ML20081E828
| ML20081E828 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/31/1991 |
| From: | Albertin L, Chicots J, Lloyd T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20081E823 | List: |
| References | |
| WCAP-12867, NUDOCS 9105020204 | |
| Download: ML20081E828 (176) | |
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WESTINGHOUSE CLASS 3 l l WCAP-12867 l ANALYSIS OF CAPSULE X FROM THE SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM J. M. Chicots T. M. Lloyd
, L. Albertin March 1991 Work Performed Under Shop Order VSGP-106 Prepared by Westinghouse Electric Corporation for the South Carolina Electric and Gas Company h
Approved by: ' bN T. A. Meyer, Mknager Structural Reliability and Plant Life Optimization h L WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 e 1991 Westinghouse Electric Corp.
PREFACE I 1
. l This report has been technically reviewed and verified.
Reviewer Sections 1 through 5, 7, and 8 E. Terek Ib - wM-Section 6 S.L.AndersondebbMablLtap)- Appendix B N. K. Ray 'A~\( ) w-c:txM/ y 4 e e i
_- __ _. . .. . . . _ _ .._ _ _ . . _ _ _ . _ . . ~ _ . _ _ . - . _ _ _ _ _ . . TABLE OF CONTENTS Section li.tle P_iLqe 1.0
SUMMARY
OF RESULTS 1-1
2.0 INTRODUCTION
2-1
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM Capsule X 5-1 5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-4
.- 5.3 Tension Test Results 5-5
! 5.4 Compact Tension Tests 5-6 6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-7 7.0 SVRVEILLANCE CAPSULE REMOVAL SCHEOULE 7-1
8.0 REFERENCES
8-1
-APPENDIX A - LOAD-TIME RECORDS
' APPENDIX B - HEAT AND C00LDOWN LIMIT CVRVES FOR NORMAL OPERATION a
11
LIST OF ILLUSTRATIONS i Fiaure Title bagg 4-1 Arrangement of Surveillance Capsules in the V. C. Summer 4-5 Unit 1 Reactor Vessel 4-2 Capsule X Diagram Showing Location of Specimens, Thermal 4-6 Monitors and Dosimeters 5-1 Charpy V-notch Impact Properties for V. C. Summer Unit 1 5-14 Reactor Vessel Shell Plate A9154-1 (Longitudinal Orientation) 5-2 Charpy V-notch Impact Properties for V. C. Summer Unit 1 5-15 Reactor Vessel Shell Plate A9154-1 (Transverse Orientation) 5-3 Charpy V-notch Impact Properties for V. C. Summer Unit 1 5-16 Reactor Vessel Weld Metal 5-4 Charpy V-notch Impact Properties for V. C. Summer Unit 1 5-17 Reactor Vessel Weld Heat Affected Zone Metal 5-5 Charpy Impact Specimen Fractura Surfaces for V. C. Summer 5-18 Unit 1 Reac+.or Vessel Shell Plate A9154-1 (Longitudinal Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for V. C. Summer 5-19 Unit 1 Reactor Vessel Shell Plate A9154-1 (Transverse Orientation) 5-7 Charpy Impact Specimen Fracture Surfaces for V. C. Summer 5-20 Unit 1 Reactor Vessel Weld Metal 4 iii
i LIST OF ILLUSTRATIONS (Cont) Fiaure Title Elag 5-8 Charpy Impact Specimen Fracture Surfaces for V. C. Summer 5-21 Unit 1 Reactor Vessel Weld Heat Affected Zone (HA2) Metal 5-9 Tensile Properties for V. C. Summer Unit 1 Reactor Vessel 5-22 ! Shell Plate A9154-1 (longitudinal Orientation) 5-10 Tensile Properties for V. C. Summer Unit 1 Reactor Vessel 5-23 Shell Plate A9154-1 (Transverse Orientation) 5-11 Tensile Properties for V. C. Summer Unit 1 Reactor Vessel 5-24 Weld Metal 5-12 Fractured Tensile Specimens from V. C. Summer Unit 1 Reactor 5-25 - Vessel Shell Plate A9154-1 (longitudinal Orientation) 5-13 Fractured Tensile Specimens from V. C. Summer Unit 1 Reactor 5-26 Vessel Shell Plate A9154-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens from V.-C. Summer Unit 1 Reactor 5-27 Vessel Weld Metal 5-15 Engineering Stress-Strain Curves for Shell Plate A9154-1 5-28 Tension Specimens CLIO and CLil 5-16 Engineering Stress-Strain Curves for Shell Plate A9154-1 5-29 Tension Specimens CL12 and CTIO 5-17 Engineering Stress-Strain Curves for Shell Plate A9154-1 5-30 Tension Specimens CTll and CT12 4 iv l
LIST OF ILLUSTRATIONS (Cont) i I
~
Fiaure Title ELqg 5 Engineering Stress-Strain Curves for Weld Tension Specimens 5-31 CW10 and CWil 5-19 Engineering Stress-Strain Curve for Weld Tension Specimen 5-32 CW12 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 6-2 Core Power Distributions Used in Transport Calculations 6-14 for V, C. Summer Unit 1 l
=
V
LIST OF TABLES Table Title EAqn 4-1 Chemical Composition of the V. C. Summer Unit 1 4-3 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the V. C. Summer Unit 1 Reactor 4-4 Vessel Surveillance Materials i 5-1 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 5-7 Shell Plate A9154-1 Irradiated et 550*F, Fluence 2.46 x 10l9 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 5-8 Reactor Vessel Weld Metal and HAZ Metal Irradiated at-550*F, Fluence 2.46 x 1019 n/cm2 (E > 1.0 MeV) .! 5-3 Instrumented Charpy Impact Test Results for V. C. Summer 5-9 [ Unit 1 Shell Plate A9154-1 Irradiated at 550*F, Fluence 2.46 x 10l9 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results for V. C. Summer 5-10 Unit 1 Weld Metal and HAZ Metal Irradiated at 550'F, L Fluence 2.46 x 1019 -n/cm2 (E >'l.0 MeV) 5-5 Effect of 550*F 1rradiation at 2.46 x 1019 n/cm 2 5-11 (E > 1.0 MeV) on Notch Toughness Properties of V. C. Summer-Unit 1 Reactor Vessel Surveillance Capsule X Materials 1 vi
LIST OF TABLES (Cont) 1 Table lLtle h_qe l
- \
l 5-6 Comparison of V. C Summer Unit 1 Surveillance Material 5-12 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 l Predictions 5 Tensile Properties for V. C, Summer Unit 1 Reactor Vessel 5-13 Surveillance Capsule X Material Irradiated at 550*F to 2.46 x 1019 n/cm2 (E > 1,0 MeV) 6-1 Calculated Fast Neutron Exposure Parameters at the 6-15 Surveillance Capsule Center
, 6-2 Calculated Fast Neutron Exposure Parameters at the 6-16 Pressure Vessel Clad / Base Metal Interface 6 Relative Radial Distributions of Neutron Flux 6-17 (E > 1,0 MeV) within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-18 (E > 1,0 MeV) within the Pressure Vessel Wall 6-5 Relative Radial Distributions of Iron Displacement Rate 6-19 (dpa) within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6 6-7 Irradiation History of Neutron Sensors Contained in 6-21 Capsule X vii
LIST OF TABLES (Cont) lablg Title Pace 6-8 Measured Sensor Activities and Reaction Rates 6-25 6-9 Summary of Neutron Oosimetry Results 6-27 6-10 Comparison of Measured and Ferret Calculated Reactio'. 6-28 Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-29 Capsule Center 6-12 Comparison of Calculated and Measured Exposure Levels 6-30 for Capsule X 6-13 Neutron Exposure Projections at Key Locations on the 6-31 Pressure Vessel Clad / Base Metal Interface for V. C. Summer Unit 1 6-14 Neutron Exposure Values for use in the Generation of 6-32 Heatup/Cooldown Curves 6-15 Updated Lead Factors for V. C. Summer Unit 1 Surveillance 6-33 Capsules O viii
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule ! X, the third capsule to be removed from the South Carolina Electric and Gas l Company V. C. Summer Unit I reactor pressure vessel, led to the following l conclusions: l o The capsule received an average fast neutron fluence (E > 1.0 MeV) of I 2.46 x 1019 n/cm 2 after 5.03 EFPY of plant operation, - I o Irradiation of the reactor vessel intermediate shell plate A9154-1 ! Charpy specimens to 2.46 x 10 19 n/cm2 (E > 1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 50'F, for specimens oriented parallel to the major working direction (longitudinal orientation).
- l. o Irradiation of the reactor vessel intermediate shell plate A9154-1 l
Charpy specimens to 2.46 x 10 19 n/cm2 (E > 1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 35 and 30'F, respectively, for specimens oriented normal to the major working direction (transverse orientation). o The weld metal Charpy specimens irradiated to 2.46 x 10 19 n/cm 2 . (E > 1.0 MeV) resulted in both a 30 and 50 ft-lb transition temperature increase of 35'F. o Irradiation of the reactor vessel weld HAZ metal Charpy specimens to 2.46 x 10 l9 n/cm2 (E > 1.0 MeV) resulted in a 30 and 50 ft-lb transition temperature increase of 45'F. l l-1 i
l SECTION
2.0 INTRODUCTION
This report presents the results of the examination of Capsule X, the third capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the South Carolina '. Electric and Gas Company V. C. Summer Unit I reactor pressure vessel materials under actual operating conditions. The surveillance program for the V. C. Summer Unit I reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented by Davidson and Yanichko (1]. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73,
" Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule X from the reactor and its shipment to the Westinghouse Science and Technology Center where the postirradiation mechanical +esting of the Charpy V-notch impact and tensile surveillance specimens were performed at the remote metallographic facility.
