L-2017-074, St. Lucie, Unit 1, Updated Final Safety Analysis Report, Amendment No. 28, Chapter 5, Reactor Coolant System
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REACTOR COOLANT SYSTEM CHAPTER 5 LIST OF TABLES Table Title Page 5.1-1 Parameters of Reactor Coolant System 5.1-4 5.1-2 Nozzle Identification 5.1-5 5.1-3 Reactor Coolant System Volumes 5.1-6 5.2-1 Reactor Coolant System Pressure Boundary Code Requirements 5.2-33 5.2-2 Loading Combinations and Primary Stress Limits 5.2-34 5.2-2a Loading Combinations and Primary Stress Limits Reactor Coolant 5.2-36 Pumps 5.2-3 Active Valves in the Reactor Coolant System Boundary 5.2-38 5.2-4 Major Component Material Specifications 5.2-40 5.2-4a Chemical Analyses of Plate Material in Reactor Vessel Beltline 5.2-43 5.2-5 Materials Exposed to Coolant 5.2-44 5.2-6 Charpy V-Notch and Drop Weight Test Values Reactor Vessel 5.2-45 5.2-7 Charpy V-Notch and Drop Weight Test Values Pressurizer 5.2-46 5.2-8 Charpy V-Notch and Drop Weight Test Values - Steam Generator 1A 5
.2-47 5.2-9 Charpy V-Notch and Drop Weight Test Values - Steam Generator 1B 5.2-48 5.2-10 Charpy V-Notch and Drop Weight Test Values Reactor Coolant 5.2-49 Piping 5.2-11 Reactor Coolant Leak Detection Sensitivity 5.2-53 5.2-12 Pressure Isolation Valves 5.2-55
UNIT 1 5-vii Amendment No. 27 (04/15)
CHAPTER 5 LIST OF TABLES (Cont'd)
Table Title Page 5.7-1 Failure Modes Effects Analysis for the Reactor Coolant Gas Vent 5.7-7 System 5A-1 Safety Valve Relief Capacities 5A-12 5C-1 Model Geometry Description (Deleted) 5C-3
5-ix Amendment No. 26 (11/13)
CHAPTER 5LIST OF FIGURES (Cont'd)FigureTitle5.5-13Steam Generator Feed Ring5.5-14Flow Diagram - Miscellaneous Systems5A-1Steam Generator Pressure Complete Loss of Turbine Generator Load with DelayedReactor Trip5A-2Optimized Safety Valve Sizing5A-3Maximum Reactor Coolant System Pressure vs. Time for Worst Case Loss of LoadIncident with Delayed Reactor Trip5A-4Maximum Reactor Power vs. Time for Worst Case Loss of Load Incident5A-5+10% Power Step5A-6+10% Power Step5A-7+10% Power Step5A-8+10% Power Step5A-9-10% Power Step5A-10-10% Power Step 5A-11-10% Power Step 5A-12-10% Power Step 5A-13Generator Trip 5A-14Generator Trip 5A-15Generator Trip 5A-16Generator Trip 5A-175% Per Minute Power Ramp (51% to 18% Power) 5A-185% Per Minute Power Ramp (51% to 18% Power)5-xiii Amendment No. 21 (12/05)
CHAPTER 5 LIST OF FIGURES (Cont'd)
Figure Title
5A-19 Level vs. Time During Loss of Load with Delayed Reactor Trip 5A-20 Normalized Pressure vs. Time During Loss of Load with Delayed Reactor Trip 5A-21 Normalized Steam Enthalpy During Loss of Load with Delayed Reactor Trip 5A-22 Normalized Safety Valve Flow Rate vs. Time During Loss of Load with Delayed Reactor Trip 5C-1 Model Volume, Junction, and Conductor Geometry (Deleted)
5C-2 Reactor Coolant Temperature vs. Time Cooldown at 30F/Hr to 325F 5C-3 Reactor Coolant Temperature vs. Time Cooldown at 50F/Hr to 325F 5C-4 Recommended Cooldown Guideline (Deleted)
5-xiv Amendment No. 26 (11/13)
TABLE 5.1-2 NOZZLE IDENTIFICATION (See Figure 5.1-1)
Code End Let.Function Pipe Size Preparation Qty.A Pressurizer Surge 12 SCH 160 Butt Weld 1 B Shutdown Cooling Outlet 12 SCH 160 Butt Weld 1 C Safety Injection and Shutdown Cooling Inlet 12 SCH 160 Butt Weld 4 D Pressurizer Spray 3 SCH 160 Butt Weld 2 E Charging Inlet 2 SCH 160 Butt Weld 2 F Letdown & Drain 2 SCH 160 Butt Weld 1 G Drain 2 SCH 160 Butt Weld 4 H Temp. Measurement, RTD Socket Weld 25 J Flow Measurement 3/4 SCH 160 Socket Weld 8 K Sampling 3/4 SCH 160 Socket WeldL Feedwater 18 SCH 80 Butt WeldM Steam Outlet 34" Butt Weld 2 N Flow Measurement 1" SCH 160 Socket Weld 8 P Q R
S Pump Middle Seal Vent & Pressure 3/4" 1500#
R.F. Flange 12 T Pump Upper Seal Vent & Pressure 3/4" 15 00#R.F. Flange 4 U Pump Lower Seal Vent & Pressure 3/4" 150#R.F. Flange 4 V Pump Intergasket Leakage Monitor 3/4" 150#R.F. Flange 4 W Pump and Motor Cooling Water Inlet 1 1/2" 150#
R.F. Flange 4 X Pump and Motor Cooling Water Outlet 1 1/2" 150#
R.F. Flange 4 Y O-Ring Seal Monitor Tube 3/4" SCH 80 Butt Weld 2 Z Safety Valve 3" 2500#R.F. Flange 3 AA Steam Generator Level 1" SCH 80 Butt Weld 16 AB Steam Generator Pressure 1" SCH 80 Butt Weld 2 AC Steam Generator Blowdown 2" SCH 80 Butt Weld 2 AD Reactor Vessel Vent 3/4 2500#R.F. Flange 1 AE Pressurizer Relief 4" SCH 160 Butt Weld 1 AF Spray Nozzle 4" SCH 160 Butt Weld 1 5.1-5
FLORIDA POWER & LIGHT COMPANYFLOW DIAGRAM REACTOR COOLANT SYSTEMAmendment No. 16, (1/98)
Pressure and temperature fluctuations resulting from the above transients are computed by means of computer simulations of the reactor coolant system. Computer output entailing time dependent physical
parameters throughout the reactor coolant system are detailed in the component specifications. The
component vendor then uses the specification transient curves as the basis for fatigue design.
Fatigue analysis for each component of the reactor coolant system is performed in accordance with the applicable ASME codes. As appropriate the combined effects of the load and thermal transients specified for
each condition of cyclic operation are evaluated as a function of time. The evaluations are performed in a
manner to yield the maximum range in stress intensity during the particular cyclic condition under
consideration. In those cases where conservative results are produced, peak stresses due to pressure may
be combined with those due to thermal transients by direct superposition. In addition, the results of analysis
obtained for the most severe transient condition in a group may be applied in evaluating the cumulative effects
of the entire group.
Pressure and thermal stress variations associated with the above design transients are included in the engineering design of each of the reactor coolant system components, piping, and supports. In addition, the
loads and moments resulting from the design transients are included in the design of equipment support
foundations and interfacing support structures for the equipment.
Pressure fluctuations associated with the transients apply to the following components and piping:
a)Four safety injection system discharge isolation check valves adjacent to reactor coolant loop and any piping and supports between these valves and reactor coolant system.
b)The two suction isolation valves on each of the two shutdown cooling lines and the piping and
supports between these valves and the reactor coolant system.
c)The reactor coolant system safety an d power-operated relief valves on the pressurizer.
d)Charging line piping and valves from and including the charging isolation valve to the reactor coolant
system.e)The pressurizer auxiliary spray line piping and valves to the reactor coolant system.
All components that are designed and fabricated as Class A vessels are analyzed in accordance with the
ASME Code requirements. It is demonstrated that the maximum stress intensities and cumulative usage
factors are in compliance with code values.
5.2-4 Amendment No. 16, (1/98)
5.2.3MATERIAL CONSIDERATIONS5.2.3.1Materials SpecificationSpecifications that are used for the principal pressure retaining ferritic and austenitic materials that formthe reactor coolant pressure boundary are given in Table 5.2-4. The chemical analysis of the plate material in the core region of the vessel and the as deposited weld material are listed in Table 5.2-4A.
NDTT is discussed in Section 5.2.3.5.5.2.3.2Material Exposed to CoolantMaterials used in the pressure containing boundary of the reactor coolant system or exposed to thereactor coolant are chosen to minimize corrosion and have shown satisfactory performance in other operating reactor plants. A listing of materials is given in Table 5.2-5. The table shows materials of component construction as well as internal surface material normally exposed to reactor coolant. Valve materials in contact with the reactor coolant are austenitic stainless steel.5.2.3.3InsulationPiping and components located inside the reactor containment vessel are insulated for thermal, personnelor anti-sweat requirements with a reflective-type, calcium silicate, or blanket-type material compatible with the temperature and functions involved. Metallic reflective insulation and calcium silicate blocks represent the large majority of NSSS insulation type, and are the original design. All insulation material used on stainless steel has a low soluble halide content to minimize the possibility of halide induced stress corrosion. The insulation of stainless steel pipe conforms completely to the requirements of Regulatory Guide 1.36. Prior to initial installation of insulation on stainless steel pipe, a coating of halide-free silicone paint is applied to the pipe as further protection. Subsequent installation(s) or reinstallation(s) of insulation on stainless steel pipe may be made without applying additional coating(s) of Halide free silicone paint. The SG channel heads and primary nozzles and the hot leg elbows are insulated with fiberglass blankets.The insulation on the elbows has a stainless steel jacket and the insulation on the SG channel heads and primary nozzles is protected with a stainless steel wire mesh. A fiberglass blanket insulation encapsulated in stainless steel is used on the flange stud area of thereactor vessel closure head to permit access to the head studs for removal and reinstallation of the head.
