a coolant loop, condenses steam in
the vessel to prevent the pressurizer pressure from reaching the set point of
the power operated relief valves. Power operated spray valves on the
pressurizer limit the pressure during load transients. In addition the spray
valves can be operated manually from the control room. A small continuous
spray flow is provided to assure that the pressurizer liquid is homogeneous
with the coolant and to prevent excess cooling of the spray and surge line
piping.
During a negative pressure surge, caused by an increase in unit load, flashing of water to steam and generation of steam by automatic actuation of
the heaters keep the pressure above the minimum allowable limit. Heaters are
also energized on high water level during positive surges to heat the
subcooled surge water entering the pressurizer from the reactor coolant loop.
4.2-4 Revised 04/17/2013 The pressurizer is constructed of carbon steel with internal surfaces clad with austenitic stainless steel. The heaters are sheathed in austenitic stainless steel. Unit 3 pressurizer #11 heater well nozzle has been modified with a half nozzle design welded to the outside of the pressurizer shell instead of the internal cladding. This change was submitted under an ASME Section XI Relief Request to the NRC and approved ( Ref. ADAMS Access1on No.:
The pressurizer vessel surge nozzle is protected from thermal shock by a thermal sleeve. A thermal sleeve also protects the pressurizer spray nozzle connection.
Each loop contains a vertical shell and U-tube steam generator.* A steam generator of this type is shown in Figure 4.2-4. Principal design parameters are listed in Table 4.1-4.
Reactor coolant enters the inlet side of the channel head at the bottom of the steam generator through the inlet nozzle, flows through the U-tubes to an outlet channel and leaves the generator through another bottom nozzle. The inlet and outlet channels are separated by a partition. Manways are provided to permit access to the U-tubes and moisture separating equipment. The general procedural activities for plugging a defective steam generator tube is outlined in Appendix 4B. Defective steam generator tubes having indications may require corrective maintenance actions such as plugging or plugging and staking. Design drawings for plugs and stakes and related procedures shall be approved in accordance with plant administrative procedures.
Feedwater to the steam generator enters just above the top of the U-tubes through a feedwater ring. The water flows downward through an annulus between the tube wrapper and the shell and then upward through the tube bundle where part of it is converted to steam.
- See Appendix 4C for description of replaced steam generator lower assemblies.
4.2-5 Revised 12/18/2015 C28 The steam-water mixture from the tube bundle passes through a steam swirl vane assembly which imparts a centrifugal motion to the mixture and separates
the water particles from the steam. The water spills over the edge of the
swirl vane housing and combines with the feedwater for another pass through
the tube bundle.
The steam rises through additional separators which limit the moisture
content of the steam to one fourth of one per cent or less under all design
load conditions.
The steam generator is constructed primarily of carbon steel. The heat
transfer tubes are Inconel. The interior surfaces of the channel heads and
nozzles are clad with austenitic stainless steel, and the side of the tube
sheet in contact with the reactor coolant is clad with Inconel. The tube to
tube sheet joint is welded.
Steam Generator Support Structure
The steam generator support structures are described in Section 5 and shown
in Figure 5.1-20.
Reactor Coolant Pumps
Each reactor coolant loop contains a vertical single stage centrifugal pump
which employs a controlled leakage seal assembly. A view of a controlled
leakage pump is shown in Figure 4.2-5 and the principal design parameters for
the pumps are listed in Table 4.1-5. The reactor coolant pump estimated
performance and NPSH characteristics are shown in Figure 4.2-6. The
performance characteristic is common to all of the higher specific speed
centrifugal pumps, and the 'knee' at about 40% design flow introduces no
operational restrictions, since the pumps operate at full speed.
4.2-6 Reactor coolant is pumped by the impeller attached to the bottom of the rotor shaft. The coolant is drawn up through the impeller, discharged through passages in the diffuser and out through a discharge nozzle in the side of the casing. The motor-impeller can be removed from the casing for maintenance or inspection without removing the casing from the piping. All parts of the pumps in contact with the reactor coolant are austenitic stainless steel or equivalent corrosion resistant materials.
The pump employs a controlled leakage seal assembly to restrict leakage along the pump shaft. The seal package is a multi-seal cartridge containing three identical stages in series. Only one stage is required to function to prevent excessive leakage from the Reactor Coolant System. Seal staging flow, designated Control Bleed Off, exits the RCP through lines to the Chemical and Volume Control seal return line that goes to the volume control tank. The flow that passes across the upper seal stage, designated Seal Leak off, is routed to the reactor coolant drain tank to minimize leakage of water and vapor into the containment atmosphere. To mitigate the effects caused by failure of all three stages, the seal package includes a Abeyance (shutdown) seal. The flow rates that result from various failure modes of the three stages vary and could result in flows that are not high enough to activate the abeyance seal. With the RCP tripped, the abeyance seal is designed to stop leakage from upper, third stage seal for an indefinite period and RCP seal leakage is limited to flow through the Control Bleed Off seal return line
A portion of the high pressure water flow from the charging pumps is injected into the reactor coolant pump between the thermal barrier (above the RCP impeller) and the controlled leakage seal in the lower pump shaft housing to serve as a buffer to keep reactor coolant from entering the upper portion of the pump. The remainder of the injection water flows along the drive shaft, through the controlled leakage seal, and finally out of the pump. A very small amount which leaks through the upper (third stage) seal is also collected and removed from the pump.
The RCP seals are provided with redundant means of cooling, seal injection via the charging system and thermal barrier cooling via the Component Cooling Water (CCW) system. When both systems are operating, either is sufficient to provide adequate seal cooling for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.2-7 Revised 07/15/2016 C28C28 For an event which occurs causing a safety injection signal with off site power available (i.e., RCPs remain running), thermal barrier cooling will continue so long as the High-High containment pressure signal setpoint of 20 PSIG is not reached. Seal injection will be lost on an S.I. signal (charging pumps tripped on S.I.). On High-High containment pressure, Phase B containment isolation is initiated and thermal barrier cooling will be automatically isolated. If seal injection is re-established, operating procedures permit continued RCP operation until upper or lower motor bearing temperatures reach 195F or RCS subcooling is lost. Upon reaching 195F the RCPs are manually stopped. If neither CCW or Seal Injection are available, the RCPs will be tripped.
The RCPs at Turkey Point are capable of operating for 30 minutes without cooling water with no significant seal degradation. The RCPs must, however, be shutdown when the RCP motor bearing temperature reaches 195F. The RCP seals will not leak excessively even if cooling water is lost for an extended period of time after the RCPs are tripped Flowserve has conducted a series of tests to explore the ability of the N-Seal RCP package to withstand pump running operation under a Loss of Seal Cooling (LOSC) situation.
Additionally, a model for failure of the three stage N-seal package with an Abeyance (shutdown) seal that addresses the probability of occurrence of activation conditions for the shutdown seal has been developed and testing evaluates the ability of the N-Seal package to withstand LOSC during pump shutdown conditions. Failure modes that result in flow rates lower than required to activate the abeyance seal will not actuate the seal. The Flowserve Dynamic tests demonstrate that the seal package can remain intact for over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with the RCS at full temperature and pressure while the pump is operating. The Abeyance seal provides assurance that after degradation to the mechanical seal package occurs with the RCPs stopped, the coping time is extended to at least 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> under the most extreme operating parameters.
The N-Seal design has been installed and operated at plants such as Surry, Oconee, and Crystal River 3 (Reference PRA Model for Flowserve 3 Stage N-Seals with Abeyance Seal, Revision 0, dated 12/20/2013).
The squirrel cage induction motor driving the pump is air cooled and has oil lubricated thrust and radial bearings. The lube oil leakage drain arrangement for protection against fire at the reactor coolant pump motor discussed in Engineering Guidelines for Fire Protection for Turkey Point Units 3 & 4 (Reference 14). A water lubricated bearing provides radial support for the pump shaft. Component cooling water is supplied to the motor bearing cooler and the thermal barrier cooling coil.
4.2-8 Revised 09/20/2016 C28C28C28 A reactor coolant pump motor vibration monitor system is provided to assist in balancing the pump-motor combination and to provide alarm and recording level capability. Two shaft vibration proximity probes are mounted 90 apart and a third probe develop the key phasor. Two velocity probes are mounted on the motor.
The Turkey Point reactor coolant pump casings were electroslag welded. The following efforts were performed for quality assurance of the components.
- 1. The electroslag welding procedure employing two and three wire technique was qualified in accordance with the requirements of the ASME B&PV Code Section IX and Code Case 1355 plus supplementary evaluations as requested by WNES-PWRSD. The following test specimens were removed from an 8 inch thick and from a 12 inch thick weldment and successfully tested for both the 2 wire and the 3 wire techniques, respectfully.
They are:
A. Two wire electroslag process - 8" thick weldment.
- 1. 6 Transverse Tensile Bars - 750F post weld stress relief
- 2. 12 Guided Side Bend Test Bars B. Three wire electroslag process - 12" thick weldment
- 1. 6 Transverse Tensile Bars - 750F post weld stress relief
- 2. 17 Guided Side Bend Test Bars
- 3. 21 Charpy Vee Notch Specimens
- 4. Full section macroexamination of weld and heat affected zone.
- 5. Numerous microscopic examinations of specimens removed from the weld and heat affected zone regions.
- 6. Hardness survey across weld and heat affected zone.
C. A separate weld test was made using the 2 wire electroslag technique to evaluate the effects of a stop and restart of welding by this process. This evaluation was performed to establish proper procedures and techniques as such an occurrence was anticipated during production applications due to equipment malfunction, power outages, etc. The following test specimens were removed from an 8 inch thick weldment in the stop-restart-repaired region and successfully tested. They are:
- 1. 2 Transverse Tensile Bars - as welded
- 2. 4 Guided Side Bend Test Bars
- 3. Full section macroexamination of weld and heat affected zone.
4.2-9 Revised 07/15/2016
D. All of the weld test blocks in (A), (B) and (C) above were radiographed using a 24 Mev Betatron. The radiographic quality
level (as defined by ASTM E-94) obtained was between one-half of
1% to 1%. There were no discontinuities evident in any of the
electroslag welds.
- 1. The casting segments were surface conditioned for 100%
radiographic and penetrant inspections. The radiographic
acceptance standards were ASTM E-186 severity level 2 except
no category D or E defectiveness was permitted for section
thickness up to 4-1/2 inches and ASTM E-280 severity level 2
for section thicknesses greater than 4-1/2 inches. The
penetrant acceptance standards were ASME B&PV Code Section
III, paragraph N-627.
- 2. The edges of the electroslag weld preparations were machined.
These surfaces were penetrant inspected prior to welding.
The acceptance standards were ASME B&PV Code Section III, paragraph N-627.
- 3. The completed electroslag weld surfaces were ground flush with the casting surface. Then, the electroslag weld and adjacent
base material were 100% radiographed in accordance with ASME
Code Case 1355. Also, the electroslag weld surfaces and
adjacent base material were penetrant inspected in accordance
with ASME B&PV Code Section III, paragraph N-627.
- 4. Weld metal and base metal chemical and physical analyses were determined and certified.
- 5. Heat treatment furnace charts were recorded and certified.
4.2-10 Revised 04/17/2013 Pump Support Structure The reactor coolant pump support structures are described in Section 5 and
shown in Figure 5.1-20.
RCP Trip Criteria
RCS pressure, RCS subcooling and secondary pressure dependent RCS pressure (RCS - P Steam Line) are the most appropriate in providing pump trip discrimination between Loss of Coolant Accident (LOCA) and Steam Generator
Tube Rupture (SGTR) or non-LOCA events.
The RCS pressure does not meet the acceptance criteria for discrimination
between LOCA and SGTR or non-LOCA events since the secondary pressure
dependent RCS pressure trip parameter requires the reactor operator to look
at two instruments (RCS pressure and Steam Generator pressure).
The RCS subcooling is the desired parameter for pump trip since it only
requires the operator to look at one instrument (subcooled margin monitor).
The Subcooled Margin Monitor (SMM), as described in Section 4.2.10, is a
fully redundant, qualified system as required by TMI Action Item II.F.2, "Instrumentation for Detection of Inadequate Core Cooling". The SMM takes
input from redundant hot leg pressure transmitters and temperature elements.
These inputs are then used in a calculational program that determines RCS
subcooling which is then displayed in the control room.
It is desirable to keep the RCPs running during a Steam Generator Tube
Rupture (SGTR) and other non LOCAs to 1) maintain normal pressure control
using pressurizer spray and thereby avoiding opening of the pressurizer
PORVs, 2) prevent the formation of a stagnant water volume in the upper head
region which may flash and form a steam bubble during subsequent cooldown and
depressurization, 3) minimize potential pressurized thermal shock challenges
and 4) minimize operator action such as tripping the RCPs and then
restarting them later.
4.2-11 Revised 04/17/2013 The LOFTRAN computer code was used to perform the alternate RCP trip criteria analyses. Both Steam Generator Tube Rupture (SGTR) and non-LOCA event were
simulated in these analyses. Results for the SGTR analyses were used to
obtain all but three of the trip parameters. LOFTRAN is a Westinghouse
licensed code used for FSAR SGTR and non-LOCA analyses.
The following are considered to have the most impact on the determination of
the RCP trip criteria:
- 1. Break flow
- 2. SI flow
- 3. Decay heat
- 4. Auxiliary feedwater flow
The effects of all these uncertainties with the models and input parameters
were evaluated and it was concluded that the contributions from the break
flow conservatism and the SI uncertainty dominate. The calculated overall
uncertainty in the WOG analyses as a result of these considerations for the Turkey Point units is +1 to +5 F for the RCS subcooling RCP trip setpoint.
Due to the minimal effects from the decay heat model and AFW input, these results include only the effects of the uncertainties due to the break flow
model and SI flow inputs.
Manual Trip
Manual trip of an RCP motor requires the availability of 125V DC power, the
motor control switch, and the motor breaker. This provides a reliable means
of tripping the RCP. With the exception of the motor and cabling, all the
components associated with the RCP motor are outside containment. Therefore, adverse environmental conditions will not prevent RCP trip when required.
4.2-12 Revised 04/17/2013 Instrumentation Uncertainties for use of the RCS Subcooling for both Normal and Adverse Containment Conditions The minimum RCS pressure for SGTRs and non-LOCAs is approximately 1135 psig.
The subcooling uncertainty for normal containment conditions at this pressure is 22.3 F. At Residual Heat Removal (RHR) system pressure, 450 psig, the subcooling uncertainty is 25.5 F. Therefore, the trip setpoint for the RCPs under normal containment conditions is less than 25.5 F subcooling.
Under the adverse containment conditions, the RCP trip setpoint at the
non-LOCA SGTR lower pressure limit of 1135 psig was determined to be subcooling 65 F. The instrument uncertainties consider uncertainties from the transmitter or
temperature sensor, through the electronics to the display itself. While the
temperature sensors associated with the SMM are not sensitive to containment
conditions, the RCS pressure transmitters exhibit higher uncertainty under
adverse containment conditions. The permissive for using the adverse containment setpoint is either 180 F containment temperature or 1.3 x 10 5 R/hr.
The design of the Subcooled Margin Monitor (SMM) software is such that
invalid or failed instrument inputs, such as might be caused by pipe whip, are not used. The arrangement of instrumentation precludes failure of the
SMM due to pipe whip or single failure.
Pressurizer Relief Tank
Principal design parameters of the pressurizer relief tank are given in Table
4.1-3.
Steam discharged from the power relief and safety valves passes to the
pressurizer relief tank which is partially filled with water at or near
containment ambient conditions. The tank normally contains water in a
predominantly nitrogen atmosphere. Steam is discharged under the water level
to condense and cool by mixing with the water. The tank is equipped with a
spray, and a drain to the Waste Disposal System, which are operated to cool
the tank following a discharge.
4.2-13 Revised 04/17/2013 The tank size is based on the requirement to condense and cool a discharge of pressurizer steam equivalent to 110 percent of the volume above the full
power water level setpoint.
The tank is protected against a discharge exceeding the design value by two
rupture discs which discharge into the reactor containment. The two rupture
discs on the pressurizer relief tank have the combined relief capacity to
relieve the total capacity of the three pressurizer safety valves. The tank
design pressure (and the rupture disc setting) is twice the calculated
pressure resulting from the maximum safety valve discharge described above.
This margin is to prevent deformation of the disc. The tank and rupture disc
holder are also designed for full vacuum to prevent tank collapse if the tank
contents cool without nitrogen being added.
The discharge piping from the safety and relief valves to the relief tank is
sufficiently large to prevent backpressure at the safety valves from
exceeding 20 per cent of the set point pressure at full flow.
The pressurizer relief tank, by means of its connection to the Waste Disposal
System, provides a means for removing any non-condensable gases from the
Reactor Coolant System which might collect in the pressurizer vessel.
The tank is constructed of carbon steel with a corrosion resistant coating on
the internal surface.
Piping The general arrangement of the reactor coolant system piping is shown on the
layout drawings in Section l. Piping design data are presented in Table
4.1-6.
The austenitic stainless steel reactor coolant piping and fittings which make
up the loops are 29 in. ID in the hot legs, 27-1/2 in. ID in the cold legs
and 31 in. ID between each loop's steam generator outlet and its reactor
coolant pump suction. The pressurizer relief line, which connects the
pressurizer safety and relief valves' outlets to the inlet nozzle flange on
the pressurizer relief tank, is constructed of carbon steel.
Smaller piping, including the pressurizer surge and spray lines, drains and
connections to other systems are austenitic stainless steel. All joints and
connections are welded except for stainless steel flange connections to the
carbon steel pressurizer relief tank and the connections at the relief and
safety valves.
4.2-14 Revised 04/17/2013 C26 The Turkey Point 90 elbows were electroslag welded. The following efforts were performed for quality assurance of these components.
- 1. The electroslag welding procedure employing one wire technique was qualified in accordance with the requirements of ASME B&PV Code Section
IX and Code Case 1355 plus supplementary evaluations as requested by
W NES-PWRSD. The following test specimens were removed from a 5 inch thick weldment and successfully tested. They are:
- a. 6 Transverse Tensile Bars - as welded
- b. 6 Transverse Tensile Bars - 2050 F, H 2 O Quench c. 6 Transverse Tensile Bars - 2050 F, H 2 O Quench + 750 stress relief heat treatment
- d. 6 Transverse Tensile Bars - 2050 F, H 2 O Quench, tested at 650 F e. 12 Guided Side Bend Test Bars
- 2. The casting segments were surface conditioned for 100% radiographic and penetrant inspections. The acceptance standards were ASTM E-186
severity level 2 (except no category D or E defectiveness was permitted)
and USAS Code Case N-10, respectively.