This report summarizes the testing of and the postirradiation data obtained from surveillance Capsule X removed from the V. C. Summer Unit I reactor vessel and discusses the analysis of these data. e 2-1
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring :afety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 533 Grade B Class 1 plate (base material of the V. C. Summer Unit I reactor pressure vessel beltline region) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation. A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile failure," !. . Appendix G to Section til of the ASME Boiler and Pressure Vessel Code (6), The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)* l RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208I73)' or the temperature 60'f less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from
'Charpy specimens oriented normal (transverse) to the major working direction of the material. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined
[ using these allowable stress intensity factors. 3-1
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure-vessel steel can be monitored by a reactor surveillance program such as the V. C. Summer Unit 1 Reactor Vessel Radiation Surveillance , Program,Ill ni which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the material to the KIR curse and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
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SECTION
4.0 DESCRIPTION
OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the V. C. Summer Unit I reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron shield pads and the vessel wall as shown in figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. Capsule X was removed af ter 5.03 effective full power years of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact tension (CT) and bend bar specimens (figure 4-2) from the intermediate shell plate A9154-1, Charpy V-notch, tensile and 1/2 T CT specimens (figure 4-2) from weld metal representative of the longitudinal and intermediate to lower shell beltline weld seams of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) material. All heat-affected zone specimens were obtained from wuhin the HAZ of plate A9154-1 of the representative weld. The chemical composition and heat treatment of the surveillance material is presented in Tables-4-1 and 4-2, respectively. All test specimens were machined from the 1/4 thickness location of the plate. Test specimens represent material Sken at least one plate thickness from the quenched end of the plate. Base metal Charpy V-notch impact and tension specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation) and also normal to the major working direction-(transverse orientation). Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding direction. The Compact Tension (CT) test specimens in Capsule X were machined such that the simulated crack in the specimen will propagate normal and parallel- to the major working direction
~
for the plate specimen and parallel to the weld direction. 4-1
Capsule X contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (g238) were contained in the capsule. Thermal . monitors made from the two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. The composition of the two alloys and their melting points are as follows: 2.5% Ag, 97.5% Pb Melting Point: 579'F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point: 590*F (310'C) The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule X are shown in Figure 4-2. l l l l l 1 4-2 L. l L___________-____--_____
TABLE 4-1 CHEMICAL COMPOSITION OF THE V. C. SUMMER UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Intermediate Shell Weld Metal (c) Plate A9154-1 Lukens Steel Co. Element Lukens Steel Co. Analy111 Analysis C 0.22 0.085 S 0.015 0.012/0.007(b) N 0.0076 0.015 Co 0.010 0.016/0.01(b) Cu 0.10 0.05/0.04(b) Si 0.24 0.48/0.42(b) Ho 0.49 0.49/0.46(b) Ni 0.51 0.91/0.95(b) Mn 1.30 1.32/1.50(b) Cr 0.08 0.14/0,12(b) -. V 0.001(a) 0.005 P 0.009 0.013/0.009(b) Sn 0.007 0.0047 A1 0.024 0.007/0.03(b) B 0.0004 0.0005 Ti 0.0002 0.001 Pb' <0.005 0.0206 Zr 0.001 0.001 As 0.006 0.006 W <0.01 0.01 (a) Westinghouse Analysis (b) Analysis performed on irradiated weld specimen CW14.
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(c)- Surveillance weld was made of the same RAC01 NMM wire heat #4P4784 and Linde 124 Flux Lot No. 3930 as the beltline welds of the reactor vessel. 4-3
l~ TABLE 4-2 HEAT TREATHENT 0F THE V. C. SUMMER UNIT 1 REACTOR VESSEL SVRVEILLANCE MATERIALS t l Material Temoerature (*F) Time (Hr) Coolant L j Shell Plate 1550/1650 1/2 hr/in., min. Water quenched i l A9154-1 1225 +/- 25 1/2 hr/in., min. Air cooled i 1150 +/- 25 43 furnaced cooled I to 600*f I i l l Weldment 1150 +/- 25 12 Furnaced cooled 1 I S t l l l l i l S 1 4-4 l l
a U O' REACTOR VESSEL Z CORE BARREL NEUTRON PAQ l Y 1 4,. 17' 1
' 90' 270'
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w4 4 Nw l l I fl 1e0' Figure 4-1. Arrangement of Surveillance Capsules in the V C Summer Unit 1 Reactor Vessel 4-5
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Aperture Cant i D hb ~ figure 4-2 Capsule X Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters 4-6
SECTION 5.0 TESTING OF SPECIMENS FROM CAPSULE X 5.1 Overvieg The post-irradiation mechanical testing of the Charpy V-notch ar.d tensile specimens was performed at the Westinghouse Science and Technology Center with consultation by Westinghouse Power Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and HI43, ASTM Specification E185-82[8], and Westinghouse Procedure MHL 8402, Revision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revisian 1. Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9234Ill. No discrepancies were found.
- Examination of the two low-melting point 304*C (579'F) and 310'C (590*F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304'C (579'F).
The Charpy impact tests were performed per ASTM Specification E23 8 sol and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology Model 500 instrumentation system. With this system, load-time and energy-timo signals can be recorded in addition to the standard measurement of Charpy energy (ED ). From the load-time curve, the load of general yielding (Pgy), the time to general yielding (tcy), the maximum load (Pg), and the time to maximum load (tg) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated is identified as the arrest load (PA )- 5-1
The energy at maximum load (Eg) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E p
) is th- difference between the total energy to fracture (E D) and the energy at mnimum load. ,
The yield stress (cy) is calculated from the three-point bend formula, having the following expression: oy Pgy L (1) B(W-a)2C where the constant C is dependent on the notch flank angle (p), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending). In three-point bending a Charpy specimen in which p - 45' and p = 0.010", Equation 1 is valid with C = 1.21. Therefore (for L = 4W), oy - Pgy L =WPgy W (2) j B(W-a)2(1.21) B(W-a)2 For the Charpy specimen, B - 0.394 in., W - 0.394 in., and a = 0.079 in. Equation 2 then reduces to ay - 33.3 x Pgy, (3) where oy_is in units of psi and Pay is in units of lbs. The flow stress is calculated from the average of the yield and maximum loads, also using the three-point bend formula. Percent shear was determined from post-fracture photographs usirtg the ratio-of-areas methods in compliance with ASTM Specification A370-89[10], The lateral expansion was measured using a dial gage rig similar to that shown in the same specification. l 5-2
Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-89Illl and E21-79(1988)ll23, and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of inconel.718 hardened to HRC45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test. Deflection measurements were made with a linear variable displacement I transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-85Il3}. Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature. l
- l. Chromol-alumel thermocouples were inserted in shallow holes in the center and !
each end of the gage section of a dummy specimen and in each grip. In the test I configuration, with a slight load on the specimen, a plot of specimen j temperature versus upper ano lower grip and controller temperatures was developed over the range of room temperature tc 550*F (288'C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtained desired specimen
-temperatures. Experiments indicated that this method is accurate to 12*F.
The-yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the
. original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using -the final diameter measurement.
5-3
5.2 Charov V-Notch Imoact Test-Results The-results of Charpy V-notch impact tests performed on the various materials
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contained in Capsule X irradiated at 550*F- to 2.46 x 1019 n/cm2 (E > 1.0 MeV) are presented in Tables 5-1 through 5-4 and are compared with . unirradiated resultsill as shown in Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule X materials are summarized in Table 5-5. Irradiation of the reactor vessel intermediate shell plate A9154-1 Charpy specimens to 2.46 x 10 19 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-1) resulted in : 30 ft-lb transition temperature increase of 50'F and a 50 ft-lb transition temperature increase of 50*F for specimens oriented perpendicular to the major working direction (longitudinal orientation). This resulted in a 30 ft-lb transition temperature of 25'F and a 50 ft-lb transition temperature of 50*F for specimens oriented perpendicular to the major working direction (longitudinal orientation). , The average upper shelf energy (VSE) of the intermediate shell plate A9154-1 . Charpy specimens (longitudinal orientation) resulted in a 7 ft-lb energy decrease after irradiation to 2.46 x 1019 n/cm2 (E > 1.0 MeV) at 550*F. This results in an llSE of 125 ft-lb (Figure 5-1). Irradiation of the reactor vessel intermediate shell plate A9154-1 Charpy specimens to 2.46 x 10 19 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-2) resulted in 30 and 50 ft-lb transition temperature increases of 35 and 30
*F, respectively, for specimens oriented parallel to the major working direction (transverse orientation). This resulted in a 30 ft-lb transition temperature of 60 *F and a 50 ft-lb transition temperature of 105 *F for specimens oriented perpendicular to the major working direction (transverse orientation).
The average upper shelf energy-(USE) of the intermediate shell plate- A9154-1 , Charpy specimens (transverse orientation) resulted in a decrease of 2 ft-lb in energy after' irradiation to 2.46 x 1019 n/cm2 (E > 1.0 MeV) at 550*F. . This results in an USE of 73 ft-lb (Figure 5-2). ( 5-4 l
trradiation of the reactor vessel core region weld metal Charpy specimens to 2.46 x 1019 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-3) resulted in 30 i and 50 ft-lb transition temperature increases of 35'F. This resulted in a 30 ft-lb transition temperature of -25'F and a 50 ft-lb transition temperature of 20*F. The average upper shelf energy (USE) of the reactor vessel core region weld metal resulted in a 6 ft-lb energy decrease after irradiation to 2.46 x 10 19 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an USE of 85 f t-lb. Irradiation of the reactor vessel weld metal Heat-Affected Zone (HAZ) specimens to 2.46 x 1019 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-4) resulted in 30 and 50 ft-lb transition temperature increases of 45'F. This results in a 30 ft-lb transition temperature of -45'F and a 50 ft-lb transition temperature of -25'F. The average upper shelf energy (USE) of the reactor vessel HAZ metal resulted l in a decrease of 13 ft-lb after irradiation to 2.46 x 1019 n/cm2 (E > 1.0
- l. MeV) at 550*F. This resulted in an USE of 117 ft-lb.