The insulation system is composed of stainless steel sheet metal, fiberglass blankets, structural stainless steel framing, latch and strike connections and quick disconnect toggle locks to form a composite system which facilitates quick removal and installation. Removable metal reflective or fiberglass blanket thermal installation is on weld areas of the reactor coolant system subject to inservice inspection. Nonremovable metal reflective type thermal insulation is on the reactor vessel.The reactor vessel head dome is primarily insulated with approximately 3" to 6" thick stainless steel metalreflective insulation. Circular openings are provided in the dome insulation to accommodate the CEDM nozzles protruding from the vessel head. The metal reflective insulation has a flat top supported approximately 3" above the highest point of the reactor vessel dome which allows sufficient clearance for inspection tooling and visual bare head inspection. Removable insulation panels are located on the outer perimeter to allow access to the space between the bottom of insulation and the dome surface.The reactor vessel below the flange is insulated with approximately 4-inch thickness of stainless steelreflective insulation with removable sections around the vessel nozzles to allow inservice inspection.5.2-11Amendment No. 21 (12/05)
The possibility of leakage of reactor coolant onto the reactor vessel head or other part of the reactorcoolant pressure boundary causing corrosion of the pressure boundary has been investigated by C-E.Metal reflective insulation is used on the pressurizer upper head, shell, and skirt. Removable blanketinsulation panels are used on the pressurizer bottom head. Flexible blanket-type insulation is used onportions of the spray lines and charging/letdown lines. The thickness of insulation is such that the exterior surface temperature is not higher than approximately 50 F above the maximum containment ambient (120 F). Exterior surface temperature of the flange stud area insulation is not higher than 80F above maximum containment ambient. All insulation support attachments were attached prior to final stress relief.5.2.3.4Coolant ChemistryControl of the reactor coolant chemistry is a function of the chemical and volume control system. Samplelines (refer to Section 9.3.2) from the reactor coolant system provide a means for taking periodic samples of the coolant for chemical analysis. All wetted surfaces in the reactor coolant system are compatible with the water chemistry. The water chemistry is to be maintained as indicated in Section 9.3.4.5.2.3.5Fracture Toughness of Ferritic MaterialsThe material toughness test requirements are as follows:
5.2.3.5.1Reactor Vessel Carbon and low-alloy steel materials which form a part of the pressure boundary meet the requirements ofthe ASME Code,Section III, Paragraph N-330 at a temperature of +40 F. It was an objective that the materials meet this requirement at +10 F. Charpy tests were performed and the results used to plot a transition curve of impact values vs. temperature extending from fully brittle to fully ductile behavior. The actual nil-ductility transition temperature of inlet and outlet nozzles, vessel, shell and head materials wasdetermined by drop weight tests per ASTM E208. NDT was established by Charpy test. Drop weight tests were conducted and are presented in Table 5.2-6. See Note 1.The maximum NDTT as obtained from the drop weight test is +20 F. The maximum temperature corresponding to the 50 ft.-lb. value of the Cv fracture energy is +37F (closure head dome plate Code No.
C-20-1). The minimum upper shelf Cv energy value for the strong direction (RW) is 104 ft.-lbs. (closure head dome plate Code No. C-20-1). The data for the weak direction was not obtained. The Charpy V-Notch results are shown in Figures 5.2-1 through 5.2-28. See Note 1.Initial RT NDT, copper and nickel values for the reactor vessel beltline plate and weld materials are alsolisted in Section 5.4.3 Tables 5.4-7 and Tables 5.4-8.Note 1: The Reactor Vessel Closure Head has been replaced and this information is for historical purposes only. Figures 5.2-24 through 5.2-28 have been marked "For historical information only". For the replacement RVCH material, see Table 5.2-6.5.2-12Amendment No. 21 (12/05)
ººº
The following are methods for detecting the resulting radiation levels.1)Blowdown line radiation - Increasing radiation levels due to dissolved andentrained fission products in the secondary side water can be detected by theradiation monitors in each of the steam generator blowdown sample lines.
Remote readout and high radiation alarms are provided.2)Off gas radiation due to gaseous and volatile fission products in the mainsteam system will be detected by the radiation monitor in the condenser off gas stream. This monitor is provided with remote readout and high radiation alarm.3)Main steam radiation monitors will detect increasing levels of radiation due togaseous and volatile fission products in the main steam system. These monitors provide remote readout and high radiation alarm. g)Reactor Vessel Head Closure LeakageThe space between the double O-ring seal is monitored by a local pressure gage (PI-1118) and pressure switch (PS-1118) to detect an increase in pressure which indicates a leak past the inner O-ring. A high-pressure alarm actuated by pressure switch PS-1118 alerts the operator to the presence of leakage past the inner seal.h)Reactor Coolant Pump Closure LeakageThis system is essentially the same as the one for the reactor vessel head closure.
The local indicators (PI-1150, 1160, 1170 and 1180) and pressure switches (PS-1150, 1160, 1170 and, 1180) provide the leak detection monitoring system with control room annunciation via the annunciator window and a Distributed Control System (DCS) driven flat panel display for the reactor coolant pump closures.i)Reactor Coolant Pump SealsInstrumentation detects abnormal seal operation. The reactor coolant pumps are equipped with three stages of seals plus a vapor backup seal as described in Section 5.5.5. During operation the reactor coolant system operating pressure is decreased through the three seals to approximately volume control tank pressure. The vapor seal prevents leakage to the containment atmosphere and allows sufficient pressure to be maintained to direct the controlled seal leakage to the volume control tank. The vapor seal is designed to withstand full reactor coolant system5.2-23Amendment No. 21 (12/05)
Changes in safety injection tank, component cooling surge tank and quench tank water levels and pressure are calculated assuming direct liquid leakage into the tanks.
5.2.4.7 Testing and Inspection Preoperational testing consists of calibrating the instruments, testing the automatic controls for activation at the proper set points and checking the operability and limits of alarm functions. Radiation
detectors can be remotely checked against a standard source during normal operation.
Normal leakage rates will be identified at the early stages of plant operation by the makeup water data.
The normal operating levels will be compared with the identified leakage and used to verify the sensitivity
of the instrumentation.
5.2.4.8 Leakage Checks During Shutdown Leakage of reactor coolant is checked during shutdowns in the following manner:
a)Prior to reactor startup following each refueling outage, pressure retaining components of the reactor coolant pressure boundary will be visually examined for evidence of reactor coolant
leakage while the system is under a test pressure of not less than the nominal system
operating pressure at rated power.
b)The visual examinations above will be conducted in conformance with the procedures of Section XI of the ASME Boiler and Pressure Vessel Code 1989 Edition subject to modification s established by 10 CFR 50.55a.
The source of any reactor coolant leakage detected by the examinations of (a) above will be located and evaluated for corrective measures as described in the ASME Code, 1989 Edition subject to modifications established by 10 CFR 50.55a.
5.2-26 Amendment No. 16, (1/98)
5.2.5 INSERVICE INSPECTIO N Provisions are made in the plant design for access to permit the conduct of preoperational and inservice inspections as specified in the ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Reactor Coolant Systems
. The design and arrangement of the components are such that space is provided to conduct examinations either from the interior or the exterior or a
combination of both.
The inservice inspection program shall be updated periodically to meet the 10 CFR 50.55a requirements.
Note: The replacement steam generators (RSGs) are designed to minimize the time required for ISI.
The RSGs comply with the ISI requirements specified by the ASME Boiler and Pressure Code,Section XI, 1986 Edition, No Addenda.
The areas selected for inspection represent those with high service factors and random areas which represent the general overall condition of the reactor coolant system. The NRC approved ISI program is
referenced in the facilities Technical Specifications.
The use of conventional nondestructive, and visual test techniques, both direct and remote, can be applied to the reactor coolant system components. The high radiation levels and remote underwater accessibility of the reactor vessel present special problems. In order to facilitate an inservice inspection of the vessel from the internal surfaces during refueling, the vessel internals and the core barrel are
removable. During refueling, the reactor vessel head, closure seal surfaces and studs may be
examined. This allows the internal parts of the vessel which are visible, including the cladding and components, to be visually examined, as well as allowing access to the vessel wall for volumetric
examinations.
The design considerations which have been incorporated into the system and plant layout to permit the required examinations are as follows:
a)Storage space is provided for the reactor vessel internals and core barrel in the refueling cavity, which will permit internal and external examinations of these components.
b)The reactor vessel head is stored dry on the containment operating floor during refueling to facilitate direct visual inspection.
c)Reactor vessel studs, nuts and washers can be removed to dry storage during refueling.
d)Limited clearance is provided around the reactor coolant piping penetrating the primary shield which permits access to at least one of the reactor vessel cold leg nozzle welds.
e)Limited access is provided to the external surface of the reactor vessel lower head through t he reactor cavity drain tunnel.
5.2-29 Amendment No. 17 (10/99)
TABLE 5.2-2ALOADING COMBINATIONS AND PRIMARY STRESS LIMITSREACTOR COOLANT PUMPSSupplementing the requirements of Section III of the ASME Boiler and Pressure Vessel Code,which specifies stress limits for all design and normal operating loading conditions, the reactor coolant pump case assembly shall be designed such that the calculated primary membrane stresses, and thecalculated primary bending stresses, , , do not exceed the following limits: for rectangular sections for annular sectionsThe design stress value, , appearing in the above expressions, shall be as follows for the loadingcombinations indicated.
LoadingPump CasingCriteria for DesignCombinationDesign Stress Valueof Supports1.Design LoadingsSection III, ASMESection III, ASME+ Design EarthquakeB & PV Code B & PV Code2.Normal Operating for ferritic steelsStresses within YieldLoadings + Maximum Earthquake , for austenitic steels 3.Normal Operating Deflections limited Loading + Pipeto maintain supportedRupture + Maximumequipment within specifiedEarthquakestress limits (See Note 4)5.2-36Amendment No. 16, (1/98)
TABLE 5.2-2A (Cont'd)
Where:=Tabulated Allowable Stress Limit at Temperature,Section III, ASME Boiler and Pressure Vessel Code.
=Tabulated Minimum Yield Strength at Temperature,Section III, ASME B &
PV Code.=Minimum Tensile Strength of Material at Temperature.
Notes: 1.The expression for the limit on for annular sections is based on a minimum "shape factor". This limit may be modified to incorporate the shape factor of the particular section
being analyzed.
2.In evaluating the effects of local loads on the vessel, as a result of loading combinations 2
and 3, the Primary Local Membrane Stress shall replace in the expressions for stress limits.
3.The stress criteria for L oading Combination No. 3 need not be applied to the piping run within which a pipe break is considered to have occurred.
4.It is not intended that the pump supports be designed to sustain pipe rupture loads applied
directly to the pumps. Suitable stops or restraints to accom m odate these loads will be furnished as part of the foundation structure.
5.2-37 Amendment No. 17 (10/99)
TABLE 5.2-3 (Cont'd)
LPSI Header Check/ V3114, V3124, V3134, V3144 Closed Open HPSI Header Check/ V3113, V3123, V3133, V3143 Closed Open SDC Return Relief/ V3469, V3482 Closed Closed (1)Valves may be open or shut during normal operation or post-accident.