- 3. The edges of the electroslag weld preparations were machined. These surfaces were penetrant inspected prior to welding. The acceptance
standards were USAS Code Case N-10.
- 4. The completed electroslag weld surfaces were ground flush with the casting surface. Then, the electroslag weld and adjacent base material
were 100% radiographed in accordance with ASME Code Case 1355. Also, the electroslag weld surfaces and adjacent base material were penetrant
inspected in accordance with USAS Code Case N-10.
- 5. Weld metal and base metal chemical and physical analysis were determined and certified.
- 6. Heat treatment furnace charts were recorded and certified.
4.2-15 Revised 04/17/2013
Thermal sleeves are installed at the following locations where high thermal stresses could otherwise develop due to rapid changes in fluid temperature
during normal operational transients:
a) Return line from the residual heat removal loop.
b) Both ends of the pressurizer surge line.
c) Pressurizer spray line connection to the pressurizer.
d) Charging lines and auxiliary charging line connections.
Valves All valve surfaces in contact with reactor coolant are austenitic stainless
steel or equivalent corrosion resistant materials. Connections to stainless
steel piping are welded. Valves that perform a modulating function may be
equipped with two sets of packing and an intermediate leakoff connection.
4.2.3 PRESSURE-RELIEVING DEVICES
The Reactor Coolant System is protected against overpressure by control and
protective circuits such as the high pressure trip and by code relief valves
connected to the top head of the pressurizer. The relief valves discharge
into the pressurizer relief tank which condenses and collects the valve
effluent. The schematic arrangement of the relief devices is shown in Figure
4.2-1, and the valve design parameters are given in Table 4.1-3. Valve sizes
are determined as indicated in Section 4.3.4. Power-operated relief valves
and code safety valves are provided to protect against pressure surges which
are beyond the pressure limiting capacity of the pressurizer spray. Each
pressurizer safety valve has an acoustic accelerometer mounted on the
discharge of the valve to provide the control room operator with positive
indication of the pressurizer safety valve position.
The pressurizer relief tank is protected against a steam discharge exceeding
the design pressure value by two rupture discs which discharges into the
reactor containment. The rupture disc relief conditions are given in Table
4.1-3.
4.2-16 Revised 04/17/2013
4.2.4 PROTECTION AGAINST PROLIFERATION OF DYNAMIC EFFECTS
Engineered Safety Features and associated systems are protected from loss of
function due to dynamic effects and missiles which might result from a
loss-of-coolant accident. Protection is provided by missile shielding and/or
segregation of redundant components. This is discussed in Section 6.1.
The Reactor Coolant System is surrounded by concrete shield walls. These
walls provide shielding to permit access into the containment during full
power operation for inspection and maintenance of miscellaneous equipment.
These shielding walls also provide missile protection for the containment
liner plate.
The concrete deck over the Reactor Coolant System also provides for shielding
and missile damage protection.
Steam generator lateral bracing is provided near the upper tube support
elevation to resist lateral loads, including those resulting from seismic
forces and pipe rupture forces. Additional bracing is provided at a lower
elevation to resist pipe rupture loads.
The NRC documents in their letter of November 28, 1988 (Reference 3) that the
leakage detection systems at Turkey Point Units 3 and 4 satisfy the
requirements of Generic Letter 84-04, and that the primary loop piping
complies with the criteria of GDC 4 from 10 CFR 50, Appendix A. GDC 4 allows
the use of plant-specific Leak-Before-Break analysis to eliminate the dynamic
effects of postulated pipe ruptures in high energy piping from the design
basis of a plant. Plants with an NRC-approved Leak-Before-Break analysis may
remove pipe whip restraints and jet impingement barriers. Turkey Point Units
3 and 4 received NRC approval (Reference 4) for elimination of the dynamic
effects of postulated pipe ruptures in reactor coolant piping from the design
basis of the plant. The Turkey Point analysis for the Leak-Before-Break
Methodology is documented in the Westinghouse report WCAP-14237 (Reference
5). Therefore, the dynamic loads associated with a rupture of the reactor
coolant piping need not be considered in the design of the reactor support
structures.
Missile protection afforded by the arrangement of the Reactor Coolant System
is illustrated in the containment structure drawings which are given in
Section 5.
4.2-17 Revised 04/17/2013 4.2.5 MATERIALS OF CONSTRUCTION
Each of the materials used in the Reactor Coolant System is selected for the
expected environment and service conditions. The major component materials
are listed in Table 4.2-1. Those pressure-containing or strength-bearing
stainless steel component parts in the reactor vessel and associated reactor
coolant systems that have become furnace sensitized during the fabrication
sequence are listed in Table 4.2-4.
All reactor coolant system materials which are exposed to the coolant are
corrosion-resistant. They consist of stainless steels and Inconel, and they
are chosen for specific purposes at various locations within the system for
their superior compatibility with the reactor coolant. During mode 1, the
chemical composition of the reactor coolant is maintained within the
specification given in Table 4.2.2. Reactor coolant chemistry is further
discussed in Section 4.2.8.
In Mode 1 the water in the secondary side of the steam generators is normally
maintained within the chemistry parameters given in Table 4.2-3 to control
deposits and corrosion inside the steam generators. Specific operating
chemistry specifications and limits are maintained as outlined in the Nuclear
Chemistry Parameters Manual and the Turkey Point Chemistry Procedures.
The phenomena of stress-corrosion cracking and corrosion fatigue are not
encountered unless a specific combination of conditions is present. The
necessary conditions are a susceptible alloy, an aggressive environment, stress, and time.
It is characteristic of stress corrosion that combinations of alloy and
environment which result in cracking are usually quite specific.
Environments which have been shown to cause stress-corrosion cracking of
stainless steels are free alkalinity in the presence of a concentrating
mechanism and the presence of chlorides and free oxygen. With regard to the
former, experience has shown that deposition of chemicals on the surface of
tubes can occur in a steam blanketed area within a steam generator. In the
presence of this environment, stress-corrosion cracking can occur in
stainless steels having the nominal residual stresses resulting from normal
manufacturing procedures. However, the steam generator contains Inconel
tubes. Testing to investigate the susceptibility of heat exchanger
construction materials to stress corrosion in caustic and chloride aqueous
solutions has indicated that Inconel alloy has excellent resistance to
general and pitting-type corrosion in severe operating water conditions.
4.2-18 Revised 04/17/2013 Considerable experience with Inconel in steam generator and heat exchanger applications has been accumulated in the industry. Since 1962, widespread
adoption of Inconel for steam generator tubes in nuclear stations is evident:
as for example, Connecticut-Yankee; San Onofre; PM-1, Sundance; PM-3A, McMurdo Sound; CVTR; NPD, and Hanford N-Reactor. In none of these plants
has there been any evidence of steam generator tube leakage. Materials with
lead traces in the overall composition were present in the secondary side of
the referenced plants. The use of lead in the materials of the secondary
side of this plant has been minimized to the practical limit of that
occurring as trace elements in metallurgical alloys and, as such, is
insignificant.
All external insulation of Reactor Coolant System components is compatible
with the component materials. The cylindrical shell exterior and closure
flanges to the reactor vessel are insulated with metallic reflective
insulation. The Unit 3 & 4 closure heads are insulated with a self
supporting panel insulation arrangements that are constructed of a metallic
reflective material and woven fiberglass blanket. All other external
corrosion-resistant surfaces in the Reactor Coolant System are insulated with
low halide or halide-free insulating material as required.
The Nil Ductility Transition Temperature (NDTT) of the reactor vessel plate
or forging material opposite the core is established at a Charpy V-notch test
value of 30 ft-1b or greater. The material is tested to verify conformity to
specified requirements and to determine the actual NDTT value. In addition, this plate was initially 100 per cent volumetrically inspected by ultrasonic
test using both longitudinal and shear wave methods.
The remaining material in the reactor vessel, and other Reactor Coolant
System components, meets the appropriate design code requirements and
specific component function.
The reactor vessel material is heat-treated specifically to obtain good notch-ductility which ensures a low NDTT, and thereby gives assurance that
the finished vessel can be initially hydrostatically tested and operated as
near to room temperature as possible without restrictions. The stress limits
established for the reactor vessel are dependent upon the temperatures at
which the stresses are applied. As a result of fast neutron irradiation in
the region of the core, the material properties will change, including an
increase in the NDTT. During fabrication initial maximum values of NDTT
have been established at 40 0 F for No. 3 vessel and 50 o F for No. 4 vessel forgings.
4.2-19 Revised 04/17/2013 C25 Framatome ANP Topical Report BAW-2308, Revisions 1A and 2A (References 10 and
- 11) provide new initial weld materials properties. The NRC approved the
exemption request to use these values in a letter dated March 11, 2010 (Reference 12).
The techniques used to measure and predict the integrated fast neutron (E > 1
Mev) fluxes at the sample locations are described in Appendix 4A. The
calculation method used to obtain the maximum neutron (E > 1 Mev) exposure of
the reactor vessel is identical to that described for the irradiation
samples. Since the neutron spectra at the sample can be applied with
confidence to the adjacent section of reactor vessel, the maximum vessel
exposure will be obtained from the measured sample exposure by appropriate
application of the calculated azimuthal neutron flux variation.
At uprated conditions, the maximum integrated fast neutron (E>1 Mev) exposure of the vessel was computed to be 6.377 x 10 19 n/cm 2 at the end of the extended license terms of 48 EFPY*, approximately (Reference 6). Under the same conditions, the maximum vessel exposure at the limiting circumferential vessel weld is predicted to be 5.739 x 10 19 n/cm 2 at the end of the extended license terms of 48 EFPY*, approximately (Reference 6)**. The predicted extended end of life RT(ndt) is less than the 10CFR50.61 screening criteria (Reference 6).
To evaluate the RT(ndt) shift of welds, heat affected zones and base material
for the vessel, test coupons of these material types have been included in
the reactor vessel surveillance program described in Section 4A.
- This value is approximate and will change from year to year based on the unit availability. Fluence prediction is acceptable in the +/-20% range, so
this value can easily vary within that limit.
- After the (hafnium) pressurized thermal shock absorbers were removed from the vessel cores in 2009, the maximum vessel exposure at the limiting
circumferential weld is predicted to be 5.739 x 10 19 n/cm 2 at the end of the extended license terms of approximately 48 EFPY (Reference 6). The NRC was
notified of this proposed change as captured in Reference 9. The NRC
approved the changes as documented in Reference 12.
4.2-20 Revised 04/17/2013 C26C26 The methods used to measure the initial NDTT of the reactor vessel base plate material are given in Appendix 4A.
4.2.6 MAXIMUM HEATING AND COOLING RATES
The reactor system operating cycles used for design purposes are given in
Table 4.1-8 and described in Section 4.1.5. During unit heatup and cooldown, the rates of temperature and pressure changes are limited. The system design heatup and cooldown rate of 100 F per hour satisfies stress limits for cyclic operation (ASME B&PV Code,Section III) and is consistent with the expected number of cycles. However, the normal system heatup and cooldown rate is administratively limited to less than or equal to 90 F per hour. Sufficient electrical heaters are installed in the pressurizer to permit a heatup rate, starting with a minimum water level, of 55 F per hour. This rate takes into account the small continuous spray flow provided to maintain the pressurizer liquid homogeneous with the coolant.
For the pressurizer, the allowable heatup rate is 100 F per hour and the maximum cooldown rate for the pressurizer is 200 F per hour. The stresses are within acceptable limits for the anticipated usage. A maximum temperature
difference (T) of 320 F between the pressurizer and reactor coolant system is specified up to a maximum pressurizer temperature of 500 F (Reference 1 and 2). This allows steam bubble formation at an earlier time during startup to reduce the chances of an overpressure event by reducing the period during which the plant is solid. At pressurizer temperature greater than 500 F, T is specified as 200 F with a minimum of 100 F. Spray actuation transients during the condition of T greater than 100 F shall be limited to those in Table 2-2, Figure 2-1 and Figure 2-5 in Reference 1.
The fastest cooldown rates which result from the hypothetical case of a break
of a main steam line are discussed in Section 14.
4.2-21 Revised 04/17/2013 4.2.7 LEAKAGE
The existence of leakage from the Reactor Coolant System to the containment
regardless of the source of leakage, is detected by one or more of the
following conditions:
a) Two radiation sensitive instruments provide capability for detection of leakage from the Reactor Coolant System. The containment air
particulate monitor is quite sensitive to low leak rates. The
containment radiogas monitor is much less sensitive but can be used as a
backup to the air particulate monitor.
b) An increase in the amount of coolant makeup water which is required to maintain normal level in the pressurizer, or an increase in containment
sump level are less sensitive means of detection leakage.
c) One radiation sensitive instrument provides capability to detect Reactor Vessel Head Leakage. The leak detection system draws a sample from the
Reactor Head Area or containment atmosphere in a skid mounted
particulate sampling system located inside containment.
Leakage detection methods are described in detail and evaluated in Section
6.5.
Leakage Prevention
Reactor Coolant System components are manufactured to exacting specifications
which exceed normal code requirements (as listed in Table 4.1-9). In
addition, because of the welded construction of the Reactor Coolant System
and the extensive non-destructive testing to which it is subjected (as
outlined in Section 4.4), it is considered that leakage through metal
surfaces or welded joints is very unlikely.
However, some leakage from the Reactor Coolant System is permitted by the
reactor coolant pump seals. Also, all sealed joints are potential sources of
leakage even though the most appropriate sealing device is selected in each
case. Thus, because of the large number of joints and the difficulty of
assuring complete freedom from leakage in each case, a small integrated
leakage is considered acceptable. Leakage from the reactor through its head
flange will leak-off between the double O-ring seal and actuate an alarm in
the control room.
4.2-22 Revised 04/17/2013 Locating Leaks Experience has shown that hydrostatic testing is successful in locating leaks
in a pressure containing system.
Methods of leak location which can be used during shutdown include visual
observation for escaping steam or water or for the presence of boric acid
crystals near the leak. The boric acid crystals are transported outside the
Reactor Coolant System in the leaking fluid and deposited by the evaporation
process.
4.2.8 WATER CHEMISTRY
The water chemistry is selected to provide the necessary boron content for
reactivity control and to minimize corrosion of reactor coolant system
surfaces.
All materials exposed to reactor coolant are corrosion resistant. During
mode 1, periodic analyses of the coolant chemical composition are performed
to monitor the adherence of the system to the reactor coolant water quality
listed in Table 4.2-2. Chemistry specifications and limits for shutdown and
startup conditions are maintained as outlined in the Florida Power and Light
Nuclear Chemistry Parameters Manual and the Turkey Point Chemistry
Procedures. Maintenance of the water quality to minimize corrosion is
accomplished using the Chemical and Volume Control System and Sampling System
which are described in Section 9.
4.2.9 REACTOR COOLANT FLOW MEASUREMENTS
Elbow taps are used in the reactor coolant system as an instrument device
that indicates the status of the reactor coolant flow. The basic function of
this device is to provide information as to whether or not a reduction in
flow rate has occurred. The correlation between flow reduction and elbow tap read out has been well established by the following equation; 2 PP where P o is the referenced pressure differential with the corresponding referenced flow rate o and P is the pressure differential with the corresponding flow rate . The full flow reference point is established during initial unit startup. The low flow trip point is then established by extrapolating along the correlation curve.
4.2-23 Revised 04/17/2013
The technique has been well established in providing core protection against low coolant flow in Westinghouse PWR plants. The expected absolute accuracy of the channel is within +/-10% and field results have shown the repeatability
of the trip point to be within +/-1%. The analysis of the loss of flow
transient presented in Section 14.1.9 assumes instrumentation error of
+ 3.7%.
4.2.10 REACTOR COOLANT SUBCOOLED MARGIN MONITOR
The reactor coolant system subcooled margin monitor system is an on-line
microcomputer based system which uses reactor coolant process signals to
provide a continuous indication of the margin from saturation conditions.
The subcooled margin monitor system also provides an alarm signal into the
main control room annunciator.
The reactor coolant system parameters monitored are the three coolant loops
hot leg temperature, and loops A and B hot leg pressure. The operator has
the choice of continuous main control board indication of either the pressure
or temperature margin from saturation.
The temperature sensors are dual RTD's installed in thermowells. These RTD's
are connected to provide the subcooling margin monitor system computing
module with a 4-20 ma dc signal.
The reactor coolant pressure transmitters also provide a 4-20 ma dc signal to
the computing module.
The computing module selects the highest temperature from those provided and
the lowest pressure and calculates the margin to saturation from those two
readings. The readings then appear on the display module in the control
room.
4.2-24 Revised 04/17/2013 4.2.11 REACTOR COOLANT VENT SYSTEM
The RCS vent system provides the operator with a means to vent non-condensable gases from the Reactor Coolant System. As shown on Figure 4.2-1 and 4.2-5, the RCS can be vented separately through the reactor vessel head vent or from the pressurizer steam space via the pressurizer relief line.
To vent system discharges to the containment Atmosphere and/or the pressurizer relief tank. A housekeeping drain is provided to the containment sump.
The RCS vent system can vent one-half of the RCS volume (gas) in one hour at operating pressure, but is sized such that the RCS mass inventory will be maintained by the charging pumps should the vent line suffer a guillotine break.
The power for the vent valves is taken from vital DC power outside the containment. The control power fuses are normally removed to prevent inadvertent operation of the vent valves under postulated fire conditions (see Engineering Guidelines for Fire Protection for Turkey Point Units 3 & 4 (Reference 14)). Valve control and position indication is located in the control room. Pressure indication is provided in the control room to assist the operator in determining leakage in the vent line. Each vent is powered from an emergency bus.
The vent system has been seismically analyzed.
4.2.12 REACTOR VESSEL DRAINDOWN LEVEL INDICATION SYSTEM
The reactor vessel drain down level indication system (see Figure 4.2-1) provides the continuous measurement of reactor coolant level during drain down operations and while in a drain down condition. The system consists of two independent and redundant level (differential pressure) transmitters with control room indication. This system provides audible and visual annunciation in the Control Room on decreasing reactor level below a preset value. Additionally, this system provides audible and visual annunciation in the Control Room on decreasing reactor level below an adjustable value or increasing reactor water above an adjustable value. This system also provides a local audio alarm (horn) and light (located at each steam generator manway) on increasing reactor level above a preset value.
4.2-25 Revised 09/20/2016 C28 4.2.13 REFERENCES
- 1. Westinghouse Electric Corporation, Report Number STC-TR-85-003 dated February 8, 1985, "Structural Evaluation - Pressurizer Surge Line and Spray Line for Pressurizer/RCS Differential Temperature of 320F," PROPRIETARY.
