ML16256A242

Waterford Steam Electric Station, Unit 3, Revision 309 to Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems, Section 5.3

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WSES-FSAR-UNIT-3 5.3-1 Revision 307 (07/13) 5.3 REACTOR VESSEL 5.3.1 REACTOR VESSEL MATERIALS

5.3.1.1 Material Specifications (EC-1020, R307)

The principal ferritic materials used in the reactor ve ssel are listed in Table 5.2-3. These materials were specified to be in accordance with the ASME Boile r and Pressure Vessel Code,Section III, 1971 Edition including Summer 1971 Addenda, except for the Replac ement Reactor Vessel Closure Head materials which were specified to be in accordance with the ASME Boiler and Pressure Vessel Code,Section III, 1998 Edition through 2000 Addenda. (EC-1020, R307)

5.3.1.2 Special Processes Used for Manufacturing and Fabrication

The reactor vessel is a right circular cylinder wi th two hemispherical heads.

No special manufacturing methods that could compromise the integrity of the vessel are used. The lower head is permanently welded to the lower end of the reactor vessel shell but the upper closure head can be removed to provide

access to the reactor vessel internals. The head flange is drilled to match the vessel flange stud bolt

locations. The stud bolts are fitted with spherical washers located between the closure nuts and the head flange. These washers maintain stud alignment during boltup when flexing of the head must be accommodated. The lower surface of the head flange is machined to provide a mating surface for the

vessel closure seals.

The vessel flange is a forged ring with a machined ledge on the inside surface to support the core support barrel, which in turn supports the reactor internals and the core. The flange is drilled and tapped to

receive the closure studs and is machined to provi de a mating surface for the reactor vessel closure seals. An externally tapered transition secti on connects the flange to the cylindrical shell.

Sealing is accomplished by using two silver-pla ted, NiCrFe alloy, self-energized 0-rings.

Nozzles are provided in the closure head for nucl ear instrumentation and control element drive mechanisms (CEDM).

The inlet and outlet nozzles are located radially on a common plane just below the vessel flange. Ample thickness in this vessel course provides most of the reinforcement required for the nozzles. Additional reinforcement is provided for the individual nozzle attachments. A boss located around the outlet nozzles

on the inside diameter of the vessel wall provides a mating surface for the core support barrel and guides the outlet coolant flow. This boss and the outlet sleeve on the core support barrel are machined to a common contour to minimize reactor coolant bypass leakage. Shell sections are joined to the nozzle region by a transition section.

Snubbers built into the lower portion of the reacto r vessel shell limit the amplitude of flow-induced vibrations in the core support barrel.

WSES-FSAR-UNIT-3 5.3-2 Revision 15 (03/07) 5.3.1.3 Special Methods for Nondestructive Examination

Prior to, during, and after 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

indicated.

All full-penetration, pressure-containing welds were 100 percent radiographed to the standards of Section III of the ASME Boiler and Pressure Vessel Code. Weld preparation areas, back-chip areas, and final weld surfaces were magnetic-particle or dye-penetrant examined. Other pressure-containing welds, such

as used for the attachments of nonferrous nickel-chromium-iron mechanism housings, vents, and instrument housings to the reactor vessel head, were inspected by liquid-penetrant tests of the root pass, the lesser of one- third of the thickness or each 1/2 in. of weld deposit, and the final surface. Additionally, the base metal weld preparation area was magnetic-particle examined prior to overlay with nickel-

chromium-iron weld metal.

All forgings were inspected by ultrasonic testing, using longitudinal beam techniques. In addition, ring forgings were tested using shear wave techniques. Rejection under longitudinal beam inspection, with calibration so that the first back reflection is at least 75 percent of screen height, was based on indications causing complete loss of back reflection (when not associated with geometrical configuration).

All carbon-steel forgings and ferrite welds are also subjected to magnetic-particle examination after stress

relief. Rejection is based on relevent indication of:

a) Any cracks and linear indications

b) Rounded indications with dimensions greater than 3/16 in.

c) Four or more rounded indications in a line separated by less than 1/16 in. edge to edge d) Ten or more rounded indications in any 6.0 in.

2 in the most unfavorable locations

Plates were subjected to ultrasonic examination using straight beam techniques. Rejection was based on areas producing a continuous total loss of back reflection with a frequency and instrument adjustment that produce a minimum of 50 to a maximum of 75 percent of full scale reference back reflection from the

opposite side of a sound area of the plate.

(DRN 06-872, R15)

Any defect that showed a total loss of back reflection that could not be contained within a circle whose diameter is the greater of three inches or one-half the plate thickness was unacceptable. Two or more defects smaller than described above, which cause a complete loss of back reflection, shall be unacceptable unless separated by a minimum distance equal to the greatest diameter of the larger

defect, unless the defects are contained within the area described above. All carbon and low-alloy steel products were magnetic-particle examined after accelerated cooling to the magnetic-particle acceptance

standard cited above. (DRN 06-872, R15)

WSES-FSAR-UNIT-3 5.3-3 Revision 15 (03/07)

(DRN 06-872, R15)

Nondestructive testing of a vessel was performed throughout fabrication. The nondestructive examination requirements including calibration methods, instrumentation, sensitivity, and reproducibility of data, are in accordance with requirements of the ASME B&PV Code,Section III. (See Table 5.2-1).

Strict quality control was maintained in critical areas such as calibration of test instruments. (DRN 06-872, R15)

All vessel bolting material received ultrasonic and magnetic-particle examination during the

manufacturing process.

The bolting material receives a straight-beam, radial-scan, ultrasonic examination with a search unit not exceeding one square in. area. The standard for rejection was 50 percent loss of first back reflection or an indication in excess of 20 percent of the height of the back reflection. All hollow material receives a

second ultrasonic examination using angle beam, radial scan with a search unit not exceeding one square in. in area. A reference specimen of the same composition and thickness containing a notch (located on the inside surface) one in. in length and a depth of three percent of nominal section thickness, or 3/8 in., whichever is less, was used for calibration.

Any indications exceeding the calibration notch amplitude are unacceptable. Use of these techniques ensures that no materials that have unacceptable flaws, observable cracks, or sharply defined linear

defects were used.

(DRN 06-911, R15)

The magnetic-particle inspection was performed both before and after threading of the studs. Axially aligned defects whose lengths were greater than one in. and nonaxial defects were unacceptable. (DRN 06-911, R15)

Upon completion of all postweld heat treatments, the reactor vessel was hydrostatically tested at 3125 psig after which all accessible ferritic weld surfaces, including those of welds used to repair material, were magnetic-particle inspected in accordance with Section III of the ASME Code.

5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels

Special controls for ferritic and austenitic stainless steels are as follows:

(DRN 00-1059, R11-A) a) Regulatory Guide 1.31, Control of Stainless Steel Welding is addressed in Subsection 5.2.3.4.

b) Regulatory Guide 1.34, Control of Electroslag Weld Properties is addressed in Subsection 5.2.3.3.

c) Regulatory Guide 1.43, Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components is addressed in Subsection 5.2.3.3.

d) Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel is addressed in Subsection 5.2.3.4. (DRN 00-1059, R11-A)

WSES-FSAR-UNIT-3 5.3-4 Revision 307 (07/13)

(DRN 00-1059, R11-A) e) Regulatory Guide 1.50, Control of Preheat Te mperature for Welding of Low-Alloy Steel is addressed in Subsection 5.2-3-3.

f) Regulatory Guide 1.71, Welder Qualification fo r Areas of Limited Accessibility is addressed in Subsection 5.2.3.3. (DRN 00-1059, R11-A) g) Regulatory Guide 1.99, Effects of Residual El ement on Predicted Radiation Damage to Reactor Vessel Materials (DRN 03-2059, R14)

Westinghouse previously took exception to t he methods and procedure for predicting radiation damage to pressure vessel steels contained in R egulatory Guide 1.99. Westinghouse's formal position on Regulatory Guide 1.99 was forw arded to the U.S. NRC in September, 1975 (1). The methods contained in Regulatory Guide 1.99 for predicting RT NDT shift and Charpy upper shelf energy decreases with irradiation are not appropr iate for determining the irradiation behavior of A533-B, Class 1, materials. The methods utilized are based on non-A533-B materials data and incorporate incorrect assumptions concerning the irradiation behavior of vessel materials.

The curve shown in Figure 5.3-1 is utilized for predicting the RT NDT shift of reactor vessel material with low copper content. The curve is based on 550 F irradiation data for A533-B materials.

The data base was collected from published works on the subject of irradiation damage in reactor vessel materials and from test data generated by a joint research program with Westinghouse (then Combustion Engineering), NRC and the Naval Research Laboratory (NRL). Table 5.3-1 lists this data. The indicated literature referenc es for the data are listed in Table 5.3-2. The RT NDT shift prediction curves are shown in relation to t he data in Figure 5.3-1. Weld and plate irradiation behavior is considered separately, because research has shown that some weld metal tends to be more sensitive to irradiation damage. The curv e is conservatively dr awn, envelopes the data, and follows trends described by the data.

Regulatory Guide 1.99 is now used without exception.