5.2-39 Amendment No. 16, (1/98)
TABLE 5.2-4 (Cont'd)Applicable Code Cases For Reactor Vessel, Steam Generator, Pressurizer1332 - Requirements for Steel Forgings1335 - Requirements for Bolting Materials 1336 - Requirements for NiCrFe Alloy 1338 - L/T Examination of Plates 1359 - L/T Examination of Forgings 1361 - Socket Welds, Section IIIASME Code Case N-432, "Repair Welding Using Automatic or Machine Gas Tungsten-Arch Welding (GTAW)Temperbead Technique Section XI, Division 1".ASME Code Case N-474-1, "Design Stress Intensities and Minimum Yield Strength Values for Alloy 690(UN06690) with a Minimum Yield Strength of 35ksi, Class 1 Components - Section III, Division 1".ASME Code Case N-474, "Design Stress Intensities and Yield Strength Values for Alloy 690 with a MinimumYield Strength of 35ksi, Class I Components, - Section III, Division 1".For the replacement steam generators, the following Code Cases are used as permitted by NRC Regulatoryguides 1.85 and 1.147:N-20-3SB-163 Ni-Cr-Fe Tubing (Alloys 600 and 690) and Ni-Cr Alloy 800,N-10Time of Examination for Class 1, 2, & 3 Section III, Division 1, N-401Eddy Current Examination Per Section XI, Division 1, 2143 F-Number Grouping for Ni-Cr-Fe, Classification UNS W86152 Welding Electrode Section IX, 2142F-Number Grouping for Ni-Cr-Fe, Classification UNS N06052 Filler Metal Section IX.For the replacement of the pressurizer instrument nozzles, the following Code Cases were used as permitted byNRC Regulatory Guides 1.85 and 1.147:2142F-Number Grouping for Ni-Cr-Fe, Classification UNS N06052 Filler Metal Section IX.N-432Repair Welding Using Automatic or Machine Gas Tungsten - Arc Welding (GTAW) Temperbead Technique,Section XI, Division 1.N-474-1Design Stress Intensities and Yield Strength Values for UNS N06690 with a Minimum Specified Yield Strength of 35 ksi, Class 1 Components,Section III, Division 1.For the replacement of RCS Hot Leg instrument and sample nozzles, the following Code Cases were used aspermitted by NRC Regulatory Guides 1.85 and 1.147:2142, -1F-Number Grouping for Ni-Cr-Fe, Classification UNS N06052 Filler Metal Section IX.N-474-1, -2Design Stress Intensities and Yield Strength Values for UNS N06690 with a MinimumSpecified Yield Strength of 35 ksi, Class 1 Components,Section III, Division 1.For the replacement of the pressurizer and surge line elbow ASME Boiler on Pressure Vessel Code 1998 with2000 Addenda was used.5.2-42Amendment No. 21 (12/05)
TABLE 5.2-4A CHEMICAL ANALYSES OF PLATE MATERIAL IN REACTOR VESSEL BELTLINE Surveillance Intermediate Shell Plates Lower Shell Plates Weld (1)Code #C-7-1C-7-2C-7-3C-8-1C-8-2C-8-3 90136 Flux 3999 Element (wt %)
Si.17.20.20.17.17.19.20 S.013.010.012.010.010.010
.012 P.004.004.004.006.006.004
.013 Mn1.281.281.331.281.291.22 1.02 C.24.23.21.28.22.22.12 Cr.03.03.06.07.07.06.06 Ni.64.64.58.56.57.58.07 (2)Mo.60.59.58.65.66.59.55 V.003.003.003.002.002.002
.006 Nb.01.01.01.01.01.01.01 B.0003.0002.0002.0003.0003.0001
.0001 Co.005.006.00 5.007.007.006
.004 Cu.11.11.11.15.15.12.2 7 (2)Al.018.020.020.027.025.022
.001 W.01.01.01..01.01.01.01 Ti.01.01.01.01.01.01.01 As.015.015.010.013.014.011
.014 Sn.006.006.006.009.010.006
.005 Zr.002.002.002.002.002.002
.001 N 2.006.007.007.008.009.006
.008 Fe (balance)(1)Identical to intermediate to lower shell course girth weld seam 9-203.
(2)This weld chemical analysis is a single actual result from the surveillance weld except that copper and nickel values are bas ed on the "best estimate" values for a specific weld wire heat using all industry available data.
The "best estimate" valures were determined in response to NRC GL 92-01, Rev. 1, Supplement 1 and were submitted to the NRC in FPL Letter L-97-2 23 (Referenc e 14 in Section 5.4)
.NOTE: Letter L-77-308 is the source document for all data.
5.2-43 Amendment No. 18, (04/01)
TABLE 5.2-5 MATERIALS EXPOSED TO COOLANT
- Reactor Vessel Cladding Weld Deposited Type 308 SS Vessel Internals 304 SS and NiCrFe Alloy, SB-168 Fuel Cladding Zirconium alloy Pipe Instrument Nozzle SA-516 Grade 70 Base Replacement with Split Nozzles Pipe Instrument Nozzles SB-166 UNS N06690 and UNS N0 66 0 0 Pipe Cladding Austenitic Stainless Steel Type 304 L Replacement Steam Generators Bottom Head Cladding Austenitic Stainless Steel, ER308L, ER309L Tube Sheet Cladding NiCr Fe Alloy, ERNiCr-3 Tubes NiCrFe Alloy, SB-163 (UNS N06690)
Divider Plate SA 240 Type 304L Seat Bars SB-166 Alloy 600; SB-168 Alloy 690 Nozzle Dam Rings SB-166 Alloy 690 Pumps Casing Austenitic Stainless Steel, Grade CF8M
Internals Austenitic Stainless Steel, Type 316 and Type 304 Pressurizer Cladding Weld Deposited Stainless Steel Type 308L, 309L
- Replacement RVCH RVCH Cladding Austenitic Stainless Steel, ER308L, ER309L CEDM Nozzles SB-167 UNS N06690 Instrument Nozzles SB-167 UNS N06690 Instrument Nozzle Adaptor SA-479 Type 304 CEDM Nozzle Adaptor SB-166 UNS N06690 Vent Nozzle SB-167 UNS N06690 Vent Pipe SA-312 Type 316 CEDM Pressure Housing ASME SA-479 TP316, SA-213 TP316, SA-182 F348, SB-166 Alloy UNS N06690, ASTM 276 Type 403 Condition T, Code Case N-2 Mechanism Austenitic Stainless Steel Types 304, 316, 348, and CF8M Martensitic Stainless Steel Types 403, 410, 440C, and 17-4 PH NiCrFe Alloy 690 NiCrFe Alloy X-750 Cobalt Alloys Stellite 36 and Stellite 6B Weld Metal ERNiCrFe-7A, ER316L, and IN316L
UNIT 1 5.2-44 Amendment No. 27 (04/15)
TABLE 5.2-5 (Cont'd)
MATERIALS EXPOSED TO COOLANT RVLMS Pressure Housing SA-479 TP316, SB-166 TP UNS N06990, SA-312 TP 316, SA-213 TP316
5.2-44a Amendment No. 26 (11/13)
Table 5.2-7Replacement Pressurizer Ferritic Material Fracture Toughness Reference - Pressure BoundaryPart NameHeat NumberLocationMaterialRTndt or Lowest ServiceTemperature Ref.
NotesUpper HeadT 1384Upper HeadSA-508 Gr 3 CI 2minus 18 degrees F.
1Upper ShellT 1386Upper ShellSA-508 Gr 3 CI 2minus 10 degrees F.Lower ShellT 1385Lower ShellSA-508 Gr 3 CI 2minus 10 degrees F.Lower HeadT 1383Lower HeadSA-508 Gr 3 CI 2minus 36 degrees F.
2Manway Cover11727.1Upper HeadSA-533 Tp B CI 2minus 9 degrees F.Manway BoltsN 9879 / U3BUpper HeadSA-540 B24 CI 3Test Temperature 40 degrees F.Cv ft-lbs 58.3, 58.3, 59MLE (mils) 28.5, 35.8, 32.8 3Manway NutsF87688Upper HeadSA-193 B7Test Temperature 40 degrees F.Cv ft-lbs 79, 81, 76, MLE (mils) 43.2, 52.4, 48.4 Test Temperature 60 degrees F.
Cv ft-lbs 81, 80, 82 MLE (mils) 45.7, 47.7, 52.7 4Replacement Pressurizer Ferritic Material Fracture Toughness Reference - Non-Pressure BoundaryPart NameHeat NumberLocationMaterialRTndt or Lowest ServiceTemperature Ref.
NotesSkirt Cylinder4-1575Lower HeadSA-508 Gr 3 CI 2Test temperature zero degrees F.
C V Ft-lbs 66, 76, 92MLE (mils) 47, 54, 56 5Skirt Flange4-1575Lower Skirt ShellSA-508 Gr 3 CI 2Test temperature zero degrees F.
C V Ft-lbs 105, 89, 107MLE (mils) 71, 63, 70 5Luft Lug10887.2Upper HeadSA-533 Tp B Cl 2minus 45 degrees F.
6Notes:
1)Ferritic portion of spray nozzle, safety nozzle, relief nozzle and manway boss integral with upper head.
2)Ferritic portion of the surge nozzle integral with lower head.
3)Lowest service temperature 40 degrees F.
4)Lowest service temperature 40 degrees F. Under 4 inch only MLE greater than 25 mils, Cv greater than45 required.5)NF component lowest service temperature is zero degrees F.
6)ANSI 14.6 lowest lift temperature equals minus 5 degrees.5.2-46Amendment No. 21 (12/05)
TABLE 5.2-10 (Cont'd)
Charpy V-Notch Values @ 10F (ft.-lbs.)