- 2. Safety Evaluation, JPE-M-85-013, dated June 13, 1985, "Increased T between Pressurizer and Reactor Coolant System to 320F for PTP Unit 3." 3. NRC Letter, from G.E. Edison (NRC) to W.F. Conway (FPL), "Turkey Point Units 3 and 4 - Generic Letter 84-04, Asymmetric LOCA Loads," dated November 28, 1988.
- 4. NRC Letter, from R. P. Croteau (NRC) to J. H. Goldberg (FPL), "Turkey Point Units 3 and 4 - Approval to Utilize Leak-Before-Break Methodology for Reactor Coolant System Piping (TAC Nos. M91494 and M91495)," dated June 23, 1995.
- 5. Westinghouse WCAP-14237, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Turkey Point Units 3 and 4 Nuclear Power Plants," dated December 1994.
- 6. Westinghouse WCAP-17152-P, "Turkey Point Units 3 and 4 Extended Power Uprate (EPU) Engineering Report," Revision 0, dated August, 2012.
- 7. DELETED
- 8. Westinghouse Electric Company, Report Number MTLS-06-0131, Revision 3, "Westinghouse Supplement to EPRI PWR Primary Water Chemistry Guidelines Rev.6," dated March 6, 2008.
- 9. Letter to NRC from William Jefferson, Turkey Point, dated March 18, 2009, "Update to NRC Reactor Vessel Integrity Database and Exemption Request for Alternate Material Properties Bases per 10 CFR 50.12 and 10 CFR 50.60(b)".
- 10. Framatome ANP Topical Report BAW-2308, Revision 1A, "Initial RTndt of Linde 80 Weld Materials", Approved August 2005.
- 11. Framatome ANP Topical Report BAW-2308, Revision 2A, "Initial RTndt of Linde 80 Weld Materials", Approved March 2008.
- 12. Letter from Jason Paige, NRC, to Mano Nazar, FPL, "Turkey Point Units 3 and 4 - Exemption from the Requirements of 10 CFR part 50, Appendix G and 10 CFR Part 50, Section 50.61 (TAC Nos. ME 1007 and ME 1008)", March 11, 2010.
- 13. EC 281319 Unit 3 Pressurizer Heater #11 Element Nozzle/Sleeve Repair and Heater Replacement. Unit 3 pressurizer modification EC 281319 replaced the #11 heater well nozzle with a half nozzle design welded to the outside of the pressurizer shell instead of the internal cladding (Ref. NRC Relief Request SER:ADAMS Access1on No.: ML 15271A325.)
- 14. STD-M-006, Engineering Guidelines for Fire Protection for Turkey Point Units 3 & 4.
4.2-26 Revised 09/20/2016 C28 TABLE 4.2-1 Sheet 1 of 2 MATERIALS OF CONSTRUCTION OF TRE REACTOR COOLANT SYSTEM COMPONENTS
Component Section Materials
Reactor Vessel Pressure Plate SA-302, Gr. B
Shell & Nozzle Forgings A-508 Class II
Cladding, Stainless Weld Rod Type 304 equivalent
Thermal Shield and Internals A-240, Type 304
Instrument Tubes Inconel
Replacement RVCHs CRDM Housing Bodies Inconel Alloy 690 Head & Flange Mono Forging SA-508, Class 3
Insulation SS-A1 Foil-SS
Steam Generator Plate (shell course) SA-533 Grade A Class 2
Tube Sheet Forging SA-508 Class 2a
Channel Head Casting SA-216 Grade WCC
Support Plates SA-240 Type 405
Channel Head Cladding Stainless Steel, Type 304 or equivalent
Tube Sheet Cladding Inconel
Tubes SB-163 Thermally
Treated
Pressurizer Shell SA-302, Gr. B
Heads SA-216 WCC
External Plate SA-302, Gr. B
Cladding, Stainless Type 304 equivalent
Internal Plate SA-240 Type 304
Internal Piping SA-376 Type 316
Pressurizer Relief Shell A-285 Gr. C
Tank Heads A-285 Gr. C
Internal Coating Vinyl
Revised 07/14/2005 TABLE 4.2-1 Sheet 2 of 2
Component Section Material
Piping Pipes A-376 Type 316
Fittings A-351, CF8M
Nozzles A-182 F316
Pump Shaft Type 304
Impeller A-351, CF8
Locknut Type 304
Casing A-351, CF8M
Bearings Stellite and
graphitar
Seals Silicon Carbide
Valves Pressure Containing Parts A-351, CF8M and A-182 F316
Shafts, stems l7-4PH or equivalent
Hard surfacing Stellite 6 or equivalent
Bushings, bearings Cast Stellite 6 or
equivalent
Springs Alloy 600 or equivalent corrosion resistant
material
Misc. Fasteners and washers 410 and 416 Series
General Note: This table represents original materials. Approved equivalents may be installed as necessary to support on-going maintenance.
Revised 07/15/2016 C28 TABLE 4.2-2
REACTOR COOLANT WATER CHEMISTRY SPECIFICATION
(Mode 1 Normal Values)
Electrical Conductivity Determined by the concentration of boric acid and alkali present.
Expected range is <1 to 40 uMhos/cm at 25 o C.
Solution pH Determined by the concentration of boric acid and alkali present.
Expected values range between 4.2 (high boric acid concentration) to 10.5 (low boric acid concentration) at 25 o C.
Oxygen, ppm, max. 0.1
Chloride, ppm, max. 0.15
Fluoride, ppm, max. 0.15
Hydrogen, cc (STP)/kg H 2 0 25 - 50 (15-50 no more than 2 days prior to shutdown)
Total Suspended Solids, ppm, max.
1.0 pH Control Agent (Li), ppm 0.2 - 3.50*
Boric Acid as ppm B Variable from 0 to approximately 4000
NOTES:* Lithium concentrations at hot zero power critical conditions can be up to 5.0 ppm, but should be less than 3.5 ppm once equilibrium xenon levels are reached (Reference 8)
04/29/2009 C24C24 TABLE 4.2-3 STEAM GENERATOR WATER (STEAM SIDE) CHEMISTRY SPECIFICATION MODE 1 AVT NORMAL VALUES
pH at 25 o C > 9.0 Cation Conductivity, mhos/cc
< 0.8 Na, ppm
< 0.020 Cl, ppm
< 0.020 Si0 2 , ppm < 0.3 Sulfate, ppm
< 0.020 Blowdown Rate As necessary to maintain steam generator chemistry. However, a continuous blow-down is recommended.
Rev. 8 7/90 TABLE 4.2-4
FURNACE SENSITIZED RCS STAINLESS STEEL COMPONENTS
- 1. Reactor Vessel
- a. CRDM housings - Inconel 690 bodies with Type 304 CRDM adapters. The replacement RVCH CRDM adapters are not sensitized.
- b. Bottom instrumentation nozzles - Inconel with 304 safe ends.
- c. Six (each vessel) primary nozzle safe ends - First layer is type 309; balance of layers is type 308 weld metal deposit.
- d. Gasket monitor tubes (not under pressure) - Type 304.
Two primary nozzle safe ends per generator - Type 309 first layer with balance of safe end of 308L weld.
- 3. Pressurizer
All nozzle safe ends in top and bottom head - Type 316, inlet
nozzle-forged (Al82), balance-pipe (A312).
NOTE: A. Reactor coolant piping field welds are Type 304 (A371) with filler of 308L (A298), and pass temperature was held <350F.
The piping was water quenched during manufacture.
B. Core support structure heat treatment:
Barrel welds - 165 o F, furnace cooled.
Other assembly welds - local heating to 75 o F, air cooled.
Revised 07/14/2005
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-1
REFER TO ENGINEERING DRAWING 5613-M-3041 , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 REACTOR COOLANT SYSTEM FIGURE 4.2-1
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-2 PART 1
REFER TO ENGINEERING DRAWING 5610-M-400-4, SHEET 1
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 ARRANGEMENT OF REACTOR VESSEL LONGITUDINAL SECTION FIGURE 4.2-2 PART 1 05/07/2007 C22
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-2
PART 2
REFER TO ENGINEERING DRAWING 5610-M-400-5, SHEET 1
05/07/2007 C22FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 ARRANGEMENT OF REACTOR VESSEL LONGITUDINAL SECTION FIGURE 4.2-2 PART 2
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-2
PART 3
REFER TO ENGINEERING DRAWINGS 5613-M-460-2, SHEET 1 5614-M-460-2, SHEET 1
05/07/2007 C22FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 SPECIFICATION DRAWING FOR REPLACEMENT REACTOR VESSEL CLOSURE HEAD SECTION FIGURE 4.2-2 PART 3
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-2 PART 4
REFER TO ENGINEERING DRAWINGS 5613-M-460-2, SHEET 2 5614-M-460-2, SHEET 2
C22FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 REPLACEMENT REACTOR VESSEL CLOSURE HEAD AS-BUILT DIMENSIONS FIGURE 4.2-2 PART 4 05/07/2007
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-9
REFER TO ENGINEERING DRAWING 5613-M-3041 , SHEET 2
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 REACTOR COOLANT SYSTEM FIGURE 4.2-9
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-10 REFER TO ENGINEERING DRAWING 5613-M-3041 , SHEET 3
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 REACTOR COOLANT SYSTEM REACTOR COOLANT PUMPS FIGURE 4.2-10
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-11
REFER TO ENGINEERING DRAWING 5613-M-3041 , SHEET 4
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 REACTOR COOLANT SYSTEM PORV CONTROL FIGURE 4.2-11
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-12 REFER TO ENGINEERING DRAWING 5614-M-3041 , SHEET 1 REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 4 REACTOR COOLANT SYSTEM FIGURE 4.2-12
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-13
REFER TO ENGINEERING DRAWING 5614-M-3041 , SHEET 2
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 4 REACTOR COOLANT SYSTEM FIGURE 4.2-13
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-14
REFER TO ENGINEERING DRAWING 5614-M-3041 , SHEET 3
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 4 REACTOR COOLANT SYSTEM REACTOR COOLANT PUMPS FIGURE 4.2-14
FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-15
REFER TO ENGINEERING DRAWING 5614-M-3041 , SHEET 4
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 4 REACTOR COOLANT SYSTEM PORV CONTROL FIGURE 4.2-15
4.3 SYSTEM DESIGN EVALUATION 4.3.1 SAFETY FACTORS
The safety of the reactor vessel and all other Reactor Coolant System pressure
containing components and piping is dependent on several major factors
including design and stress analysis, material selection and fabrication, quality control and operations control.
Reactor Vessel
A stress evaluation of the reactor vessel has been carried out in accordance
with the rules of Section III of the ASME Nuclear Vessel Code. The evaluation
demonstrates that stress levels are within the stress limits of the Code.
Table 4.3-1 presents a summary of the results of the stress evaluation.
Figures 4.3-1, 2, and 3 illustrate the areas of the pressure vessel that are
analyzed in detail through systematic analytical procedures. The maximum
thermal stress due to gamma ray heating occurs in the cylindrical portion of
the vessel adjacent to the core and its value is about 2200 psi and is
considered negligible.
A summary of fatigue usage factors for components of the reactor vessel is
given in Table 4.3-2. The effect of gamma ray heating on the cumulative usage
factor is negligible.
The cycles specified for the fatigue analysis are the results of an evaluation
of the expected station operation coupled with experience from nuclear power
plants now in service, such as Yankee-Rowe. These cycles include five heatup
and cooldown cycles per year, a conservative selection when the vessel may not
complete more than one cycle per year during normal operation.
The vessel design pressure is 2485 psig while the normal operating pressure
will be 2235 psig. The resulting operating membrane stress is therefore amply
below the code allowable membrane stress to account for operating pressure
transients. To preclude the possibility of brittle failure the stresses allowed in the vessel in relation to operation below NDTT and DTT (NDTT+60 F) are:
4.3-1 Rev. 3-7/85
- 1. At DTT; a maximum stress of 20% yield. 2. From DTT to DTT minus 200 F; a maximum stress decreasing from 20% to 10%
yield. 3. Below DTT minus 200 F; a maximum stress of 10% yield.
These limits are based on the data reported (1,2), which show that if the
stresses are maintained within the above limits, brittle fracture does not
occur. These stress limits are maintained by prescribing operating procedures
which rely upon administrative pressure and temperature control during heatup
and cooldown as described in Reference 3, and by actuation of the overpressure
mitigating system (OMS). The OMS is enabled/disabled by the operator during
heatup and cooldown. The OMS is enabled by procedure when the RCS is less
than 300 0 F and prior to operating without a bubble in the pressurizer. The OMS varies the setpoint of the power operated relief valve as the RCS temperature
varies. Above 285 o F, the OMS setpoints increase from 440 psig at 285 o F to 2335 psig at 554 o F as a segmented curve. Below 285 o F, the setpoint is maintained at a constant 440 psig, while above 554 0 F (up to 750 0 F) the setpoint is maintained at a constant 2335 psig (Ref. 10). Technical Specifications LCO 3.4.9.3a
indicates a PORV lift setting of <
448 psig, however, the field device for OMS setpoint actuation will be set at 440 psig to provide buffer from the
Technical Specifications value.
The actual shift in RT(ndt) will be established periodically during unit
operation by testing of vessel material samples which are irradiated
cumulatively by securing them near the inside wall of the vessel in the core
area. To compensate for any increase in the RT(ndt) caused by irradiation, the limits given in the unit operating manual on the pressure-temperature
relationship are periodically changed to stay within the stress limits, which
will be stated above during heatup and cooldown.
The vessel closure contains fifty-eight, 6-inch diameter studs. The stud
material is ASTM A-540 with minimum yield strength of 104,400 psi at design
temperature. Combined membrane stresses at design conditions of maximum
calculated stress intensities are below the Code allowables. The membrane
stress in the studs, when they are at the steady state operational conditions, is less than one half of the minimum yield strength.
4.3-2 Revised 04/17/2013 C26 The normal operating temperature always exceeds even the highest anticipated
DDT during the life of the unit. Thus, the emphasis of conservative operation
is placed on heatup and cooldown because long term irradiation of the vessel
raises the DDT and thereby limits the heatup or cool down rates. The
conservatism in setting up the temperature-pressure relationship limits stated
above are:
- 1. Use of a stress concentration factor of 4 on assumed flaws in calculating the stresses. 2. Use of nominal yield of material instead of actual yield.
- 3. Neglecting the increase in yield strength resulting from radiation effects.
The factor of four in Item 1 is not an actual stress concentration factor such
as described in Article 4 Design of Section III but is a margin of
conservatism based on the Fracture Analysis Diagram in ASTM E208 as well as
the stress limits maintained by the prescribed operating procedures which rely
upon administrative pressure and temperature control during heatup and
cooldown as described in ASTM Paper No. 63-WA-100 "Reactor Vessel Design
Considering Radiation Effects", L. Porse. At the DTT the stresses are 20% of
the yield strength versus a prescribed upper limit of 80% of the yield
strength; therefore at this point there is a margin of four (80%/20%).
Since the Fracture Analysis Diagram is based on a plot of nominal stress
versus temperature and different size flaws (cracks) are assumed, the use of
actual stress concentration factors do not apply.
As part of the operator training program Westinghouse instructs supervisory
and operating personnel in reactor vessel design, fabrication and testing as
well as present and future precautions necessary for pressure testing and
operating modes. The need for record keeping is stressed, such records being
helpful for future summation of time at power level and temperature which
tends to influence the irradiated properties of the material in the core
region. These instructions are incorporated in the operating manuals.
4.3-3 Revised 04/17/2013 Steam Generators
- The Westinghouse analysis of the steam generator tubesheets is included as part of the Stress Report requirement for Class I Nuclear Pressure Vessels.
The evaluation must be based on the stress and fatigue limitations outlined in
Article-4 Design of Section III.
Calculations confirm that the steam generator tube sheet will withstand the loading (which is a quasi-static rather than a shock loading) by loss of
reactor coolant. The maximum primary membrane plus primary bending stress in
the tube sheet under these conditions is less than 35,550 psi per design
stress. This is well below ASME Section III allowable stress of 45,000 psi at 650 F. Because the pressure in the primary channel head would drop to zero under the condition postulated, no damage will result to the channel head.
The rupture of primary or secondary piping has been assumed to impose a maximum pressure differential of 2485 psi across the tubes and tube sheet from
the primary side or maximum pressure differential of 1100 psi across the tubes
and tube sheet from the secondary side, respectively. Under these conditions
there is no rupture of the primary to secondary boundary (tubes and tube
sheet). This criterion prevents any violation of the containment boundary.
The tube sheet has been designed to accept a primary to secondary pressure differential of 1700 psi, which is the specified primary side design pressure differential. Under this pressure differential the stress criteria for design are a), the primary membrane stresses in the tube sheet ligaments, averaged across the ligament and through the tube sheet thickness, do not exceed the material S m Value at the design temperature; and b), the primary membrane plus primary bending stress in the tube sheet ligaments, averaged across the ligament width at the tube sheet surface location giving maximum stress, do not exceed 150% of the material S m value at the design temperature.
- See Appendix 4C for description replaced steam generator lower assemblies.
4.3-4 Revised04/17/2013C26C26 This use of these stress criteria for this abnormal operation is consistent with the ASME Boiler and Pressure Vessel Code,Section III Nuclear Vessels, paragraph N-414.1, N-414.2, and N-414.3 stress criteria. The stresses and
stress factors in the actual tube sheet, obtained using the above stress
criteria, are given in Table 4.3-3.
The tube sheet designed on the above basis meets code allowable stresses for a
primary to secondary differential pressure of 1700 psi. The maximum normal
operating differential pressure is 1549 psi.
The tubes have been designed to the requirements (including stress
limitations) of Section III for normal operation, assuming 2485 psi as the
normal operating pressure differential. Hence, the secondary pressure loss
accident condition imposes no extraordinary stress on the tubes beyond that
normally expected and considered in Section III requirements.
No significant corrosion of the Inconel tubing is expected during the lifetime
of the unit. The corrosion rate reported in Reference (4) shows "worst case" rates of 15.9 mg/dm 2 in the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> test under steam generator operating conditions. Conversion of this rate to a 60-year unit life gives a corrosion
loss of less than 2.25 x 10
-3 inches which is insignificant compared to the nominal tube wall thickness of 0.050 inches.
In the case of a primary pressure loss accident, the secondary-primary
pressure differential can reach 1100 psi. This pressure differential is less
than the primary-secondary pressure differential capability (1549 psi) for
normal operating conditions. Hence, no stresses in excess of those covered in
Section III rules for normal operation are experienced on the tube sheet for
this accident case. For the tubes, actual pressure tests of 3/4 in. O.D./.058
inch wall Inconel tubing show collapse under external pressure of 5700-5900
psi. Extrapolating these data to 7/8 in. O.D./.050 inch wall tubes, collapse would occur at about 2630 psi at 650 F. This gives a factor of safety of 2.4 against collapse under the 1100 psig accidental application of external pressure to tubes. The ASME Section VIII design curves for
Iron-Chromium-Nickel Steel cylinders under external pressure indicate a
predicted collapse pressure for the tubes of 2310 psi, which checks closely
with the extrapolated value for the experimental results.