\ (DRN 03-2059, R14)

5.3.1.5 Fracture Toughness (EC-1020, R307)

The reactor vessel materials were ordered to the ASME 1971 Code Section III, Summer 1971 Addenda, specification, except for the R eplacement Reactor Vessel Closure Head materials which were ordered to the ASME Boiler and Pressure Vessel Code,Section III, 1998 Edition through 2000 Addenda. The

materials meet the Charpy impact requirements of Subsection NB-2300 (three tests at a temperature to verify 30 ft.-lbs. of absorbed energy). Longitudinal (strong direction) Charpy test data was used to develop RT NDT's as per Branch Technical Position MTEB 5-2, "Fracture Toughness Requirements". The highest MTEB 5-2 RT NDT value for the Waterford 3 reactor vessel beltline plate material is 22 F (lower shell plate M-1004-2). Testing of vessel weld and heat affected zone materials was not required by the applicable code year and addenda and the materials were not available. (EC-1020, R307)

Transverse (weak direction) Charpy impact data on plate M-1004-2, weld and heat-affected-zone (HAZ) material is reported in Subsection 5.3.

1.6-1, as results of the baseline su rveillance testing. This testing, which establishes an RT NDT in a manner consistent with Appendix G 10CFR50, yields an RT NDT for plate M-1004-2 of -20 F. The RT NDT for the weld and HAZ material are shown in Table 5.3-3 and the Charpy data is plotted in Figures 5.3-2, -3 and -4.

WSES-FSAR-UNIT-3 5.3-5 Revision 14 (12/05

)The lowest reported Charpy upper shelf energy, longitudinal tests, for the reactor vessel beltline material is 140 ft.-lbs., (intermediate shell plate M-1003-1). Using Branch Technical Position MTEB 5-2 and reducing that value to 65 percent, to reflect the difference between longitudinal and transverse testing, yields a conservative Charpy upper shelf energy of 91 ft.-lbs. This is well in excess of the 75 ft.-lb. requirement of 10CFR50, Appendix G. 5.3.1.6 Material Surveillance The irradiation surveillance program for Waterford 3 will be conducted to assess the neutron-induced changes in the RT NDT (reference temperature) and the mechanical properties of the reactor vessel materials. Changes in the impact and mechanical properties of the material will be evaluated by the comparison of pre- and post-irradiation test specimens. The capsules containing the surveillance test specimens used for monitoring the neutron induced property changes of the reactor vessel materials will be irradiated under conditions which represent, as closely as practically possible, the irradiation conditions

of the reactor vessel.

ASTM E-185-82, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, and 10CFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, present criteria for monitoring changes in the fracture toughness properties of reactor vessel beltline materials through surveillance programs. This reactor vessel surveillance program for Waterford 3 adheres to all of the

requirements of ASTM E-185-82 and satisfies 10CFR50, Appendix H.5.3.1.6.1 Test Materials Selection (DRN 03-2059, R14)Regions of both the intermediate and lower shells of the reactor vessel are nearest to the reactor core and, therefore, sustain the greatest neutron exposure. The material from which surveillance test specimens were manufactured were cut from that plate in the core region which would become the limiting plate with respect to reactor operation during its lifetime. This material (lower shell plate M-1004-2) was selected on

the basis of highest initial RT NDT , chemical composition and fluence.The test materials were processed so that they are representative of the materials in the completed reactor vessel. A record of chemical analyses, fabrication history and mechanical properties of the shell plate from which the surveillance test materials were prepared is maintained. The results of mill test chemistries for the six plates of the beltline

region of the vessel are presented in Table 5.2-5. (DRN 00-1059, R11-A) (DRN 03-2059, R14)

Three metallurgically different materials representative of the reactor vessel were used for test specimens. These include base metal, weld metal and heat affected zone (HAZ) materials. In addition to the materials from sections of reactor vessel shell plate, material from a standard heat of ASTM A533-B Class I manganese-molybdenum-nickel steel made available by the USAEC sponsored Heavy Section Steel Technology (HSST) program is also included. This(DRN 00-1059, R11-A)

WSES-FSAR-UNIT-35.3-6reference material has been fully processed and characterized, and is used for Charpy impact specimencorrelation monitors to permit comparisons among the irradiation data from operating power reactors and irradiation data from experimental reactors. Compilation of data generated from post-irradiation tests of the correlation monitors will be carried out by the HSST program.5.3.1.6.1.1Base Metal Base metal test material was manufactured from sections of lower shell plate M-1004-2 which was foundto have the combination of RT NDT, chemical composition (Cu and P), and neutron fluence during service,which would first appear to limit the vessel operating lifetime. The unirradiated RT NDT, of each plate inthe intermediate and lower shells was determined from drop weight and Charpy data as required in NB-2331 of the ASME Boiler Code, 1971 Edition Summer 1971 Addenda,Section III and is shown in Table 5.2-6. All base metal test material is cut from one shell.The section of shell plate used for the base metal test material is adjacent to the test material used forASME Code Section III tests and is at a distance of at least one plate thickness from any water-quenched edge. This material was heat-treated to a metallurgical condition which is representative of the final metallurgical condition of the base metal in the completed reactor vessel.5.3.1.6.1.2Welded PlatesWeld metal and HAZ material were produced by welding together sections from the selected base metalplate and another intermediate plate of the reactor vessel. The HAZ test material was manufactured from a section of the same shell plate used for base metal test material.The sections of shell plate used for weld metal and HAZ test material are adjacent to the test materialused for ASME Code Section III tests and are at a distance of at least one plate thickness from any water-quenched edge. The procedure used for making the intermediate-to-lower shell girth weld in the reactor vessel was followed in the manufacture of the weld metal and HAZ test materials. The welded plates were heat-treated to metallurgical conditions that are representative of the final metallurgical conditions of similar materials in the completed reactor vessel.The test specimens used in establishing the unirradiated RT NDT temperature of the base metal wereobtained from 1/4 T (where T is plate thickness) locations of sections of the plate used in the core region.

The heat-affected-zone samples were taken from the inner region of the deposited weld metal. Theimpact properties of the specimen locations are representative of the material through the entirethickness. Use of the RTNDT values obtained from samples taken from the inner regions of the test materials represent a conservative approach for establishing the initial minimum operating temperature and the base for the predicted minimum operating temperature after irradiation, because the advantages of the more favorable RT NDT properties of the surface regions are not taken into consideration.

WSES-FSAR-UNIT-3 5.3-7 Revision 11-A (02/02)5.3.1.6.2T est Specimens 5.3.1.6.2.1 Type and Quantity(DRN 00-1059)

The magnitude of the neutron-induced property changes of the reactor vessel materials is determined by comparing the results of tests using irradiated impact and tensile specimens to the results of similar tests usingunirradiated specimens. The changes in RT NDT of the vessel materials are determined by adding to the reference temperature (RT NDT) the amount of the temperature shift in the Charpy test curves between the unirradiated material and the irradiated material, measured at the 50 ft.-lb. level or that measured at the 35 mils. lateral expansion level, whichever temperature shift is greater. The new values of RT NDT are known as adjusted reference temperature.(DRN 00-1059)Drop weight, Charpy impact, and tensile test specimens were provided for unirradiated tests. Drop weighttests were conducted in accordance with ASTM E-208.Charpy impact tests were conducted in accordance with ASTM E-3. Tensile tests were conducted in accordance with ASTM E-8 and E-21. Correlation of drop

weight and Charpy impact tests to establish RT NDT were made in accordance with NB-2300 of the ASMECode,Section III.Charpy impact and tensile test specimens are provided for post-irradiation tests.

The total quantity of specimens furnished for carrying out the overall requirements of this program is presented in Table 5.3-4. A sufficient amount of base metal, weld metal, and HAZ test material to provide

two additional sets of test specimens has been obtained with full documentation and identification for future evaluation should the need arise. Each of the test materials has been chemically analyzed for

approximately 21 elements, including all those listed in ASTM E-185-82.

5.3.1.6.2.2 Unirradiated Specimens The type and quantity of test specimens provided for establishing the properties of the unirradiated reactor vessel materials are presented in Table 5.3-5. The data from tests of these specimens provide the basis for

determining the neutron-induced property changes of the reactor vessel materials.a)Drop Weight Test Specimens Twelve drop weight test specimens of base metal (longitudinal and transverse), weld metal, and HAZ material are provided for establishing the NDTT of the unirradiated surveillance materials. These data form the basis for RT NDT determination. RT NDT is the reference temperature from which subsequent neutron-induced changes are determined.b)Charpy Impact Test Specimens Thirty test specimens each of base metal (longitudinal and transverse), weld metal, and HAZmaterial are provided. This quantity exceeds the minimum number of test specimens

recommended by ASTM E-185 for developing a Charpy impact energy transition WSES-FSAR-UNIT-35.3-8curve and is intended to provide a sufficient number of data points for establishing accurateCharpy impact energy transition temperatures for these materials. This data, together with the drop weight NDTT, is used to establish an RT NDT for each material.c)Tensile Test SpecimensEighteen tensile test specimens each of base metal (longitudinal and transverse), weld metal andHAZ materials are provided. This quantity also exceeds the minimum number of test specimensrecommended by ASTM E-185 and is intended to permit a sufficient number of tests for accurately establishing the tensile properties for these materials at a minimum of three test temperatures (e.g., ambient, operating and design).5.3.1.6.2.3Irradiated SpecimensBoth tensile and impact test specimens are used for determining changes in the static and dynamicproperties of the materials due to neutron irradiation. A total of 288 Charpy impact and 54 tensile test specimens is provided. The type and quantity of test specimens provided for establishing the properties of the irradiated materials over the lifetime of the vessel are presented in Table 5.3-6. The attachment of the capsule assemblies to the inside wall of the reactor vessel is described in CENPD-155P. (2)5.3.1.6.3Specimen Irradiation5.3-1.6.3.1Encapsulation of Specimens The test specimens are placed within corrosion-resistant capsule assemblies:

a)To prevent corrosion of the carbon steel test specimens by the primary coolant duringirradiation.b)To physically locate the test specimens in selected locations within the reactorc)To provide a means by which the irradiation conditions (fluence,(a) flux spectrum,temperature) can be determinedd)To facilitate the removal of a desired quantity of test specimens from the reactorwhen a specified fluence has been attained.