Piece No.Code No.Location Material 0 180 502-12-1 C-4304-1 PIPE SEGMENT SA-516 GRADE 70 50-48-50 (49.3) 502-12-2 C-4304-2 PIPE SEGMENT
" 52-51-55 (52.7) 502-08-l C-4308-1 ELL SEGMENT
" 40-30-28 (32.7) 502-08-2 C-4308-1 ELL SEGMENT
" 40-30-28 (32.7) 507-02-2 C-4314-2 NOZZLE FORGING SA-105 GRADE 2 49-50-31 (43.3) 507-07-2 C-4315-2 NOZZLE FORGING
" 49-50-31 (43.3) 508-08-2 C-4316-2 NOZZLE FORGING SA-182 GRADE F1 92-71-91 (84.7) 508-08-2 C-4316-2 NOZZLE FORGING SA-182 GRADE F1 91-66-84 (80.3) 502-12-3 C-4304-3 PIPE SEGMENT SA-516 GRADE 70 51-51-55 (52.3) 502-12-4 C-4304-4 PIPE SEGMENT
" 42-58-50 (50) 502-08-3 C-4308-1 ELL SEGMENT
" 40-30-28 (32.7) 502-08-4 C-4308-1 ELL SEGMENT
" 40-30-28 (32.7) 507-02-1 C-4314-1 NOZZLE FORGING SA-105 GRADE 2 49-50-31 (43.3) 508-02-4 C-4316-4 NOZZLE FORGING SA-182 GRADE F1 116-104-85 (101.7) 502-14-1 C-4304-5 PIPE SEGMENT SA-516 GRADE 70 60-54-48 (54) 501-14-2 C-4304-6 PIPE SEGMENT
" 36-36-36 (36) 502-10-1 C-4309-1 ELL SEGMENT "39-44-48 (43.7) 502-10-2 C-4309-1 ELL SEGMENT
" 39-44-48 (43.7) 507-10-1 C-4313-1 NOZZLE FORGING SA-105 GRADE 2 49-50-31 (43.3) 502-14-1 C-4304-5 PIPE SEGMENT SA-516 GRADE 70 60-54-48 (54) 502-14-2 C-4304-6 PIPE SEGMENT
" 36-36-36 (36) 502-14-2 C-4304-6 PIPE SEGMENT
" 34-37-60 (43.7) 502-10-1 C-4309-1 ELL SEGMENT "34-37-60 (43.7) 502-10-2 C-4309-1 ELL SEGMENT
" 34-37-60 (43.7) 507-10-2 C-4313-2 NOZZLE FORGING SA-105 GRADE 2 49-50-31 (43-3) 5.2-50 TABLE 5.2-10 (Cont'd)
PIECE NO.CODE NO.DESCRIPTION MATERIAL CHARPY V-NOTCH VALUES @ 10F (FT.-LBS.), 0 502-06-1 C-4307 ELL SEGMENT SA-516 GRADE 70 29-40-30 (33) 502-06-2 C-4307 ELL SEGMENT
" 29-40-30 (33) 502-16-1 C-4303-1 PIPE SEGMENT
" 53-42-46 (47) 502-16-2 C-4303-2 PIPE SEGMENT
" 53-42-46 (47) 502-06-3 C-4307-1 ELL SEGMENT SA-516 GRADE 70 29-40-30 (33) 502-06-4 C-4307-1 ELL SEGMENT
" 29-40-30 (33) 502-16-3 C-4303-1 PIPE SEGMENT
" 53-42-46 (47) 502-16-4 C-4303-2 PIPE SEGMENT
" 53-42-46 (47) 502-06-5 C-4307-1 ELL SEGMENT SA-516 GRADE 70 29-40-30 (33) 502-06-6 C-4307-1 ELL SEGMENT
" 29-40-30 (33) 502-16-5 C-4303-1 PIPE SEGMENT
" 53-42-46 (47) 502-16-6 C-4303-2 PIPE SEGMENT
" 53-42-46 (47) 502-06-7 C-4307-1 ELL SEGMENT SA-516 GRADE 70 29-40-30 (33) 502-06-8 C-4307-1 ELL SEGMENT
" 29-40-30 (33) 502-16-7 C-4303-1 PIPE SEGMENT
" 53-42-46 (47) 502-16-8 C-4303-2 PIPE SEGMENT
" 53-42-46 (47) 5.2-51 TABLE 5.2-10 (Cont'd)
PIECE NO.CODE NO.DESCRIPTION MATERIAL CHARPY V-NOTCH VALUES @ 10F (FT.-LBS.), 0 502-14-3 C-4304-7 PIPE SEGMENT SA-516 GRADE 70 60-54-48 (54) 502-14-4 C-4304-8 PIPE SEGMENT
" 41-56-54 (50.3) 502-10-5 C-4309-1 ELL SEGMENT "34-37-60 (43.7) 502-10-6 C-4309-1 ELL SEGMENT "34-37-60 (43.7) 507-10-3 C-4313-3 NOZZLE FORGING SA-105 GRADE 2 49-50-31 (43.3) 502-14-3 C-4304-7 PIPE SEGMENT SA-516 GRADE 70 60-54-48 (54) 502-14-4 C-4304-8 PIPE SEGMENT "41-56-54 (50.3) 502-10-7 C-4309-2 ELL SEGMENT "37-42-34 (37.7) 502-10-8 C-4309-2 ELL SEGMENT "37-42-34 (37.7) 507-10-4 C-4313-4 NOZZLE FORGING SA-105 GRADE 2 49-50-31 (43.3) 502-04-7 C-4306-2 ELL SEGMENT SA-516 GRADE 70 43-32-31 (35.3) 502-04-8 C-4306-2 ELL SEGMENT "43-32-31 (35.3) 502-04-1 C-4306-1 ELL SEGMENT SA-516 GRADE 70 33-29-34 (32) 502-04-2 C-4306-1 ELL SEGMENT "33-29-34 (32) 502-04-3 C-4306-1 ELL SEGMENT SA-516 GRADE 70 33-29-34 (32) 502-04-4 C-4306-1 ELL SEGMENT "33-29-34 (32) 502-04-5 C-4306-2 ELL SEGMENT SA-516 GRADE 70 43-32-31 (35.3) 502-04-6 C-4306-2 ELL SEGMENT "43-32-31 (35.3) 5.2.52
<
- 140 130 o Test Value 120 D Average 110 100 V') 90 .c I = 80 . e;; '-70 w -60 c.. E -50 40 30 20 10 0 -100 -50 FLORIDA POWER & LIGHT CO. ST. LUCIE PLANT UNIT 1 0 0 0 0 50 Test Temp, °F Charpy_ V-Notch Impact Curve -item Code C-1-1 Vessel Flange 100 150 Figure 5.2-1
- 140 130 o Test Value 120 o Average 110 100 V') 90 ..c I 80 .. """ 70 LU -ii 60 E -50 40 30 20 10 0 -100 -50 FLORIDA POWER & LIGHT CO. ST. LUCIE PLANT UNIT 1 0 0 0 0 50 Test Temp, °F CharpY. V-Notch Impact Curve Item Code C-1-1 Vessel Flange 100 150 Figure 5. 2-la
- 150 140 130 o Test Value 120 o Average 110 100 VI 90 ...0 I ..... -.. 80 .... 70 LI.I t:) fg_ 60 E -50 40 30 20 10 0 -100 FLORIDA POWER.& LIGHT CO. St. Lucie Plant -40 0 40 Test Temp, OF Charpy V-Notch Imppct Curve Item Code C-6-l Upper Shell Plate 100 160 Figure 5.2-2
- 150 140 130 o Test Value 120 o Average 110 100 0 Jg 90 I --0 .. 80 >. L. CV i5 70
- t::> 60 E -50 0 40 30 20 0 10 0 0 -100 -40 0 40 Test Temp, °F 100 160 FLORIDA CharpY. V-Notch Impact Curve POWER & LIGHT co. Item Code C-6-2 st. Lucie Plant Upper Shell Plate 5.2-3 150 140 130 o Test Value 120 o Average llO 100 a 90 I --80 !.-Q,) ..5 70 u ro E" 60 .-i 50 40 30 20 10 0 -100 FLORIDA POWER & LIGHT CO. St. Lucie Plant--40 0 40 Test Temp, °F Charp'[ V-Notch Imoact Curve 1tem Code C-6-3
- Upper Shell Plate 0 100 160 Figure 5. 2-4
- w .2 w ro > ro ..... "--V> CL> Cl.> > ...... <C 0 0 0 0 8 0 ("'t"\ N ..... ..... ..... .,.... ..... FLORIDA POWER & LIGHT CO. St. Lucie Plant. 0 0 0 0 0 Lt'\ 1111:::7'
("'t"\ 00 ....... 5Ql*U 'A0Jau3 p2dw1 Charpy V-Notch Curve Item Code C-7-1 Intermediate Shell Plate 0 0 0 N ..... 8 ..... u.. 0 0.. E C> Q.) ...... C> 0 I I -VI CL> I-8 o'I' Figure 5.2-5
- 150 140 130 o Test Value 120 o Average 110 0 100 :a 90 0 I --.. 80 lo.. Q,)
- 70 0 t) 60 E 0 -50 40 0 0 30 20 0 10 0 0 -100 40 0 40 Test Temp, °F 100 160
- FLORIDA Charpy V-Notch Impact Curve Figure POWER & LIGHT CO. Item Code C-7-2 St. Lucie Plant Intermediate shell Plate 5.2-6
- *
..... 0 Q) :::> I'll Q) 0 01 > I'll I ..... L. II'> CIJ Q) > ..... <( 0 0 8 ..... 0 0 0 8 0 55 0 0 0 0 0 o' M N ..... ....... Lt\ rt'\ N ..... ...... ...... ..... ..... SQl-U 'A6Jau3 i=>edw1 Figure FLORIDA POV/ER & LIGHT CO. St. Lucie Plant Charpy V-Notch Impact Curve Item Code C-8-1 Lower Shell Plate 5.2-8
- 150 140 130 o Test Value 120 o Average 110 100 90 I 0 .. 80 >, C'> .... Q,) 70
- i:; 60 E 0 ...... 50 40 30 20 10 0 -100 -40 .. 0 40 100 Test Temp, °F
- Figure FLORIDA Charpy V-Notch Curve POWER & LIGHT CO. Item Code C-8-2 St: Lucie Plant Lower Shell Plate 5.2-9
- Q,) :J Q,) ro O') ro > r.... ...... Q,) V'I > Q,) <C ....... 0 D 0 0 8 ("I"\ ..... ..... ..... ..... ..... * ' FLORIDA POWER & LIGHT CO. St. Lucie Plant c Q 0 0 0 0 00 ,..... '° U"\ ('f'\ sq1-u 'A6Jau3 pedw1 Charpy V-Notch Impact Curve Item Code C-8-3 Lower Shell Plate 0 0 I 8 -I 0 0 0 N ..... Figure 5.2-10 150 140 130 o Test Value 120 D Average 110 100 90 V'l .0 -I 80 --. >. e> 70 Cl.) c: LU u 60 re c.. E 50 -40 30 20 10 0 0-100 FLORIDA POWER & LIGHT CO. St. Lucie *Plan-f 0 -40 0 40 Test Temp, °F CharpY. V-Notch Imoact Curve Item Code C-9-1 Bottom Head 1 100 Figure 5.2-11
- 150 140 130 o Test Value 120 o Average 0 llO 100 0 90 V"l .c 80 I --. 70 L.. Q)
- c: L.i.J 60 u E 50 1-4 I 40 I I 30 I I 20 l I 10 I I 0 I 0 -100 -40 0 40 100 160 Test Temp, °F
- ____________
______ .......,-;.-Fig-u-rel POWER & LIGHT co. !tern Code C-9-2 5. 2._12 st. Luci* Plant Bottom Head
- 150 140 130 0 Test Value 120 0 Average 110 100 90 V'l .0 I 80 --*;:; 70 !.... Q.)
- c: l..LJ n 60 ro 0.. E 50 0 -40 0 30 20 0 10 0 -40 0 40 100 -100 Test Temp, °F CharpY. V-Notch Impact Curve Figure FLORIDA POWER & LIGHT CO. St. Lucie Plant Item Code C-10-1 Bottom Head 5.2-14
- 150 140 o Average 130 o Test Value 120 110 ' 100 I c,I) I == el .!. 90 -I 80 I I Q) I I i5 70 j I I u I ::g_ 60 E I e-t I 50 I 40 30 20 10 0-140 -100 -40 0 60 100 150 Test Temp, °F
- Charpy V-Notch Impact Curve Figure FLORIDA POWER & LIGHT CO. Item Code C-4-1 St. Lucie Plant In let Nozzle 5.2-15
- 150 140 0 130 0 Test Value 120 o Average 110 100 90 V') .0 I 80 :::::: .