4.3-5 Revised 04/17/2013 C26C26C26 Consideration has been given to the superimposed effects of secondary side
pressure loss and the maximum potential earthquake loading. The fluid dynamic
forces on the internal components affecting the primary-secondary boundary (tubes) have been considered as well. For this condition the criterion is
that no rupture of primary to secondary boundary (tubes and tube sheet)
occurs.
For the case of the tube sheet, the maximum hypothetical earthquake loading
will contribute an equivalent static pressure loading over the tube sheet of
less than 10 psi (for vertical shock). Such an increase is small when
compared to the pressure differentials (up to 2485 psi) for which the tube
sheet is designed. Under horizontal shock loading of the maximum hypothetical
earthquake the stresses are less than those for the 1.0g loading experienced
by a steam generator when in a horizontal position, which the design can
readily accept. The fluid dynamic forces on the internals under secondary
steam break accident conditions indicate, in the most severe case, that the
tubes are adequate to constrain the motion of the baffle plates with some
plastic deformation but boundary integrity is maintained.
4.3-6 Revised 04/17/2013 The ratio of the allowable stresses on various tube sheet-head-shell components to the computed stresses for an abnormal primary to secondary pressure differential of 2485 psi are summarized in Table 4.3-4. The allowable stress limits for abnormal loading are yield and 150% of yield for membrane and membrane plus bending, respectively.
The steam generators were analyzed in accordance with Section III, N-415.1 of
the ASME code (1965 Edition). Based upon this analysis, it was concluded
that the only areas that required a fatigue analysis were the tubesheet and
the mist extractor support. Results of these analyses give a fatigue usage
factor of 0.41 for the mist extractor support and very low usage factors for
the tubesheet, the greatest being 0.2976 at the secondary shell to tubesheet
intersection.
Reactor Coolant Pumps
The casing, main flange and main flange bolts of the reactor coolant pump
were analyzed in accordance with Article 4 Section III, ASME Code. The
analysis included pressure, thermal and cyclic stresses.
Mathematical models of the parts were prepared and used in the analysis:
- 1) The design was checked against the design criteria of the ASME Code for pressure stresses. The shells were profiled to attain optimum metal
distribution.
- 2) The interactivity forces needed to maintain geometric capability between the various components were determined at design pressure and
temperature, and applied to the components along with the external
loads, to determine the final stress state of the components. These
were within the Code allowable values.
4.3-7 Revised 04/17/2013 C26 4.3.2 RELIANCE ON INTERCONNECTED SYSTEMS
The principal heat removal systems which are interconnected with the Reactor
Coolant System are the Steam and Power Conversion, Safety Injection and
Residual Heat Removal Systems. The Reactor Coolant System is dependent upon
the steam generators, and the steam, feedwater, and condensate systems for
stored and residual heat removal from normal operating conditions down to a reactor coolant temperature of approximately 350 F. The layout of the system ensures the natural circulation capability to permit unit cool down following a loss of all reactor coolant pumps.
The Steam and Power Conversion System is described in Section 10.2. In the
event that the condensers are not available to receive the steam generated by
residual heat, the water stored in the feedwater system may be pumped into
the steam generators and the resultant steam vented to the atmosphere. The
auxiliary feedwater system will supply water to the steam generators in the
event that the main feedwater pumps are inoperative.
The Safety Injection System is described in Section 6. The Residual Heat
Removal System is described in Section 9.
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 filled.
With the reactor vessel head removed and at least 23 feet of water above the
reactor pressure vessel flange, the heat sink and backup decay heat removal
capability afforded by the large volume of water above the vessel flange will
provide adequate time to initiate emergency procedures to cool the core in the
event of a failure of the operating residual heat removal loop. The 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 assumes that: (1) the reactor has been subcritical for at least
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and (2) the reactor coolant system temperature is not more than
140°F. 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, stable natural circulation patterns occur such that
adequate heat transfer capability is maintained to prevent fuel damage. For
more detailed information regarding the analysis, see Reference 11.
4.3-8 Revised 04/17/2013
For the upper internals assembly installed case, the natural circulation flow path modeled is up from the core to the vessel upper plenum to the refueling
cavity via the holes in the upper support plate, the CRDM guide tubes, the
head spray flow nozzles, the upper internals hold-down spring gap, and the hot
leg gap, with return flow to the core via the downcomer and barrel/baffle
bypass. A direct flow path for natural circulation from the core to the vessel
upper plenum to the refueling cavity and back to the core via the downcomer
and barrel/baffle bypass exists for the upper internals assembly removed case.
The presence of these 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.
4.3.3 SYSTEM INTEGRITY
A complete stress analysis which reflects consideration of all design
loadings detailed in the design specification has been prepared by the
manufacturer. The analysis shows that the reactor vessel, steam generator, pump casing and pressurizer comply with the stress limits of Section III of
the ASME Code. A similar analysis of the piping shows that it complies with
the stress limits of the applicable USASI Code.
As part of the design control on materials, Charpy V-notch toughness test
curves are run on all ferritic material used in fabricating pressure parts of
the reactor vessel, steam generator and pressurizer to provide assurance for
hydrotesting and operation in the ductile region at all times. In addition, drop-weight tests were performed on the reactor vessel plate material. Refer
to Table 4.3-5.
As an assurance of system integrity, all components in the system are
hydrotested at 3107 psig prior to initial operation.
4.3-9 Revised 04/17/2013 4.3.4 OVERPRESSURE PROTECTION
The Reactor Coolant System is protected against overpressure by safety valves
located on the top of the pressurizer. The safety valves on the pressurizer
are sized to prevent system pressure from exceeding the design pressure by
more than 10 per cent, in accordance with Section III of the ASME Boiler and
Pressure Vessel Code. The capacity of the pressurizer safety valves is
determined from considerations of: (1) the reactor protective system, and (2)
accident or transient conditions which may potentially cause overpressure.
The combined capacity of the safety valves is equal to or greater than the
maximum surge rate resulting from complete loss of load without a direct
reactor trip or any other control, except that the safety valves on the
secondary system are assumed to open when the steam pressure reaches the
secondary system safety valve setting.
Details of the analysis are reported in Section 14.1.8. Experience has shown
that the safety valve capacity so determined is adequate for all the other
transients as the results of Section 14.1 show.
4.3.5 SYSTEM ACCIDENT POTENTIAL
The potential of the Reactor Coolant System as a cause of accidents is
evaluated by investigating the consequences of certain credible types of
components and control failures as discussed in Sections 14.1 and 14.2.
Reactor coolant pipe rupture is evaluated in Section 14.3.
4.3.6 REDUNDANCY
Each loop of the Reactor Coolant System contains a steam generator and a
reactor coolant pump. Operation at reduced reactor power is possible with
one loop out of service (Section 14.1.9). The normal power supply to the
reactor coolant pumps is from two electrically separate buses, as shown in
Figure 8.2-2.
4.3-10 Revised 04/17/2013 REFERENCES 1.Kibara and Masubichi, "Effect of Residual Stress on Brittle Fracture," (Welding Journal Volume 38), April 1959.2.Robertson, "Propagation From Brittle Fracture in Steel", (Journal of the Iron and Steel Institute), 1953.
3.L. Porse, "Reactor Vessel Design Considering Radiation Effects," ASME Paper No. 63-WA-100.4.W.E. Berry and F.W. Fink, "The Corrosion of Inconel in High Temperature Water," Battelle Memorial Institute, April, 1958.5.W. A. Van der Sluvs, "An Investigation of Mechanical Properties and Chemistry within a Thick MnMoNi Submerged Arc Weldment," EPRI NP-373, February 1977.6.Westinghouse WNET-168, "Model 44F Replacement Steam Generator Stress Report for Florida Power and Light Company Turkey Point Units 3 & 4,"
January 1982.
7.Westinghouse WCAP-14291, "Turkey Point Units 3 and 4 Uprate Engineering Report," December 1995.8.WCAP-15092 Revision 3, Turkey Point Units 3 and 4 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for
Normal Operation, Westinghouse, May 2000.9.Low Temperature Overpressure Protection System Setpoints, Florida Power
& Light Company, 32 EFPY and 48 EFPY for Turkey Point, Units 3 & 4.
Westinghouse Electric Company LLC, June 2000.10.PCM 00-027, Rev. 0, "Cold Overpressure Mitigation System (COMS)
Setpoint Change."
11.Engineering Evaluation PTN-ENG-SENS-07-032, "RHR System Operation with the Reactor Cavity Filled and the Vessel Upper Internals In Place," Revision 0.12.Framatome ANP Topical Report BAW-2008, Revision 2A, "Initial RTndt of Linde 80 Weld Materials", Approved March 2008.
4.3-11 Revised04/17/2013C26 TABLE 4.3-1
SUMMARY
OF PRIMARY PLUS SECONDARY STRESS INTENSITY FOR COMPONENTS OF THE REACTOR VESSEL
Allowable Stress 3 S m Stress Intensity (psi) Area (psi) (Operating Temperature)
CRDM Housing 56,440 69,900 Head Flange 66,500 80,100 Vessel Flange 56,100 80,100 Closure Studs 84,700 104,400 Outlet Nozzles 46,142 80,100 Inlet Nozzles 59,686 80,100 Core Support pad(s) 21,078 80,100 Shell at Core Support Pads 35,637 80,100 Bottom head to shell juncture 34,785 80,100 Bottom instrumentation 70,300(1) 69,900 Shell to shell juncture 45,644 80,100 (1) This value is greater than the allowable value. The ASME code year of construction specifies that the 3S m limit can be exceeded based on the occurrence of shakedown, but does not specify a procedure. The simplified elastic plastic procedure from the 1998 Edition of ASME Code through the 2000 Addenda was used to demonstrate shakedown, and it was concluded that the Code Criteria was met.
Revised 04/17/2013 C26 TABLE 4.3-2
SUMMARY
OF CUMULATIVE FATIGUE USAGE FACTORS FOR COMPONENTS OF THE REACTOR VESSEL
Item Usage Factor
- Control Rod Housing 0.73 Head Flange 0.083 Vessel Flange 0.531 Stud Bolts 0.81 Outlet nozzles 0.063 Inlet nozzles 0.066 Core support pad 0.020 Shell at Core Support Pads 0.509 Bot. head to shell juncture 0.023 Bot. instrumentation 0.002 Shell to shell juncture 0.034
- As defined in Section III of the ASME Boiler and Pressure Vessel Code, Nuclear Vessels.
TABLE 4.3-2a
SUMMARY
OF CUMULATIVE FATIGUE USAGE FACTORS FOR PRESSURE BEARING COMPONENTS OF THE REACTOR COOLANT PUMPS
Item Usage Factor
Casing < 0.001 Main Flange 0.025
Main Flange Studs 0.29
Revised 04/17/2013 C26C26 TABLE 4.3-3 MAXIMUM LIGAMENT STRESSES DUE TO STEAM GENERATOR TUBE SHEET DESIGN PRESSURE DIFFERENTIAL (1700 PSI)
Stress Computed Value Allowable Value Primary Membrane Stress 8.9 ksi 30 ksi Primary Membrane plus 40.7 ksi 45 ksi Primary Bending Stress In addition to the foregoing evaluation, elasto-plastic limit analysis of the tube sheet-head-shell combination indicates an allowable limit differential pressure of 3100 psi at 650 F, which compares to the 1700 psi primary to secondary design differential pressure.
Revised 04/17/2013 C26 TABLE 4.3-4
RATIO OF ALLOWABLE STRESSES TO COMPUTED STRESSES FOR A STEAM GENERATOR TUBE SHEET ABNORMAL PRESSURE DIFFERENTIAL OF 2485 PSI at 668 F
Component Part
Yield (Membrane)
150% Yield (Membrane + Bending)
Channel head 3.38 4.29 Channel head-tube sheet joint 1.48 1.14 Shell 3.91 2.82 Tube sheet 7.36 Max. Avg. Ligament 1.59
Revised 04/17/2013 C26 TABLE 4.3-5
SUMMARY
OF RESULTS OF CHARPY V-NOTCH AND DROP WEIGHT TESTS FOR REACTOR VESSEL PLATES AND FORGINGS AND BELTLINE WELDS
30ft-lb Fix Min. Curve Drop Weight
Component Grade (Temp/
o F) NDT (o F)
Unit 3 Unit 4 Unit 3 Unit 4
Replacement RVCH Mono Forging SA508 Class 3 **** **** -50 -50
Vessel Flange " " -62 -36 - -
Upper Shell " " 20 -25 50 40 Inlet Nozzle " " -18 - -
" " " " - -
" " " " -19 - -
Outlet Nozzle " " -60 -63 - -
" " " " -20 -20 - -
" " " " 8 8 - -
Intermediate Shell " " -8 42 40 50 Lower Shell " " -26 -22 30 40 Lower Transition Ring " " 18 - -
Bottom Head Dome A302 Grade B 6 0 -10 0
Drop. Wt. 50 Ft.-Lb. Upper Shelf
Vessel NDT Temp. (o F) Energy (Ft.-Lbs.)
Unit Component Temp. (o F) Long. Trans.
Long. Trans.
No. 3 Upper Shell 50 20 40** -- 99**
No. 3 Core Region Weld -60*** -- 70*** -- 65 No. 4 Inter. Shell 50 50 70 -- 88 No. 4 Core Region Weld -60*** -- 70*** -- 65
- This information represents the vessel components having the highest drop weight NDTT, the highest 50 ft.-lb. temperature and the lowest upper shelf
energy level.
- Estimated
and the minimum lateral expansion was 79 mils. Which is above the
required 35 mils. minimum. Based on these results, RT NDT = -50°F.
Revised 04/17/2013 C26C26 TABLE 4.3-6
SUMMARY
OF ESTIMATED STRESS INTENSITIES FOR AREAS OF CONCERN IN THE STEAM GENERATORS
[DELETED]
Rev. 16 10/99
4.4 TESTS AND INSPECTIONS 4.4.1 REACTOR COOLANT SYSTEM INSPECTION
Non-Destructive Inspection of Materials and Components (See Note 1)
Table 4.4-1 summarizes the quality assurance program for all Reactor Coolant
System components. In this table all of the non-destructive tests and
inspections which are required by Westinghouse specifications on Reactor
Coolant System components and materials are specified for each component.
All tests required by the applicable codes are included in this table.
Westinghouse requirements, which are more stringent in some areas than those
requirements specified in the applicable codes, are also included.
Westinghouse requires, as part of its reactor vessel specification, that
certain special tests which are not specified by the applicable codes be
performed. These tests are listed below:
- 1) Ultrasonic Testing - Westinghouse requires that a 100% volumetric ultrasonic test of reactor vessel plate for both shear wave and
longitudinal wave be performed.Section III Class A vessel plates are
required by code to receive only a longitudinal wave ultrasonic test on
a 9 in. x 9 in. grid. The 100% volumetric ultrasonic test is a severe
requirement, but it assures that the plate is of the highest quality.
- 2) Radiation Surveillance Program - This program monitors the effects of neutron irradiation on the reactor vessel beltline. Irradiation damage
is based on pre- and post- irradiation testing of charpy V-notch and
tensile test specimens. The program evaluates the effect of irradiation
on the fracture toughness of reactor vessel steels and weldments by
measuring transition temperature shift and using a fracture mechanics
methodology. The program is in accordance with ASTM E185, "Recommended
Practice For Surveillance Tests on Structural Material in Nuclear
Reactors," required by 10 CFR 50 Appendix H.
Note 1: The Unit 3 AND Unit 4 Reactor Vessel Closure Heads (RVCH) have been replaced. See the discussion at the end of this heading for the NDE information pertaining to the replacement RVCHs.
4.4-1 Revised 07/14/2005 The original program used eight surveillance capsules in each vessel which are mounted on the thermal shield about 3 inches from the wall opposite the core midplane. Capsules are withdrawn according to a
schedule and the surveillance materials are tested. The capsules
contain reactor vessel steel machined specimens from forgings, welds, and heat affected zones as well as correlation monitor material.
Dosimeters are included to permit the evaluation of the flux seen by the
specimens. Dosimetry includes Ni, Cu, Fe, Co, Al, Cd shielded Co-Al, Cd
shielded Np-237, and Cd shielded U-238. Thermal monitors of low melting
point alloys are included to monitor the temperature range seen by the
specimens. Wedge open loading fracture toughness specimens are included
in the capsules but have not been tested because they were not yielding
meaningful data. They will be tested in the future.
The circumferential girth weld of both vessels is the limiting weld material and is a high Copper (0.23%) Linde 80 flux submerged arc weld (SA1101) for both. This weld appears as surveillance material in only 3
capsules for each vessel. Since this would not supply enough data
throughout life on an individual vessel basis, the surveillance programs
were integrated in accordance with the provisions of Appendix H. The
integrated program approach has been approved for use at PTN by the NRC.
The surveillance program shown in Table 4.4-2 reflects the integrated
program and meets all requirements.
There are two (2) supplemental capsules which contain the limiting weld SA 1101 being irradiated in the Babcock and Wilcox Owners' Group master
integrated surveillance program. When these capsules are removed and
tested, the data will be evaluated and considered as appropriate.
4.4-2 Revised 03/01/2011 C25 Table 4.4-1 summarizes the quality assurance program with regard to inspections performed on primary system components. In addition to the
inspections shown in Table 4.4-1, there are those which the equipment
supplier performs to confirm the adequacy of material he receives, and those
performed by the material manufacturer in producing the basic material. The
inspections of reactor vessel, pressurizer, and steam generator are governed
by ASME code requirements. The inspection procedures and acceptance
standards required on pipe materials and piping fabrication are governed by
USAS B31.1 and Westinghouse requirements and are equivalent to those
performed on ASME coded vessels.
Procedures for performing the examinations are consistent with those
established in the ASME Code Section III and are reviewed by qualified
Westinghouse engineers. These procedures have been developed to provide the
highest assurance of quality material and fabrication. They consider not
only the size of the flaws, but equally as important, how the material is
fabricated, the orientation and type of possible flaws, and the areas of most
severe service conditions. In addition, the surfaces most subject to damage
as a result of the heat treating, rolling, forging, forming and fabricating
processes, receive a 100% surface inspection by Magnetic Particle or Liquid
Penetrant Testing after all these operations are completed. All reactor
coolant plate material is subject to shear as well as longitudinal ultrasonic
testing to give maximum assurance of quality. (All forgings receive the same
inspection.) In addition, 100% of the material volume is covered in these
tests as an added assurance over the grid basis required in the code.