______________________________________(a) Time integrated neutron flux.

WSES-FSAR-UNIT-35.3-9A typical capsule assembly, illustrated in Figure 5.3-5, consists of a series of seven specimencompartments, connected by wedge couplings, and a lock assembly. Each compartment enclosure of the capsule assembly is internally supported by the surveillance specimens and is externally pressure tested to 3125 psia during final fabrication. The wedge couplings also serve as end caps for the specimen compartments and position the compartments within the capsule holders which are attached to the reactor vessel. The lock assemblies fix the locations of the capsules within the holders by exerting axial forces on the wedge coupling assemblies which cause these assemblies to exert horizontal forces against the sides of the holders preventing relative motion. The lock assemblies also serve as a point of attachment for the tooling used to remove the capsules from the reactor.Each capsule assembly is made up of four Charpy impact test specimen (Charpy impact) compartmentsand three tensile test specimen - flux/temperature monitor (tensile-monitor) compartments. Each capsule compartment is assigned a unique identification so that a complete record of test specimen location withineach compartment can be maintained.a)Charpy Impact CompartmentsEach Charpy impact compartment (Figure 5.3-6) contains 12 impact test specimens. Thisquantity of specimens provides an adequate number of data points for establishing a Charpy impact energy transition curve for a given irradiated material. Comparison of the unirradiated and irradiated Charpy impact energy transition curves permits determination of the RT NDT changesdue to irradiation for the various materials.The specimens are arranged vertically in four 1 x 3 arrays and are oriented with the notch towardthe core. The temperature differential between the specimen and the reactor coolant is minimized by using spacers between the specimens and the compartment and by sealing the entireassembly in an atmosphere of helium.b)Tensile - Monitor CompartmentsEach tensile-monitor compartment (Figure 5.3-7) contains three tensile test specimens, a set offlux spectrum monitors and a set of temperature monitors, for estimating the maximum temperature to which the specimens have been exposed. The entire tensile-monitor compartment is sealed within an atmosphere of helium.The tensile specimens are placed in a housing machined to fit the compartment. Split spacersare placed around the gage length of the specimen to minimize the temperature differential between the specimen gage length and the coolant.5.3.1.6.3.2Flux and Temperature Measurement The changes in the RT NDT of the reactor vessel materials are derived from specimens irradiated tovarious fluence levels and in different neutron energy spectra. In order to permit accurate predictions of

the RT NRT of the vessel materials, complete information on the neutron flux, neutron energy spectra, andthe irradiation temperature of the surveillance specimens must be available.

WSES-FSAR-UNIT-35.3-10a)Flux MeasurementsFast neutron flux measurements are obtained by insertion of threshold detectors into each of thesix irradiation capsules. Such detectors are particularly suited for the proposed application, because their effective threshold energies lie in the low Mev range. Selection of threshold detectors is based on the recommendations of ASTM E-261, "Method of Measuring Neutron Flux by Radioactive Techniques".These neutron threshold detectors and the thermal neutron detectors, listed in Table 5.3-7, can beused to monitor the thermal and fast neutron spectra incident on the test specimen. These detectors possess reasonable long half-lives and activation cross sections covering the desired neutron energy range.One set of flux spectrum monitors is included in each tensile monitor compartment. Eachdetector is placed inside a sheath which identifies the material and facilitates handling. Cadmium covers are used for those materials (e.g., uranium, nickel, copper and cobalt) which havecompeting neutron capture activities.The flux monitors are placed in holes drilled in stainless steel housings as shown in Figure 5.3-7at three axial locations in each capsule assembly (Figure 5.3-5) to provide an axial profile of the level of fluence which the specimens attain.In addition to these detectors, the program also includes correlation monitors (Charpy impact testspecimens made from a reference heat of ASTM A533-B, Class 1, manganese-molybdenum-nickel steel) which are irradiated along with the specimens made from reactor vessel materials.

The changes in impact properties of the reference material provide a cross-check on the dosimetry in any given surveillance program. These changes also provide data for correlating the results from this surveillance program with the results from experimental irradiations and other reactor surveillance programs using specimens of the same reference material.b)Temperature EstimatesBecause the changes in mechanical and impact properties of irradiated specimens are highlydependent on the irradiation temperature, it is necessary to have knowledge of the temperature of specimens as well as the pressure vessel. During irradiation, instrumented capsules are notpractical for a surveillance program extending over the design lifetime of a power reactor. Themaximum temperature of the irradiated specimens can be estimated with reasonable accuracy by including within the capsule assembly small pieces of low melting point alloys or pure metals. The compositions of candidate materials with melting points in the operating range of power reactors are listed in Table 5.3-8. The monitors are selected to bracket the operating temperature range ofthe reactor vessel.The temperature monitors consist of a helix of low melting alloy wire inside a sealed quartz tube.A stainless steel weight is provided to destroy the integrity of the wire when the melting point of the alloy is reached. The compositions and therefore the melting temperatures of the temperature monitors are differentiated by the physical lengths of the quartz tubes which contains the alloy wires.

WSES-FSAR-UNIT-35.3-11Revision 10 (10/99)A set of temperature monitors is included in each tensile-monitor compartment. The temperaturemonitors are placed in holes drilled in stainless steel housings as shown in Figure 5.3-7 and are also placed at three axial locations in each capsule assembly (Figure 5.3-5) to provide an axial profile of the maximum temperature to which the specimens were exposed.5.3.1.6.3.3Irradiation Locations The encapsulated test specimens are irradiated at approximately identical radial positions about themidplane of the core. The test specimens are enclosed within six capsule assemblies the axial positions of which are bisected by the midplane of the core. A summary of the specimens contained in each of these capsule assemblies is presented in Table 5.3-9.The test specimens contained in the capsule assemblies are used for monitoring the neutron-inducedproperty changes of the reactor vessel materials. These capsules, therefore, are positioned near the inside wall of the reactor vessel so that the irradiation conditions (fluence, flux spectrum, temperature) of the test specimens resemble as closely as possible the irradiation conditions of the reactor vessel. Theneutron fluence of the test specimens is expected to be approximately 50 percent greater than that seen by the adjacent vessel wall.The RT NDT changes resulting from the irradiation of these specimens closely approximate the RT NDTchanges in the materials of the reactor vessel.The capsule assemblies are placed in capsule holders positioned circumferentially about the core atlocations which include the regions of maximum flux. Figure 5.3-8 shows the location of the capsule assemblies.All capsule assemblies are inserted into their respective capsule holders during the final reactor assemblyoperation.5.3.1.6.3.4Capsule Assembly RemovalThe capsule assemblies remain within their capsule holders until the test specimens contained thereinhave attained desired levels of exposure (EFPY). At that time, selected capsule assemblies are removed.

The distribution of target exposures for removal of capsule assemblies is presented in Table 5.3-10.The target exposure levels for the surveillance capsules are based on the time intervals indicated in thewithdrawal schedule in ASTM E-185-82, referenced in 10CFR50, Appendix H.Withdrawal schedules may be modified to coincide with those refueling outages or plant shutdowns whichmost closely approach the withdrawal schedule.During unit start-up and shutdown, the rates of temperature and pressure changes are limited. Thedesign number of cycles for heatup and cooldown is based upon a rate of 100

° F/hr and for cyclicoperation.