- 70 Q) c::: 1.1.J u 60 "' 0... E -50 40 30 20 10 0-100 -40 0 40 100 160 Test Temp, °F
- CharpY-V-Notch Impact Curve Figure FLORIDA POWER.& LIGHT CO. Item Code C-4-2 St. Luc:.ie Plant Inlet Nozzle 5.2-16 00 Q> ::s re > -Vl Q) I-0 fa 0 8 ...... ..... ..... ...... FLORIDA POWER & LIGHT CO. St. Lucie Plant 0 Q,) re '-< 0 0 0 0 0 s 0 Si 0 fa ...... "" ('t'\ SQl-U 'A6Jau3 pedw1 Charpy V-Notch Impact Curve Item Code C-4-3 In let Nozzle 0 0 '111:3" 1..1.. 0 c: E 0 Q,) I-..... Vl Q.) I-0 I 0 0 8 ...... I 0 ...... I 0 0 ...... Figure 5. 2-17 150 140 130 0 120 0 110 100 90 VI 80 --. E;:l 70 L... Q,) c: LJ.J u 60 ro c.. 50 40 30 20 10 0 0 -120 FLORIDA POWER & LIGHT CO. St. Lucie Plant Test Value Average 0 0 0 -40 0 40 Test Temp, °F Charpy V-Notch Impact Curve* Item Code C-4-4 Inlet Nozzle 0 0 100 140 5.2-18 150 140 130 OTest Value 120 DAverage 110 100 90 V'l :e 80 I ::::::: C'\ i.... Ql 5 60 u ('ti E"so 1-1 40 30 20 10 0 -100 FLORIDA POWER & LIGHT CO. St. Lucie Plant -40 0 0 0 0 0 40 Test Temp, °F Charpy V-Notch Impact Curve Item Code C-16-1, 2. 3. 4 Inlet Nozzle Extension 0 0 Figure 5.2-19
- 150 140 130 o Test Value 0 120 D Average 110 10 0-100 -40 0 40 100 160 Test Temp, °F "
Charpy V-Notch lm_pact Curve Figure FLORIDA POWER & LIGHT CO. St. Lucie Plant Item Code C-17-1, 2 Outlet Nozzle Extension
- 5. 2-20
- * """"' V') Q) ...... 0 0 8 .-I ..... .-1 .-1 FLORIDA POWER & LIGHT CO. St. Lucie Plant Q) en f'c:I ._ <t D 0 $ 0 0 0 ...... U"\ rt'\ SQl-U 1 A6Jau3 pedWI Charpy V-Notch Imoact Curve Item Code C-33-1, 2 3 Vessel Support Forg(ng 0 ...... 8 -u.. 0 0 I -V') 8 ........ I 0 Figure . 2-21
- D 0 0 0 SQl-U ',\6J9U3 pedw1 0 u. 0 FlOIUDA Charpy V-Notch Curve Figure POWER & LIGHT co. Item Code C-3-1 st. Luc:s.e :nanc Out let Nozzle 5.2-22
- e! I
- 0 0 0 FLORIDA POWER & LIGHT CO. St. Lucie Plant 0 0 sq1-u 'A0Jau3 pedwr Charpy V-Notch Impact Curve Item Code C-3-2 Outlet Nozzle 8 --Vl 0 Q.> '° I-0 Figure 5.2-23 Florida Power & Light CompanySt. Lucie Plant Unit 1Charpy V-Notch Impact CurveItem Code C-2Closure Head Flange Fi gure 5.2-24Amendment No. 21 (12/05)
Florida Power & Light CompanySt. Lucie Plant Unit 1Charpy V-Notch Impact CurveItem Code C-21-1Closure Head Torus Fi gure 5.2-25Amendment No. 21 (12/05)
Florida Power & Light CompanySt. Lucie Plant Unit 1Charpy V-Notch Impact CurveItem Code C-21-2Closure Head Torus Fi gure 5.2-26Amendment No. 21 (12/05)
Florida Power & Light CompanySt. Lucie Plant Unit 1Charpy V-Notch Impact CurveItem Code C-21-3Closure Head Torus Fi gure 5.2-27Amendment No. 21 (12/05)
Florida Power & Light CompanySt. Lucie Plant Unit 1Charpy V-Notch Impact CurveItem Code C-20-1Closure Head Torus Fi gure 5.2-28Amendment No. 21 (12/05)
³
- 920 910. 900 890 .. 880 *;; ll.
- tF 1170 .. :I Ill Ill f! Cl. 860. 850. 1140 830 820 0
- ST. LUCIE 1 BEST ESTIMATE SG PRESSURE VERSUS POWER LEVEL Original Design, 2570 MWI -8485 Tubes/SG (Actual Tubes= 6519) , Projected Cycle 15 } New BWI SGa * --_,..... -.. ,..... -_,.....
*---__ ,__ ,_-----"\ " 10 -,._ ------_,__ -__ ,.... ;::__ s1re1ch Power curve, AOB Cc Lener f*Cc-751S1, 101101111
- -.. .... .. -.. -. .. ...... Projected Cycle 14 Condlllona, 2713.84 MWI, Average Tubes a 6346/SG___}f" .... " ..... .... 20 30 40 50 60 70 60 90 100 Amendment No. 16, (1/98) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT OU:CT 1 SECONDARY PRESSURE PROGRAM FIGURE 5.3-2 5.4REACTOR VESSEL AND APPURTENANCES5.4.1 REACTOR VESSEL DESCRIPTIONThe reactor vessel (Figure 5.4-1) integral supports consist of three pads welded to the underside of oneoutlet and two inlet reactor vessel nozzles, in turn supported by graphite lubricate bearing plates. The arrangement of the vessel supports, allows radial growth of the reactor vessel due to thermal expansion while maintaining it centered and restrained from movement caused by seismic disturbances. Departure from levelness of not more than 0.002 inch per foot of flange diameter is maintained during construction to facilitate proper assembly of reactor internals. The design parameters for the reactor vessel are given in Table 5.4-1. The vessel bottom heads, cylindrical shell courses and head lifting lugs are made ofSA 533-65, Grade B, Class I material. The vessel top head is made from a one piece forging of SA-508Class 3 material.The vessel closure flange is a forged ring with a machined ledge on the inside surface to support thereactor internals. No ring forgings are used for reactor vessel shell sections. The flange is drilled and tapped to receive the closure studs and is machined to provide a mating surface for the reactor vessel closure seal. The vessel closure contains 54 studs, 7 inches in diameter, with eight threads per inch. The stud material is Parkerized ASTM A-540, Grade B24, with a specified minimum yield strength of 130,000 psi. Test results on the material used for these studs demonstrates a minimum yield strength not less than 141,000 psi at 78F. The tensile stress in each stud when elongated for operational conditions is approximately 40 ksi.Six radial nozzles on a common plane are located just below the vessel closure flange. Extra thickness inthis vessel-nozzle course provides the reinforcement required for the nozzles. Additional reinforcement is provided for the individual nozzle attachments. A boss located around each outlet nozzle on the inside diameter of the vessel wall provides a mating surface for the internal structure which guides the outlet coolant flow. This boss and the outlet sleeve on the core support barrel are machined to a common contour to minimize core bypass leakage. A hemispherical head forms the lower end of the vessel shell.
There are no penetrations in the lower head.The removable top closure head is hemispherical. The head flange is drilled to match the vessel flangestud bolt locations. The 54 stud bolts are fitted with spherical crowned washers located between the closure nuts and head flange, to maintain stud alignment during head flexing due to boltup. To ensure uniform loading of the closure seal the studs are hydraulically bolt-tensioned.A detailed analysis has been performed which shows that 36 evenly distributed studs can fail before theremaining stud stresses reach yield and the closure separates at design pressure, and that 16 adjacent studs can fail before the closure will fail by "zippering".Flange sealing is accomplished by a double-seal arrangement utilizing two silver jacketed Ni-Cr-Fe alloy,spring-energized O-ring seals. The space between the two rings is monitored (see Section 5.2.4.5) to allow detection of any inner ring leakage. The control element drive mechanism nozzles terminate with threaded and canopy seal-welded ends at the top. There are eight instrumentation nozzles with mechanical seal connections. In addition to these nozzles there is a 3/4 inch vent connection utilized bythe Reactor Coolant Gas Vent System (RCGVS). See Section 5.7.5.4-1Amendment No. 21 (12/05)
)
The differences between the plant surveillance program and the requirements presented in Appendix H are the following:
a)Appendix H,Section II A - Sample Materials
The weldments for the weld metal samples and heat-affected-zone (HAZ) samples are prepared in a reactor vessel girth seam weld rather than being an extension of a longitudinal seam weld.
Following the same procedures as used for the reactor vessel welds and the use of the same
lots/heats of filler wire, rod and flux, this sample weldment is representative of the girth seam weld
which passes through the areas of highest radiation flux in the core region of the reactor vessel.
b)Appendix H,Section II B - Attachments to Reactor Vessel In adhering to the requirement of placing the surveillance specimens as close as possible to the reactor vessel wall, the capsule holders are attached to the cladding of the reactor vessel and are
not major load bearing components. By such placement, temperature, flux spectrum, and fluence
differences between the surveillance specimens and the reactor vessel are minimized, thereby
permitting more accurate assessment of the changes in the reactor vessel properties.
c)Appendix H,Section II B - Capsule Replacements The surveillance capsule holders will permit installation of replacement capsule assemblies during those shutdown periods when the reactor internals are removed.
d)Appendix H,Section II A - Specimen Orientation Charpy impact specimens of base metal oriented in the strong (RW) direction are included in each surveillance capsule assembly. Similar specimens oriented in the weak (WR) direction are included
in 4 of the 6 capsule assemblies. The unirradiated baseline properties will be thoroughly
established for both the strong (RW) and weak (WR) directions permitting RW/WR correlations to
be established.
Sufficient quantities of WR oriented base metal specimens are included in the program to adequately verify the unirradiated RW/WR correlations after irradiation. In this way the weak (WR)
direction properties of the irradiated plate materials in the reactor vessel can be established.
Three metallurgically different materials representative of the reactor vessel are investigated. These are base metal, weld metal, and weld heat-affected-zone (HAZ) material. In addition to the materials from the reactor
vessel, materials from a standard heat of SA533B, made available through the Heavy Section Steel
Technology (HSST) Program, are also used. This reference material is fully processed and heat treated and
is used for Charpy impact specimens so that a comparison may be made between the irradiations in various
operating power reactors and in experimental reactors. A complete record of the chemical analysis, fabrication history and mechanical properties of all surveillance test materials is maintained.
5.4-4 Amendment No. 18, (04/010)
The results of detailed chemical analyses on the plates which comprise the beltline region of the vessel and on a sample weld with identical materials as the girth weld passing through the maximum flux region are
presented in Table 5.2-4.
The exposure locations and a summary of the specimens at each location is presented in Table 5.4-2. The pre-irradiation NDT temperature of each plate in the intermediate and lower vessel shell courses is determined from the drop weight tests and correlated with Charpy impact tests.
Base metal test specimens are fabricated from sections of the lower shell plate material since it exhibits the highest unirradiated NDT temperature. All base mate rial test specimens are cut from the same shell plate. This material is heat treated to a condition which is representative of the final heat treated condition of the base metal in the completed reactor vessel.
Weld metal and HAZ material are produced by welding together two plate sections from the lo wer shell course of the reactor vessel. All HAZ test materials are also fabricated from the plate which exhibits the highest unirradiated NDT temperature.
The material used for weld metal and HAZ test specimens is adjacent to the test material used for ASME Code,Section III tests and is at least one plate thickness from any water-quenched edge. The procedures
used for making the shell girth welds in the reactor vessel welds are followed for inspection of the welds in
the test materials. The welded plates are heat treated to a condition representative of the final heat treated
condition of the completed reactor vessel.
The test specimens are contained in six irradiation capsule assemblies. The axial position of the capsules is bisected by the midplane of the core. The circumferential locations include the peak flux regions.
The location of the in-vessel surveillance capsule assemblies is shown in Figure 5.4-2. A typical surveillance capsule assembly is shown in Figure 5.4-3. A typical Charpy impact compartment assembly is
shown in Figure 5.4-4. A typical tensile monitor compartment assembly is shown in Figure 5.4-5.
Fission threshold detectors (U-238) are inserted into each surveillance capsule to measure the fast neutron flux. Threshold detectors of Ni, Ti, Fe, S, and Cu with known Cobalt content have been selected for this
application to monitor the fast neutron exposure. Cobalt is included to monitor the thermal neutron
exposure.5.4-5 Amendment No. 16, (1
/98)
The selection of threshold detectors is based on the recommendations of ASTM E-261, "Method for Measuring Neutron Flux by Radioactive Techniques". Activation of the specimen material will also be
analyzed to determine the amount of exposure.
The maximum temperature of the encapsulated specimens will be monitored by including in the surveillance capsules small pieces of low-melting point eutectic alloys individually sealed in quartz tubes.
The temperature monitors provide an indication of the highest temperatures to which the surveillance specimens are exposed but not the temperature history or the variance between the temperature history of
different specimens. These factors, however, affect the accuracy of the estimated vessel material NDTT to
only a small extent.