Westinghouse Quality Control engineers monitor the supplier's work, witnessing key inspections not only in the supplier's shop but in the shops
of subvendors of the major forgings and plate material. Normal surveillance
includes verification of records of material, physical and chemical
properties, review of radiographs, performance of required tests and
qualification of supplier personnel. Florida Power and Light Company, using
Bechtel and others as consultants has reviewed the quality control methods
and results of vendors and has found them to be satisfactory.
4.4-3 Field erection and field welding of the reactor coolant system are performed such as to permit exact fit-up of the 31" I.D. closure pipe subassemblies
between the steam generator and the reactor coolant pump. After installation
of the pump casing and the steam generator, measurements are taken of the
pipe length required to close the loop. Based on these measurements, the 31" I.D. closure pipe subassembly is properly machined and then erected and field
welded to the pump suction nozzle and to the steam generator exit nozzle.
Cleaning of RCS piping and equipment is accomplished before and during erection of various equipment. Stainless steel piping is cleaned in sections
as specific portions of the systems are erected. Pipe and units large enough
to permit entry by personnel are cleaned by locally applying approved
solvents (acetone or alcohol), and demineralized water, and by using a rotary
disc sander or 18-8 wire brush to remove all trapped foreign particles.
Standards for final physical and chemical cleanliness are defined in Section
- 13. Equipment specifications for fabrication require that suppliers submit the manufacturing procedures (welding, heat treating, etc.) to Westinghouse where
they are reviewed by qualified Westinghouse engineers. This also is done on
the field fabrication procedures to assure that installation welds are of
equal quality.
Section III of the ASME B&PV Code requires that nozzles carrying significant external loads shall be attached to the shell by full penetration welds.
This requirement has been carried out in the reactor coolant piping, where
all auxiliary pipe connections to the reactor coolant loop are made using
full penetration welds.
4.4-4 The Reactor Coolant System components are welded under procedures which require the use of both preheat and post-heat. Preheat requirements, non-mandatory under Code rules, are performed on all weldments, including P1
and P3 materials which are the materials of construction in the reactor
vessel, pressurizer and steam generators. Preheat and post-heat of weldments
both serve a common purpose: the production of tough, ductile metallurgical
structures in the completed weldment. Preheating produces tough ductile
welds by minimizing the formation of hard zones whereas post-heating achieves
this by tempering any hard zones which may have formed due to rapid cooling.
Replacement RVCH Non-Destructive Inspection Table 4.4-1 summarizes the quality assurance program inspections for the replacement RVCHs. In this table are identified all of the nondestructive test and inspections required by the RVCH design specification. All tests required by the applicable Code (ASME Section III, 1989 Edition, no Addenda) are included in the table as well as any additional test or more stringent acceptance criteria as may have been specified in the design specification for the RVCH.
In addition to the inspections summarized in Table 4.4-1, there are those inspections which the equipment supplier performs to confirm the adequacy of material he receives, and those performed by the manufacturer of the materials in producing the basic materials. Procedures for performing all of the examinations are consistent with those established in the ASME Code SectionIII and are reviewed by qualified FPL and Owner's Agent representatives. These procedures have been developed to provide the highest assurance of quality in the materials and fabrication. They consider not only the size of flaws, but equally as important, how the material is fabricated, the orientation and type of possible flaws, and the areas of most severe service conditions. The volumetric inspections (Ultrasonic Testing) of the forging were done using both the straight beam and the angle beam techniques. In addition, the surfaces most subject to damage as a result of forging, heat treating, forming, fabricating, and hydrostatic testing received 100% surface inspections by Magnetic Particle or Liquid Penetrante Testing at various stages during the processes and after final completion of the hydrostatic test of the RVCH.
The RVCH requires welding and weld cladding performed under procedures which require the use of both preheat and post heat treating. Preheat of weld areas and post-heat treating are performed on all welds on the replacement RVCH. Preheat and post-heat of weldments both serve the common purpose of producing tough, ductile metallurgical structures in the weldment.
Preheating produces tough ductile welds by minimizing the formation of hard non-ductile zones whereas post weld heat-treating achieves this by tempering any hard zones which may have formed due to rapid cooling. 4.4-4a Revised 07/14/2005 FPL and the Owner's Agent reviewed the manufacturer's quality control methods and results of the vendor and subvendor of the RVCH and have found them to be acceptable. FPL and the Owner's Agent Quality Control engineers monitored the supplier's work, witnessing key inspections not only in the supplier's shop but in the shops of the subvendor of the major forging. Normal surveillance includes verification of records of material, physical and chemical properties, review of radiographs, performance of the required tests and qualification of supplier personnel. FPL and the Owner's Agent reviewed the manufacturing quality control results and records of the vendor and subvendors of the RVCH and have found them to be complete and acceptable.
In-Service Inspection Capability
During the design phase of the Reactor Coolant System, careful consideration
is given to provide access for both visual and non-destructive in-service
inspection of primary loop components. The following components and areas
are available for visual and/or non-destructive inspection.
- 1) Reactor Vessel - The entire inside surface.
- 2) Reactor Vessel Nozzles - The entire inside surface.
- 3) Closure Head - The entire inside and outside surface.
- 4) Reactor Vessel Studs, Nuts and Washers.
- 5) Field Welds between the Reactor Vessel, Steam Generators, and Reactor
Coolant Pumps and the Main Coolant Piping.
- 6) Reactor Internals
- 7) Reactor Vessel Flange Seal Surface
- 8) Fuel Assemblies
- 9) Rod Cluster Control Assemblies
- 10) Control Rod Drive Shafts
- 11) Control Rod Drive Mechanism Assemblies
- 12) Main Coolant Pipe External Surfaces (except for the five foot
penetration of the primary shield)
4.4-5 Revised 04/29/2005
- 13) Steam Generator - The internal surface, the internal surfaces of the steam drum, and channel head.
- 14) Pressurizer - The internal and external surfaces.
- 15) Reactor Coolant Pump - The external surfaces; motor and impeller.
The design considerations which have been incorporated into the primary
system design to permit the above inspections are as follows:
- 1) All reactor internals are completely removable. The storage space
required to permit these inspections is provided.
- 2) The closure head is stored dry on the reactor operating deck during
refueling to facilitate visual inspection.
- 3) All reactor vessel studs, nuts and washers are removed to dry storage
during refueling.
- 4) Removable plugs are provided in the primary shield just above the coolant nozzles, and the insulation covering the nozzle welds may be
removed. 5) Access holes are provided in the lower internals barrel flange to allow remote access to the reactor vessel internal surfaces between the
flange and the nozzles without removal of the internals. 6) A removable plug is provided in the lower core support plate to allow
access for inspection of the bottom head without removal of the lower
internals.
- 7) The storage stands provided for storage of the internals allow for
inspection access to both the inside and outside of the structures.
- 8) The station provided for changeout of control rod clusters from one fuel assembly to another is specially designed to allow inspection of
both fuel assemblies and control rod clusters.
- 9) The control rod mechanism is designed to allow removal of the mechanism assembly from the reactor vessel head.
4.4-6
- 10) Manways are provided in the steam generator, steam drum and channel head to allow access for internal inspection. 11) A manway is provided in the pressurizer top head to allow access for internal inspection.
- 12) All insulation on primary system components (except the reactor vessel) and piping (except for the penetration in the primary shield) may be removed.
The use of non-destructive, direct visual and remote visual test techniques
can be applied to the inspection of primary loop components other than the
reactor vessel. The reactor vessel requires special consideration because of
the radiation levels and remote underwater accessibility to this component.
Because of these limitations on access to the reactor vessel, several steps
have been incorporated into the design and manufacturing procedures in
preparation for non-destructive test techniques which may be available in the
future. These are:
- 1) Shop ultrasonic examinations are performed on all internally clad surfaces to an acceptance and repair standard to assure an adequate
cladding bond to allow later ultrasonic testing of the base metal.
Size of cladding bonding defect allowed is 3/4 inch, which permits
subsequent UT of the base metal through the clad surface.
- 2) The design of the reactor vessel shell in the core area is a clean, uncluttered cylindrical surface to permit future positioning of test
equipment without obstruction.
- 3) During the manufacturing stage, selected areas of the reactor vessel were ultrasonic tested and mapped to facilitate possible future
in-service inspection.
The areas which were ultrasonic mapped include:
4.4-7 Rev. 3-7/85 a) Vessel flange radius, including the vessel flange to upper shell weld. b) Middle shell course. c) Lower shell course above the radial core supports.
d) Exterior surface to the closure head from the flange knuckle to the cooling shroud. e) Nozzle to upper shell weld.
f) Middle shell to lower shell weld.
g) Upper shell to middle shell weld.
The pre-operational ultrasonic testing of these areas was performed
after shop hydrotest.
Plans for inservice inspection of the reactor coolant system pressure
envelope are currently being developed. The applicability of ultrasonic
testing
techniques is also being evaluated.
Various tests are currently underway to determine the effect of cladding
surface finish on ultrasonic inspectability of vessel material.
For the Unit 3 and Unit 4 replacement RVCHs, a baseline UT examination was performed on all of the nozzle penetration to RVCH forging welds after final PT of the welds. The acceptance criteria for the PT was no indications (PT White). The baseline UT was performed using the best technique and practices available and the results are achieved for future ISI comparative reference.
4.4-8 Revised 07/14/2005 TABLE 4.4-1 Sheet 1 of 3 REACTOR COOLANT SYSTEM QUALITY ASSURANCE PROGRAM Component RT* UT*PT*MT*ET* 1.Steam Generator1.1 Tube Sheet 1.1.1 Forging yes yes1.1.2 Cladding yes yes 1.2 Channel Head 1.2.1 Casting yes yes 1.2.2 Cladding yes1.3 Secondary Shell & Head 1.3.1 Plates yes 1.4 Tubes yes yes 1.5 Nozzles (forgings) yes yes 1.6 Weldments 1.6.1 Shell, longitudinalyes yes 1.6.2 Shell, circumferential yes yes 1.6.3 Cladding yes1.6.4 Nozzle to shell yes yes 1.6.5 Support brackets yes 1.6.6 Tube-to-tube sheet yes 1.6.7 Instrument connections yes (primary and secondary) 1.6.8 Temporary attachments after removal yes1.6.9 After hydrostatic test (all welds) yes1.6.10 Nozzle safe ends yes yes (if forgings) 1.6.11 Nozzle safe ends yes (if weld deposit) 2.Pressurizer2.1 Heads2.1.1 Casting yes 2.1.2 Cladding yes yes 2.2 Shell 2.2.1 Plates yes yes 2.2.2 Cladding yes 2.3 Heaters 2.3.1 Tubing yes yes 2.3.2 Centering of element yes 2.4 Nozzle yes yes 2.5 Weldments 2.5.1 Shell, longitudinal yes yes 2.5.2 Shell, circumferential yes yes 2.5.3 Cladding yes 2.5.4 Nozzle Safe End yes yes (if forging) 2.5.5 Nozzle Safe End yes (if weld deposit) 2.5.6 Instrument Connections yes 2.5.7 Support Skirt yes 2.5.8 Temporary attachments after removal yes Revised 04/29/2005
TABLE 4.4-1 Sheet 2 of 3
Component RT* UT* PT* MT* ET*
2.5.9 All welds after yes
hydrostatic test
- 3. Piping 3.1 Fittings (Castings) yes yes 3.2 Fittings (Forgings) yes yes 3.3 Pipe yes yes 3.4 Weldments 3.4.1 Longitudinal yes yes 3.4.2 Circumferential yes yes 3.4.3 Nozzle to run pipe yes yes 3.4.4 Instrument Connections yes yes
- 4. Pumps 4.1 Casting yes yes 4.2 Forgings yes yes 4.3 Weldments 4.3.1 Circumferential yes yes 4.3.2 Instrument connections yes
- 5. Reactor Vessel 5.1 Forgings 5.1.1 Flanges yes yes 5.1.2 Studs yes yes 5.1.3 Head Adapters yes yes 5.1.4 Head Adapter Tube yes yes 5.1.5 Instrumentation Tube yes yes 5.1.6 Main Nozzles yes yes 5.1.7 Nozzle Safe Ends (If forging is employed) yes yes 5.2 Plates yes yes 5.3 Weldments 5.3.1 Main Seam yes yes 5.3.2 CRD Head Adapter Connection yes 5.3.3 Instrumentation Tube Connection yes 5.3.4 Main Nozzles yes yes 5.3.5 Cladding yes* yes 5.3.6 Nozzle Safe Ends yes yes (if forging) 5.3.7 Nozzle Safe Ends yes (if weld deposit) 5.3.8 Head adapter forging yes yes
to head adapter tube 5.3.9 All welds after hydrotest yes
- 6. Valves
6.1 Castings yes yes 6.2 Forgings yes yes
Revised 04/29/2005 TABLE 4.4-1 Sheet 3 of 3 Component RT* UT* PT* MT* ET* 7. Unit 3 and Unit 4 Replacement Reactor Vessel Closure Head 7.1 Head Mono-block Forging 7.1.1 After Rough Machining yes yes 7.1.2 After Final Machining yes 7.1.3 Machined Surfaces to be Clad yes (4) 7.1.4 External Un-clad Surfaces yes (1) 7.1.5 All Clad Surfaces yes (2&3) yes (5) 7.1.6 Final Machined O-Ring Groove yes (6) 7.2 Inconel Alloy 690 CRDM Housing yes (10) yes (7) 7.3 SA-182 F304 CRDM Nozzle Adapter yes (10) yes (7) 7.4 Weldment 7.4.1 All Weld Prep Areas yes (8) 7.4.2 Root Pass of All Welds yes 7.4.3 Final Surface of All Welds yes yes (9) 7.4.4 Nozzle to Forge Weld Areas yes (11) 1. All accessible ferritic surfaces after final hydrostatic test. 2. Sealing and bearing surfaces of the head examined for defects and bond. 3. Non-sealing and non-bearing surfaces examined for bond. 4. After machining and prior to cladding. 5. After post weld heat treatment. 6. The bottom sealing surfaces must be free of indications (PT White). 7. After final machining. 8. After final weld prep machining but prior to root pass welding. 9. Final surfaces of all CRDM nozzle attachment and vent piping welds must be free of indications (PT White). 10. After rough machining. 11. This is a baseline for future examinations.
__________
MT - Magnetic Particle; ET - Eddy Current
UT of clad bond-to-base metal.
Revised 07/14/2005 Table 4.4-2 Surveillance Capsule Withdrawal Schedule Turkey Point Units 3 & 4 (e)
(a) Capsule X 3 and Capsule X 4 were moved from the 50° location to the 270° location in 1990. (b) Effective Full Power Years (EFPY) from plant startup. (c) Capsule X 4 should be removed at 38.1 EFPY to fulfill the requirements of the "5 th Capsule" to be withdrawn. This EPFY will yield a capsule fluence that is approximately equivalent to the 80-year (67 EFPY) peak vessel fluence of 9.297 x 10 19 n/cm 2 (E> 1.0 MeV). (d) The lead factors listed for Capsule x 4 and the standby capsules are 48 EPFY projections and pertain to the most limiting core design case (lowest lead fa ctor). The lowest lead factor is considered most limiting to prevent premature capsule withdraw al. Turkey Point Unit 3 and 4 operate under an integrated surveillance program. Therefore, the Unit 4 standby capsule lead factors are approximated to be equivalent to the Un it 3 standby capsule lead factors. (e) Capsule removal changes require NRC approval per 10 CFR 50 Appendix H.
Revised 04/17/2015 Capsule (Unit shown as subscript)
Capsule Location (Degree) Withdrawal EFPY(b) Lead Factor(d)
Fluence (n/cm 2, E > 1.0 MeV) T 3 270° 1.15 2.736 0.5990 x 10 19 T 4 270° 1.17 2.74 0.649 x 10 19 S 3 280° 3.46 1.997 1.272 x 10 19 S 4 280° 3.41 2.03 1.29 x 10 19 V 3 290° 8.06 0.891 1.223 x 10 19 X 3 270° / 50°(a) 19.85 1.129 2.897 x 10 19 X 4 270° / 50°(a) 38.1(c) 2.088 9.297 x 10 19 V 4 290° Standby 1.015 ---
U 3 30° Standby 0.767 ---
U 4 30° Standby 0.767 ---
W 3 40° Standby 0.523 ---
W 4 40° Standby 0.523 ---
Y 3 150° Standby 0.767 ---
Y 4 150° Standby 0.767 ---
Z 3 230° Standby 0.523 ---
Z 4 230° Standby 0.523 --- C27C27C27C27 APPENDIX 4A DETERMINATION OF REACTOR PRESSURE VESSEL REFERENCE NIL-DUCTILITY TRANSITION TEMPERATURE (RT NDT)
- 1. MEASUREMENT OF INTEGRATED FAST NEUTRON (E > 1.0 MEV) FLUX AT THE IRRADIATION SAMPLES
The use of passive neutron sensors such as those included in the internal surveillance capsule dosimetry sets does not yield a direct measure of the
energy dependent neutron flux level at the measurement location. Rather, the
activation or fission process is a measure of the integrated effect that the
time-dependent and energy-dependent neutron flux has on the target material
over the course of the irradiation period. An accurate assessment of the
average flux level and, hence, time integrated exposure (fluence) experienced
by the sensors may be developed from the measurements only if the sensor
characteristics and the parameters of the irradiation are well known. In
particular, the following variables are of interest:
1 - The measured specific activity of each sensor 2 - The physical characteristics of each sensor 3 - The operating history of the reactor 4 - The energy response of each sensor 5 - The neutron energy spectrum at the sensor location
In this section the procedures used to determine sensor specific activities, to
develop reaction rates for individual sensors from the measured specific
activities and the operating history of the reactor, and to derive key fast
neutron exposure parameters from the measured reaction rates are described.
Determination of Sensor Reaction Rates
The specific activity of each of the radiometric sensors is determined using
established ASTM procedures. Following sample preparation and weighing, the
specific activity of each sensor is determined by means of gamma-ray
spectrometry utilizing a high purity germanium gamma spectrometer. In the case
of the surveillance capsule multiple foil sensor sets, these analyses are
performed by direct counting of each of the individual wires; or, as in the
case of U-238 and Np-237 fission monitors, by direct counting preceded by
dissolution and chemical separation of cesium from the sensor.