WSES-FSAR-UNIT-35.3-12The maximum allowable reactor coolant system pressure at any temperature is based upon the stresslimitations for brittle fracture considerations. These limitations are derived by using the rules contained in Section III of the ASME Code including Appendix G, Protection Against Nonductile Failure and the rules contained in 10CFR50, Appendix G, Fracture Toughness Requirements. Compliance with the criteria in 10CFR50, Appendix H is discussed in Subsection 5.3.1.6.5.3.1.7Reactor Vessel FastenersThe stud material for the reactor vessel closure head is fabricated from SA 540 Grade B24 Class IIImaterial. The nuts and washers for the reactor vessel fasteners are made from SA 540 Grade B24 or B23 material. These materials were ordered prior to the issuance of 10CFR50, Appendix G and Regulatory Guide 1.65, "Materials and Inspections for Reactor Vessel Closure Studs". Material tests for Waterford 3 stud material demonstrates adequate toughness in accordance with ASME Code Section III, 1971 Edition through Summer 1971 Addenda, and an acceptable level of ultimate tensile strength which is consistent with the recommendations of Reg Guide 1.65 (see Table 5.3-11). Test results demonstrate that the studmaterial meets the 25 mil lateral expansion, 45 ft-lb criteria of 10CFR50, Appendix G at 10

°F and that theydo not exceed 170 KSI ultimate tensile strength (UTS).Testing adequate to establish compliance with the ASME Code Section III, 1971 Edition through Summer1971 Addenda, was done for the nut and washer material. 10CFR50, Appendix G, Section IV, "Fracture Toughness Requirements", Paragraph 4 requires that material for bolting and other fasteners with nominal diameters exceeding one in. shall meet the minimum requirements of 25 mils lateral expansion and 45 ft-lbs in terms of Charpy V-notch tests conducted at the preload temperature or at the lowest service temperature, whichever is lower.In order to determine whether the nuts and washers met the 10CFR50 Appendix G requirements, allCharpy test data for SA540 Gr B-24 steel (Southern California Edison Electric Co.'s San Onofre Generating Station Units 2 and 3 and Waterford Unit 3 reactor vessel studs, washers and nuts) was accumulated. Waterford Unit 3 data is presented in Table 5.3-12 and all the data is graphically presented in Figure 5.3-9 (Charpy Energy Absorbed) and Figures 5.3-10 (Mils Lateral Expansion). It can be seen that the available data plotted in Figures 5.3-9 indicates that a lower bound curve through minimum pointsyields a Charpy absorbed energy value of approximately 46 ft-lbs at +60

°F. Further, a similar lower boundcurve in Figure 5.3-10 yields a value of 27 mils lateral expansion at +60

°F. Since both these curves areminimum point curves for data points from six separate heats, the temperature necessary to meet10CFR50, Appendix G requirements (C v = 45 ft-lbs, lateral expansion = 25 mils) is 60

°F.

WSES-FSAR-UNIT-3 5.3-13 Revision 14 (12/05

)5.3.2 PRESSURE TEMPERATURE LIMITS 5.3.2.1 Limit Curves The reactor vessel beltline material consists of six plates. The nil ductility transition temperatures (TNDTT) of each plate was established by drop weight test. Charpy tests were then performed to determine at what

temperature the plates exhibited 50 ft-lb absorbed energy and 35 mils lateral expansion. From this testing a reference temperature for transverse direction (RT NDT) of 22F was established.

For the remaining material in the RCS, a limiting RT NDT of 90F was established based upon SA 105 Class 2 material used to fabricate the lower driver mount flanges of the reactor coolant pumps. (DRN 00-1059, R11-A)As a result of fast neutron irradiation in the region of the core, RT NDT will increase with operation. The techniques used to analytically and experimentally predict the integrated fast neutron (E 1 MeV) fluxes of the reactor vessel are described in Subsections 5.3.1.4 and 5.3.1.6. Extent of compliance with Regulatory

Guide 1.99 is discussed in Subsection 5.3.1.4. (DRN 03-2059, R14)

Since the neutron spectra and flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured reference transition temperature shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron flux variation. The peak end-of-license (32 EFPY) neutron fluence (E >1.0 MeV) at the core midplane for the Waterford Unit 3 reactor vessel is 2.48 x

10 19 n/cm 2. That is the fluence corresponding to the clad-base metal interface. Projections of neutron fluence beyond Cycle 11 were based on a 1.5% uprate (3441 MWt) at the start of Cycle 12 and a 8%

uprate (3716 MWt) at the start of Cycle 14. The highest predicted Adjusted Reference Temperature (ART) at 32 EFPY is 50F(7). This corresponds to an integrated fast neutron fluence (E > 1.0 MeV) at the 1/4 thickness of 1.48 x 10 19 n/cm 2 and was determined using the methodology of Regulatory Guide 1.99, Revision 2. The actual shift in RT NDT will be established periodically during plant operation by testing of reactor vessel material samples which are irradiated cumulatively by securing them near the inside wall of

the reactor vessel as described in Subsection 5.3.1.6 and shown in Figure 5.3-8. To compensate for any

increase in the RT NDT caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown. During the first 10 years of reactor

operation, a conservatively high fluence of 9.2 x 10 18 n/cm 2 is assumed which corresponds to 3580 Mwt and 80 percent load factor. The corresponding RT NDT is 75F based on the curve shown in Figure 5.3-1.

Thus, for this interval, the upper limit to the RT NDT is (initial + shift) or 22F + 75F = 97F. This is greater than (and conservatively bounds) the 40 years Adjusted Reference Temperature at 88F calculated from the methodology of Regulatory Guide 1.99, Revision 2.The limit lines identified in Technical Specification

16.3/4.4 are based on the following: (DRN 00-1059, R11-A; 03-2059, R14)

WSES-FSAR-UNIT-35.3-14a)Heatup and Cooldown Curves (from Section III of the ASME Code Appendix G-2215)

K IR = 2 K IM + K IT K IR = Allowable stress intensity factor at temperatures related to RT NDT (ASME III Figure G-2110-1)

K IM=Stress intensity factor for membrane stress (pressure)The 2 represents a safety factor of 2. on pressure K IT =Stress intensity factor for radial thermal gradientThe above equation is applied to the reactor vessel beltline.For plant heatup the thermal stress varies from compressive at the inner wall to tensile at the outer wall.These thermal induced compressive stresses tend to alleviate the tensile stresses induced by internalpressure. Therefore, a pressure-temperature curve based on steady-state conditions (i.e. no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.For plant cooldown thermal and pressure stress are additive. The design cooldown rate of 100

°F/hr isused for calculation.

K IM = M M PR t M M =ASME III, Figure G-2214-1P =Pressure, psi.R =Vessel radius - in.t =Vessel wall thickness - in.

K IT =MT T W M T =ASME III, Figure G-2214-2 T W =Highest radial temperature gradient through wall at end of cooldown.

K IT is therefore calculated at a maximum- gradient and is considered a constant = A for cooldown andzero for heatup. M MR/T is also a constant = B.therefore:

WSES-FSAR-UNIT-3 5.3-15 Revision 15 (03/07)

K IR = BP + A P KA B IR (DRN 06-842, R15)

K IR is then varied as a function of temperature from Figure G-2110-1 of ASME III and the allowable pressure calculated. Other regions of the reactor vessel have also been analyzed. With the exception of

the vessel flange during heatup, the beltline region is controlling after considering RT NDT shifts. The limit curve for heatup is a composite of the limitations imposed by the vessel flange and beltline region.

Instrumentation errors and hydrostatic head corrections are considered when implementing the curves.

Whenever the core is critical, an additional 40°F is added to these curves as required by 10CFR50, Appendix G. (DRN 06-842, R15) b) System In-service Testing

The in-service testing curve is developed in the same manner as in a) above with the exception

that a safety factor of 1.5 is allowed by ASME III in lieu of 2.

c) Lowest Service Temperature

As indicated previously, an RT NDT for all material with the exception of the reactor vessel beltline was established at 90°F. ASME III, Art. NB-2332 (b) require a lowest service temperature of

RT NDT + 100°F for piping, pumps and valves. Below this temperature, a pressure of 20 percent of the system hydrostatic test pressure cannot be exceeded.

d) Maximum Pressure for Shutdown Cooling

This pressure is established by considering the design pressure of the Shutdown Cooling

System, shutoff head of the low pressure safety injection (LPSI) pumps, elevation head from the pressurizer to the LPSI pumps and the design temperature of the Shutdown Cooling System.

The pressure-temperature limitation curves are predicted for 40-year life. During plant life operation, the surveillance capsules (refer to Subsection 5.3.3.7) will be removed from their location in the reactor vessel for testing. The data obtained will be compared to that used to develop the predicted limitation

curves presented in the Technical Specifications.

If this information indicates anomalies to the existing predictions, the curves will be redrawn as previously

indicated to reflect actual data.

WSES-FSAR-UNIT-3 5.3-16 Revision 15 (03/07) 5.3.2.2 Operating Procedures

(DRN 06-911, R15)

Pressure-temperature limitations and additional information are described in the Technical Specifications. The pressure-temperature limit curves provided in Section 3/4.4 have been prepared in accordance with Appendix G, ASME Code Section III. Maintenance of reactor coolant system (RCS) pressure and

temperature within these prescribed limits ensure that the integrity of the reactor coolant pressure

boundary (RCPB) is maintained. (DRN 06-911, R15)

5.3.2.3 Fracture Toughness for Pressurized Thermal Shock Events

(DRN 00-1059, R11-A; 03-2059, R14)

An evaluation was performed of the Waterford Unit 3 reactor vessel beltline materials relative to the Pressurized Thermal Shock (PTS) screening criteria of 10CFR50.61

[3] and the upper shelf screening criteria of 10 CFR Part 50, Appendix G

[4]. The PTS values are calculated in accordance with 10CFR50.61

[3]. The predicted upper shelf energy values are evaluated using the methods of Regulatory Guide 1.99, Revision 2

[5], for each beltline material. The calculation of the PTS and upper shelf energy values represents power uprate conditions, including a 1.5% uprate (3441 MWt) at the start of Cycle 12

and a 8% uprate (3716 MWt) at the start of Cycle 14.