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile i or tougher appearance with increasing test temperature. A comparison of the 30 ft-lb transition temperature increases and %USE decreases for the various V. C. Summer Unit 1- surveillance materials with values predicted using the methods of NRC Regulatory Guide 1.99, Revision 2[5]ispresentedinTable5-6. This comparison indicates that the transition-temperature-increases and USE decreases resulting from irradiation to 2.46 x 1019 n/cm2 (E > 1.0 MeV) are less than the Guide predictions. I 5.3 Tension Test Results l, The results of tension tests performed on shell plate A9154-1 (longitudinal and 19 2 l transverse orientation) and the weld metal irradiated to 2.46 x 10 n/cm 5-5
(E > 1.0 MeV) sre shown in Table 5-7 and are compared with unirradiated resultsIll as shown in Figures 5-9, 5-10 and 5-11. Plate A9154-1 test results are shown in figures 5-9 and 5-10 and indicated that irradiation to
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2.46 x 1019 n/cm2 (E > 1.0 MeV) caused a less than 11 ksi increase in the 0.2 percent offset yield strength and ultimate tensile strength. Weld metal , tension tests results shown in Figure 5-11, show that the ultimate tensile strength and the 0.2 percent offset yield strength increased by less than 7 ksi with irradiation. The small increases in 0.2% yield strength and tensile strength exhibited by the plate material and weld metal indicate that these materials are not highly sensitive to radiation at 2.46 x 1019 n/cm2 (E > 1.0 MeV), as is also indicated by the Charpy impact test results. The fractured tension specimens for the plate material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14. Engineering stress-strain curves for the tension specimens are shown in Figures 5-15 through 5-19. 1 5.4 Compact Tension Tests Per the surveillance capsule testing program with the South Carolina Electric and Gas Company, 1/2 T-compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center flot Cell . -
+
5-6
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNil 1 SHELL PLATE A9154-1 1RRADIATED AT 550'f, FLUENCE 2.46 x 10 19 n/cm2 (E > 1.0 MeV) Temperature Impact Energy Lateral Expansion Shear Sample No. W'F ('C) (ft-lb) IJ1 _(ell s) imm1 (%) Longitudinal Orientation CL49 -25 - 8.0 11.0 8.0 0.20 5 CL60 -10 - 23.0 31.0 12.0 0.30 10 CL53 0 - 32.0 43.5 26.0 0.66 15 CL56 10 - 13.0 17.5 10.0 0.25 10 CL58 25 - 41.0 55.5 30.0 0776 20 CL47 55 50.0 68.0 32.0 0.81 25 CL59 70 64.0 87.0 40.0 1027 35 CL50 85 78.0 100.0 50.0 1.27 50 CL51 125 96.0 130.0 62.0 1,57 65
- CL52 150 99.0 134.0 69.0 1.75 95 CL57 175 109.0 148.0 69.0 1.75 95 CL48 190 115.0 156.0 78.0 1.98 100 CL54 200 120.0 162.5 82.0 2.08 100 CL48 225 133.0 180.5 86.0 2.18 100 CL55 250 133.0 180.5 74.0 1.88 100 Transverse Orientation CT58 -50 -4 3.0 4.0 1.0 0.03 0 0752 -15 -2 10.0 13.5 7.0 0.18 10 CT57 15 -
17.0 23.0 17.0 0.43 15 CT47 25 - 19.0 26.0 18.0 0.46 15 CT56 40 34.0 46,0 26.0 0.66 25 CT54 50 32.0 43.5 27.0 0.69 25 CT60 65 38.0 51.5 29.0 0.74 30 CT49 85 35.0 47.5 29.0 0.74 35 CT46 125 45.0 61.0 39.0 0.99 40 CT55 150 50.0 68.0 46.0 1.17 90 CT59 160 64.0 87.0 53.0 1.35 90 CT53 175 77.0 104.5 58.0 1,47 100 CT51 200 68.0 92.0 53.0 1.35 100 0748 225 1 78.0 106.0% 57.0 1.45 85 0750 250 1 75.0 101.5h 59.0 1.50 100 5-7 j 1
- . ~ ~ -
l 0 TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNil 7 C CILA VESSEL WELD METAL AND HAZ METAL 1RRADIATED A1
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550*F, FLUENCE 2.46 x 10 19 n/cm2 IE > 1.0 MeV) Temperature Impact Energy Lateral Expansion Shear Sample No. ('F) ('C) (ft-lb) QJ (ells) ,(yq), (%) Weld Watal CW50 -100 - 4.0 5.4 4.0 0.10 5 CW53 - 75 - 15.0 20.5 11.0 0.28 10 CW58 - 50 - 15.0 20.5 17.0 0.43 15 CW47 - 35 - 27.0 36.5 20.0 0.51 25 CW60 - 25 - 30.0 40.5 28.0 0.71 30 CW46 - 10 .- 34.0 46.0 25.0 0.64 35 CW49 0 - 57.0 77.5 44.0 1.21 50 CW56 10 - 65.0 88.0 53.0 1.35) 70 CW48 25 MACHINE FUNCTION (a) - - CW59 60 72.0 97.5 55.0 1.40 95 C151 75 62.0 84.0 49.0 1.24 95 . CW57 100 72.0 97.5 59.0 1.50 100 CW55 125 86.0 116.5 69.0 1.75 100 CW52 150 95.0 129.0 73.0 1.85 100 . CW54 200 85.0 115.0 71.0 1.80 100 BAZ Wetal CH58 -100 - 12.0 16.5 6.0 0.15 5 CH50 - 75 - 13.0 17.5 8.0 0.20 10 CH60 - 50 - 37.0 50.0 23.0 0.58 25 CH49 - 25 - 28.0 38.0 24.0 0.61 25 CR51 0 - 31.0 42.0 20.0 0.51 30 0856 25 - 86.0 116.5 50.0 1.27 45 CH48 30 - 104.0 141.0 77.0 1.96 100 CH46 50 98.0 133.0 65.0 1.65 100 CH53 75 112.0 152.0 70.0 1.78 95 CH57 50 3f 0 51.5 30.0 0.76 35 0H55 100 78.0 106.0 52.0 1.32 90 CH52 115 118.0 160.0 76.0 1.93 100 CH47 135 113.0 153.0 76.0 1.93 100 CB59 200 111.0 150.5 75.0 1.91 100 CB54 225 1 124.0 168.0 70.0 1.78 100 (a) Machine loading malfunction, specimen incorrectly loaded in tester. Specimen did not fracture. 5-8
TABLE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE V. C. SUMMER UNIT I SHELL PLATE I9 n/cm2 (E > 1.0 MeV) A9154-1 IRRADIATED AT 550*F, FLUENCE 2.46 x 10 Normalized Enernice Time Wasimum Time en Fracture Arrest Yield Flow Test Charpy Charpy Max imumEp/A Prop Load Yield to Yield Lead Maximum Load Load Stre== Stre== Sample Temp Energy- Ed/A Ee/A (p. ,c) (kipe) (peec) (kipe) {Lipe) (kei) _(keil Number (*F) (ft-lb) (ft-lb/in 23 (kip ) Longitudinal Orientation 125 3.75 0.01 119 122 22 3.60 90 3.75 CL49 -25 8.0 64 42 4.20 350 4.20 0.25 106 122 144 41 3.2 100 132 CL60 -10 23.0 185 85 4.50 435 4.50 0.01 IIS 32.0 258 202 56 3.45 0.25 103 109 CL53 0 3.10 95 3.45 16C 3.45 13.0 los 51 54 0.01 los 127 CL56 10 3.2 95 4.50 505 4.45 25 41.0 330 228 102 675 4.30 0.01 103 126 CL58 307 E6 3.10 105 4.50 119 CL47 55 50.0 403 4.35 655 4.25 1.15 95 294 221 2.85 90 129 CLEO 70 64.0 515 105 4.75 695 4.25 1.15 102 628 336 292 3.10 104 125 CL50 85 78.0 90 4.40 665 3.70 1.50 96.0 773 302 471 3.15 1.65 106 123 CLSI 125 481 3.2 175 4.55 750 3.55 CL52 150 99.0 797 316 4.15 700 3.25 2.05 94 116 878 277 600 2.85 130 95 115 CL57 175 109.0 2.9 110 4.05 605 . -. 190 115.0 926 253 673 0.25 0.25 95 118 CL48 676 2.9 80 4.25 655 CL54 200 120.0 966 290 4.15 685 -. -. 95 116 298 783 2.9 110 86 108 CL46 225 123.0 1071 100 3.90 675 -. -. 1071 273 799 2.60 CLES 250 133.0 Traneverse Orientation 60 2.60 0.01 85 86 12 12 2.50 55 2.60 119 CT58 -50 3.0 24 3.60 130 3.60 0.01 118 42 39 3.55 95 124 CT52 -15 10.0 81 4.00 280 4.00 0.01 114 l 107 30 3.45 IIS 120 CT57 15 17.0 137 80 4.00 260 4.00 0.01 108 153 107 46 3.25 97 117 CT47 25 19.0 2.95 ISO 4.15 485 4.15 0.55 34.0 274 179 95 0.40 100 119 CT56 40 3.00 85 4.15 435 4.15 32.0 258 186 72 0.65 101 120 CT54 50 3.05 210 4.20 520 4.15 65 38.0 306 181 125 4.10 0.60 104 120 CTSO 185 97 3.15 85 4.10 435 CT49 85 35.0 282 460 3.90 2.10 89 110 183 179 2.70 90 3.90 CT46 125 45.0 382 435 3.90 2.50 90 109 172 231 2.75 85 3.90 83 CTS 5 ISO 50.0 403 3.20 450 2.9 2.10 61 153 362 1.85 40 111 CT59 160 64.0 515 4.00 505 -. . 90 207 413 2.75 80 108 CT53 175 77.0 620 3.80 500 -. _. 8g 197 351 2.70 85 129 CT51 200 68.0 548 4.55 655 4.40 2.65 108 628 312 316 3.25 80 102 CT48 225 78.0 2.50 95 3.70 600 -. -. 82 75.0 604 188 416 CT50 250
. Fully ductile fracture, no arrest load.
5-9
l l L TABLE 5-4 i INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE V. C. SUMMER UNIT I l WELD METAL AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 2.46 x 10I9 n/cm2 (E > 1.0 MeV) ! l +
+
e Normalised Eriermies Test Charpy Charpy Manimum Prop Yield Time Maximum Time to Freettre Arrent Yield Flow i Sample Temp Energy Ed/A Em/A Ep/A Load to Yicld Load . Maximum Load Load Streem Stree. l Number M (ft-Ib) (ft-Ib/ inh (hipe) (sarec) (kipe) __ (usec) (hipel (kipe) (heil (hel) ! i
- Weld wet.1 1
l i CW50- -.100 4.0 32 COMPUTER MAIEUNCTION (D) - - - - - - - CW63 - 76 16.0 121 76 43 3.80 llo 4.30 200 4.16 0.01 126' 134 l CW68 - SO 16.0 121 80 41 3.66 90 4.00 200 3.96 0.01 120 127 . CW47 - 36 27.0 217 ISO 67 3.26 96 4.26 360 4.26 0.76 107 123 1 CW60 - 26 30.0 242 190 61 3.60 90 4.46 410 4.40 0.26 119 134 I ! CW46 - 10 34.0 274 187 87 3.10 96 4.26 436 4.26 1.20 102 121 l CW49 O 67.0 469 242 217 3.60 90 4.60 606 4.3 1.80 116 134 g i CW66 10- 66.0 623 234 289 3.40 90 4.60 606 3.76 1.60 113 131 , CW48 26 MACHINE MALFUNCTION - - - - - - - - l CW69 60 72.0 680 227 [a ] 363 3.26 86 4.30 SOS 3.90 2.70 107 126 r CW61 76 62.0 499 207 292 2.90 96 4.06 600 3.80 2.60 96 116 i CW67 100 72.0 680 226 366 3.20 86 4.30 606 -. -. 106 124 i CW66 126 86.0 692 221 472 2.96 80 4.26 606 -. -. 98 119 CW62 160 96.0 766 281 484 2.86 40 4.30 610 -. -. 96 118 [ CW64 200 86.0 684 260 434 2.60 86 4.10 600 -. -. 82 109 ' l RAZ Metal ; i CH68 -100 12.0 97 69 27 4.16 96 4.46 170 4.46 O.01 138 143 CH60 - 76 13.0 106 77 27 4.20 100 4.40 186 4.40 0.01 139 142 s CBSO - 60 37.0 '298 207 90 3.76 100 4.TO 436 4.66 0.01 126 140 ' ! CH49 - 26 28.0 226 128 98 3.80 110 4.60 296 4.40 0.65 126 137 CH61 0 11.0 260 119 131 3.50 90 4.16 280 4.10 1.36 II6 127 - CH66 26 86.0 692 308 384 3.90 116 4.90 600 4.60 2.7 128 146 i CH48 30 104.0 837 324 514 3.36 100 4.70 676 - - 110 132 ? CH67 SO 38.0 306 142 163 3.30 86 4.20 336 4.16 1.26 109 124 l' CH46 60 98.0 789 214 676 2.60 125 4.25 630 - - 86 113 CH63 76 112.0 903 COMPUTER MALFUNCTION [D] - - - - .. - - j CH66 100 '78.0 628 309 319 3.25 80 4.66 660 4.40 2.66 108 129
; CH62 116 118.0 960 291 669 3.00 130 4.30 696 -. -. 98 120 l CH47 136 113.0 910 268 e42 3.15 166 4.50 630 -. -- 104- 127 l CH69 200 111.0 894 293 601 3.16 90 4.25 666 -. -. 104 122 i
! CH64 226 124.0 998 310 688 2.7 90 4.00 740 -. -. 89 til l
- Fully ductile fracture, no arrest loed. (a) Mom ine toediws selfunction, oreeleen 1icerrectly leed.d in tester- specimen did ret fracture.