4A-1 Revised 04/17/2013 C26 The irradiation history of the reactor over its operating lifetime is obtained from NUREG-0020, "Licensed Operating Reactors Status Summary Report" or from
other plant records. In particular, operating data are extracted on a monthly
basis from reactor startup to the end of the capsule irradiation period. For
the sensor sets utilized in the surveillance capsule irradiations, the
half-lives of the product isotopes are long enough that a monthly histogram
describing reactor operation has proven to be an adequate representation for
use in radioactive decay corrections for the reactions of interest in the
exposure evaluations.
Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to
full power operation are determined from the following equation:
where:
A = measured specific activity (dps/gm)
R = reaction rate averaged over the irradiation period and referenced to operation at a core power level of P ref (rps/nucleus). N 0 = number of target element atoms per gram of sensor. F = weight fraction of the target isotope in the sensor material.
Y = number of product atoms produced per reaction.
P j = average core power level during irradiation period j (MW).
P ref = maximum or reference core power level of the reactor (MW).
C j = calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period.
= decay constant of the product isotope (sec
-1). t j = length of irradiation period j (sec).
t d = decay time following irradiation period j (sec).
and the summation is carried out over the total number of monthly intervals
comprising the total irradiation period.
4A-2 Revised 04/17/2013 dt et-e - 1C P P FYN A Rj ref j j j In the above equation, the ratio P j/P ref accounts for month by month variation of power level within a given fuel cycle. The ratio C j is calculated for each fuel cycle and accounts for the change in sensor reaction rates caused by
variations in flux level due to changes in core power spatial distributions
from fuel cycle to fuel cycle. Since the neutron flux at the measurement
locations within the surveillance capsules is dominated by neutrons produced in
the peripheral fuel assemblies, the change in the relative power in these
assemblies from fuel cycle to fuel cycle can have a significant impact on the
activation of neutron sensors. For a single-cycle irradiation, C j = 1.0. However, for multiple-cycle irradiations, particularly those employing low
leakage fuel management, the additional C j correction must be utilized in order to provide accurate determinations of the decay corrected reaction rates for
the dosimeter sets contained in the surveillance capsules.
Corrections to Reaction Rate Data
Prior to using the measured reaction rates in the least squares adjustment
procedure discussed above, additional corrections are made to the U-238
measurements to account for the presence of U-235 impurities in the sensors as
well as to adjust for the build-in of plutonium isotopes over the course of the
irradiation.
In addition to the corrections made for the presence of U-235 in the U-238
fission sensors, corrections are also made to both the U-238 and Np-237 sensor
reaction rates to account for gamma ray induced fission reactions occurring
over the course of the irradiation.
Least Squares Adjustment Procedure
Least squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as neutron fluence (E > 1.0 MeV) or iron atom displacements (dpa) along with their uncertainties are then easily obtained from the adjusted spectrum. The use of measurements in combination with the analytical results reduces the uncertainty in the calculated spectrum and acts to remove biases that may be present in the analytical technique. In general, the least squares methods, as applied to pressure vessel fluence evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties.
4A-3 Revised 04/17/2013 C26C26 Values of key fast neutron exposure parameters are derived from the measured reaction rates using the FERRET least squares adjustment code (Reference 1).
The FERRET approach uses the measured reaction rate data, sensor reaction
cross-sections, and a calculated trial spectrum as input and proceeds to adjust
the group fluxes from the trial spectrum to produce a best fit (in a least
squares sense) to the measured reaction rate data. The "measured" exposure
parameters along with the associated uncertainties are then obtained from the
adjusted spectrum.
In the FERRET evaluations, a log-normal least squares algorithm weights both
the trial values and the measured data in accordance with the assigned
uncertainties and correlations. In general, the measured values, f, are
linearly related to the flux, , by some response matrix A:
where i indexes the measured values belonging to a single data set s, g
designates the energy group, and delineates spectra that may be simultaneously adjusted. For example,
relates a set of measured reaction rates, R i, to a single spectrum, g , through the multigroup dosimeter reaction cross-section, ig , each with an uncertainty . The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the least squares adjustment, the continuous quantities (i.e., neutron
spectra and cross-sections) are approximated in a multi-group format consisting
of 53 energy groups. The trial input spectrum is converted to the FERRET 53
group structure using the SAND-II code (Reference 2). This procedure is
carried out by first expanding the 47 group calculated spectrum into the SAND-
II 620 group structure using a SPLINE interpolation procedure in regions where
group boundaries do not coincide. The 620 point spectrum is then re-collapsed
into the group structure used in FERRET.
4A-4 Revised 04/17/2013
)(g A (s)ig g = f)(s, i ))( (g=R i g g ig ig i R C26C26 The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry file[3], are also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded to 620 groups, is
employed as a weighting function in the cross-section collapsing procedure.
Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix
for each sensor reaction are also constructed from the information contained on
the ENDF/B-VI data files. These matrices include energy group to energy group
uncertainty correlations for each of the individual reactions.
Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation is obtained from plant
specific calculations for each dosimetry location. While the 53 x 53 group
covariance matrices applicable to the sensor reaction cross-sections are
developed from the cross-section data files, the covariance matrix for the
input trial spectrum is constructed from the following relation:
where R n specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set of values. The fractional uncertainties R g specify additional random uncertainties for group g that are correlated with a correlation matrix given by:
where: 2 2 2)g-(g = H 4A-5 Revised04/17/2013 Pgg R g R g + R 2 n = Mgg eH- + gg ]-[1 = Pgg The first term in the correlation matrix equation specifies purely random
uncertainties, while the second term describes short range correlations over a
group range ( specifies the strength of the latter term). The value of is 1 when g = g' and 0 otherwise.
The use of least squares adjustment methods in LWR dosimetry evaluations is not new. The American Society for Testing and Materials (ASTM) has addressed the use of adjustment codes in ASTM Standard E944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" and many industry workshops have been held to discuss the various applications. For example, the ASTM-EURATOM Symposia on Reactor Dosimetry holds workshops on neutron spectrum unfolding and adjustment techniques at each of its bi-annual conferences.
The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement. The analytical method alone may be deficient because it inherently contains uncertainty due to the input assumptions to the calculation.
Typically these assumptions include parameters such as the temperature of the water in the peripheral fuel assemblies, by-pass region, and downcomer regions, component dimensions, and peripheral core source. Industry consensus indicates that the use of calculation alone results in overall uncertainties in the neutron exposure parameters in the range of 15-20% (1). The application of the least squares methodology requires the following input:
- 1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.
- 2. The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set.
- 3. The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.
For a given application, the calculated neutron spectrum is obtained from the results of plant specific neutron transport calculations applicable to the irradiation period experienced by the dosimetry sensor set.
4A-6 Revised 04/17/2013 C26 This calculation is performed using the benchmarked transport calculational methodology described in the next section of this Appendix. The sensor reaction rates are derived from the measured specific activities obtained from the counting laboratory using the specific irradiation history of the sensor set to perform the radioactive decay corrections. The dosimetry reaction cross-sections and uncertainties that are utilized in LWR evaluations comply with ASTM Standard E1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)".
The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectra are input to the least squares procedure in the form of variances and covariances. The assignment of the input uncertainties also follows the guidance provided in ASTM Standard E 944.
- 2. CALCULATION OF INTEGRATED FAST NEUTRON (E > 1.0 MEV) FLUX AT THE IRRADIATION SAMPLES Calculation and Dosimetry Measurement Procedures A generalized set of guidelines for performing fast neutron exposure calculations within the reactor configuration, and procedures for analyzing measured irradiation sample data that can be correlated to these calculations, has been promulgated by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 8). Since different calculational models exist and are continuously evolving along with the associated model inputs, e.g., cross-section data, it is worthwhile summarizing the key models, inputs, and procedures that the NRC staff finds acceptable for use in determining fast neutron exposures within the reactor geometry. This material is highlighted below. The selection of a particular geometric model, the corresponding input data, and the overall methodology used to determine fast neutron exposures within the reactor geometry are based on the needs for accurately determining a solution to the problem that must be solved and the data/resources that are currently available to accomplish this task. Based on these constraints, engineering judgment is applied to each problem based on an analyst's thorough understanding of the problem, detailed knowledge of the plant, and due consideration to the strengths and weaknesses associated with a given calculational model and/or methodology. Based on these conditions, Regulatory Guide 1.190 does not recommend using a singular calculational technique to determine fast neutron exposures. Instead, Regulatory Guide 1.190 suggests that one of the following neutron transport tools be used to perform this work.
4A-7 Revised 04/17/2013 C26
- Discrete Ordinates Transport Calculations
- 1. Adjoint calculations benchmarked to a reference-forward calculation, or stand-alone forward calculations.
- 2. Various geometrical models utilized with suitable mesh spacing in order to accurately represent the spatial distribution of the material compositions and source.
- 3. In performing discrete ordinates calculations, Regulatory Guide 1.190 also suggests that a P3 angular decomposition of the scattering cross-sections be used, as a minimum.
- 4. Regulatory Guide 1.190 also recommends that discrete ordinates calculations utilize S 8 angular quadrature, as a minimum.
- 5. Regulatory Guide 1.190 indicates that the latest version of the Evaluated Nuclear Data File, or ENDF/B, should be used for determining the nuclear cross-sections; however, cross-sections based on earlier or equivalent nuclear data sets that have been thoroughly benchmarked are also acceptable.
- Monte Carlo Transport Calculations A complete description of the Westinghouse pressure vessel neutron fluence methodology along with the SER documenting NRC staff approval of the method and computer codes are provided in Reference 9.
Plant-Specific Calculations The most recent fast (E > 1.0 MeV) neutron fluence evaluations for each of the Turkey Point reactor pressure vessels was based on a 2D/1D synthesis of neutron fluxes that were obtained from a series of plant- and cycle-specific forward discrete ordinates transport calculations run in R-, R-Z, and R geometric models. The set of calculations, which assessed dosimetry as part of the reactor vessel surveillance program and pressure vessel neutron fluences, were conducted in accordance with the guidelines that are specified in Regulatory Guide 1.190.
4A-8 Revised 04/17/2013 C26
- 3. MEASUREMENT OF THE INITIAL NIL-DUCTILITY TRANSITION (NDT) TEMPERATURE OF THE REACTOR PRESSURE VESSEL WELDMENTS, BASE PLATE AND FORGINGS MATERIAL The unirradiated or initial nil-ductility transition temperature of the
pressure vessel weldments, base plate and forgings material was measured by two
methods. These methods are the drop weight test per ASTM E208 and the Charpy
V-notch impact test (Type A) per ASTM E23. The nil-ductility transition (NDT)
temperature is defined in ASTM E208 as "the maximum temperature where a
standard drop-weight specimen breaks when tested according to the provisions of
this method". Using the Charpy V-notch test, the NDT temperature was defined
as the temperature at which the energy required to break the specimen is a
certain "fixed" value. For SA 302B and A508 Class 2 steel the ASME III Table
N-421 specified an energy value of 30 ft-lb. This value was based on a
correlation with the drop weight test and referred to as the "30 ft-lb-fix". A
curve of the temperature versus energy absorbed in breaking the specimen was
plotted. To obtain this curve, 15 tests were performed which included three
tests at five different temperatures. The intersection of the energy versus
temperature curve with the 30 ft-lb ordinate was designated as the NDT
temperature.
For weld materials, Framatome ANP Topical Report BAW-2308, Revisions 1A and 2A (References 6 and 7) give the initial (Linde 80) material properties (RT NDT and i). These properties were obtained by performing fracture toughness testing based on the application of the "Master Curve" evaluation procedure. The Master Curve evaluation procedure permits data obtained from sample sets tested at different temperatures to be combined, as the basis for redefining the initial, unirradiated material properties of Linde 80 welds. Guidelines for the application of the Master Curve evaluation methodology used in Topical Report BAW-2308, Revision 1A were given in the 1997 and 2002 Editions of ASTM Standard Test Method E 1921 (ASTM E 1921) "Standard Test Method for Determination of Reference Temperature, T o , for Ferritic Steels in the Transition Range". Revised initial values were given in Topical Report BAW-2308, Revision 2A (supplemental). Additional guidance on the application of reference temperature values based on Master Curve evaluation to the establishment of reactor pressure vessel material properties for regulatory applications was provided by ASME Code Case N-629, "Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials of Section III, Division 1, Class 1". The B&W Owner Group's motivation for pursuing this option of using a Master Curve based approach to evaluate Linde 80 welds is related to the fact that, due to their generally low Charpy V-notch upper shelf energy behavior, the testing specified in ASME Code,Section III, Paragraph NB-2331 has shown to be overly conservative when used to predict the transition from ductile to brittle failure in Linde 80 welds.
4A-9 Revised 04/17/2013 C26C26 The available data indicate differences as great as 40 F between curves plotted through the minimum and average values respectively. The determination of the NDT temperature from the average curve was considered representative of the
material and was consistent with procedures as specified in ASTM E23. In
assessing the NDT temperature shift due to irradiation, the translation of the
average curve was used.
As part of the Westinghouse surveillance program referred to above, Charpy
V-impact tests, tensile tests, and fracture mechanics specimens were taken from
the core region plates and forgings, and core region weldments including
heat-affected zone material. The test locations are similar to those used in
the tests by the fabricator at the plate mill.
The uncertainties of measurement of the NDT temperature of the base plate were:
- 1. Differences in Charpy V-notch foot pound values at a given temperature between specimens.
4A-9a Rev. 14 2/97
- 2. Variation of impact properties through plate thickness.
The fracture toughness technology for pressure vessels and correlation with service failures based on Charpy Vnotch impact data were based on the averaging of data. The Charpy Vnotch 30 ftlb "fix" temperature was based on
multiple tests by the material supplier, the fabricator, and by Westinghouse as
part of the surveillance program. The average of sets of three specimens at
each test temperature was used in determining each of five data points (total of 15 specimens). In the review of available data, differences of 0F to approximately 40F were observed in comparing curves plotted through the minimum and average values, respectively. The value of the NDT temperature derived from the average curve was judged to be representative of the material
because of the averaging of at least 15 data points, consistent with the
specified procedures of ASTM E23.
In the case of the assessment of RT NDT shift due to fast neutron flux, the displacement of transition curves is measured. The selection of maximum, minimum, or average curves for this assessment is not significant since like
curves would be used.
There are quantitative differences between the RT NDT at the surface, 1/4 thickness, or the center of a plate.
The 1/4T location is considered conservative, since the enhanced metallurgical
properties of the surface are not used for the determination of RT NDT. In addition, the limiting RT NDT for the reactor vessel after operation will be based on the RT NDT shift due to irradiation. Since the fast neutron dose is highest at the inner surface, usage of the 1/4T RT NDT criterion is conservative.
To assess any possible uncertainties in the consideration of the RT NDT shift for welds heat affected zone, and base metal, test specimens of these three "material types" have been included in the reactor vessel surveillance program.
4A10 Rev. 15 4/98 REFERENCES
- 1. Schmittroth, E. A., "FERRET Data Analysis Code," HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979.
- 2. McElroy, W. N., et. al., "A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation," AFWL-TR-67-41, Volumes I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 3. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium," July 1994.
- 4. RSIC Computer Code Collection CCC-650 DOORS 3.2, "ORNL RSICC CCC-650 DOORS 3.2 One, Two and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," Radiation Safety Information Computational Center, Oak Ridge National Laboratory, April 1998.
- 5. RSICC Data Library Collection DLC-185, "BUGLE-96 Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," Radiation Safety Information Computational Center, Oak Ridge National Laboratory, July 1999.
- 6. Framatome ANP Topical Report BAW-2308, Revision 1A, "Initial RT NDT of Linde 80 Weld Materials", Approved August 2005.
- 7. Framatome ANP Topical Report BAW-2308, Revision 2A, "Initial RT NDT of Linde 80 Weld Materials", Approved March 2008.
- 8. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.
- 9. Andrachek, J. D., "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," WCAP-14040-A, Revision 4, May 2004.
4A-11 Revised 04/17/2013 C26C26 APPENDIX 4B PROCEDURE FOR PLUGGING A TUBE IN A STEAM GENERATOR Inspection and repair of defective steam generator tubes is governed by approved plant procedures. A typical sequence is as follows: 1.The reactor is shutdown and taken to cold shutdown condition; i.e., both primary and secondary sides are depressurized and cold. Decayheatis removed via the residual heat removal system. 2.The reactor coolant level is lowered until the level is between the bottom of the steam generator and the hot leg elbow, thus maintaining the remainder of the hot leg between the elbow and the vessel full of
water.Lowering of water to this level does not affect operation of the residual heat removal system because the residual heat removal suction
line is connected to the hot leg of loop C for Unit 3 and loop A for
Unit 4, and the return line is connected to the cold legs of all three
loops. 3.The steam generator is entered via the two manways, one on either side of the channel head partition plate. Prior to the performance of any
work, the area around the steam generator is monitored to determine the
radiation level. In the event of high radiation levels, biological
shielding is installed around the coolant channel head, and portable respiratory apparatus is used if required. Temporary nozzle covers are placed over the inlet and outlet reactor coolant legs to the steam
generator to prevent any debris from entering the reactor coolant
system.4.The defective tube is located and plugged. Remotely operated equipment may be used to locate and plug the defective tube. Tubes with indications may be plugged or plugged and staked. Remotely operated equipment may also be used to perform an in-situ pressure test on the defective tube prior to performing the repair activity.5.The temporary nozzle covers are removed and the manway covers are replaced thus resealing the system.6.The reactor coolant level is raised to its normal cold shutdown level, and the air which has been introduced into the steam generator is
vented n the normal manner i.e., in the same way as following a
refueling shutdown.
4B-1 Rev. 17 APPENDIX 4C
REPLACEMENT STEAM GENERATOR DESIGN
In 1982 and 1983 the Unit 3 and Unit 4 steam generator lower assemblies were
replaced. The new assemblies match the design performance of the original
assemblies. However, several design improvements have been made. This
Appendix 4C describes the design parameters of the new assemblies. The pages
herein are taken from Chapter 2 of the Steam Generator Repair Report as
amended, which was submitted to the NRC under FPL letter number L-77-296, dated September 20, 1977.
4C-1 Rev. 1-11/83 1.0 REPLACEMENT COMPONENT DESIGN
Westinghouse has fabricated new steam generator lower assemblies as
illustrated by Figure 4C-1. The design of the lower assemblies matches the
design performance of the lower assemblies being replaced. However, several
design improvements that do not alter mechanical, performance and FSAR
parameters are included in the design. These design features will improve
flow distribution, improve tube bundle access and reduce secondary side
corrosion. This section discusses the design and manufacture of the lower
assemblies.
1.1 COMPARISON WITH EXISTING COMPONENT DESIGN
1.1.1 Parametric Comparison
The steam generators for the Turkey Point plants, upon completion of the
repair, have physical, mechanical and thermal characteristics consistent with
the original design and safety analysis as currently documented in the FSAR.