Tests specimens removed from the surveillance capsules will be tested in accordance with ASTM Standard Test Methods for Tension and Impact Testing. The data obtained from testing the irradiated specimens will
be compared with the unirradiated data and an assessment of the neutron embrittlement of the reactor vessel material will then be made. This assessment of the NDTT shift is based on the temperature shift in the average Charpy curves, the average curves being considered representative of the material. This shift, when measured at the 30 ft.-lb. Level between initial and irradiated Charpy curves, is referred to as "the shift" or delta RT N DT.The periodic analysis of the surveillance samples permit the monitoring of the neutron radia tion effects upon the vessel materials. The Charpy curve RT NDT shift results from the irradiated surveillance samples are compared to predicted RT NDT shifts for each material at the same fluence to determine if the values are consistent with pertinent ra diation effects data studies accumulated under similar conditions.
If, with due consideration for uncertainties in RT NDT determination, the surveillance samples exhibit a higher than anticipated RT NDT shift, then appropriate limitations would be imposed on permissible operating pressure-temperature combinations and transients to insure that existing reactor vessel stresses are low
enough to preclude brittle fracture failure.
5.4-6 Amendment No. 16, (1/98)
5.4.4.2 Post Thermal Shield Removal Surveillance Following the 1983 removal of the thermal shield, one in-vessel dosimetry capsule, one ex-vessel dosimetry capsule and twelve ex-vessel wire dosimeters were installed. The in vessel capsule was installed in the W-
97 position. The ex-vessel capsule was placed at the same azimuthal and axial position as the W-97 capsule but in the annular space between the vessel and primary shield wall. The ex-vessel wire dosimeters
were placed in the annular space in positions to give a detailed azimuthal flux profile in a 45 octant as well
as to evaluate neutron flux symmetry around the vessel.
The in-vessel replacement dosimetry capsule has the same external configuration as the existing St. Lucie Unit 1 surveillance capsules with minor design modifications to facilitate remote installation. The capsule compartments contain three sets of nine neutron dosimeters and one set of four temperature monitors (See
Figure 5.4-3b). No mechanical test specimens are included. The replacement dosimetry capsule data will
complement the irradiated W-97 surveillance capsule evaluation since neutron flux measurements at the
same azimuthal position will be available both before and after removal of the thermal shield.
The ex-vessel capsule contains flux monitor sets identical to the replacement in-vessel capsule.
There are six ex-vessel iron wire dosimeters and six nickel wire dosimeters located per Table 5.4-6.
These capsules are stamped at each end for identification with the designations of FE 1, 2, 3, 4, 5, and 6, and NI 1, 2, 3, 4, 5 and 6.
5.4.5 NONDESTRUCTIVE TESTS During fabrication of the reactor vessel, nondestructive tests based upon Section III of the ASME Boiler and Pressure Vessel Code were performed on all welds, forgings and plates as follows:
5.4-8 Amendment No. 16, (1/98
)
The detection of flaws in irregular geometries was facilitated because most nondestructive testing of thematerials was completed while the material was in its simplest form. Nondestructive inspection during fabrication was scheduled so that full penetration welds were capable of being radiographed to the extent required by Section III of the ASME Boiler and Pressure Vessel code.Each of the vessel studs received one ultrasonic test and one magnetic particle inspection during themanufacturing process.The ultrasonic test was a radial longitudinal beam inspection. Rejection was based on any discontinuitywhich caused an indication which exceeded 20 percent of the height of the adjusted first back reflection.
Any discontinuity which prevents the production of a first back reflection of 50 percent of the screen height was also cause for rejection.The magnetic particle inspection was performed on the finished studs. Linear axially aligned defectswhose lengths are greater than 1 inch long and linear nonaxial defects were unacceptable.The vessel studs are stressed as they are elongated by the stud tensioners during the initial installation ofthe vessel head and at each refueling. The amount of elongation versus hydraulic pressure on the tensioner is compared with previous readings to detect any significant changes in the elongation properties of the studs.During fabrication of the other components of the reactor coolant system, such as the steam generator,pressurizer and piping, nondestructive testing based upon the requirements of Section III of the ASME Boiler and Pressure Vessel Code is used to determine the acceptance criteria for various size flaws. The requirements for the Class A vessel are the same as the reactor vessel. Vessels designated as Class C were fabricated to the standards of Subsection C, Article 21 of Section III of the ASME Code.During the manufacture of the reactor vessel, in addition to the areas covered by the ASME Boiler andPressure Vessel Code,Section III, quality control included:a)preparation of detailed purchase specifications which included cooling rates for testsamplesb)requiring vacuum degassing for all ferritic plates and forgingsc)specification of fabrication instructions for plates and forgings to provide control ofmaterial prior to receipt and during fabricationd)use of written instructions and manufacturing procedures which enabled continual reviewbased on past and current manufacturing experiences5.4-9Am. 2-7/84 e)performance of chemical analysis of welding electrodes, welding wire, and materials forautomatic welding, thereby providing continuous control over welding materialsf)the determination of NDT temperature through use of drop weight testing methods as wellas Charpy impact testsg)test programs on fabrication of plates up to 15 inches thick to provide information aboutmaterial properties as thickness increasesh)documentation of tests and inspections and formal retention of records for futuresurveillance comparisonsi)and longitudinal wave ultrasonic testing was performed on 100 percent of all platematerial.Cladding for the reactor vessel is a continuous integral surface of corrosion resistant material, 5/16 inchnominal thickness. The procedure used specified the type of weld rod, welding position, speed of welding, nondestructive testing requirements, and was in compliance with the ASME Boiler and Pressure Vessel Code. The cladding is ultrasonically inspected for lack of bond at intervals not to exceed 12 inches transverse to the direction of welding. Unbonded areas equal to, or in excess of calibration require additional scanning of the surrounding material until the boundary of the discontinuity is established. An area of unbonded clad in excess of acceptance standards is repaired.Upon completion of all postweld heat treatments, the reactor vessel was hydrostatically tested, after whichall weld surfaces, including those of welds used to repair material, were magnetic particle inspected in accordance with Section III, paragraph N-618 of the ASME Boiler and Pressure Vessel Code.Table 5.4-4 summarizes the component inspection program during fabrication and construction.
Periodic tests and inspections of the reactor coolant system are conducted after startup on a regularbasis.For preoperational and inservice inspection of the reactor coolant system, refer to Section 16.4. Tests forreactor coolant system integrity after a shutdown following refueling, modification or repair are specified in Chapter 16.Replacement RVCH Non-Destructive InspectionTable 5.4-4 summarizes the quality assurance program inspections for the replacement RVCH. In thistable are identified all of the non-destructive test and inspections required by the RVCH Certified Design Specification. All tests required by the applicable Code (ASME Section III, Applicable Edition and Addenda) are included in the table as well as any additional test or more stringent acceptance criteria as may have been specified in the Certified Design Specification for the RVCH.In addition to the inspections summarized in Table 5.4-4, there are those inspections which the equipmentsupplier 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 Section III 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 Penetrant Testing at various stages during the processes and after final completion of the hydrostatic test of the RVCH.5.4-10Amendment No. 21 (12/05)
The RVCH requires welding and weld cladding performed under procedures which require the use of bothpreheat and post-weld heat treating. Pre-heat of weld areas and post-weld heat treating are performed on all welds on the replacement RVCH. Pre-heat and post-weld heat treat of weldments both serve the common purpose of producing tough, ductile metallurgical structures in the weldment. Pre-heating 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.FPL and the Owner's Agent reviewed the manufacturer's quality control methods and results of the vendorand subvendors 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 also 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 found them to be complete and acceptable.5.4.6ADDITIONAL TESTS During design and fabrication of the reactor vessel additional tests were performed as summarized inTable 5.4.5.5.4-10aAmendment No. 21 (12/05)
TABLE 5.4-1REACTOR VESSEL PARAMETERSDesign Pressure, psig 2485Design Temperature, F 650Nozzles Inlet (4 ea.), ID, in.30Outlet (2 ea.), ID, in.42 CEDM (65), ID, in.
2,728Instrumentation (8), nominal in.
4,625Vent (1), nominal, in.
3/4DimensionsInside Diameter, nominal, in.
172Overall Height, Including CEDM Nozzles 503 3/4Height, Vessel without Head 408 9/16Wall Thickness, minimum, in.8 5/8Upper Head Thickness, minimum, in.
7 3/8Lower Head Thickness, minimum, in.
4 3/8Cladding Thickness, Bottom Head, minimum in.3/16Cladding Thickness, Remainder of vessel, minimum, in.1/8 Cladding Thickness, nominal, in.5/16Material (Note 1)
ShellSA-533-65 Grade B, Class 1 SteelNo. of Shell Courses 2ForgingsSA-508-64, Class 2 CladdingsType 308 Stainless Steel*Dry Weights Head, lb.158,400Vessel, without flow skirt, lb.
682,000Studs, Nuts, and Washers, lb.