i l [b} Socceesftal test, bewever, ce*pu*er selfunctioned, therefore
- computer d_sta net availabfL. (
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( a 1 A S l n r R . e a p er r D I I C 5 x o 3 E t a d, F Y eg l a e 0 5 5 F O a r e t r e wta
. i S v a i d 0 5 5 0 F E A L T a 5 3 6 O I r - - ,
T i e T R n C E u E P F O F R - E P 5 5 T 5 5 5 E. 0 M i A 5 3 3 4 ri d t 0 e T t b ) a 5 0 5 5 M l F i 2 6 2 4 C -
- d - -
T t ( a O f r N 0 e r i r N 3 u - D t
'de e
g a r a e t
- p ia ev a 5 5 0 0 e d 5 9 A T a 2 2 r.
r
- i n
U
)
1 l 1
+ - a - )
l l 4 5 i n . l l 4 5 e s l e 1 d eh 1 r a l a l h 9 u . 3 e t a S A t S A v a t i i s M e e r r e g r e n N t e ew t a n o w t a a r d l Z a o l l o l T e A M t P ( t P ( W N L
- i1'1! ,le[. i; . ,' j i I 3 , 3!j i!; ,iI!It;
TABLE 5-6 COMPARISON OF-V. C. SUMMER UNIT I SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS f' AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS i i 30 ft-1b Transition Temp. Shift Upper Sheh Energy Decrease l 1- R.G. 1.99 Rev. 2 R.G. 1.99 Rev. 2 Fluence (Predicted)' Measured (Predicted) Measured j Material Capsule IQ I9 n/cm2 g.p3 g.p3 gg) gg) ; [ Int. Plate A9154-1 U 0.639 57 40 17 1 i (Longitudinal) V 1.47 72 60 21 8 i X 2.46 81 50 23 5 Int. Plate A9154-1 0 0.639 57 30 17 0 (Transverse) V I.47 72 40 21 0 X 2.46 81 35 23 3 , Weld Metal U 0.639 59 30 18 4 2 i V 1.47 75 45 22 7 l' X 2.46 102 35 25 7 i HAZ Metal U 0.639 -- 30 -- 7 [
- V 1.47 --
45 -- 15 ; !' X 2.46 -- 45 -- 10 l i i I I 5-12 l l
TABLE 5-7 TENSILE PROPERTIES FOR V. C. SUMMER UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE X MATERIAL I9 n/cm2 (E > 1.0 MeV) IRRADIATED AT 550*F TO 2.46 x 10 Total Redection Test 0.2% Yield Ultimmt Frmeture Fracture Fracture Uniform Streee Strength Elenzation Elonzation in Are. Smeple Temp. Strength Strength Load (%) (kie) (kei) (kell (%) (%) Material Number f*F1 (kol) (kei) 202.0 et.1 10.5 23.4 70 Plate CLIO
- 76 80.0 95.7 3.00 169.8 61.1 0.0 19.8 64 A-9154-1 CL11 300 73.3 89.6 3.00 3.20 164.2 es.2 9.8 19.4 60 (Long. CL12 660 70.3 88.6 Orient.)
3.45 156.6 70.3 12.s 23.3 65 Plate CT10 75 69.8 92.7 68 84.6 3.36 161.5 63.2 12.0 18.9 A-9154-1 CT11 300 62.6 19.8 60 88.6 3.50 141.4 71.3 11.1 (Tranev. CT12 ESO 61.1 Orlant.) 208.8 63.2 12.0 23.7 70 Weld CW10 75 80.5 93.7 3.10 69 2.95 191.6 00.1 10.5 21.5 CW11 300 72.s se.e 21.5 53 90.7 3.16 151.9 64.2 10.5 CW12 550 71.3 l 5-13
( c)
-100 -50 0 50 100 150 200 250 i i 100 i
ig 3 --- i i i GW s - h @0 - - 0
' ' ' ' I -
lit i i i i i i i 2.5 55 - 2.0 2 W - [ - 1.5 ^$ 40 - 1.0 - g 20 -
' ', s*r -
O ' 0 0 . 160 i i i i i i i - 1@ - e_ 2W t e 120 - ND 160 g- .
- W
~
1 N m n M m*r 3 60 - 5,F 2.# x to# n/an _ g 3*r 20 0 0
-200 -103 0 100 200 300 400 500 BPERATURE (* F )
Figure 5-1. Charpy V-Notch Impact Properties for V. C. Summer Unit 1 Reactor , Vessel Shell Plate A9154-1 (l.ongitudinal Orientation) 5-14 l l
( 0)
-100 -50 0 50 100 150 200 250 , g -
_:-). 80 H o 60 g - o . _ 20 - 0 i M 2 e i i i _,, 100 i i i i i i i i 2.5 x8 2 - - 10 60 - #*f a- .
. o 1.5^g
$5 # -
. 2 -
1.0 -
- 8 -
20 0.5 0 - 0 90 i i i i i i i 120 80 .. useaAn : - 3 - o . 100 Y ~
*o -
52 - E# -
. . 8f \. - meannr
,2 5 30 8 2.46 x to" n/an2 -
g 20 3r - 20 10 I I f I I I 0 0
-200 -100 0 100 200 300 #0 500 TEMPERATURE (* F )
Figure 5-2. Charpy V-Notch Impact Properties for V. C. Summer Unit 1 Reactor Vessel Shell Plate A9154-1 (Transverse Orientation) l 5-15 l
(*C )
-100 - 50 0 50 100 150 200 250 i '
I i 1 2 12 '3 I 100 - 6 d-- - G 80 , 60
$e -
j 20 - 0 V ' 1 I 1 I 100 i i i i i i i 2.5 aE 80 o
,2 2 -
2.0 2s 60 1.57 25 40 - 1.0 W 20 2ff - 0.5 5 0 I i I I I 0 IN i i i i i i 1 200 WRROAD - 160 120 1 2 2 100 -
,. L -
120 ^, g g _ o ;
~
g .
~
C 60 - 3fr * - 80 o IRRADIAE AT 550' f E ~ 2 3rr 2.46 X 10" n/cm2 _ 20 - 2/ V i I ' ' I 0 0
-200 -100 0 100 200 300 400 500 BPERATURE (* F )
Figure 5-3. Charpy V-Notch Impact Properties for V. C. Summer Unit 1 Reactor Vessel Weld Metal 5 16
( *C )
- 50 0 50 100 150 200 250
_ -100 i I i i- i i i 2 13 100 _ , t l' 1 g 80 - -
- o 8 -
60 - c4 w @ - m
- 20 -
2
' ' ' ' I 0
100 2.5 I i i i2 12 i i 4 e 80
. - ./ -
F 2.0
- 60 -
2 95 w 40 - 70*r _ 1'5^! 1.0 - d 20 0.5 5 0 0 160 i i i i i i i 140 - UNmAotAn o
~2 2M e
- 120 -
, , .o; -
160
= -
- 100 -
120 ^ 6,f EADIAE AT 550* f 3 "s' 60 - o*r 2.46 x 10" n/crn2 _ g 0 -
/. *
~
20 -
.),l * ~ ' ' ' ' '
0 0
-200 -100 0 100 200 300 400 500 TEMPERATURE (* F )
Figure 5-4. Charpy V-Notch Impact Properties for V. C. Summer Unit 1 Reactor . Weld Heat Affected Zone Metal ' 5-17
i l
, :.n -
c,., g 3 4 , . . f,l p.j . ea {&c.' A lr 'i(M;, llb
-C) Mi @h (tkbY $
CL49 CL60 CL53 CL56 CL5B
- ~ w W::
e ., _ .4
'j- ,
.hkb '*! b -
CL47 CL59 CL50 CL51 CL52
?"*'* vm, ~;;; ^
- s % ?
,. y ..
p- -. (5 - , d- '
'e, ~
.y 'q.: gu,4 }
-r;p .@
a w b y:7,t . CL57 CL48 CL54 CL46 CL55 Figure 5-5. Charpy Impact Specimen Fracture Surfaces for V. C. Summer Unit 1 Reactor Vessel Shell Plate A9154-1 (longitudinal Orientation) 5-18
t t meme ,, CT47 CT56 CT52 CT57 CT58 I l _,, ,u 7 ,
/
CT46 0T55 CT60 CT49 l CT54 l I
.num= = + - d*"Y
- m, .; *= e -
gyy f CT48 CT50 CT53 CT51 !. C759
. Figure 5-6. Charpy Impact Specimen Fracture Surfaces for V. C. Summer Unit 1 Reactor Vessel Shell Plate A9154-1 (Transverse Orientation) 5-19
l t 4
; . ; < c,
.t, p; I -
El .af :..f '
. . , 9 -
Sk '
/ ,'T . . ,, ; l-
't maammmed CW50 CW53 CW58 CW47 CW60 m, ** 4"5
,.M -u d '
,hj;.
~
MACHINE - MALFUNCTION 1,2,,,,_,
"yk 1, (see Noto (a) j, j{ jy tv-^)' Table 5-2) l$h,p ? 3S 3 me CW46 CW49 CW50 CW59 CW48 yy. ,
p_ pm '" ~~g , g CW51 CW57 CW55 CW52 CW54 . Figure 5-7, Charpy impact Specimen Fracture Surf aces for V. C. Summer Unit 1 , Reactor Vessel Weld Metal 5-20
- =s.,,, ~r , 7 "'
< ,. . , +'.j l .. , ;;, , '. hie fr.
s , t; ?.l;r . C ,jff,. _.;9, h
;r x ! :, . .
e
^l
'U (g_ _ _
.y um
- f. ,' '
CH58 CH50 CBSO CH49 CH51
+" . we,
,w une.. ., ,
i; -
^
f' ,,
- s. - .,. a f
$' ~ s g', _
CH56 CH48 CH46 Clf53 0H57
'ttenV p~
ww .7 *G ,). s , .,<
; 3.