The existing steam generators were built to the 1965 edition of the ASME
Boiler and Pressure Vessel Code (ASME Code); the new component parts of the
steam generators are designed and fabricated based upon the 1974 edition of the ASME Code, including all addenda through Summer 1976. The Stress Report l
is based upon the 1965 edition of the ASME Code, including all addenda
through Summer 1965. The replacement lower assemblies were fabricated and
analyzed to standards equivalent to the original units.
The replacement lower assembly incorporates a number of refinements in design
which are discussed in Section 1.2. During 1975 several modifications were
made to the installed steam generators to increase performance and promote
reliability. These modifications (described and noted in the text) were
retained or improved with the replacement lower assemblies. The
modifications accomplished at that time consisted of removing the downcomer
resistance plate, improving the moisture separators, modifying the blowdown
arrangement inside the steam generators, installing tube lane blocking
devices and modifications to the feedring to improve performance. These
modifications increased the circulation ratio and improved the units' ability
to resist sludge build-up.
Design data for the steam generators is presented in Table 4C-1 allowing
comparison between the present steam generators and the repaired units.
Improvements have been made for increased access to the secondary side of the
steam generators incorporating six 6-inch hand holes around the bundle in the
tube sheet area. The thermal data for each steam generator remains the same
as the original steam generator.
Since the replacement lower assemblies have been designed to incorporate
changes based on field experience, a number of minor changes in specific
components have been made which could affect the thermal hydraulic
performance of the unit. In order to maintain the original thermal and
hydraulic conditions, adjustment of heat transfer surface parameters was
necessary; changes in the support plate configuration and desire to improve
the circulation ratio resulted in a decrease in the number of tubes. These
4C-2 Rev. 17 modifications resulted in the reactor coolant water volume in the steam generator being reduced slightly, the secondary side volume being increased
slightly, a slight decrease in the amount of heat transfer surface area, as
well as a slight increase in the heat transfer coefficient. Imposing closer
manufacturing tolerances on the tube wall thickness results in an increase
in the overall heat transfer coefficient (approximately 2.5%) for the
repaired units. This increase in heat transfer coefficient offsets the
decrease in heat transfer area (approximately 2.2%) so that steam generator
heat transfer remains essentially unchanged.
Materials used in the fabrication of the replacement lower assemblies were procured to the requirements of the 1974 edition of the ASME Code, including
all addenda through Summer 1976. These materials are identical to those used l in the original steam generators except where specific design changes have
been recently incorporated or fabrication practice has changed. Specific
examples of these occurrences are enumerated as follows: plate material used
in the secondary shell formation has been changed to SA-533 Grade A Class 2
from SA302 Grade B Class 1 as a result of fabrication practices; support
plate material has been changed to SA-240 Type 405 from SA-285 Grade C as a
result of design changes to prevent corrosion. Material changes due to
design improvements do not degrade the physical, mechanical and thermal
properties of the steam generators. Further discussion is provided in
Section 1.2 and Table 4C-2 enumerates past and present applications of
materials. 1.1.2 Physical Compatibility With Existing Steam Generators and Systems New steam generator lower assemblies (see Figure 4C-1 were provided. These lower assemblies are designed to be identical physical replacements for the
existing units. Outside overall dimensions are the same as are the location
of the nozzles and support attachments. Interfaces between the steam
generators and plant components and systems are maintained. Dry and wet
weights of the steam generators remain approximately the same as are the
center of gravity; therefore, no changes to the present supports or their
configuration are necessary. 1.1.3 ASME Code Application The present operating steam generators were designed and constructed to the requirements of the 1965 edition of the ASME Code,Section III, Summer 1965
addenda. The replacement assemblies have been fabricated to the requirements of the 1974 edition of the ASME Code including all addenda through Summer l
1976. Design of the steam generators is consistent with the original design
of the reactor coolant system as well as the upper shell assembly of the
steam generators which were not replaced. Materials to be used in
fabrication were procured to the requirements of the current codes to
facilitate construction. All material certification tests were performed and
recorded as required by
current versions of the code. None of the requirements imposed on the
replacement assemblies inhibit the capability of the steam generators to meet
performance and FSAR safety requirements.
4C-3Rev.17 1.1.4 Regulatory Guide Application The compilation below addresses Regulatory Guides considered applicable to
the fabrication of the replacement lower assemblies. It must be noted that
these guides were issued subsequent to construction and operation of this
facility. The intent was to accommodate, consistent with facility design and
repair program objectives, the guidance by these regulatory guidelines.
1.26 Quality Group Classifications and Standards for Water, Steam and Radioactive-Waste-Containing Components of Nuclear Power Plants
(Rev.3, February 1976).
Westinghouse utilizes the classification system ANSI N18.2A-1975 for water and steam containing components. This classification
method assigns safety-related components to safety classes.
Assignment of the primary side of the steam generator to Safety
Class 1 and the secondary side to Safety Class 2 is consistent
with the quality groupings which would result from this
regulatory guide and 10 CFR 50.55a.
1.28 Quality Assurance Program Requirements (Design and Construction)
(Safety Guide 28, June 1972)
Westinghouse position on Regulatory Guide 1.28 is presented in WCAP-8370, "WRD Quality Assurance Plan". For activities which
occurred during the period from January 1, 1975 to September 30, 1977, the position is presented in WCAP-8370, Revision 7A. For
activities occurring on or after October 1, 1977, the position is
presented in WCAP-8370, Revision 8A.
1.31 Control of Stainless Steel Welding (Rev. 1, June 1973)
The Westinghouse production weld verification program, as described in WCAP-8324-A, was approved by the NRC as a
satisfactory substitute for following the recommendations of the
NRC Interim Position on Regulatory Guide 1.31 (4/74). The
results of the verification program support the hypothesis
presented in WCAP-8324-A; these results have been summarized and
documented in WCAP-8693, which has been submitted to the NRC for
information.
1.34 Control of Electroslag Weld Properties (December 28, 1972)
Where electroslag welding is used, Westinghouse requires its suppliers to follow the recommendations of this guide.
1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (March 16, 1973)
4C-4 Rev. 1-11/83 The Westinghouse position on Regulatory Guide 1.37 is presented in WCAP-8370, "WRD Quality Assurance Plan." For activities which occurred during the period from January 1, 1975 to September 30, 1977, the position is presented in WCAP-8370, Revision 7A. For
activities occurring on or after October 1, 1977, the position is
presented in WCAP-8370, Revision 8A.
1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving Storage, and Handling for Nuclear Power Plants (Rev. 2, May 1977)
The Westinghouse position of Regulatory Guide 1.38 is presented in WCAP-8370, "WRD Quality Assurance Plan". For activities which
occurred during the period from January 1, 1975 to September
30, 977, the position is presented in WCAP-8370, Revision 7A.
For activities occurring on or after October 1, 1977, the
position is presented in WCAP-8370, Revision 8A.
1.43 Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components (May 1973)
The Westinghouse Tampa Division uses materials made to fine-grain practice or which are not susceptible to underclad cracking.
These materials do not require the controls listed in the guide.
1.44 Control of the Use of Sensitized Stainless Steel (May 1973)
All of the unstabilized austenitic stainless steels used for component parts of the reactor coolant pressure boundary are
utilized in the final heat treated condition required by the
respective ASME Code,Section II, material specification for the
particular type or grade of alloy. Processing and fabrication
are performed using established methods and techniques to avoid
sensitization. Westinghouse has verified that these practices
will prevent sensitization by performing corrosion tests on
as-received wrought materials, as well as on production and
qualification weldments. In addition, the water chemistry in the
reactor coolant system is controlled to prevent intergranular
attack of unstabilized stainless steels; the effectiveness of
these controls has been demonstrated by both laboratory tests and
operating experience.
1.48 Design Limits and Loading Combinations for Seismic Category I Fluid System Components (May 1973)
Westinghouse meets the requirements of General Design Criterion 2 and will thereby satisfy the concerns of Regulatory Guide 1.48.
The loading combinations and design limits used in the code
stress analysis of the steam generator are the same as those in
the Turkey Point FSAR.
4C-5 Rev. 1-11/83 1.50 Control of Preheat Temperature for Welding of Low-Alloy Steel (May 1973)
Westinghouse practices are in agreement with Regulatory Positions C.1.a, C.3 and C.4. For Regulatory Position C.1.b, Westinghouse qualifies welding procedures within the preheat temperature
ranges required by Section IX of the ASME Code. For Regulatory
Position C.2, Westinghouse uses the methods documented in
WCAP-8577-A, which has been accepted by the NRC. 1.58 Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel (August 1973)
The Westinghouse position on Regulatory Guide 1.58 is presented in WCAP-8370, "WRD Quality Assurance Plan". For activities which
occurred during the period from January 1, 1975, to September
30, 1977, the position is presented in WCAP-8370, Revision 7A.
For activities occurring on or after October 1, 1977, the
position is presented in WCAP-8370, Revision 8A. 1.64 Quality Assurance Requirements for the Design of Nuclear Power Plants (Rev. 1, February 1975) The Westinghouse position on Regulatory Guide 1.64 is presented in WCAP-8370, WRD Quality
Assurance Plan". For activities which occurred during the period
from January 1, 1975 to September 30, 1977, the position is
presented in WCAP-8370, Revision 7A. For activities occurring on
or after October 1, 1977, the position is presented in WCAP-8370, Revision 8A. 1.66 Nondestructive Examination of Tubular Products (October 1973)
Steam generator nozzles are either radiographed or ultrasonically tested in the circumferential and axial directions in accordance
with the guides' positions. Steam generator tubing receives eddy
current, circumferential ultrasonic testing, and hydrostatic
testing to satisfy the guides' recommendations. 1.71 Welder Qualification for Areas of Limited Accessibility (December 1973)
Westinghouse practice does not require qualification of welders for areas of limited accessibility. Shop welds are repetitive and closely supervised and the ASME Code, Sections III and IX
requirements are followed. 1.83 Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes (Rev. 1, July 1975)
Westinghouse steam generators are designed to permit access to tubes for inspection and plugging. A pre-service inspection of
the steam generators was conducted to establish baseline
conditions.
4C-6 Rev. 1-11/83 1.84 Code Case Acceptability - ASME Section III Design and Fabrication (June 1974) (Rev. 1, April 1975) (Rev. 2, June 1975) (Rev. 3, September 1975) Rev. 4, November 1975) (Rev. 5, February 1976)
(Rev. 6, May 1976) (Rev. 7, August 1976) (Rev. 8, November 1976)
(Rev. 9, March 1977)
1.85 Code Case Acceptability - ASME Section III Materials (June 1974) (Rev. 1, April 1975) (Rev. 2, June 1975) (Rev. 3, September
1975) (Rev. 4, November 1975) (Rev. 5, February 1976) (Rev. 6, March 1976) (Rev. 7, August 1976) (Rev. 8, November 1976) (Rev.
9, March 1977)
- 1. Westinghouse controls its suppliers to:
- a. Limit the use of code cases to those listed in Regulatory Position C.1 of the applicable guide
revision in effect at the time the equipment is
ordered, except as allowed in item 2 below.
- b. Identify and request permission for use of any code cases not listed in Regulatory Position C.1 of the
applicable guide revision in effect at the time the
equipment is ordered, where use of such cases is
needed by the supplier.
- c. Allow continued use of a code case considered acceptable at the time of equipment order, where such
code case was subsequently annulled or amended.
- 2. Westinghouse seeks NRC permission for use of code cases needed by suppliers and not yet endorsed in Regulatory
Position C.1 of the applicable guide revision in effect at
the time the equipment is ordered and permits supplier use
only if NRC permission is obtained or is otherwise assured (e.g., a later version of the regulatory guide includes
endorsement)
1.88 Collection, Storage and Maintenance of Nuclear Power Plant Assurance Records (Rev. 2, October 1976)
The Westinghouse position on Regulatory Guide 1.88 is presented in WCAP-8370, Revision 7A. For activities
occurring on or after October 1, 1977, the position is
presented in WCAP-8370, Revision 8A.
1.123 Quality Assurance Requirements for Control of Procurement of Items and Service for Nuclear Power Plants (Revision 1, July
1977)
The Westinghouse position on Regulatory Guide 1.123 is presented in WCAP-8370, "WRD Quality Assurance Plan". For activities which
occurred during the period from January 1, 1975 to September 30, 1977, the position is presented in WCAP-8370, Revision 7A. For
activities occurring on or after October 1, 1977, the position
is presented in WCAP-8370, Revision 8A.
4C-7 Rev. 1-11/83 2.2 COMPONENT DESIGN IMPROVEMENTS
As noted, the physical, thermal and hydraulic characteristics of the steam
generators essentially duplicate those of the original units. However, design changes which do not alter FSAR safety requirements have been
incorporated in the design. These changes increase the operating
availability and improve resistance to corrosion of the secondary side
thereby minimizing the potential for future repair efforts. Figure 4C-1
illustrates some of these improvements. It should be noted that some of
these features have been installed in the in situ units (see Section 1.1.1).
Research, development and testing have been utilized to select design
parameters, material and component configurations which will prevent
degradation of the repaired steam generators. Confirmatory tests in model
boilers and other tests on the material and component configuration are
continuing.
1.2.1 Design Refinements To Prevent And Inhibit Corrosion
1.2.1.1 Increased Circulation Ratio
Circulation ratio is defined as the total tube bundle flow divided by the
feedwater flow and is inversely proportional to the steam quality exiting the
tube bundle. As the circulation ratio increases, certain parameters of the
steam generator, such as lateral velocity sweeping across the tubesheet, steam quality, void fraction and number of tubes exposed to sludge, change in
a favorable direction. Low steam quality in the bundle reduces tube exposure
to local steam blanketing. This also reduces the number of potential areas
of concentration for chemical impurities. In addition, higher circulation
ratios increase the fluid velocity sweeping across the tubesheet to the
center of the bundle. Specific design changes, such as the quatrefoil plates (See Subsection 1.2.1.8), modification in the tube bundle size and wrapper to
shell distance, influence the circulation ratio.
1.2.1.2 Flow Distribution Baffle
A flow distribution baffle has been provided 18 inches above the tubesheet.
This baffle has a cut out center section and oversized drilled tube holes.
The increased circulation ratio provides a greater lateral flow across the
tubesheet surface. The baffle plate will assist in directing this flow
across the tubesheet then up the center of the bundle through the center
cutout. The design is sized to minimize the number of tubes exposed to
sludge. Consistent with this purpose, the design causes the sludge to
deposit in and near the
center of the bundle at the blowdown intake. The flow distribution baffle
plate material is ferritic stainless steel. Figure 4C-2 illustrates the flow
distribution baffle.
While the baffle will direct flow toward the center of the bundle, the
average velocity around the tubes will be sufficient to prevent sludge from
settling. In addition, as noted, access holes have been provided to allow
sludge lancing of the baffle plate.
4C-8 Rev. 1-11/83 1.2.1.3 Improved Internal Blowdown Design
Each steam generator was designed to have two 2-inch schedule 40 Inconel
internal blowdown pipes. The blowdown rate from the steam generators is
varied as required by chemistry conditions in the feedwater and as monitored
in the blowdown. Maintenance of the steam side water chemistry is assisted
through the use of the blowdown system. Continuous blowdown of the steam
generator provides a dynamic system which is constantly removing impurities
from the steam generator. During hot standby and hot functional testing, blowdown is employed, as needed, to maintain the steam generator chemistry
within specification. The blowdown intake location is coordinated with the
baffle plate design so that the maximum intake is located where the greatest
amount of sludge is expected to deposit. The improved blowdown system allows
higher capacity blowdown in comparison with the present blowdown arrangement.
1.2.1.4 Tube Expansion in Tubesheet
Following insertion into the tubesheet hole, tack rolling, welding and gas
leak testing, the tubes are expanded to the full depth of the tubesheet hole.
Full-depth expansion prevents crevice boiling. In addition it prevents a
buildup of impurities from forming in the crevice region. The present steam
generator tubes were only partially expanded in the tubesheet.
1.2.1.5 Thermally Treated Inconel 600 Tubing
Research by Westinghouse has determined that significant improvement in the
stress corrosion resistance of Inconel 600 tubing can be achieved by
modification of the metallurgical structure through thermal treatment. The
primary objective of this treatment is to develop an improved metallurgical
structure, associated with grain boundary precipitate morphology, which
provides increased margin with respect to stress corrosion performance.
Several benefits result from this treatment such as improved resistance to
stress corrosion cracking in NaOH, resistance to intergranular attack in
oxygenated environments, resistance to intergranular attack in sulphur-
containing species and reduction of residual stress imparted by tube
processing.
Studies conducted at Westinghouse and elsewhere have indicated that certain
heat treatments can improve caustic stress corrosion resistance but result in
a chromium-depleted grain boundary layer (sensitization) which is not as
resistant to off-chemistry environments, should they be experienced.
However, analysis of available data also indicates that there is a broad band
of temperature and time within the typical sensitization range for Inconel
600 which provides improved resistance to stress corrosion cracking in both
caustic and pure water environments. Thermal treatment in this time-
temperature band avoids formation of the chromium depleted grain boundary
layer. The thermal treatment to be used was within this time-temperature
band.
4C-9 Rev. 1-11/83 1.2.1.6 Offset Feedwater Distribution
Previously, feedwater flow within the steam generators was modified so that
80 percent of the flow is directed to the hot leg side of the bundle and 20
percent of the flow is directed to the cold leg side of the bundle. This
reduces the steam quality in the hot leg side of the bundle and raises the
steam quality in the cold leg side of the bundle. The effect of these
changes in steam quality is to shift the point of highest steam quality at
the tubesheet elevation toward the center of the bundle. The point of
highest steam quality has the lowest density and is, therefore, a likely
region for chemical concentration and sludge deposition. This area is
utilized for location of blowdown intake. Feedwater flow distribution is
accomplished by providing a greater number of flow paths on the portion of
the feedwater ring which traverses the hot leg side of the tube bundle.
These modifications were maintained in the replacement assemblies.
1.2.1.7 Corrosion Resistant Support Plate Material
Corrosion in the crevice between the tube and tube support plate has led to
denting of the tubing in that area and in some cases affected the steam
generator performance in general. Alternative support plate materials have
been evaluated, and SA-240 Type 405 ferritic stainless steel has been
selected as the optimum material for this application. This material is ASME
Code approved and is resistant to corrosion with the chemistry expected
during the operation of the steam generator. In addition, SA-240 has a low
wear coefficient when paired with Inconel and has a coefficient of thermal
expansion similar to carbon steel. Corrosion of SA-240 results in an oxide
which has approximately the same volume as the parent material, whereas
corrosion of carbon steel results in oxides which have approximately two
times the volume of the parent material. Type 405 also has material
properties such as machinability and weldability which are comparable to
carbon steel. In addition to the tube support plates, the baffle plate (discussed in Subsection 1.2.1.2) was constructed of SA-240 Type 405.