The determination of the chemistry factor values per Position 1.1 and 2.1 of Reference 5 is detailed in Reference 6 and summarized in Table 5.3-14. (In this table, the Chemistry Factor Basis refers to values

from Table 1 and 2 of 10CFR50.61

[3] for weld and plate, respectively, and to surveillance data for the values derived in Reference 6.) The neutron fluence is the peak value for the corresponding plate and

weld for 32 EFPY. RTPTS was determined for each material in the beltline region is given by the following expression:

RTPTS = Initial RT NDT + RTPTS + Margin

Initial RT NDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. Measured values of initial RTNDT are available for each of the materials. RTPTS is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:

RTPTS = CF

• f (0.28 - 0.10 log f) where f is the vessel fluence at the clad-base metal interface given in units of 10 19 n/cm 2. Margin is determined based on the uncertainty in initial RT NDT and the uncertainty in the RTPTS prediction. Margin is calculated as:

M = 2 2 2i The initial RTNDT values are based on measured values and, therefore, i is equal to 0F. The uncertainty in the RTPTS prediction, is 28F for welds and 17F for plates. However, the value of 2 does not have to exceed RTPTS. (Note: The value of for the two materials for which credible surveillance data are available does not have to exceed 14F for welds and 8.5F for base metal). The values of i , , and the total margin are given for each material in Table 5.3-14. Total margin in these cases is 2 or RTPTS , whichever is smaller. Values of RTPTS are given for each material in Table 5.3-14. The highest value is 53F for lower shell plate M-1004-2 at 32 EFPY. All the projected values for the Waterford Unit 3 reactor vessel beltline materials are well below the Pressurized Thermal Shock (PTS) screening criteria of 270F for axial welds and plates, and 300F for circumferential welds. (DRN 00-1059, R11-A; 03-2059, R14)

WSES-FSAR-UNIT-3 5.3-17 Revision 307 (07/13)

(DRN 03-2059, R14)

The predicted upper shelf energy values for each of the Waterford 3 Unit beltline material were evaluated in accordance with 10 CFR Part 50, Appendix G

[4] using Position 1.2 of Regulatory Guide 1.99, Revision 2[5]. The predictions were based on the predicted fluence

[6] at the vessel 1/4T location at 32 EFPY. (Note: The peak beltline fluence at 1/4T was used for all the beltline materials rather than taking credit for the relatively small variation in the axial fluence betw een the intermediate and lower shells.) The projected upper shelf energy values far exceed the 50 ft-lb screening criterion of 10CFR50, Appendix G

[4] at 32 EFPY.

The plate and weld data from the surveillance capsule analyses

[6] were also evaluated using Position 2.2 of Regulatory Guide 1.99, Revision 2

[5]. The projections of upper s helf energy decrease were based on the upper bound of the measurements extended parallel to the lines in Figure 2 of Reference 5. The projected shelf decrease for plate M-1004-2 based on the measurements is 14% at 32 EFPY. The projected shelf decrease for weld 101-171 based on the measurements is 9.5% at 32 EFPY. The projected upper shelf energies are given below at 32 EFPY. The corresponding values based on Position

1.2 are also shown below. This comparison demonstrat es the low radiation sens itivity of the Waterford Unit 3 beltline materials, and it adds confidence to t he expectation that those materials will far exceed the 50 ft-lb screening criterion of 10CFR50, Appendix G

[4] at 32 EFPY.

Comparison of Upper Shelf Energy Decrease Predictions Material 32 EFPY USE 32 EFPY USE (ft-lbs) using (ft-lbs) using Position 1.2 Position 2.2 Plate M-1004-2

117 121 Weld 101-171

124 141 (DRN 03-2059, R14)

5.3.3 REACTOR VESSEL INTEGRITY

(EC-1020, R307)

C-E designed and fabricated the reactor vessel for Wa terford 3 C-E has been involved in reactor vessel design and fabrication since the late 1950's, and this pr oven expertise is reflected in the Waterford 3 reactor vessel and the satisfactory performance of a large number of reactor vessels in operating plants.

Westinghouse designed the Replacement Reactor Vessel Closure Head (RRVCH). The RRVCH was fabricated by Doosan. (EC-1020, R307)

Vessel integrity is ensured by the use of proven fabrication techniques and well characterized steels which exhibit uniform properties and consistent behavio

r. The characterization of these materials was established through industrial and governmental studies which examined the prefabrication material properties through to irradiated service operation. Inservice inspection and material surveillance

programs are also conducted during the service life of the vessel, which further ensures that vessel integrity is maintained.

5.3.3.1 Design

Applicable design codes are found in Table 5.2-1. A schematic of the reactor vessel is shown on Figure 5.3-11. Additional information c an be found in Subsection 5.3.1.2.

WSES-FSAR-UNIT-3 5.3-18 Revision 307 (07/13) 5.3.3.2 Materials of Construction The reactor vessel shell is fabric ated from SA-533, Grade B, Class 1, material. This material has a minimum tensile strength of 80 ksi and a minimum yield strength of 50 ksi. This shell material responds well to quench and tempered heat treatment, which in combination with fine-grain melting practice produces high quality plate with excellent fractu re toughness properties. The nozzles, also having excellent toughness properties, are fabricated from SA508, Class 2 forgings. The welding materials used include Mil Spec B-4 wire for submerged arc processes and E-8018C-3 material for manual arc

processes. The stainless steel cladding utilized is nominal 19Cr-9Ni.

5.3.3.3 Fabrication Methods (EC-1020, R307)

The reactor vessel is constructed of formed plat es welded into cylinders and hemispherical heads. The closure head, upper shell, and nozzles are forgings. This typifies construction of the reactor vessel in the preceding introductory material. No special fabr ication methods were used in the reactor vessel fabrication. The basic design and fabrication of the reactor vessel are as follows.

The reactor vessel is a vertically mounted cylindric al vessel with a hemispherical lower head welded to the vessel and a removable hemispherical upper closur e head. The pressure vessel is approximately 520 in. high (overheads) by 172 in. inside diameter and is all welded manganese molybdenum nickel

steel plate and forging construction. Except for the Replacement Reactor Vessel Closure head which is clad with 3/16 in. minimum. Type 308L (and 309L used for base layer) stainless steel, the internal surfaces that are in contact with the reactor c oolant are clad with 1/8 in. minimum Type 304 austenitic stainless steel and have a finish of 250 micro in.

or better. The closure head flange and reactor vessel shell flange provide the structural rigidity necessary for bolting the head to the shell. (EC-1020, R307)

The reactor vessel fabrication is begun with an upper vessel assembly which consists of the upper shell, intermediate shell, nozzles, and reactor vessel she ll flange. Both the upper and intermediate shells consist of three 120 degree segments formed from plate material and welded together to form cylindrical shells. Once the shells are welded, the upper she ll is welded to the reactor vessel shell flange. The intermediate shell is then welded to form the upper vessel assembly. Four inlet nozzles and two outlet

nozzles are then welded to complete the upper vessel assembly.

The lower vessel assembly consists of the lower shell and the bottom head. The lower shell is formed from three plates into 120 degree segments and welded t ogether to form a cylindrical shell. The bottom head is constructed of six peel segments and a dome sect ion, all formed from plate material. These are welded together to form a hemispherical head. The lower shell and bottom head are then welded

together to complete the lower vessel assembly.

The closure head is fabricated separatel

y. It is bolted to the reactor vessel only for hydrostatic testing.

(EC-1020, R307)

The closure head assembly consists of the closur e head forging, control element drive mechanism (CEDM) housings, and instrument nozzles. Penetra tions are then machined in the closure head for 87 control rod mechanisms, 10 instrumentation nozzles, and one vent pipe. Attachment of these complete the closure head assembly. The closure head is attached to the reactor vessel by 54 seven in. diameter studs which are threaded into the vessel flange and extend through the closure head flange. (EC-1020, R307)

Previous experience using the above procedures in fabricating other r eactor vessels is summarized in Subsection 5.3.3.

WSES-FSAR-UNIT-3 5.3-19 Revision 307 (07/13) 5.3.3.4 Inspection Requirements (EC-1020, R307) Inspection requirements of ASME Code,Section III, 1971 Edition including Summer 1971 Addenda and, for the replacement closure head,Section III, 1998 Edition through 2000 Addenda, are discussed in Subsection 5.3.1.3. (EC-1020, R307) 5.3.3.5 Shipment and Installation (EC-1020, R307) The reactor vessel is shipped by barge to the site mounted on the shipping skid used for installation. The

vessel is protected by closing all openings (including t he top of the vessel) with metal shipping covers and pressurizing with inert gas. The replacement closure head is shipped on a separate skid. During shipment, the environment within the replacement closure head is maintained clean and dry, and is protected from external humidity and atmosphere by its shipping skid, shrink wrap and the use of desiccants. Vessel surfaces and covers are spra yed with a strippable coating for protection against corrosion during shipping and installation. Prior to the welding of inter-connecting piping and installation

of insulation, the temporary protecti ve coating is removed by peeling. (EC-1020, R307) 5.3.3.6 Operating Conditions

Operating parameters are provided in Subsecti on 4.4.3. Design transient information is supplied in Subsection 3.9.1.1.

5.3.3.7 In-service Surveillance

5.3.3.7.1 Irradiated Ma terials Surveillance

This program is described in Subsection 5.3.1.6.

5.3.3-7.2 In-service Inspection

This program is described in Subsection 5.2.4.