38,900Flow skirt, lb.5,030 Total 884,330 Weld deposited austenitic stainless steel Type 308, which is equivalent to SA-240, Type 304 in contact withcoolant.Note 1: Material for Replacement RVCH5.4-12Amendment No. 21 (12/05)
TABLE 5.4-1 (cont.)Replacement RVCHForgingSA-508, Class 3Lifting LugsSA-533 Gr B, Class 1CEDM NozzlesSB-167 690 (UNS N06690)Code Case N4792CEDM AdaptersSB-166 690 (UNS N06690)Code Case N4792Instrumentation NozzlesSB-167 690 (UNS N06690)Code Case N4792Instrumentation Nozzle AdapterASME SA-479 Type 304Austentic Stainless Steel CladdingER-309L/ER-308LWeldsBase MaterialWeld MaterialSA-508 to SB-167ERNiCrFe-7 or ENiCrFe-7SB-167 to SB-166ERNiCrFe-7 or ENiCrFe-7SB-167 to SA-479ERNiCrFe-7SA-508, Class 3 to SA-533 Gr B, Class 1SFA 5-5 or E80185.4-12aAmendment No. 21 (12/05)
TABLE 5.4-2 SURVEILLANCE SPECIMENS PROVIDED FOR EACH EXPOSURE LOCATION Capsule Location on Vessel Wall Base Metal Weld Metal HAZ Reference Total Specimens (See Figure 5.4-2)
Imact Tensile Impact Tensile Impact Tensile Impact (c) Impact Tensile L (a)
T (b)83 12 12 3 12 3123--489 97 12 123123123--489 104 12 -3123123 12489 263 12 -3123123 12489 277 12 123123123--489 284 12 123123123489__ ____________________72 481872187218 2428854 (a)L = Longitudinal (b) T = Transverse (c) Reference Material Correlation Monitors 5.4-13 Amendment No. 16, (1/98)
TABLE 5.4-4REACTOR COOLANT SYSTEM NON-DESTRUCTIVE TESTS1.Reactor Vessel (See Heading 6 for Replacement (RVCH)ForgingsFlangesUT, MT StudsUT, MT CladdingUT, PTNozzlesUT, MT PlatesUT, MT CladdingUT, PTWeldsMain SeamsRT, MTMain Nozzles to ShellRT, MT CladdingUT, PTNozzle Safe EndsRT, MTVessel Support BuildupUT, MTAll Welds - After Hydrostatic TestMT or PT2.Replacement Steam GeneratorsTube SheetForgingUT, MT CladdingUT, PTPrimary HeadForgingUT, MT CladdingUT, PTSecondary Shell and HeadPlates and ForgingsUT, MTTubesUT, ETNozzles (Forgings)UT, MT (or PT)
StudsUT, MTWeldsSecondary Shell, LongitudinalRT, MT, UTShell, CircumferentialRT, MT, UT CladdingUT, PTNozzles to ShellRT, MT, UTTube-to-Tube Sheet PTInstrument Connections PTTemporary Attachments after Removal MTAll Welds - After Hydrostatic TestMT or PTNozzle Safe EndsRT, (MT or PT), UTLevel Nozzles PT5.4-15Amendment No. 21 (12/05)
TABLE 5.4-4 (Cont.)3.Pressurizer HeadsForgingsUT, MT CladdingUT, PT ShellForgingsUT, MT CladdingUT, PTHeatersTubingUT, PTSafe EndsUT, MT StudsUT, MTWeldsShell, CircumferentialRT, MT CladdingUT, PTNozzle Safe EndsRT, PTInstrument Connections PTSupport SkirtRT, MTTemporary Attachments after RemovalMT or PTAll Welds after Hydrostatic TestMT or PTHeater Assembly PTEnd Plug Weld PT4.PumpsCastingsRT, PTForgingsUT, PTWeldsCircumferentialRT, PTInstrument Connections PTAll Welds after Hydrostatic Test PT5.PipingFittings (castings)RT, PTPipe (cast)RT, PTNozzles (carbon steel forgings)UT, MTPipe and elbowsCarbon steel plateUT, MTRoll bond cladUT, PT5.4-16 Amendment No. 21 (12/05)
TABLE 5.4-4 (Cont.)WeldsCircumferentialRT, MTNozzle to Run PipeRT, MT, UTInstrument Connections PT CladdingUT, PTSafe ends to NozzlesRT, PTRTUTPTMTET6.Replacement Reactor Vessel Closure Head6.1Head Mono-block Forging6.1.1After Rough Machiningyesyes6.1.2After Final Machiningyes6.1.3Machined Surfaces To Be Cladyes (4)6.1.4External Un-clad Surfacesyes (1)6.1.5All Clad Surfacesyes(2&3)yes (5)6.1.6Final Machined O-Ring Grooveyes (6)6.2SB-167 UNS N0 6690 CEDM & ICIyes (10)yes (7)6.3SB-166 UNS N0 6690 CEDM Nozzlesyes (10)yes (7)6.4SA-479 Type 304 ICI Nozzle Adapteryes (10)yes (7)6.5Weldment6.5.1All Weld Prep Areasyes (8)6.5.2Root Pass of All Weldsyes6.5.3Final Surface of All Weldsyes (9)6.5.4Nozzle to Forging Weld Areayes (11)6.5.5CEDM & ICI Nozzle to Adapter Butt Weldyes6.5.6Vent Pipe Nozzle to Vent Pipe Butt Weldyes1.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 surface of all CEDM nozzle attachment and vent pipe welds must be freeof indications (PT White).10.After rough machining.
11.This is a baseline for future examinations.
Legend:RT = RadiographicUT = Ultrasonic PT = Dye Penetrant MT = Magnetic Particle ET = Eddy Current5.4-17Amendment No. 21 (12/05)
TABLE 5.4-5REACTOR COOLANT SYSTEM ADDITIONAL INSPECTIONSReactor VesselInspectionCode RequirementUltrasonic Testing Of weld clad for bond NoneDye PenetrantTest root, each 1/2 inch andPT test of each 1/2 weldfinal layer of welds for partialthroat or 1/2" whicheverpenetration welds to controlis lesselement drive mechanism head adapters and instrument tube connectionsSteam GeneratorUltrasonic TestDefects in tube sheet clad NoneWeld clad for bond NonePressurizerUltrasonic TestingClad for bond None5.4-18Amendment No. 21 (12/05)
TABLE 5.4-6EX-VESSEL DOSIMETER CAPSULE LOCATIONS FLUX FLUX FLUXAZIMUTHMONITORNICKEL WIREIRON WIRE(DEGREES)CAPSULE CAPSULE CAPSULE-1-2 3-219 0-329 61--9 7- 4 3 105- - 4 108- - 5 135 207- - 6 284 2865.4-19Am. 3-7/85 TABLE 5.4-7 REACTOR VESSEL BELTLINE PLATES LOCATION HEAT NO.CODE NO.%Cu%Ni RT(NDT)Intermediate Shell A-4567-1 C-7-1 0.11 0.64 0 F Intermediate Shell B-9427-1 C-7-2 0.11 0.64-10 F Intermediate Shell A-4567-2 C-7-3 0.11 0.58 10 F Lower Shell C-5935-1 C-8-1 0.15 0.56 20 F Lower Shell C-5935-2 C-8-2 0.15 0.57 20 F Lower Shell C-5935-3 C-8-3 0.12 0.58 0 F 5.4-20 Amendment No. 18, (04/01)
Florida Power & Light CompanySt. Lucie Plant Unit 1Reactor VesselFigure 5.4-1Amendment No. 21 (12/05)
- m z "T1 i r r 0 0 m :JJ z n "' 6 -I n?t z :-t )> p :!! "'o r -o co cZ c: 0 Cl r"' n :E ;J -m -c: mO m ::u ::0 >"'11 .,, m "" J> r m::u z-:p.. 3:< -I Cl N IXlm c: :c r-z -I -r -n mr -I 0 .... :!: -0 n )> m z -< I I ( "'\ \ \ \. \ REACTOR VESSEL l I ' ' ' ) I I I , __ ---___ ,,
- oo ENLARGED PLAN VIEW ... ....--_,..*'"\ INLET ' \NOZZLE \ ,... I I ,J I ' ) CORE MIDPLANE I ") 284° I I ' I '_ ................
,,,1 * ...__VESSEL CAPSULE ASSEMBLY ELEVATION VIEW
- * * *
- Tensile -Monitor---
.........
Compartment Tensile -Monitor Compartment
.....
Tensile -Monitor---*
Compartment , ___ Lock Assembly } Wedge Coupling Assembly Charpy Impact Compartments C harpy Impact Compartments Am. 3-7 /85 FLORIDA POWER &
Typical Surveillance Capsule Assembly Figure St. Lucie Plant 5.4-3a
- * * * .. Temperature, Flux Capsule Assembly --..1 Flux Capsule Assembly ---lock Assembly } Wedge Coupling Assembly filler Capsule Assemblies Fi lier Capsule Assemblies Am. 3-7 /85 Figure FLORIDA POWER & LIGHT CO
- St. Luc:ie Plant Unit 1 REPLACEMENT SURVEILLANCE CAPSULE ASSEMBLY 5.4-31
- Wedge Coupling -End Cap Charpy Impact Specimens Rectangular Tubing Wedge Coupling -End Cap FLORIDA POWER & LIGHT CO. Typical Charpy Impact Compartment Assembly St. Lucie Plant Figure 5.4-4 Wedge Coupling -End Stainless Steel Tubing Threshold Detector Flux Spectrum Monitor Temperature Monitor Temperature Monitor------{JJ Housing Tensile Specimen -Split Spacer -
I Tensile Specimen Housing __ ...,.. Flux Spectrum Monitor Cadmium Shielded Stainless Steel Tubing Cadmium Shield Threshold Detector Quartz Tubing Weight Low Melting Alloy Rectangular Tubing Wedge Coupling -End Cap FLORIDA POWER & LIGHT CO. Typical Tensile-Monitor Compartment Assembly Figure St. Lucie 5.4-5
Refer to drawing 8770-898 , FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 PRESSURIZER QUENCH TANK-ARRANGEMENT FIGURE 5.5-5 Amendment No. 15 (1/97)
- *
- REFER TO DRAWING 8770-G-078, Sheets lllA, B, c & D FLORlDA
& LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FLOW DIAGRAM REACTOR COOLANT PUMP FIGURE 5.5-7 Amendment No. 16, (1/98)
- Rotating Element
- Stationary Element
- FigL.:ra FLORIDA 5.5-9 POWER & LIGHT CO. ANTI-ROTATIONAL DEVICE St. Lucie Plant
-0 m I:"°'"" i:: pa. r-:1 ,_o _:iu "'G'>6 I-' :c )> t ..... rt n p Al (I) w Q. 0 """I\ n 8 -:J -.,, c: 3 -0 .,, (1) :t 0 """I\ 3 w :J £ Vi Vi I ...... 0 ., co c .... (1) Tltll HHd in Feet NPSH in Feet of Water Percent B. H.P. at SP. Gr. 111 1. 0 B.H.P. at SP.GR.* 0.76 s "'---* 2m -90 100 LIO 0 6000 40CNJ 2000 70 60 50 40 30 20 10 0 flywheel Inspection Ports Fan Guide Bearing Housing Stator Yoke Support Rmgs Backstop Coil Support Rings
- T R PUMP IC AUX. Ft>W . I-IJ-0'}-r \_ -HISTORICAL
- .
- SPARGER (10/99) Amendment 17,
- *
- FEEOWATER HEADER I I I 1* j , A , ___ J 'J'TUBE SECTION ',\-:\' TYP. Amendment No. 16, (1/98) FLORIDA POWER & LIGHT COMPANY ST.
PLANT UNJ:T l STEAM GENERATOR FEED RING Ji':tatllU!:
- s. 5-l.3 FLOW DIAGRAMFIGURE 5.5-14Amendment No. 21 (12/05)
A high temperature alarm is provided on this channel to alert the operator to a high temperature condition.
The temperature from this measurement channel is indicated in the control room in addition to being
recorded. The other hot leg temperature channels are also displayed in the control room.
5.6.1.2 Cold Leg Temperature Each of the four cold legs contains three temperature measurement channels. The cold leg RTDs are located downstream of the reactor coolant pumps. Two channels from each cold leg (four per heat transfer 5.6-1a Am. 4-7/86 loop) are used to furnish a cold leg coolant temperature signal to the reactor protective system. All eight of these cold leg temperatures are indicated in the control room.
The remaining cold leg temperature measurement channels, one on each cold leg, are routed to a channel selector switch (One per heat transfer loop). This selector switch enables either cold leg temperature to be recorded on a wide range temperature recorder in the control room. The remaining channel of each loop provides a signal to the average temperature calculator in the reactor regulating system and to the automatic
CEA withdrawal prohibit subsystem of the control element drive system.
For other reactor coolant system temperature measurement channels refer to Table 5.6-1.
5.6.2 PRESSURE 5.6.2.1 Pressurizer Pressure Four independent narrow pressure channels are provided for initiation of protective system action. The pressure transmitters are connected to the upper portion of the pressurizer via the upper level
measurements nozzles and measure pressurizer vapor pressure. All four channels are indicated in the
control room and actuate separate high, low, or low -low pressure alarms in the control room.
The protection actions these pressure signals initiate are:
a)reactor trip on high primary system pressure. The reactor trip signals are also used to open the power operated relief valves at 2385 psig
- b)safety injection system actuation on low-low primary system pressure; c)reactor trip on low reactor coolant system pressure. The set point is a function of the
coolant temperatures in the hot and cold legs. The variable set point has high and low limits
alarmed in the control room and is not allowed to decrease below 1887 psia.Two independent pressure channels provide narrow range pressure signals for controlling the pressurizer
heaters and spray valves. The output of one of these channels is manually selected to perform the control
function. During normal operation, a small group of heaters are proportionally controlled to offset heat
losses. If the pressure falls below a low pressure set point, all of the heaters are energized. If the pressure increases above a high-pressure set point, the spray valves are proportionally opened to increase the spray
flow rate as pressure rises. These two channels are also used to provide 5.6-2 Amendment No. 17 (10/99)
between the four hot leg nozzles and the steam generator nozzles, resulting in four steam generator differential pressures.