1.2.1.8 Quatrefoil Tube Support Plates
The quatrefoil tube support plate design, illustrated by Figure 4C-3, consists of four flow lobes and four support lands. The lands provide
support to the
tube during all operating conditions, while allowing flow around the tube.
This design has a lower pressure drop than the most current circulation hole
designs. This low secondary pressure drop increases the circulation ratio
which, when combined with other improvements, translates into higher sweeping
velocities and fewer tubes exposed to a low steam quality at the tubesheet.
This design directs the flow along the tubes which limits steam formation and
chemical concentrations at the tube-to-tube support plate intersections. The
quatrefoil support place design results in higher average velocities along
the tubes, preventing sludge deposition. The combination of higher
velocities in the support plate region and corrosion resistant material will
minimize the
possibility of support plate corrosion.
4C-10 Rev. 1-11/83 1.2.2 Design Refinements To Improve Performance In the course of evolution of the steam generator design, as derived from operating experience and ongoing research and development programs, certain
improvements and refinements have been incorporated in recent designs to
improve performance of thermal hydraulic characteristics. These improvements
are included in the FPL design and are discussed below. They do not alter
FSAR safety requirements. 1.2.2.1 Recessed Tube to Tubesheet Weld
The tubes on the replacement lower assemblies were recessed slightly into the tubesheet holes and then welded to the tubesheet cladding. Elimination of
the protruding tube stub of the original design results in lower entry
pressure
losses and, therefore, a lower pressure drop in the primary loop. In
addition, a possible point of crud buildup and corrosion is likely avoided
with this design. This is illustrated in Figure 4C-4. 1.2.2.2 Tube Lane Blocking Device
Recirculating water exiting at the bottom of the wrapper will tend to preferentially channel to the tube lane and bypass part of the tube array.
In order to prevent this tube bundle bypass, a series of plates were
installed in the tube lane during prior modifications. These plates are
arrayed so that
there will be minimal interference with sludge lancing. These blocking
devices were retained in the replacement units. 1.2.2.3 Moisture Separator Improvements
Since the circulation ratio in the steam generator has increased, the duty for the moisture separator equipment will increase. To accommodate this
increase, several improvements were incorporated. New demister vanes were
installed to increase the efficiency of the moisture separators. Perforated
plates were
installed on the face of the demister vane housing to distribute the flow
evenly through the demisters and provide better moisture separating. The
swirl vane barrels previously modified with optimized orifice plates were
realigned. These improvements are shown in Figure 4C-5. 1.2.3 Design Changes To Improve Maintenance And Reliability Operational experience, including necessary maintenance and repair, has resulted in certain changes in design which are directed to improving the
maintainability and ultimately the reliability of the units. Other changes
have been incorporated to prevent occurrences of operational problems which
have been experienced. These changes are discussed below and do not alter
performance or FSAR safety requirements.
4C-11 Rev. 1-11/83 1.2.3.1 Access Ports
The lower assemblies were constructed with additional access ports. Four
6-inch access ports are located slightly above the tube sheet, approximately
90 degrees apart, with two located on the tube lane. Two 6 inch access ports
are located on the tube lane, between the flow distribution baffle and the
first tube support plate. The addition of these access ports improve and
promote inspection of the tube sheet and flow distribution baffle and assist
in sludge lancing.
1.2.3.2 Wet Layup Nozzle
A 2-inch nozzle was added to the upper shell to facilitate the wet layup of
the steam generators during periods of inactivity. The wet layup nozzle can
be used for addition of chemicals during these periods to prevent any
excursions of the water quality in the steam generator. The nozzle can also
be used in conjunction with other systems to circulate water through the
steam generator during periods of layup.
1.2.3.3 Primary Shell Drain
A 3/8 inch primary shell drain is included in the channel head to improve
drainage of the channel head. The improved drainage will lessen downtime and
facilitate any maintenance or inspection to be conducted in the channel head.
1.2.3.4 Primary Closure Rings
Closure rings were welded inside the channel head at the base of each primary
nozzle so that closure plates can be installed during primary chamber
maintenance. This design allows the plates to be bolted to the rings for
quick installation and removal. Closure plates allow maintenance or
inspection to be conducted in the channel head with the reactor cavity
flooded.
1.3 SHOP TESTS AND INSPECTIONS
The tests and inspections required by the ASME Code,Section III were
conducted during the fabrication of the steam generator lower assembly. In
addition to the ASME requirements, further tests and inspections were
conducted at the fabrication facility. The primary side of the steam
generator was hydrotested at the shop in accordance with approved procedures.
Each tube was individually hydrotested prior to use in fabrication. After
the tube bundle installation is completed, a gas leak test was performed to
demonstrate the integrity of the tube-to-tubesheet welds.
4C-12 Rev. 1-11/83 1.4 Onsite Storage Facility A temporary storage facility provided for the storage of the steam generator lower assemblies. The lower assemblies are stored in this
area until they can be shipped offsite to a licensed land burial site
or decommissioned with the plant. Prior to removal from the
containment, the openings in the lower assemblies were sealed to
prevent the release of radioactivity during transfer and subsequent
on-site storage.
The only radiological consideration associated with storage is the direct radiation from the steam generators. Shielding is provided to
ensure acceptable radiation levels external to the storage facility.
There are no accident considerations associated with on-site storage.
Based on the above considerations, the required storage facility design criteria are:
- a. Appropriate shielding for direct dose
- b. Provisions for periodic surveillance of steam generator seal
integrity
- c. Total enclosure of the sealed steam generators is not required.
The facility is founded on engineered fill at finished grade elevation +17'-6" MLW in the area approximately 150 feet south of the
ash disposal pits and 290 feet east of the Radwaste Building (Figure
4C-6). The elevation of the area ranges from +6 to +9 feet MLW. At
the storage facility location, the former surface layer consisted of
4 feet of limerock fill, underlain by about 6 feet of muck. Beneath
the muck, Miami limestone extends 20 feet, underlain by Key Largo
limestone to about elevation -100 feet.
Prior to construction of the facility, the existing limerock fill and muck were removed from below the potential zone of influence of the
building foundation, and replaced with compacted crushed limerock fill
up to elevation +17'-6" MLW. The existing muck and fill was excavated
to a minimum distance of 15 feet beyond the edge of the building. The
building is at least 65 feet back from the top of the compacted crushed
limerock fill boundary slope. This slope is 1-vertical on
3-horizontal.
The crushed limerock fill which supports the facility was quarried from local Miami and Key Largo limestone formations. Maximum size is about
6 inches, with up to 20 percent passing the No. 200 sieve. The crushed
limerock was stockpiled on-site to drain effectively to the optimum moisture range between 7 percent and 14 percent. The fill was placed
using a maximum loose thickness of 12 inches. The fill was compacted
with a vibrating drum roller to obtain a minimum dry density of 110 lb.
per cubic foot. An extensive series of laboratory tests on the crushed
limestone compacted to 110 lb. per cubic foot dry density has indicated
effective strength parameters of conservatively 3 kips
4C-13 Rev. 1-11/83 per square foot for cohesion and 39 degrees for internal angle of friction.
The compacted crushed limerock fill provides an allowable bearing capacity
over 5 kips per square foot, including a factor of safety of 3.
The Miami and Key Largo limestone formations underlying the compacted crushed
limerock fill have allowable bearing capacities over 30 kips per square foot
(including a factor of safety of 3).
The fill area was designed for a 100 year flood level of +12.7 feet MLW as
the design storm surge required for buildings in southern Florida. This 100
year flood level is per the Code of Metropolitan Dade County Florida.
The steam generator storage facility was designed in accordance with the
following current codes and standards:
South Florida Building Code
Code of Metropolitan Dade County Florida
Building Code Requirements for Reinforced Concrete (ACI 318)
American Institute of Steel Construction Manual of Steel Construction
and Specification of the Design, Fabrication and Erection of Structural
Steel for Buildings
American Welding Society Structural Welding Code (AWS D1.1)
The facility was designed for a hurricane wind velocity of 120 miles per hour
with application of shape factors in accordance with the American Society of
Civil Engineers Paper No. 3269.
The structure consists of 2'0" thick reinforced exterior concrete shield
walls sized to maintain a direct gamma dose rate of less than or equal to 2.5
mr/hr at exterior wall surfaces. The facility is approximately 130 feet by
42 feet with centerline along its length oriented in the East to West
direction. A 2'-0" thick reinforced interior concrete wall was provided
through the center of the structure for the full length of the facility. The
interior wall provides roof support and separation between the Unit 3 and
Unit 4 steam generator lower assemblies. All walls are founded on continuous
strip footings with bases at approximately elevation +15'-6" MLW, 2 feet
below the finished grade elevation. A maze shielded entrance with a door
through the exterior wall is provided to allow for periodic surveillance of
lower assembly seal integrity. Each lower assembly weighing approximately 186 tons (Specific Gravity 1.88) with two steel support saddles is stored in the facility on reinforced concrete bearing pads. A 6 inch thick reinforced concrete floor is provided. Top of floor and pad elevation is
+18'-0" MLW.
4C-14 Rev. 1-11/83 Wall footings and bearing pads were designed to maintain a maximum allowable
soil bearing pressure of 5000 pounds per square foot.
Other design loads for the structure are in accordance with the South Florida
Building Code. Concrete has a minimum compressive strength of 3000 psi at 28
days. Reinforcing steel and structural steel is in accordance with ASTM A6l5 (Grade 60) and ASTM A36 respectively. Concrete and steel allowable stresses
are in accordance with Building Code Requirements for Reinforced Concrete (ACI 318) and Specification for the Design, Fabrication and Erection of
Structural Steel for Buildings (AISC).
The design of the wall thickness was determined using a point-kernal computer
code which used semi-empirical methods developed by Rockwell (1) for calculating the direct gamma dose rates from a homogeneous volumetric
cylindrical source through slab shields.
The values of the source terms for the analysis were based on the results of
a field survey of a steam generator in a drained condition one month after
shutdown. For conservatism it was assumed that all short-lived isotopes had
decayed away and the sole contributor to the measured dose rate was
cobalt-60, which has the highest average gamma ray energies and is therefore
the most
difficult to shield for a given curie level. The results of the conservative
analysis indicate that 24-inch concrete walls are required to meet the dose
criteria of 2.5 mr/hr at the exterior wall surfaces. However the dose at the
exterior wall surface is expected to be at or below 0.25 mr/hr.
The skyshine analysis was performed with an industry-recognized computer code
G 3(2) based upon the same field survey of a steam generator, assuming the average energy of Co 60 as that of the source. The source strength of the isotropic point source was determined by calculating a normalization constant
equal to the total photon leakage from the steam generator. The skyshine
contribution, without taking credit for a shielding roof, will not increase
the dose rate outside the compound over 0.25 mr/hr.
The resulting dose equivalent to an individual at the north site boundary
location for a full year was calculated assuming 2.5 mr/hr at the outside
surfaces of the storage compound, plus the skyshine contribution assuming no
roof on the storage facility. The calculated dose was 5.2 x 10
-3 mrem which is considered an insignificant contribution of the offsite dose. The
presently proposed facility location (see Figure 4C-6) was assumed for the
aforementioned analysis.
(1) T. Rockwell, Reactor Shielding Design Manual , D. Van Nostrand Co., New York (1956).
(2) R. E. Malefant G 3: A General Purpose Gamma-Ray Scattering Program , Los Alamos Scientific Laboratory, LA 5176 (June 1973).
4C-15 Rev. 1-11/83 The facility roof was designed to be watertight. The major roof components consist of precast concrete roof panels with concrete topping for a thickness
of approximately 1 foot. The elevation of top of roof is approximately
+39'-6"MLW. The lower assemblies were lowered into the facility by cranes
with subsequent installation of the roof. A center wall in the facility
allows for storage of Unit 4 assemblies on one side of the facility and
storage of the Unit 3 assemblies on the opposite side of the facility.
Since the steam generators are welded in addition to being in a facility
having a watertight concrete roof and reinforced concrete walls, there are no
potential means to transport the surface contamination from the lower
assembly surfaces. Therefore a floor, sumps and/or air filtration units are
not required. However, as previously stated, a floor is provided.
An evaluation was performed to determine the most man-rem effective and
cost-beneficial method for disposition of the removed lower assemblies. It
was concluded that the lowest cost man-rem burdens are associated with (1)
long-term, on-site storage and disposal during decommissioning and (2)
immediate barge shipment to a licensed land burial facility.
Figure 4C-7 is a typical decay curve of percent steam generator gamma
activity versus time following reactor shutdown. The initial activities used
to generate the decay curve are given in Table 4C-3. With this decay curve, the effect of lower assembly radioactive decay can be directly related to the
time of ultimate steam generator disposal and to the associated man-rem
exposures.
4C-16 Rev. 1-11/83 TABLE4C-1STEAM GENERATOR DESIGN DATA (PER STEAM GENERATOR) Original (44) RSG (44F) at Uprate (2) RSG at EPU Power Level (MWt) 736769.3 (5)884 (6) Design Pressure, Reactor Coolant/Steam, psig 2485/1085N.C.
(1)(4)N.C.(1) Reactor Coolant Hydrostatic Test Pressure (tube side), psig 3107N.C.(1)N.C.(1) Hydrostatic Test Pressure, Shell side, psig 1356N.C.(1)N.C.(1) Design Temperature, Reactor Coolant/Steam, o F 650/556N.C.
(1)N.C.(1) Steam Conditions at 100% Load and 0%
Tube Plugging (per SG), Outlet Nozzle: N.C.(1)N.C.(1) Steam Flow, lb per hr. 3.39 x 10 63.39 x 106(4) 3.87 x 106 (6) Steam Temperature, o F 522.8516.3514.8 (6) Steam Pressure, psig 817772 (4)762 (6) Feedwater Temperature at 100% Load o F 443443(4)440 (6) Overall Height, ft-in 63-1.6N.C.(1)N.C.(1) Shell OD, upper/lower, in. 166/127N.C.
(1)N.C.(1) Shell Thickness, upper/lower 3.5/2.63N.C.
(1)N.C.(1) U-tube OD, in. 0.875N.C.(1)N.C.(1) Tube Wall Thickness, (nominal) in. 0.050N.C.(1)N.C.(1) Number of Manways/ID, in. 4/16N.C.(1) N.C.(1) Number of Handholes/ID, in. 2/66/6 6/6Number of U-tubes 32603214 (5)3214 (6) Tube length (largest U-bend), in. 397.5N.C.(1)N.C.(1) Total Heat Transfer Surface Area, ft 2 44,430 43,467 (5)43,467 (6) Reactor Coolant Water Volume, ft.
3 925N.C.(1)(5)N.C.(1)(6) Reactor Coolant Flow, lb/hr 32.1 x 10 632.4 x 106(4) 33.0 x 10 6 (6)Secondary Side Volume, ft.
3 45804682 (5)4698 (6) Secondary Side Mass No Load, lbs 134,000N.C.
(1)N.C.(1) Secondary Side Mass at 100% Power, lbs 76,30081,500 (5)81,775 (6) Center of Gravity (from support pads), ft./in. 25/4N.C.(1)N.C.(1) NOTE: (1) N.C. means there was No Change to design data. (2) Reflects Replacement Steam Generator (RGS) at thermal power uprate conditions.
(3) Reflects Replacement Steam Generator (RGS) at thermal power EPU uprate conditions. (4) According to PCWG-2779, at lowest full power operating T avg of 571.2 oF. (5) According to WTD-TH-79-001, Rev.5.
(6) PCWG-08-34 Case at 400 oF feedwater and lowest full power operating T avg of 577.0 o F 4C-17 Revised 04/17/2013 C26 TABLE4C-2 STEAM GENERATOR MATERIALS Original Repaired Plate (shell courses) SA-302 Grade B SA-533 Grade A Class 2 Tube Sheet Forging SA-336 SA-508 Class 2a Channel Head Casting SA-216 Grade WCC SA-216 Grade WCC Support Plates SA-285 Grade C SA-240 Type 405 Channel Head Cladding Stainless Steel, Stainless Steel Type 304 or equiv. Type 304 or equiv.
Tube Sheet Cladding Inconel Inconel Tubes SB-163 SB-163 Thermally Treated 4C-18Rev.1-11/83 TABLE4C-3 ESTIMATED CORROSION PRODUCT ACTIVITIES ON STEAM GENERATOR PRIMARY SIDE PLENUM (1) (2) (3) (4) (5) (6)
Activity Activity Isotope (Ci/cm2)Isotope (Ci/cm2) Notes (1) The activities are based on actual Turkey Point data.
(2) Activities listed are extrapolated to 9 years of commercial operation.
(3) For Unit 4 (approximately 7 years of commercial operation) activities are bounded by those for Unit 3 (approximately 9 years of commercial operation).
(4) The activities shown are for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> after shutdown.
(5) Multiplication Factors for Isotopic Concentration for Components in the steam generator:
Relative Component Concentration Factor Area (cm2)
Tubes 0.12 4.1 x 10 7 Divider Plate 2.0 7.2 x 10 4 Tube Sheet 2.0 3.8 x 10 4 Rolled Tube Ends 45/tube end 6520 tube ends Channel Head Bowl 1 1.5 x 10 5 (6) The amount of each isotope, in curies, can be obtained by decaying the isotope for 80 days (the estimate for the
earliest that the generators can be removed from the
containment); by multiplying the surface area for each
component by the primary side concentration and the relative
concentration factor; and by summing for all components.
4C-19Rev.1-11/83 TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR REPAIR REPORT STEAM GENERATOR LOWER ASSEMBLY FIGURE 4.C-1
TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR REPAIR REPORT FLOW DISTRIBUTION BAFFLE AND BLOWDOWN FIGURE 4.C-2
TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR REPAIR REPORT QUATREFOIL TUBE SUPPORT PLATES FIGURE 4.C-3
TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR REPAIR REPORT TUBE TO TUBESHEET JUNCTUNE FIGURE 4.C-4
Rev. 1-11/83 TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR REPAIR REPORT MOISTURE SEPARTOR IMPROVEMENTS FIGURE 4.C-5
Rev. 1-11/83 TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR REPAIR REPORT STEAM GENERATOR STORAGE COMPOUND Revision 7 March 199 FIGURE 4.C-6
Rev. 1-11/83 TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR REPAIR REPORT LOWER ASSEMBLY GAMMA RADIOACTIVE DECAY VS TIME Revision April 1978 FIGURE 4.C-7