SECTION 5.3: REFERENCES

1. Letter from A. E. Scherer (C-E) to Se cretary of the Commission (NRC), LD-75-655, September 26, 1975.

2. "C-E Procedure for Design, Fabrication, Inst allation and Inspection of Surveillance Specimen Holder Assemblies," Combustion Engineering Topical Report, CENPD-155P, September 1974.

(DRN 03-2059, R14) 3. 10 CFR 50.61, "Fracture Toughne ss Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243, dated December 19, 1995.

4. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fractur e Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No.243, dated December 19, 1995.

5. Regulatory Guide 1.99, Revision 2, "Radiation Em brittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

6. WCAP-16002, Revision 0, "Analysis of Capsule 263 from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation Surveillance Program," March 2003.

7. WCAP-16088, Revision 1, "Waterford Unit 3 R eactor Vessel Heatup and Cooldown Limit Curves for Normal Operation," September 2003. (DRN 03-2059, R14)

WSES-FSAR-UNIT-3 TABLE 5.3-1 (Sheet 1 of 5)DATA POINTS USED TO ESTABLISH C-E RTNDT SHIFT vs. FLUENCE AND PERCENT CUPPER SHELF DROP vs. (FLUENCE) 1/2 DESIGN CURVES FOR A533-B REACTOR VESSEL MATERIALS C v Post-PercentPlateThick t (t)1/2 C v InitialIrradiationC v UpperC vUpperDataLitTypeorness Cu p S(n/cm 2 (n/cm 2)1/2RT NDT (b) Upper ShelfUpper ShelfShelf DropShelf DropPointRef (h)MtlWeld(in.)(Wt%)(Wt%)(Wt%)x 10 19)(a) x 10 9 (°F)(ft-lb)(ft-lb)(ft-lb)(%)

_______________________________________________________________________________________________________________________________

________________________________________________321A533-B1Plate40.140.0090.0222.34.79120100 772323331A533-B1Plate80.140.0100.0232.34.79 95116 981815.5341A533-B1Plate8-1/80.190.0100.0171.74.12190101 8615 14.8 351A533-B1Plate6-3/80.090.0080.0150.21.41 0137137 0 0 361A533-B1Plate6-3/80.090.0080.0152.04.47 80137130 7 5.1 371A533-B1Plate6-3/80.090.0080.0152.04.47 90137 75 - -381A533-B2Plate6-3/80.090.0080.0150.52.23 35120 - - -391A533-B2Plate6-3/80.090.0080.0152.04.47 75120 7050 -

401A533-B1Plate7-1/20.120.0080.0151.74.12 7013610036 -

411A533-B1Plate7-1/20.110.0080.0191.74.12 85126121 5 3.9 421A533-B1Plate5-3/40.120.0080.0181.84.24 5014811533 22.3432A533-B1Plate80.090.0080.0140.52.23 0135135 0 0442A533-B1Plate80.090.0080.0142.44.89 8513512510 7.4 453A533-B1Plate6-1/40.090.0030.0142.55.0 60123128- -

481A533-B1Weld-7-1/20.220.0150.0111.74.1200109 634642.2S/A (d)491A533-B1Weld-5-3/40.190.0080.0141.64.0165 82 79 33.6E/S (e)a.Fluence: Neutron energies > I Mev, irradiation temperature - 550

°Fb.Based on NDTT measured at the C v 30 ft-lb levelC.The weld from which these specimens were taken (S/A) was back chipped and rewelded probably with a manual arc.The specimens were taken from different areas of the weld so the chemistries of the specimen could tend to vary greatly.d.S/A = Submerged arce.E/S = Electroslagf.Percent C v , upper shelf drop as reported in referenceg.Irradiation temperature - 530

°Fh.Refer to Table 5.3-2 WSES-FSAR-UNIT-3 TABLE 5.3-1 (Sheet 2 of 5)DATA POINTS USED TO ESTABLISH C-E RT NDT SHIFT vs. FLUENCE AND PERCENT C vUPPER SHELF DROP vs. (FLUENCE) 1/2 DESIGN-CURVES FOR A533-B REACTOR VESSEL MATERIALS C v Post-PercentPlateThick t (t)1/2 C v InitialIrradiationC v UpperC vUpperDataLitTypeorness Cu p S(n/cm 2 (n/cm 2)1/2RT NDT (b) Upper ShelfUpper ShelfShelf DropShelf DropPointRef (h)MtlWeld(in.)(Wt%)(Wt%)(Wt%)x 10 19)(a) x 10 9 (°F)(ft-lb)(ft-lb)(ft-lb)(%)

_______________________________________________________________________________________________________________________________

_______________________________________________

50(c)2A533-B1Weld-80.090.0100.0140.52.23 0 145 145 0 0S/A (d)51(c)2A533-B1Weld-80.090.0100.0142.44.89 90 145 1252013.8S/A (d)52(c)2A533-B1Weld-80.140.0100.0140.52.2105 105 703533.3S/A (d)53(c)2A533-B1Weld-80.140.0100.0142.44.8210 105 654038S/A (d)543A533-B1Weld-6-1/40.090.0020.0122.55.0100 88 781011.3E/S (e)584A533-B1Plate120.180.0080.0081.0-101 - - - -(surf.)594A533-B1Plate120.180.0080.0081.0-126 - - - -(1/2 T)604A533-B1Plate120.180.0080.0081.0- 70 - - - -(3/8 T)615A533-B1Plate120.140.0080.0164.52-7.0 80 120 1002016.6 5.59625A533-B1Plate120.140.0080.0163.64-6.32135>120~108 - -

4.24635A533-B1Plate120.140.0080.0161.18-3.6 85>120~110 - -

1.33645A533-B1Weld 3/40.22 0.0190.132.73-5.91256 115 ~605547.8(S/A)(d)4.25 WSES-FSAR-UNIT-3 TABLE 5.3-1 (Sheet 3 of 5)DATA POINTS USED TO ESTABLISH C-E RT NDT SHIFT vs. FLUENCE AND PERCENT C vUPPER SHELF DROP vs. (FLUENCE) 1/2 DESIGN CURVES FOR A533-B REACTOR VESSEL-MATERIALS C v Post-PercentPlateThick t (t)1/2 C v InitialIrradiationC v UpperC vUpperDataLitTypeorness Cu p S(n/cm 2 (n/cm 2)1/2RT NDT (b) Upper ShelfUpper ShelfShelf DropShelf DropPointRef (h)MtlWeld(in.)(Wt%)(Wt%)(Wt%)x 10 19)(a) x 10 9 (°F)(ft-lb)(ft-lb)(ft-lb)(%)

_______________________________________________________________________________________________________________________________

______________________________________________656A533-B1Plate60.030.0080.0082.8 - 65 - - - -666A533-B1Plate60.030.0080.0082.8 - 40 - - - -677A533-B1Plate120.180.0080.0080.47 - 70 - - - -

687A533-B1Plate120.180.0080.0080.94 3.06 95104 100 4 3.8 697A533-B1Plate120.180.0080.0081.05 -130 - -- -

708A533-B1Weld-120.230.0110.0082.5 5.0270125 705544S/A (d)718A533-B1Plate120.180.0080.0082.8 5.29200104 733129.8729A533-B1Plate60.130.0080.0082.8 5.29125135 1003525.9 739A533-B1Plate60.130.0080.0072.8 5.29140110 902013.18749A533-B1Plate60.030.0080.0083.1 5.56 70145 138 7 4.87510A533-B1Plate120.140.0080.0160.5 -- -(3/8 T)778A533-B1Plate120.140.0080.0162.7 5.19170122~1022014 (f)788A533-B1Plate120.140.0080.0162.6 5.09165 99~851414 (f)7911A533-B1Plate8-100.170.0090.0152.1 4.58145115932219.1 8011A533-B1Plate8-100.240.0080.0113.7 6.08165110842623.6 8111A533-B1Weld-8-100.360.0150.0123.4 5.83315107565147.6(S/A)(d)

WSES-FSAR-UNIT-3 TABLE 5.3-1 (Sheet 4 of 5)DATA POINTS USED TO ESTABLISH C-E RT NDT SHIFT vs. FLUENCE AND PERCENT C vUPPER-SHELF DROP vs. (FLUENCE) 1/2 DESIGN-CURVES FOR A533-B REACTOR VESSEL MATERIALS C v Post-PercentPlateThick t (t)1/2 C v InitialIrradiationC v UpperC vUpperDataLitTypeorness Cu p S(n/cm 2 (n/cm 2)1/2RT NDT (b) Upper ShelfUpper ShelfShelf DropShelf DropPointRef (h)MtlWeld(in.)(Wt%)(Wt%)(Wt%)x 10 19)(a) x 10 9 (°F)(ft-lb)(ft-lb)(ft-lb)(%)

_______________________________________________________________________________________________________________________________

__________________________________________________8211A533-B1Weld8-100.200.0160.013 3.4 5.83 95129 983124.03S/A (d)8311A533-B1Plate8-100.170.0090.015 6.7 8.18210115 942118.28411A533-B1Plate8-100.240.0080.011 6.1 7.8185110 862421.8 8511A533-B1Weld-8-100.360.0150.012 6.1 7.8350107 565147.6S/A (d)8611A533-B1Weld-8-100.200.0160.013 6.1 7.8125129 983124.03S/A (d)8711A533-B1Plate8-100.090.0090.017 4.4 6.63 35104107 - -8811A533-B1Plate8-100.090.0090.017 5.7 7.54 55104111 - -