,;.s f g,. ,;4
)
CH55 CH52 CH47 CH59 CH54 Figure 5-8. Charpy impact Specimen fracture Surfaces for V. C. Summer Unit 1 Reactor Vessel Weld Heat Affected Zone (HAZ) Metal 5-21
._..._..T-. - - - - . . - . . . . - . . . . - - - - - . - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' ' - ' ' ' ' i
- ('C )
0 50 100 150 200 250 300 120 i i i i i i i - 800 11 0 - 100 - ULT u t 10 at s u cm
~
~3 90 -
~ 2 A I _ M~ *T 70 -
.[ ' -
2 b 60 _ g 50 - 0.n mD smam
~
g i i i i i E CODE: OPEN P0(NTS - UNIRRADIATED CLOSED POINTS - IRRADIATED AT 550*fn TO 2.46 X 10n /cm2 . N 1 1 l l 1 i i
~
N ; A2 60 i ~ _ REDuctim N AREA O - TOTAL EONGATXN E 30 t _- 20 - s J' -- 10
- ! E 1 -
uwww amcAixN 0 0 100 200 300 @ 500 @ TEMPERATURE (* f ) , Figure 5-9. Tensile Properties for V. C. Summer Unit 1 Reactor Vessel Shell , Plate A9154-1 (Longitudinal Orientation) 5-22
1 i ( *C ) 0 50 100 150 200 250 300 120 , i i i i i , 800 110 - 2 100 - Utan Eat smom
~
~
g
^ 2 f
70 - - 500 - g- _ % 6- :
-e 400 50 -
0.2x ma secu g l l l l l M CODE: OPEN POINTS - 1)NIRRADIATED CLOSED POINTS - IRRADIATED AT 550*F 10 2.46 X 10"n/cm2 , 80 i i i i i i i 70 RcMCM N AREA - g -- 0 : _ 550 - - M 40 - - TotAt. noNGAM 5 M - w ,2
'- ,.2 ~
- i 20 i -- a -
10 P2 - UHroW a0NGA M 0 0 100 200- 300 400 500 600 TEMPERATURE (* F ) Figure 5-10.- Tensile Properties for V. C. Summer Unit 1 Reactor Vessel Shell Plate A9154-1 (Transverse Orientation) 5-23 wae,,,- ,,-_-n.,a-r ---r-,.,,w,,..,,..,-,-mo~.-w, e .,v~m,e--+-wr---<m-m,-- new ,,vv-
, n -n, ery-,--y , -,->.,e.g, -
i (*0 ) 120 0 50 100 150 200 250 2
' ' ' ' i i l_ g 100
- 700 g vaut Tema sTracm N ~
2-3 j- g7 h83 3 g 70 -
'N_ - .
500 -
+
M - 0.2% ED STPBGN _ g 50 -
~
e i i i i i E CODE: OPEN PO!NTS - UN1RRADIATED CLOSED POINTS - IRRADIATED AT 550*f TO 2.46 X 10"n/cm2 ' N I I i i i i i
- +-
70 _ e 2
,,,,, x z ,, _
". 50 D 0 - _
h TOTAL EWCATKN
~
2t "
/
,' 2 _
20 10
! -T'_
Usors amcATKN 0 0 100 200 300 W 500 E TEMPERATURE (* f ) . Figure 5-11. Tensile Properties for V. C. Summer Unit 1 Reactor Vessel Weld , Metal 5-24 !
l 1
+ , , - .r :, w. +;
l 5 s,- g ~I. t. no: c m n; 1 : J J g; s7 , a3 3'^-- p; ut i
, ' ~rd t ,
a
. ;% g; -t f ,,y *;?9 % **
I 4' 24 Specimen CL10 75'F
,% y ,
d,i D, ,o, . - i ( l I - i t.- I Specimen CL11 300'F 1 V
- l e
ff
. ., i
- %9
[ '; ' f i v-l . Specimen CL12 550'P l, i i
- , Figure 5-12. Fractured Tensile Specimens from V. C. Summer Unit 1 Reactor Vessel Shell Plate A9154-1 (longitudinal Orientation) s-as l
l
i i i ? i - i Specimen CT10 75'F l l Speciaen CT11 300'F i . 4 l Specimen CT12 550'F 1 l l Figure 5-13. Fractured Tensile Specimens from V. C. Summer Unit 1 Reactor ' l Vessel Shell Plate A9154-1 (Transverse Orientation) 5-26 l ____ - _ _ __
Specimen CW10 75'P Specimen CW11 300'F s ' yp,. # m
*> r r
dk. Specimen CW12 550'F figure 5-14. Fractured Tensile Specimens from V C. Summer Unit 1 Reactor Vessel Weld Metal 5-27
100 ! g 90.- . b 60-
': N -
50- : 40-30-20-SPEC CLIO . 10- ^ 75 F 0 0 0.b5 0 '1 0. 1' 5 0.'2 0.25 0.3 i STRAN, N/N J
.90 .;
80-70-a60-
- w:
,50-m
" 40 h-M 30-20-10 SPEC CL11 300F 0 -
0 0.b5 0.' 1 0.'15 0.'2 0.25 STRAN, N/N Figure 5-15. Engineering Stress-Strain-Curves for Shell Plate A9154-1 Tension . Specimens CLIO and CLil,
.5-28
. - _ -. . - - _. . -._-_-- _ -__ _ _ _ _ -- .-----<,.--,.a-.. -
----a-,...w,<-e.~+---- --.--ve, -------,--,.--re- ..
~
100 . 90-0- /
$ 60-p 50-E 40-G 30-20- SPEC CL12 10- 550 F 0- ' '
0.225 0 0.025 0.b5 0.d75 0 '1 0.125 0.'5 0.175 1 0.'2 STRAN, N/N
. 100 90-80-70-M x 60-p 50-f 40-30-
~
SPEC CT10 10- 75 F
- 0. . , ,
0 0.05 0.1 0.15 0.2 0.25 STRAN, N/N
. Figure 5-16. Engineering Stress-Strain Curves for Shell Plate A9154-1 Tension Specimens CL12 and CTIO.
5-29
90 - 80- . 70- . m 60-x 50-m y40-m 30-20-10-300 F 0 0 0.05 0.1 0.15 0.2 0.25 STRAN, N/N 100 - 90-80-70-G x 60-d 50-d g 40-30-20-SPEC CT12 10- 550 F 0 . . . . . . . . 0 0.025 0.05 0.075 0.1 0.125 0.15 0.175 0.2 0.225 STRAN, N/N , Figure 5-17. Engineering Stress-Strain Curves for Shell Plate A9154-1 Tension . Specimens CTil and CT12. 5-30
..___..__._.q l
100-80-p. 70-M x 60- ) mf $ 40-30-20- SPEC CW10 10- 75 T 0 , , , 0 0.05 0.1 0.15 0.2 0.25 STRAN, N/N
'- 90 80-70-m 60-x.
50-m
- 40-b to30-20-SPEC CW11 10 300 F 0 , , ,
0 0.05 0.1 0.15 0.'2 0.25 STRAN, N/N
- Figure 5-18.. Engineering Stress-Strain Curves for Weld Tension Specimens-CW10 and CWil.
5-31 _2.__ _ - . . . . _ _ _ . _ . _ . _ . _ . . _ . _ _ _ _ _ _ . _ _ _ - . . _ ._. _._. . . . . . _ _ _ . _ . _ . _ _ _ . . _ _ . _ _ _ _
i 100 90-80-m 70- ( x 60-y 50-w p 40-M ' 30-20- SPEC CW12 10- 550 F 0 i . . . . . . . 0 0.025 0.05 0.075 0.1 0.125 0.15 0.175 0.2 0.225 STRAN, N/N Figure 5-19. Engineering Stress-Strain Curve for Weld Tension Specimens CW12. 5-32
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 10.troduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance prcgrams for two reasons, first, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (onergy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens._ The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with-passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor aoplications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as. to a more accurate evaluation of damage gradients through the pressure vessel wall. Because _of this potential shift away fre a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice ~ E853(20], " Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with 6-1
fluence (E > l 0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693(18), " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of , the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99[5), " Radiation Damage to Reactor Vessel Materials." This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in
- surveillance Capsule X. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history.
The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surseillance capsule and with the projected exposure of the , pressure vessel are provided. 6.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached-to the neutron pads are included.in the reactor design _to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 107*, 110', 287*, 290', 340', and 343' relati m to the core cardinal axes as shown in Figure 4-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1. The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core. O 6-2
From a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pad ano reactor vessel, in order to properly determine the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters (d(E > 1.0 Mev), p(E > 0.1 Nev), and dpa} chrough tne vessel wall . The neutron spectral information was required for the interpratation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/p(E > 1.0 MeV), within the pressure vessel geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations. The second set of calculations consisted of a series of adjoint analyses
, relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for the first 5 cycles of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles, it is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects 6-3
of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased. The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the , forward calculation provided the means to:
- 1. Evaluato neutron dosimetry obtained from surveillance capsule locations,
- 2. Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.
- 3. Enable a direct comparison of analytical prediction with measurement.
- 4. Establish a mechanism fr projection of pressure vessel exposure as the design of each new fuel cycle evolves. ,
The forward transport calculation for the reactor model summarized in Figures . 4-1 and 6-1 was carried out in R, 8 geometry using the DOT two-dimensional discrete ordinates codell43 and the SAILOR cross-section library [15]. The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water-reactor applications. In these analyses anisotopic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an S8 order of angular quadrature. The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 3-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2a 6-4
level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results. All adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source locations were cho::en at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, 9 geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, p (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as: R(r,0)=fr[0[E 1(r, 0, E) S (r, 0, E) r dr d0 dE where: R (r , 0) =[(E>1.0MeV)atradiusrandazimuthalangle0 1 (r, 0, E) = Adjoint importance function at racius, r, azimuthal angle 0, and neutron source energy E. S (r, 0, E) = Neutron source strength at core location r, 0 and energy E. Although the adjoint importance functions used in the V. C. Summer Unit I analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/d (E > 1.0 MeV) is insensitive to changing core source distributions. In the application of these adjoint importance functions to the V. C. Summer Unit I reactor, therefore, the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/4 (E > 1.0 MeV) and 4 (E > 0.1 MeV)/d (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific
& (E > 1.0 MeV) solutions from the individual adjoint evaluations.
6-5 l
The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design report for the first operating cycle of V. C. Summer Unit 1[33-38). The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of the presture vessel and surveillance capsules are summarized in Figure 6-2. , For comparison purposes, the core power distribution (design basis) used in the reference forward calculation is also illustrated in Figure 6-2. Selected results from the neutron transport analyses performed for the V. C. Summer Unit I reactor are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule trradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall. In Table 6-1, the calculated exposure parameters (p (E > 1.0 MeV), d(E
> 0.1 MeV), and dpa] are given at the geometric center of the two surveillance ,
capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport , analysis, are meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycle 1 through 6 plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself. Radial gradient information for neutron flux (E > 1.0 MeV), neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the , vessel inner radius to the gradient data given in Tables 6-3 through 6-5, 6-6
For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by:
= ((199.95, 45') F (204.95, 45')
41/4T(45') where: = Projected neutron flux at the 1/4T position on 41/4T(45') the 45' azimuth ( (199.95,45') Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth. F (204.95, 45') Relative radial distribution function from Table 6-3. Similar expressions apply for exposure parameters in terms of 4 (E > 0.1 MeV) and dpa/sec, The DOT calculations were carried out for a typical octant of the reactor. However, for the neutron pad arrangement in V. C. Summer Unit 1, the pad extent
, for all octants is not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron pad extending from O' to 15' which produces the maximum vessel flux. Other octants have neutron pads extending from 0* to 26' which provide more shielding.