The output of the transmitters are sent to four analog summing devices in the low total flow trip logic. Each summer receives two differential pressure signals with the summation of these signals corresponding to the total core flow at all times.
The summers provide four independent total flow signals. The four signals are indicated separately in the control room and activate separate low flow alarms. In the reactor protective system, they are compared with
the low flow reactor trip setpoint. If two channels indicate a flow which is less than the flow setpoint, the reactor is tripped.
5.6.5 SUBCOOLED MARGIN MONITORING SYSTEM The subcooled margin monitoring system is described in Section 7.5.4.2.
5.6-4 Amendment No. 14, (6/95)
PAGE LEFT INTENTIONALLY BLANK 5.6-5 Ame ndment No. 14, (6/95)
TABLE 5.6-1 REACTOR COOLANT SYSTEM INSTRUMENTATION Indication Alarm 1 System Parameter Contr Normal 3 Inst.
& Location Local Room High Low Rec 1 Control Function Inst. Range 4 Operating Accuracy 4 Pressurizer Temperature
- 653F Spray Line Temperature
- 520 - 548.5F Surge Line Temperature
- TAVG/TREF 574F Loop 1B Hot Leg Temp.***RRS, TM/LP 594F 5.6-6 Amendment No. 18, (04/01)
TABLE 5.6-1 (Cont'd) Indication Alarm 1 System Parameter Contr Normal 3 Inst.
& Location Local Room High Low Rec 1 Control Function Inst. Range 4 Operating Accuracy 4 Loop 1B1 Cold Leg Temp.
- TAVG/TREF 574F Quench Tank Temp.**85 - 104F Pressurizer Pressure****Proportional Heaters, 2250 PSIA Backup He aters, Spray
Valves, Protective
Functions (Safety Related)
- Shutdown Cooling &
Safety Injection Tank
Interlocks Reactor Vessel Leakage
- 0 PSIA Quench Tank Pressure**16-18 PSIA 5.6-7 Amendment No. 18, (04/01)
1)The system permits remote (control room) venting from the reactor vessel head orthe pressurizer.2)The vent system is operable following all design basis events except those requiringevacuation of the control room.3)Positive open/close control room position indication is provided for all solenoidoperated valves. This indication is provided by reed switches which directly sense thevalve stem position. The switches are environmentally qualified to the same requirements as the valves.4)During normal plant operation, power is removed from the solenoid valves tominimize the probability of inadvertent operation of the RCGVS. Administrative procedures insure reconnection of power in the event that operation of the RCGVS is required.5)The RCGVS is designed for a single active failure with active components poweredfrom their respective redundant emergency power sources. Parallel vent paths with valves powered from alternate power sources are provided. The solenoid operated valves are powered from safety grade 125V dc power supplies. Power is removed from the fail closed valves, by utilizing key-locked control switches, to minimize the possibility of inadvertent operation during normal operation.c)Piping and Arrangement1)The vent path is safety grade and meets the same qualifications as the RCS.Redundance in the vent path is provided and essential piping and components are seismic Category 1, Safety Class 2.2)The system is designed not to interfere with refueling maintenance actions. Systempiping is flanged where required to facilitate removal of components that might interfere with refueling operation.3)Vent paths are provided to both the quench tank and containment atmosphere. Thequench tank path allows for cooling of gases and condensing water vapor by releasing the vented gases below the water level in the tank. The containment vent path terminates in the area where good air mixing and maximum cooling properties exist.4)The vent system materials are designed to be compatible with superheated steam,steam/water mixtures, water, fission gases, helium, nitrogen, and hydrogen as high as 2500 psia and 700F.5.7-2
TABLE 5.7-1 (Cont'd)
InherentFailureSymptoms and Local EffectsMethod ofCompensatingRemarks andNo.NameModeCauseIncluding Dependent FailuresDetectionProvisionOther Effects
- 12. Leakage Detection a. Fails Open Mechanical binding, seat leakage. Inability to isolate leakage Valve position indication N one Leakage detection system Valve V1449detection system from RCGVS.in C.R.represents another path tocontainment, though notrecommended to be used
as such.b. Fails ClosedMechanical failure, loss of power.No impact on System operation.Valve position indication NoneLoss of ability to measure leakagein C.R.remotely.13. Position Indicator for V1449 False indication of valve position Electro-mechanical failure.
Loss of ability to detect valve pos-Drain from leakage detectionNoneition in leakage detection line.system to graduated sump,increase in sump level shown valve is open.
5.7-9
²
- *
- U"\ C'J f./l .c O"l=:l 0 *-c.n u :r: c.n l.i.l c: Q) Vl oit Ll"I Cl . .!2-i....
N VJ ,-j 0 .-I i.... Q) / t-N H \ *-u < I ,._ i.... w 0 =i C/l ....... c.n N J Vl . 0) ' . Q) a,; --::t c:i... .. , a::. i-t;: G :/ a.. > O"l =i Lf"'I r-1 c: w... c: Vl Q.> Q) 0... > 0 ("\J c:: > Cl.. O"l > i... _, Q.> OJ I-c::o -C'O :i... ......, Vi Q.> Q.> >-I $: 0 > 0 1.-,.f a.. C'O ro I , I > w 1.->-c I 0 0 -u u a.> Q.) I C"t:J -Cl) ro ..fl/ I 0::. Vi I' >-r I ..c: i.... 01 n::l I :r: "O ' c: 0 I u Q.> Vl _,,.,, v---)! ;i.._. _ _, Cl N .-I 0 0-o0 ,....; ,....; ,....; c::) d aJnssaJd u51saa oi pez11ewJON a..1nssaJd wea+s FLORIDA POWER & LIGHT CO. St. Lucic Plant Steam Generator Pressure . Complete Loss of Turbine -Generator Load With DelJyed Reactor Trip u Q.) Vl -G) E I-Figure 5A-l
-I < t"I"\ Ir\ . Q) >. ...... .0 i ro I-t ra (.) VJ "' I >-, . -ra u > >. ro u -c:: ra (,f') -O"l -VI E -Q) ::J 0 E 0 -c "C Q.) N 0 ro ,_.... E !.-0 z * >, -*-<:::I'-u ro 0 c.. ro u Q.) > C'O > 00 >. . ..... 0 Cl,) -t'tl (,f') >. "-l't'I E '-Q.. a * ..... . .....
Figure FLORIDA POWER & LIGHT CO. St. Lucie Plant Optimized Safety Valve S izl ng SA-2
- * * . ..... Q. *c t-Cl) .... ::J "" "" Cl) / .... 0.. / .... / Cl) .!:::! / ... I ::J "" I "" .I G.> ... / 0.. .i= / CJ" / *-/ ::c I / I ,, ' .I .I I I .I I / I / .I I I I I I .... " og. ' , 't>.-I C'Q ... ' Cl.>t-I 0:: ... \ * .c CL> ' :co.. Q 0-_. 0 aJnssaJd uo1saa Ol pazuewJON aJnssaJd wa\SAS iue100J wnw1xew / / Q N ""' LI'\ ""' ...... u Cl) "" . Cl) E *-..... Q -LI'\ FLORIDA POWER & LIGHT CO. St. Lucie Plant Maximum Reactor Coolant Pressure vs Time for Worst Case Loss Of Load Incident With Delayed Reactor Trip Figure 5A-3 CL> .... ::l LI'\ Cl.. VI *-VI .... ..... -CL> .... .... c. u ... CL> o...._ CL> QJ! ........ Vl !o... CL> ..... .. o..."C) NI-CL> *-o E .... QJ '-i:; i= Cfti ::l rg uo:: VI CL> .... 0
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____ ...._ __
ti'\ . ...... . . . . ...-4 ...... ...... ..... JaMOd Ol pazuewJON JiMOd 9JOJ Maximum Reactor Power vs Time Figure FLORIDA POWER & LIGHT CO. St. Lucie.Plant for Worst Case Loss of Load Incident
- *
- 60 lfe. 56 M .. 0:: 1.1.J 0 6 Q.. 1..1.J 0:: 52 0 u 6 48 {\ e::,6 6 6 6 ilS:i 0 0 6 0 6 6 0 (:{) t'.(:; 6 0 0 0 C;, 0 0 0 0 PALISADES TEST o SURGE CODE SIMULATION c, SURGE CODE WITH PLANT DESIGN INPUT DATA 6 0 0 50 I 00 150 200 250 300 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 TIME, SECONDS +10'/o POWER STEP Figur*e 5A -5 555 6666 6 " 6 6 L.l l") 6 6 6 550 6 LL-0 ... LLl 545 0:::
- UJ > c( _) I-_.----,,-*-* ---"" '-J -I ,,,....,,.....0
'--' --540 0/ 0// q,_/ 0/ 0 0 .* 0 00 -PALISADES TEST 535 0 oSURGE CODE SIMULATION
!\SURGE CODE WITH PLANT L DESIGN INPUT DATA 5300 50 100 150 200 50 300 TIME, SECONDS FLORIDA Figure POWER & LIGHT CO. St. Lucie Plant Unit 1 +10% POWER STEP 5A -6
- 900 ]
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100 200 300 400 500 TIME, SECONDS -PALISADES TEST o SURGE CODE SIMULATION 6 c. SURGE CODE WITH PLANT DESIGN INPUT DATA 6 Ci 6 6 6 I 100 200 300 400 500 TIME, SECONDS Figure 5% PER MINUTE POWER RAMP (51% TO 18% POWER) SA -17
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- POWER & HT co. 5% PER MINUTE POWER RAMP (51% TO 18% POWER) SA _ is St. Lucie Plant l
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- REFERENCE LEVEL 30 TIME -SECONDS 40 50 60 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 LEVEL VS. TIME DURING LOSS OF LOAD WITH DELAYED REACTOR TRIP FIGURE 1 5A-19
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YO.L::>'o'3 l:I (/) 0 z 8 UJ Cl'.l I UJ :ii in ... Q .... FLORIDA POWER & LIGHT COMP ANY ST. LUCIE PLAt..iT UNIT 1 NORMALIZED PRESSURE VS. TIME DURING LOSS OF LOAD WITH DELAYED REACTOR TRIP FIGURE 5A-20 * *
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- Ad1'dHJ.N3 W\t3J.S 03Z11VWl:ION FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 NORMALIZED STEAM ENTHALPY DURING LOSS OF LOAD WITH DELAYED REACTOR TRIP FIGURE SA-21
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- RATE VS. TIME DURING LOSS OF LOAD WITH DELAYED REACTOR TRIP FIGURE SA-22 APPENDIX 5BLow TemperatureReactor Coolant SystemOverpressure MitigationForSt. Lucie Unit 15B-iAmendment No. 21 (12/05)
APPENDIX 5CANALYSIS OF NATURAL CIRCULATION COOLDOWNWITHOUT UPPER HEAD VOIDINGFOR ST. LUCIE UNIT 1 DECEMBER 1980NUCLEAR ANALYSIS DEPARTMENTFLORIDA POWER & LIGHT COMPANY5C-i
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