8911A533-B1Plate8-100.090.0110.018 4.0 6.32 45119130 - -9011A533-B1Plate8-100.090.0110.018 5.4 7.34 85119130 - -9111A533-B1Weld-8-100.070.0100.010 4.9 7.0 35157155 2 1.27S/A (d)9211A533-B1Weld-8-100.070.0100.010 5.07.07 50157155 2 1.27S/A (d)9311A533-B1Weld-8-100.050.0040.004 4.9 7.020144147 - -S/A (d)9412A533-B1Plate60.030.0080.00815.8 (g)12.56260145 9946 31.7(same material as Pt. 74)

WSES-FSAR-UNIT-3 TABLE 5.3-1 (Sheet 5 of 5)DATA POINTS USED TO ESTABLISH C-E RT NDT SHIFT vs. FLUENCE AND PERCENT C vUPPER SHELF DROP vs. (FLUENCE) 1/2 DESIGN CURVES FOR A533-B REACTOR VESSEL MATERIALS C v Post-PercentPlateThick t (t)1/2 C v InitialIrradiationC v UpperC vUpperDataLitTypeorness Cu p S(n/cm 2 (n/cm 2)1/2RT NDT (b) Upper ShelfUpper ShelfShelf DropShelf DropPointRef (h)MtlWeld(in.)(Wt%)(Wt%)(Wt%)x 10 19)(a) x 10 9 (°F)(ft-lb)(ft-lb)(ft-lb)(%)

_______________________________________________________________________________________________________________________________

____________________________________________9512A533-B1Plate60.030.0080.00824.1(g)15.52335145707551.7(same material as Pt. 74)9612A533-B1Plate60.030.008 0.00827.816.67295145994631.7(same material as Pt. 74)9713A533-B1Plate60.130.0080.0070.095 - 0 - - - -(same material as Pt. 73)9813A533-B1Weld8-100.360.0150.0180.095 - 55 - - - -(same material as Pt. 85)A14A302-BWeld10-1/20.22 - -0.2 - 95 - - - -B15A302-BWeld 60.270.0140.0120.7 -140 - - - -C16A508-2Weld6-1/20.230.0120.0160.49 -140 - - - -D 1A533-2Weld40.270.0160.0151.4 -205 - - - -

WSES-FSAR-UNIT-3 TABLE 5.3-2 (Sheet 1 of 2)LITERATURE REFERENCES FOR TABLE 5.3-11.Hawthorne, J. R. and Potapovs, U., "Initial Assessment of Notch Ductility Behavior ofA533 Pressure Vessel Steel with Neutron Irradiation," NRL Report 6772, Naval Research Laboratory,Washington, D.C., December 30, 1971.2.Steele, L. E., et al., "Irradiation Effects on Reactor Structural Materials," NRLMemo-1937, NavalResearch Laboratory, Washington, D.C., November 1968.3.Steele, L. E., et al., "Irradiation Effects on Reactor Structural Materials," NRLMemo-2027, NavalResearch Laboratory, Washington, D.C., August 1969.4.Berggren, R. G., et al., "Radiation Studies on HSST Plate and Welds," Paper given atORNL for HSST Program Information Meeting, April 1, 1970.5.Mager, T. R. and Thomas, F. O., "Heavy Section Steel Technical Report No. 5,(November 1969), Evaluation by Linear Elastic Fracture Mechanics of Radiation Damage to Pressure Vessel Steels," October 1969.6.Hawthorne, J. R., "Improved Radiation Embrittlement Resistance in CommerciallyProduced A533-B Plate and Weld Metal," given at ORNL for HSST Program Information Meeting, April 1, 1970.7.Witt, F. J., Program Director, Heavy Section Steel Technology Program, Semi-AnnualProgress Report for Period Ending February 28, 1969, ORNL 4463, January 1970.8.Hawthorne, J. R., "Post-Irradiation Dynamic Tear and Charpy-V Performance of 12 In.Thick A533-B Steel Plates and Weld Metal," Nuclear Engineering and Design 17, pp 116-130, 1971.9.Steele, L. E., et al., "Irradiation Effects on Reactor Structural Materials," NRLMemo-Report 2088, Naval Research Laboratory, Washington, D.C., February 1970.10.Witt, F. J., Program Director, Heavy Section Steel Technology Program Semi-AnnualReport for Period Ending February 28, 1970, ORNL-4590, October 1970.11.Hawthorne, J. R. (NRL), Koziol, J. J., and Groeschel, R. C. (Combustion Engineering),"Evaluation of Commercial Production A533-B Plates and Weld Deposits Tailored for Improved Radiation Embrittlement Resistance," ASTM STP 570, American Society for Testing and Materials, January 1976.

WSES-FSAR-UNIT-3 TABLE 5.3-2 (Sheet 2 of 2)LITERATURE REFERENCES FOR TABLE 5.3-112.Steele, L. E., Editor, "Irradiation Effects on Reactor Structural Materials," 1August 1974 - 31 January 1975, NRL Memo-3010, Naval Research Laboratory, Washington, D.C.,February 1975.13.Naval Research Laboratory, unpublished data.

14.Ireland D. R., and Scotti, V. G., "Final Report on Examination and Evaluation ofCapsule A for the Connecticut Yankee Reactor Pressure Vessel Surveillance Program," Battelle (Columbus) Memorial Institute, Docket No. 50-213, 1970.15.Serpan, C. Z., Jr. and Watson, H. F., "Mechanical Properties and NeutronSpectra Analyses of the Big Rock Point Reactor Pressure Vessel," Nuclear Engineering and Design, Vol.11, pp 363-415, 1970.16.Mager, T. R., et al., "Analysis of Capsule V from the Rochester Gas and Electric R.E. Ginna Unit No. 1, Reactor Vessel Radiation Surveillance Program," FP-RA-1, Westinghouse ElectricCorporation Nuclear Energy Systems, April 1, 1973.17.Smidt, F. A., Jr. and Sprague, J. A., "Property Changes Resulting from ImpurityDefect Interaction in Iron and Pressure Vessel Steel Alloys," Effects of Radiation on Substructure andMechanical Properties of Metals and Alloys, ASTM STP 529, pp 78-91, 1973.

WSES-FSAR-UNIT-3TABLE 5.3-3

SUMMARY

OF SURVEILLANCE MATERIALS TESTING30 ft-lb50 ft-lb35 Mils Lat.RT Yield C v Upper ShelfFit (a)Fit (b)Exp. Fit (b)NDTTRT NDTStrength (ksi)Material and Code(ft-lb)(°F)(°F)(°F)(°F)(°F)Static Dynamic

__________________________________________________________________________________________________________________________Base Metal Plate136- 30 18 20-2069 97M-1004-2 (WR)Base Metal Plate169.5 0 48 36 0--70103M-1004-2 (RW)Weld Metal146- 76-46-46-80-8085113M-1004-1/M-1004-3HAZ Metal163.5-106-72-76-50-5070113M-1004-2SRM HSST130 28 70 48 0------Plate O1MY-RW

______________________________________________________(a)Determined from average impact energy curve.(b)Determined from lower bound curve.

WSES-FSAR-UNIT-3 TABLE 5.3-4 Revision 15 (03/07)

TOTAL QUANTITY OF SPECIMENS

Type of Base Weld Specimen Orientation Metal Metal HAZ SRM (a) Totals ________________________________________________________________________________________________

Drop Weight Longitudinal 12 ____ ____ ____ 12 (DRN 06-897, R15)

(DRN 06-897, R15)

Transverse 12 12 12 ---- 36

Charpy Impact Longitudinal 78 --- --- 39 117 (DRN 06-897, R15) (DRN 06-897, R15)

Transverse 102 102 102 --- 306

Tensile Longitudinal 18 --- --- --- 18 (DRN 06-897, R15)

(DRN 06-897, R15)

Transverse 36 36 36 --- 108 (DRN 06-897, R15)

Totals 258 150 150 39 597 (DRN 06-897, R15)

________________________________

(a) Standard Reference Material WSES-FSAR-UNIT-3 TABLE 5.3-5 Revision 15 (03/07)

TYPE AND QUANTITY OF-SPECIMENS FOR UNIRRADIATED TESTS Quantity of Specimens

Type of Base Weld Specimen Orientation Metal Metal HAZ SRM (a) Totals ______________________________________________________________________________________________________

Drop Weight Longitudinal 12 --- --- --- 12 (DRN 06-897, R15)

(DRN 06-897, R15)

Transverse 12 12 12 --- 36

Charpy Impact Longitudinal 30 --- --- 15 45 (DRN 06-897, R15)

(DRN 06-897, R15)

Transverse 30 30 30 --- 90

Tensile Longitudinal 18 --- --- --- 18 (DRN 06-897, R15)

(DRN 06-897, R15)

Transverse 18 18 18 --- 54 (DRN 06-897, R15)

Totals 120 60 60 15 255 (DRN 06-897, R15)

(a) Standard Reference Material characterized by HSST Program, specimens provided only for correlation with characterization tests.