6.3 Neutron Dosimetry The passive neutron sensors included in the V. C. Summer Unit I surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaiuation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest [p (E > 1.0 Mev), p (E > 0.1 MeV), dpa]. 6-7
~ . .__
.The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes-drilled in spacers at several axial levels within
~
the capsules. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule. _ The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest. , Rather, the activation or fission process is a measure of the integrated effect 1 that the time- ano energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: o The specific activity of each monitor, o The operating history of the reactor. , o The energy response of the monitor, o The neutron energy spectrum at the monitor location. .) o The physical characteristics of the monitor. The specific activity of each of the neutron monitors was determined using established ASTM procedures [16 through 29). Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the V. C. Summer Unit I reactor during cycle I was obtained from NUREG-0020,
" Licensed Operating Reactors Status Summary Report" for the applicable period.
The irradiation history applicable to Capsule X is given in Table 6-7. Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7. e 6-8 I
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code (30). The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the center of the survelliance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra. In the FERRET evaluations, a log normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly related to the flux & by some response matrix A: ff8'"I=I g A ) d g")
- where i indexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, Rj - Ig ojg d g relates a set of measured reaction rates R$ to a single spctrum pg by the multigroe: cross section ogg . (In this case, FERRET also adjusts the cross-sections.) The lognormal approach aatomatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.
6-9
In the FERRET analysis of the dosimetry data, the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the IERRET group structure using the SAND-Il code (31]. This procedure was carried out by first expanding the a priori spectrum into the SAND-Il 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed to the group scheme used in FERRET. The cross-sections were also collapsed into the 53 energy-group structure using SAND 11 with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/8-V dosimetry file. Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant. For each set of data or a priori values, the inverse of the corresponding , relative covariance matrix M is used as a statistical weight, in some cases, as for the cross sections, a multigroup covariance matrix is used. More often, . a simple parameterized form is used: M gg,=Rh+Rg R,P g gg, where RN specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties Rg specify additional random uncertainties for group g that are correlated with a correlation matrix: Pgg, = (1 - 0) Igg, + 0 exp [- ] 4 6-10
The first term specifies purely random uncertainties while the second term . describes short-range correlations over a range a (0 specifies the strength of the latter term). For the a priori calculated fluxes, a short-range correlation of a - 6 groups was used. This choice implies that neighboring groups are strongly correlated when r is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E. Maerker[323 Maerker's results are closely duplicated when B = 6. For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties. l Results of the FERRET evaluation of the Capsule X dosimetry are given in Table i 6-9. The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 2.46 x 10 19 n/cm2 (E > 1.0 MeV) with an associated uncertainty of 8L Also reported are capsule exposures in terms of fluence _( E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the adjusted spectrum are provided in Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy groa p structure, l A summary of the measured and calculated neutron exposure of Capsule X is presented in Table 6-12. The agreement between calculation and measurement falls within 10% for all fast neutron exposure parameters listed. Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (5.03 EFPY) exposure derived from the Capsule X measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EFPY). In order to account- for recent loading philosophies, The calculated cycle 5 exposure rates given in Table 6-2 were used to perform projections beyond the end of cycle 5. l I 6-11
l in the calculation of exposure gradients for use in the development of heatup and cooldown curves for the V. C. Summer Unit I reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were employed. Data based on both . a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14. In order to access RTNDT vs. fluence trend , curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations doa d' (1/4T) = & (Surface) {gp r a e)} do d' (3/4T) - 4 (Surface) {dpa u a e)) Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Table 6-15 updated lead factors are listed for each of the V. C. Summer Unit I surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules. 6-12
i
- 16,94 DEG. - 19.72 L EG.
3 Y
, - 73.31 IN.
l l h_ a. ! \ s
- h! _
Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13
Design cyc 5 0.93 0.77 0.52 0.44 0.95 1.07 1.12 0.80 0.92 1.13 0.98 0.49 1.11 0.97 1.02 1.04 0.85 1.30 1.11 1.26 1.12 0.47 1.11 1.11 1.13 1.03 0.92 1.04 1.05 1.09 1.23 0.81 0.97 1.16 1.00 1.11 1.02 1.04 1.28 1.06 Figure 6-2, Core Power Distributions Used in Transport Calculations for V. C. Summer Unit 1 6-14
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS
~
AT THE SURVEILLANCE CAPSULE CENTER 17.0* LQ O' 4 (E > 1.0 MeV) 2 (n/cm-sec) fxla 1 1.63 x 10 11 1.41 x 10 11 2 1.39 x 10 11 1.26 x 10 11 3 1.35 x 10 ll 1.18 x 10 ll 4 1.20 x 1011 1.07 x 10 11 5 1.44 x 1011 1.30 x 10 11 CRSD. 2.09 x 10 11 1.82 x 10 11
. 4 (E > 0.1 MeV) 2 l (n/cm-sec)
Cycle 1 8.63 x 10 11 7.31 x 10 11 2 7.34 x 10 11 6.50 x 1011 3 7.18 x 1011 6.12 x 1011 4 6.36 x 10 11 5.54 x 10 ll 5 7.65 x 10 11 6.72 x 1011 CRSD 1.11 x 10 12 9.41 x 10ll dpa/sec CEla
.1 3.52 x 10-10 3.01 x 10-10 2 2.99 x 10-10 2.68 x 10-10 3 2.93 x 10-10 2.52 x 10-10 4 2.59 x 10-10 2.28 x 10-10 5 3.12 x 10-10 2 77 x 10-10 CRSD 4.51 x 10-10 3.88 x 10-10 6-15
TABLE 6-2 CALCVLATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 2 ((E > 1.0Mev) (n/cm -sec) Cycle 0 dea _ 12 dea 20.5 dea 30 dea 45 dea 1 5.367E+10 3.506E+10 2.947E+10 2.318E+10 1.654E+10 2 3.891E+10 2.761E+10 2.612E+10 1.987E+10 1.230E+10 3 3.902E+10 2.767E+10 2.462E+10 1.850E+10 1.252E+10 4 3.389E+10 2.450E+10 2.247E+10 1.764E+10 1.225E+10 5 3.804E+10 2.833E+10 2.700E+10 2.040E+10 1.314E+10 Design Basis 6.600E+10 4.450E+10 3.780E+'O 2.912E+10 2.078E+10 p(E > 0.lMov) (n/cm2-sec) Cyclg O dea 12 dea _20.5 dea 30 dea 45 dea 1 1.353E+11 8.624E+10 6.248E+10 4.822E+10 3.407E+10 2 9.806E+10 6.792E+10 5.538E+10 4.134E+10 2.533E+10 3 9.834E+10 6.807E+10 5.220E+10 3.848E+10 2.579E+10 4 8.541E+10 6.027E+10 4.763E+10 3.669E+10 2.524E+10 5 9.586E+10 6.968E+10 5.723E+10 4.244E+10 2.708E+10 Design Basis 1.663E+11 1.095E+11 8.014E+10 6.057E+10 4.281E+10 dpa/sec Cycle O dea 12 dea _20.5 dea 30 dea 45 dea 1 8.59E-Il 5.61E-Il 4.57E-Il 3.59E-ll 2.58E-11 2 6.23E-11 4.42E-11 4.05E-11 3.08E-Il 1.92E-Il 3 6.24E-ll 4.43E-Il 3.82E-Il 2.87E-ll 1.95E-11 4 5.42E-Il 3.92E-ll 3.48E-ll 2.73E-11 1.91E-Il 5 6.09E-ll 4.53E-11 4.18E-ll 3.16E-11 2.05E-11 Design Basis 1.06E-10 7.12E-Il 5.86E-11 4.51E-ll 3.24E-ll 6-16
TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0' 12' Z_0. 5
- 30' 45' 199.95(1) 1.00 1.00 1.00 1.00 1.00 200.91 0.926 0.924 0.926 0.927 0.927 202.30 0.799 0.796 0.801 0.801 0.803 203.74 0.667 0.670 0.672 0.673 0.675 205.13 0,554 0.560 0.559 0.562 0.564 206.52 0.456 0.466 0.463 0.465 0.468 207,91 C.374 0.385 0.381 c.384 0.387 209.30 0.306 0.317 0.312 0.315 0.318 210.69 0.249 0.260 0.255 0.258 0.261 212.07 0.202 0.212 0.209 0.211 0.214 213.46 0.164 0.173 0.170 0.172 0.175 214.85 0.132 0.140 0.138 0.140 0.143 216.24 0.105 0.113 0.111 0.113 0.116 217.63 0.0828 0.0904 0.0883 0.0900 0.0934 218.86 0.0650 0.0728 0.0708 0.0726 0.0766 219.95(2) 0.0510 0.0586 0.0568 0.0586 0.0628 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius g
6-17 l
TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF HEUTRON FLUX (E > 0.1 MeV) WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0' 12' 20.5* 30' 45* 199.95(l) 1.00 1.00 1.00 1.00 1.00 200.91 1.00 1.00 1.00 1.00 1.00 202.30 0.965 0.970 0.978 0.978 0.984 203.74 0.905 0.917 0.926 0.925 0.935 205.13 0.839 0.858 0.867 0.866 0.878 206.52 0.771 ^ 795 0,803 0.803 0.817 207.91 0.704 0.732 0.740 0.739 0.754 , 209.30 0.639 0.668 0.676 0.676 0.693 210.69 0.575 0.606 0.614 0.614 0.632 212.07 0.513 0.546 0.554 0.554 0.573 213.46 0.454 0.486 0.495 0.495 0.515 214.85 0.396 0.429 0.438 0.438 0.458 216.24 0.339 0.372 0.382 0.382 0.403 217.63 0.283 0.316 0.327 0.327 0.350 218.86 0.230 0.265 0.277 0.279 0.303 219.95(2) 0.187 0.222 0.235 0.238 0.262 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-18
TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0' _1ti_ 22211 _J0' 45' l 199.95(l) 1.00 1.00 1.00 1.00 1.00 l 200,91 0.939 0.939 0.938 0.938 0.938 202.30 0.841 0.841 0.837 0.835 0.836 203.74 0.739 0.744 0.732 0.731 0.732 l 205.13 0.648 0.657 0.640 0.639 0.641 f 206.52 0.567 0.579 0.558 0.557 0.560 207,91 0.495 0.509 0.487 0.486 0.490 l 209.30 0.432 0.447 0.425 0.423 0.428 ! I
. 210.69 0.375 0.391 0.370 0.369 0.374 212.07 0.325 0.341 0.321 0.320 0.326 213.46 0.280 0.296 0.278 0.277 0.283 214.85 0.239 0.255 0.240 0.239 0.245 216.24 0.202 0.217 0.204 0.204 0.211 217.63 0.166 0.182 0.172 0.171 0.180 218.86 0.134 0.152 0.145 0.145 0.155 219.95(2) 0.108 0.127 0.122 0.123 0.133
- NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius k
6-19
F' TABLE 6-6 L
. NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Weight Response Product Yield Material . Interest Fraction Rance Hal f-Life (%)
Copper Cu63(n,a)Co60 0.6917 E > 4.7 MeV 5.272 yrs Iron Fe54(n.p)Mn54 0.0582 E > 1.0 MeV 312.2 days Nickel NiS8(n,p)CoS8 0.6830 E > 1.0 MeV 70.90 days ! Uranium-238* U238(n,f)Cs137 1.0 E > 0.4 MeV 30.12 yrs 5.99 Neptunium-237* Np237(n,f)Csl37 1.0 E > 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum
- CoS9(n,0)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobalt-Aluminum CoS9(n,8)Co60 0.0015 E > 0.015 MeV 5.272 yrs
- Denotes that monitor is cadmium shielded.
l l l l 6-20
TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE X Irradiation P P 3 Irradiation Decay 3 Period Time (days) Time (days) (MWt ) Pref. 399.4 0.1439 15 2878 11/1982 875.1 0.3153 31 2847 12/1982 1311.0 0.4724 31 2816 1/1983 1256.8 0.4529 28 2788 2/1983 761.8 0.2745 31 2757 3/1983 0.0 0.0000 30 2727 4/1983 356.3 0.1284 31 2696 5/1983 2324.5 0.8377 30 2666 6/1983 2504.7 0.9026 31 2635 7/1983 8/1983 2363.0 0.8515 31 2604 2429.7 0.8756 30 2574 9/1983 2433.1 0.8768 31 2543 10/1983 11/1983 2108.5 0.7598 30 2513 12/1983 1076.1 0,3878 31 2482 2684.1 0.9672 31 2451 1/1984 2/1984 2394.1 0.8627 29 2422 1962.0 0.7070 31 2391 3/1984 4/1984 165.3 0.0596 30 2361 5/1984 2290.6 0.8255 31 2330 6/1984 2565.7 0.9246 30 2300 7/1984 1120.0 0.4036 31 2269 8/1984 2479.5 0.8935 31 2238 9/1984 1951.5 0.7032 30 2208 10/1984 0.0 0.0000 31 2177 11/1984 0.0 0.0000 30 2147
~
12/1984 612.7 0.2208 31 2116 6-21
- .- . . . ._ .. _ ~ _ _
TABLE 6-7 (Continued) IRRADIATION HISTORY OF NEUTRON SENSORS-CONTAINED IN CAPSULE X Irradiation Pj Pj Irradiation Decay , Period (MW)t Pref. Time (days) Time (days) 1/1985 2037.1 0.7341 31 2085 2/1985 2348.6 0.8464 28 2057 ' 3/1985 2548.8 0.9185 31 2026-4/1985 -2336.3 0.8419 30 1996 5/1985 1689.1 0.6087 31 1965 6/1985 2763.3 0.9958 30 1935 7/1985 2759.6 0.9944 31 1904 8/1985 2341.9 0 E439 31 1873 9/1985 2550.9 0.9193 30 1843 1 10/1985- 393.6 0.1418 31 1812 ,j 11/1985 0.0 0.0000 30 1782-12/1985 888.9 0.3203 31 1751 1/1986 2740.8 0.9877 31 1720 2/1986 2438.1 0.8786 28 1692 3/1986- 2770.5 0.9984 31 1661 4/1986 2636.3 0.9500 30 1631 5/1986 2716.8 0.9790 31 1600 6/1986 2029,4 0.7313 30 1570 7/1986 2552.2 0.9197 31 1539 8/1986 2738.2 0.9867 31 1508 9/1986 2719.3 0.9799 30 1478 10/1986 2467.4 0.8892 31 1447 11/1986 2587.2 0.9323 30 1417 12/1986 2479.6 0.8935 31 1386 1/1987 2657.5 0.9577 31 1355 2/1987 2735.4 0.9857 28 1327 6-22
=,
e TABLE 6-7 (Continued) IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSVLE X Irradiation Pj P 3 Irradiation Decay Period Time (days) Time (days) (MWt ) Pref. 448.5 0.1616 31 1296 3/1987 0.0 0.0000 30 1266 4/1987 0.0 0.0000 31 1235 5/1987 1658.1 0.5975 30 1205 6/1987 7/1987 2711.0 0,9769 31 1174 t 2725.5 0.9822 31 1143 8/1987 1546.4 0.5572 30 1113 9/1987 2460.4 0.8866 31 1082 10/1987 2772.1 0.9990 30 1052 11/1987 2771.4 0.9987 31 1021
^
12/1987 2772.3 0.9990 31 990 1/1988 2560.2 0.9226 29 961 2/1988 3/1988 2772.9 0.9992 31 930 4/1988 2768.3 0,9976 30 900 5/1988 2433.6 0.8770 31 869 6/1988 1771.7 0.6384 30 839
,g 7/1988 2545.9 0.9174 31 808 8/1988 2772.7 0.9992 31 777 1472.9 0.5308 30 747 9/1988 10/1988 0.0 0.0000 31 716 11/1988 0.0 0.0000 30 686 12/1988 81.4 0.0293 31 655 1/1989 2499.4 0.9007 31 624 2/1989 1641.1 0.5914 28 596 6-23
TABLE 6-7 (Continued) 1RRADIAT10N HISTORY Of NEUTRON SENSORS CONTAINED IN CAPSULE X Irradiation Pj Pj trradiation Decay , Period (MWt ) Pref. Time (days) Time (days) i 3/1989 427,9 0.1542 31 565 4/1989 2101.2 0.7572 30 535 5/1989 2421.2 0.8725 31 504 ; 6/1989 1764.2 0.6358 30 474 l 7/1989 2573.6 0.9274 31 443 l 8/1989 2183.4 0 7868 31 412 9/1989 1484,7 0.5350 30 382 l 10/1989 1736.6 0.6258 31 351 11/1989 2772.4 0.9991 30 321 12/1989 2512.2 0.9053 31 290 1/1990 2764.8 0.9963 31 259 2/1990 2770.4 0.9983 28 231 3/1990 2526.6 0.9105 24 207 NOTE: Reference Power - 2775 MW t I I i '( G l 6-24
TABLE 6 MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated Reaction Monitor and Activity Activity Rate Axial location (dis /sec-um) (dis /sec-am) IR_PS/ NUCLEUS) Cu-63 (n o) Co-60 Top 2.01 x 10 5 5.06 x 10 5 Middle 1.99 x 10 5 5.01 x 10 5 ] Bottom 2.08 x 10 5 5.24 x 10 5 l Average 2.03 x 10 5 s,33 x jo5 7.79 x 10-17 fe-54(n,p) Mn-54 l Top 2.31 x 10 6 4,99 x jo 6 l Middle 2.26 x 10 6 4.88 x 10 6 j I Bottom 2.40 x 10 6 5.18 x 10 6 Average 2.32 x 10 6 5.02 x 10 6 8.02 x 10-15 Ni-58 (n,p) Co-58 Top 9.22 x 10 6 7.67 x 107 Middle 9.01 x 10 6 7.50 x 10 7 Bottom 9.73 x 10 6 8.10 x 10 7 Average 9.32 x 10 6 7.75 x 10 7 1.11 x 10-14 V-238 (n,f) Cs-137 (Cd) Middle 1.05 x 10 6 g,99 x jo6 6.59 x 10-14 6-25
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd-Measured Saturated Reaction
-Monitor and Activity Activity Rate Axial location (dis /sec-om) (dis /see-am) (RPS/ NUCLEUS)
Np-237(n,f) Cs-137 (Cd) Middle 8.41 x 106 8.00 x 107 4.84 x 10-13 Co-59 (n,6) Co-60
- l t
Top 3.72 x 107 9.37 x 107 Middle 3.71 x 107 9.34 x 107 Bottom 3.53 x 107 8.89 x 107 I Average 3.68 x 107 9.27 x 107 6.04 x 10-12 l Co-59 (n,a) Co-60 (Cd) Top 2.30 x 107 5.79 x 107 Middle 2.30 x 107 5.79 x 107 Botton 2.22 x 107 5.59 x 107 Average 2.27 x 107 5.73 x 107 3.74 x 10-12 S i. 6-26
TABLE 6-9
SUMMARY
OF NEUTRON 00SIMETRY RESVLTS TIME AVERAGED EXPOSVRE RATES 2
& (E > 1.0 MeV) {n/cm -sec) 1.55 x 1011 8%
p (E > 0.1 MeV) (n/cm2-sec) 7.76 x 1011 16% dpa/sec 3.22 x 10-10 1 11% 2
$ (E < 0.414 eV) {n/cm -sec) 8.87 x 1010 i 23%
INTEGRATED CAPSULE EXPOSURE 2 4 (E > 1.0 MeV) {n/cm } 2.46 x 1019 i 8% !, 2 4 (E > 0.1 MeV) {n/cm ) 1.23 x 1020 16% dpa 5.11 x 10-2 g 33g l 2 4 (E.< 0.414 eV) (n/cm } 1.41 x 1019 23% NOTE: Total Irradiation Time - 5.03 EFPY 6-27
l TABLE 6-10 l COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE cf. "'f e U!CE CAPSULE CENTER Adjusted Reaction tig.asured Lalculation [jj Cu-63 (n,a) Co-60 7.79x10-17 7.93x10-17 1.02 Fe-54 (n,p) Mn-54 8.02x10-15 7.92x10-15 0.99 Ni-58 (n,p) Co-58 1.lix10-14 1.10x10'l4 0.99 U-238 (n,f) Cs-137 (Cd) 4.56x10-14 4.58x10'l4 1.00 Np-237 (n,f) Cs-137 (Cd) 4.84x10'l3 5.04x10-13 1.04 Co-59 (n,0) 00-60 (Cd) 3.74x10-12 3.77x10-12 1.01 Co-59 (n,B) Co-60 6.05x10-12 5.97x10-12 0.99 S 9 6-28
TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy Energy AdjusgedFlux Adjusged Flux (n/cm -sec) Group (Mev) (n/cm -secj Group (Mev) 1 1.73x10 I 1.12x10 7 28 9.12x10-3 2.96x1010 2 1.49x10 I 2.51x10 7 29 5.53x10-3 3.45x10 10 3 1.35x10 1 9.48x10 7 30 3.36x10-3 1.16x1010 4 1.16x10 1 2.15x108 31 2.84x10-3 1.17x10 10 5 1.00x10 1 4.80x10 8 32 2.40x10-3 1,18x10 10 8 2.04x10-3 3.46x1010 6 8.61x100 8.26x10 33 7 7.41x100 1.92x109 34 1.23x10-3 2.94x1010 8 6.07x10 0 2.76x109 35 7.49x10-4 2.57x10 10 9 4.97x100 5.90x109 36 4.54x10-4 2,44x10 19 10 3.68x10 0 7.87x109 37 2.75x10-4 2.59x10 10 11 2.87x10 0 1.63x10 10 38 1.67x10~4 2.89x1010 12 2.23x10 0 2,28x10 10 39 1,0lx10-4 2.88x1010 13 1.74x10 0 3,26x1010 40 6.14x10-5 2.84x10 10 14 1.35x10 0 3.82x10 10 41 3.73x10-5 2.74x10 10 15 1,lix10 0 7.33x10 10 42 2.26x10-5 2.62x1010 lci 8,21x10-1 8,73x10 10 43 1.37x10-5 2.49x1010 17 6.39x10~l 9.27x10 10 44 8.32x10-6 2.30x10 10 18 4.9Bx10'l 7.23xlL40 45 5.04x10-6 2.00x10 10 19 3.88x10-1 1.06x10 10 46 3.06x10-6 1.74x10 10 20 3.02x10-I 9.63x1010 47 1.86x10-6 1.50x1010 21 1,83x10'I 1,0lx10ll 48 1.13x10-6 1.ilx10 10 22 1.11x10~1 8.42x10 10 49 6.83x10-7 1.27x10 10 23 6,74x10-2 5.33x1010 50 4,14x10-7 1,48x10 10 24 4,09x10-2 2.74x10 10 51 2.51x10-7 1,48x10 10 25 2,55x10-2 4.24x1010 52 1, .52x10-7 1.40x10 10 26 1.99x10-7 1.50x1010 53 9.24x10~8 4,50x10 10 27 1.50x10-2 1,58x10 10 NOTE: Tabulated energy levels represent the upper energy of each group. 6-29 l
TABLE 6-12 COMPARISON OF CALCVLATED AND MEASVRED
~
EXPOSURE LEVELS FOR CAPSULE X (pleulated Measured (fB , 2 f(E > 1.0 MeV) {n/cm ) 2.22 x 1019 2.46 x 1019 0.90 2 f(E > 0.1 MeV) {n/cm ) 3,33 x io20 1.23 x 1020 0.96 dpa 4.81 x 10-2 5.11 x 10-2 0.94 2 f(E < 0.414 eV) {n/cm } 4.11 x 1018 1.41 x 1019 0.29 o 6-30
TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCA110NS i ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR V.C. SUMMER UNIT 1 A7IMUTHAL ANGLE 0*(a) _12 ' . _29_d'._ _30' 45' 5.03 EFPY 4-(E>1.0 MeV) 6.61 x 10 4.63 x 10 4.17 x 10 3.21 x 10 2.17 x 10 2 (n/cm ) f(E>0.1 MeV) 1.67 x 10 ' l.14 x 10 8.84 x 10 6.67 x 10 4.47 x 10 ' 2 (n/cm )
~ ~ ~
dpa 1.06 x 10' 7.42 x 10 6.47 x 10~ 4.97 x 10 3.39 x 10 14.0 EFPY 4-(E>1.0 MeV) 1.73 x 10 " 1.26 x 10 ' 1.17 x 10 8.93 x 10 5.86 x 10 2