WSES-FSAR-UNIT-3 TABLE 5.3-6 Revision 15 (03/07)

TYPE AND QUANTITY OF SPECIMENS FOR IRRADIATION EXPOSURE AND IRRADIATED TESTS

Quantity of Specimens

Type of Base Weld Specimen Orientation Metal Metal HAZ SRM (a) Totals (DRN 06-897, R15)

Charpy Impact Longitudinal 48 -- -- 24 -- 72 (DRN 06-897, R15)

Transverse 72 72 72 -- 216

Tensile Longitudinal -- -- -- -- -- (DRN 06-897, R15)

(DRN 06-897, R15)

Transverse 18 18 18 -- 54 (DRN 06-897, R15)

Totals 138 90 90 24 342 (DRN 06-897, R15)

__________________________________

a. Standard Reference Material WSES-FSAR-UNIT-3TABLE 5.3-7CANDIDATE MATERIALS FOR NEUTRON THRESHOLD DETECTORSMaterialReactionThreshold Energy (MeV)Half-LifeUraniumU 238 (n, f) Sr 900.728 yearsSulfurS 32 (n, p) P 322.914.3 daysIronFe 54(n, p) Mn 544.0314 daysNickelNi 58 (n, p) Co 585.071 daysCopperCu 63(n,) Co 607.05.3 yearsTitaniumTi 46(n, p) Sc 468.084 daysCobaltCo 59(n,) Co 60Thermal5.3 years WSES-FSAR-UNIT-3TABLE 5.3-8COMPOSITION AND MELTING POINTS OFCANDIDATE MATERIALS FOR TEMPERATURE MONITORSCompositionMelting Temperature(wt%)(°F)80 Au, 20 Sn536 90.0 Pb, 5.0 Sn, 5.0 Ag558 97.5 Pb, 2.5 Ag580 97.5 Pb, 0.75 Sn, 1.75 Ag590 WSES-FSAR-UNIT-3 TABLE 5.3-9 Revision 13-A (09/04)

TYPE AND QUANTITY OF SPECIMENS CONTAINED IN EACH IRRADIATION CAPSULE ASSEMBLY

(DRN 04-1049, R13-A)

Base Metal Weld Metal HAZ Reference Total Specimens Capsule Impact Location L (b) T(c) Tensile Impact Tensile Impact Tensile Impact (a) Impact Tensile Vessel-97 12 12 3 12 3 12 3 -- 48 9 Vessel-104 -- 12 3 12 3 12 3 12 48 9 Vessel-284 12 12 3 12 3 12 3 -- 48 9 Vessel-263 -- 12 3 12 3 12 3 12 48 9 Vessel-277 12 12 3 12 3 12 3 -- 48 9 Vessel-83 12 12 3 12 3 12 3 -- 48 9 Totals 48 72 18 72 18 72 18 24 288 54 (DRN 04-1049, R13-A)

a. Reference material correlation monitors b. L = Longitudinal c. T = Tranverse WSES-FSAR-UNIT-3 TABLE 5.3-10 Revision 14 (12/05)

CAPSULE ASSEMBLY REMOVAL SCHEDULE(DRN 03-2059, R14)

Capsule No./ID Azimuthal Location (deg)Lead Factor Removal Time (EFPY)*Target Fluence (n/cm 2)1/W-8383 1.18 26 2.47 x 10 192/W-9797 1.18 4.44** 6.47 x 10 18**3/W-104104 0.83 Standby --

4/W-263263 1.18 13.83** 1.45** x 10 19**5/W-277277 1.18 Standby --

6/W-284 284 0.83 Standby -- (DRN 03-2059, R14) *EFPY - Effective Full Power Years, withdrawal time may be modified to coincide with those refueling outages or plant shutdowns most closely approaching the withdrawal schedule.

• • - Values represent actual data on removed capsule (DRN 04-1049, R13-A)NOTE: As required by 10CFR Appendix H,Section III.B.3, submit a proposed withdrawal schedule with technical justification as specified in 10CFR50.4 for NRC approval prior to implementation. (DRN 04-1049, R13-A)

WSES-FSAR-UNIT-3TABLE 5.3-11WATERFORD UNIT 3 REACTOR VESSEL CLOSURE STUDS DATAUltimate Tensile StrengthFracture ToughnessPieceDrawingCodeHeatTest TempStrengthTest Temp Charpy EnergyMils LateralNumberNumberNumberNo.Material (°F) (KSI) (°F) (ft-lbs.)Expansion

_______________________________________________________________________________________________________________________________

__________________________________________98E74170-161-03M-1028-180751SA-540 Grade B24 701681049-48-4728-27-25 98-1E74170-161-03M-1028-180751SA-540 Grade B24 70162.51050-51-5125-28-29 69E74170-161-03M-1028-180751SA-540 Grade B24 701631050-48-4926-25-29 69-1E74170-161-03M-1028-180751SA-540 Grade B24 701571057-56-5738-35-3770E74170-161-03M-1028-180751SA-540 Grade B24 701561056-54-5330-34-3470-1E74170-161-03M-1028-180751SA-540 Grade B24 701641050-50-4834-29-31 72E74170-161-03M-1028-180751SA-540 Grade B24 701581056-56-5538-34-32 72-1E74170-161-03M-1028-180751SA-540 Grade B24 701591050-50-5126-25-28 74E74170-161-03M-1028-180751SA-540 Grade B24 701571050-52-5031-32-27 74-1E74170-161-03M-1028-180751SA-540 Grade B24 701581050-51-5025-25-30 76E74170-161-03M-1028-180751SA-540 Grade B24 701541054-54-5327-31-33 76-1E74170-161-03M-1028-180751SA-540 Grade B24 701611050-50-5125-30-29 W3SES-FSAR-UNIT-3TABLE 5.3-12WATERFORD UNIT 3 REACTOR VESSEL NUTS AND WASHERS DATA UltimateTensile Strength Fracture Toughness PieceDrawingCodeHeatTest TempStrengthTest Temp.Charpy EnergyMils Lateral NumberNumberNumberNo. Material (F)(KSI) (F) (Ft-lbs) Expansion -41E74170-161-03M-1029-118551SA-540 Grade B23 70163.5 10 38-40-38 19-21-1841-1E74170-161-03M-1029-118551SA-540 Grade B23 70164.5 10 42-40-38 20-22-18 48E74170-161-03M-1029-118551SA-540 Grade B23 70170.0 10 37-39-38 18-19-21 48-1E74170-161-03M-1029-118551SA-540 Grade B23 70165.0 10 43-45-42 25-27-24 WSES-FSAR-UNIT-3 TABLE 5.3-13 Revision 7 (10/94)WATERFORD UNIT 3REACTOR VESSEL MATERIALSProductMaterialDrop WeightInitial d Chemical Content %Form IdentificationNDTT (°F)RT NDT (°F)NickelCopperPhosphorusPlateM-1003-1-30-30 0.710.020.004PlateM-1003-2-50-50 0.670.020.006 PlateM-1003-3-50-42 0.700.020.007PlateM-1004-1-50-15 0.620.030.006PlateM-1004-2-20 22 0.580.030.005 PlateM-1004-3-50-10 0.620.030.007 Weld101-124 A,B,& C a-60-60 0.960.020.010Weld101-142 A,B,& C b-80-80< 0.200.030.007Weld101-171 c-70 to-80-70 to-80 0.160.050.008a.Intermediate shell course longitudinal seam weldb.Lower shell course longitudinal seam weld c.Intermediate - lower shell girth weld d.Plate RT NDT determined using Branch Technical Position MTEB 5-2; weld RT NDT determined in accordance with ASME Code,Section III, NB-2300 (DRN 03-2059, R14) W3SES-FSAR-UNIT-3 TABLE 5.3-14 Revision 14 (12/05)

CALCULATION OF THE WATERFORD UNIT 3 RTPTS VALUES FOR 32 EFPY Material Chemistry Factor Basis CF (F) Fluence (x10 19 n/cm 2) RTNDT (a) (F) PTS (b) (F) i (F) (F) Margin (F) RTPTS (c) Intermediate Shell Plate M-1003-1 Table 2 20 2.48 -30 24.9 0 12.4 24.9 20 Intermediate Shell Plate M-1003-2 Table 2 20 2.48 -50 24.9 0 12.4 24.9 0 Intermediate Shell Plate M-1003-3 Table 2 20 2.48 -42 24.9 0 12.4 24.9 8 Lower Shell Plate M-1004-1 Table 2 20 2.47 -15 24.9 0 12.4 24.9 35 Lower Shell Plate M-1004-2 Surveillance Data 12.4 2.47 22 15.4 0 7.7 15.4 53 Lower Shell Plate M-1004-3 Table 2 20 2.47 -10 24.9 0 12.4 24.9 40 Intermediate Shell Longitudinal Weld Seams 101-124 A,B,C Table 1 27 2.48 -60 33.6 0 16.8 33.6 7 Lower Shell Longitudinal Weld Seams 101-142 A,B,C Table 1 35 2.47 -80 43.5 0 21.8 43.5 7 Intermediate to Lower Shell Girth Weld Seam 101-171 Surveillance Data 16.2 2.47 -70 20.1 0 10.1 20.1 -30 Notes: (a) Initial reference temperature (RTNDT) values are measured. Thus, i equal to 0F. PTS = CF

• FF (c) RTPTS = Initial RTNDTPTS + Margin (F) (DRN 03-2059, R14)