. (n/cm )
4-(E>0.1 MeV) 4.36 x 10 3.10 x 40 2.49 x 10 1.86 x 10 1.21 x 10 2 (n/cm )
~ ~
dpa 2.7' x 10 2.01 x 10' l.82 x 10 1.38 x 10~ 9.14 x 10~ 32.0 [FPY 4-(E>1.0 MeV) 3.87 x 10 ' 2.85 x 10 ' 2.69 x 10" 2.04 x 10 1.33 x 10 2 (n/cm ) 4>(E>0.1 MeV) 9.75 x 10 7.02 x 10 ' 5.71 x 10 4.25 x 10 2.73 x 10" 2 (n/cm )
~ ~ ~ ~
dpa 6.19 x 10 4.57 x 10' 4.18 x 10 3.17 x 10 2.07 x 10 (a) Maximum point on the pressure vessel 6-31
5 [ TA8LE 6-14 ' NEUTRON EXPOSURE VALUES.FOR USE IN THE GENERATION OF HEATUP/000LDOWN CURVES 14 EFPY , NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE [ 2 (n/cm ) (equivalent n/cm2) i ! Surface 1/4 T 3/4 T Surface 1/4 7 3/4 T [ [ 0*(a) 1.73 x 10I9 9.84 x 10 I8 2.25 x 1018 1.73 x 10I9 1.14 x 10 I9 4.08 x 10 18 ; !' 12* 1.26 x 10I9 7.23 x 1018 1.74 x 1018 1.26 x 10 I9 8.42 x 1018 3.18 x 10 18 ,
- 20.5* 1.17 x 10I9 6.72 x 10 18 1.59 x 1018 1.17 x 10 I9 7.63 x 10I8 2.77 x 10 I8 l 30* 8.93 x 10 18 5.14 x 1018 1.23 x.10 I8 8.93 x 10 18 5.81 x 10 18 2.11 x 10 18 45* 5.86 x 1018 3.39 x 1018 8.26 x 10I7 5.86 x 1018 3.83 x 1018 1.42 x 1018 l !
32 EFPY j NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dea SLOPE ! 2 (n/cm ) (equivalent n/cm2) ( l Surface IL4J 3/4 T Surface 1/4 T 3/4 T [ i l l 0*(a) 3.87 x 10 I9 2.20 x 10I9 5.03 x 10 I8 3.87 x 10 I9 2.55 x 10 I9 9.13 x 10 18 l l 12* 2.85 x 10 I9 1.64 x 10 I9 3.94 x.10 18 2.85 x 10 I9 1.91 x 10 19 7.19 x 10 18 l 20.5 2.69 x 10 I9 1.55 x 10I9 3.67 x 1018 2.69 x 10I9 1.76 x 10 I9 6.39 x 10I8 l 30* 2.04 x 10 l9 1.18 x 10 l9 2.82 x 10 18 2.04 x 10 19 1.33 x 10 I9 4.82 x 10 18 ; ! 45* 1.33 x 10 19 7.66 x 1018 1.87 x 10 18 1.33 x 10 I9 8.66 x 10 I8 3.22 x 10 I8 I L l l (a) Maximum point on the pressure vessel 6-32 .
. . . 8 w ---
i TABLE 6-15 t UPDATED LEAD FACTORS FOR V.C. SUMMER UNIT 1 SURVEILLANCE CAPSULES Capsule Lead Faciqr V 3.11 V 3.11 X 3.72 W 2.98 Z 2.98 Y 2.98 0 9 6-33
1 SECTION 7.0 ! , SVRVElLLANCE CAPSVLE REMOVAL SCHEDULE . l l The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the V. C. Summer Unit I reactor vessel: 1 Capsule Estimated Location Lead Fluen e Capsule (deg.) Factor Removal Time (a) (n/cm ) ) V 343 3.11 1.13 (Removed) 6.39 x 1018 V 107 3.11 2.93 (Removed) 1.47 x 1019 X 287 3.72 5.03 (Removed) 2.46 x 1019
. W 110 2.98 10.75 3.87 x 10l9(b)
Y 290 2.98 16.00 5.77 x 1019 2 340 2.98 Standby - (a) Effective full power years from plant startup. (b) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY). 4 7-1
SECTION 8.0 T<EFERENCES
~
- 1. - Yanichko and Davidson, " South Carolina Electric and Gas Company Virgil C.
, Summer Nuclear Plant Unit No.1. Reactor Vessel Radiation Surveillance Program," WCAP-9234, January 1978.
- 2. R. S. Boggs, et. al., " Analysis of Capsule U from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program", WCAP-10814, June 1985.
- 3. D. J. Colburn, et. al., " Analysis of Capsule V from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program", WCAP-ll726, January 1988.
- 4. Code of Federal Regulations,10CFR50, Appendix G, " fracture Toughness Requirements", and Appendix H, %eactor vessel Material Surveillance i
Program Requirements," U.S. Nuclear Regulatory Commission, Washington,
. D.C.
- 5. Regulatory Guide 1.99, Revision 2. " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988.
- 6. Section 111 of the ASME Boiler and Pressure Vessel Code, Appendix G,
" Protection Against Nonductile failure."
- 7. ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of ferritic Steels."
- 8. ASTM E185-82, " Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)."
- 9. ASTM E23-88, " Standard Test Methods for Notched Bar Impact Testing of-Metallic Materials."
8-1
REFERENCES continued
- 10. ASTM A370-89, " Standard Test Methods and Definitions for Mechanical Testing of Steel Products.*
- 11. ASTM E8-89, " Standard Test Methods of Tension Testing of Metallic Materials."
- 12. ASTM E21-79(1988), " Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."
- 13. ASTM E83-85, " Standard Practice for Verification and Classification of Extensometers."
- 14. R. G. Soltesz, R. K. Disney, J. Jedruch, tr.c ' L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, todating end input Data Preparation. Vol. S--Two-Dimensional Discrete Ordinates Transport ,
Technique", WANL-PR(LL)-034, Vol, S. August liitu.
- 15. "0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors *,
- 16. ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984,
- 17. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
8-2
REFERENCES continued
- 18. ASTM Designation E693-79, "Stardard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards. Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 19. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 20. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 21. ASTM Designation E261-77, Standard Method for Determining Neutron Flux, l= Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Mattrials, philadelphia, PA, 1984.
- 22. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques *, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 23. ASTM Designation E263-82, " Star.dard Method for Determining fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 24. ASTM Designation E264-82, " Standard Method for Determining f ast-Neutron Flux Density by Radioactivation of Nickel", in ASiM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
8-3
REf[RENCES continued
- 25. AS1H Designation E481-78, " Standard Method for Measuring Neutron-flux Density by Radioactivation of Cobalt and Silver *, in ASlM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, ,
1984. 26, ASTM Designation E523-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 27. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 29. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron flux Density by Radioactivation of Neptuntum-237", in ASTM Standards, Section ,
12 American Society for Testing and Materials, Philadelphia, PA,1984,
- 29. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance", in AS1H Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 30. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
- 31. W. N. McElroy, S. Berg and T. Crocket, elqtpputer- Automated Iterative Method of Neutron Flux Spectra Detennined by foil Activation, AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 32. EPRl-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981.
8-4
REFERENCES continued
- 33. R. J. Fabean, "The Nuclear Design of the Virgil C. Summer Power Plant -
Unit 1", WCAP-9685, . March,1980. (Proprietary)
- 34. R. M. Smith, "The Nuclear Design and Core Management of the Virgil C.
Summer Power Plant - Cycle 2", WCAP-10663, March,1985. (Proprietary)
- 35. M. A. Petrunyak, et al., "The Nuclear Design and Core Management of the Virgil C. Summer Power Plant - Cycle 3", WCAP-10874, March,1986.
(Proprietary)
- 36. M. A. Petrunyak, et al., "The Nuclear Design and Operations Package for l the Virgil C. Summer Power Plant - Cycle 4", WCAP-ll430, June,1987.
(Proprietary)
, 37. M. A. Petrunyak, et al., "The Nuclear Design Report for the Virgil C.
Summer Power Plant - Cycle 5", WCAP-11990, October,1988. (Proprietary)
- 38. R. M. Smith, et al., "The Nuclear Design Report for the Virgil C. Summer Power Plant - Cycle 6", WCAP-12564, June,1990. (Proprietary) 9 8-5
APPENDIX A LOAD-TlHE RECORDS This appendix contains the load-time records for the individual instrumented charpy specimens. Load-time records were not available for specimens CW50, CW48 and CH53 due to computer and machine malfunctions. This has no effect on the final results of this report. O A-0
jl ' i!!I{t ff -
!! [1!<t!!;it[ [t.L!i :f! ii e { ;t!
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=
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- p P,
D - A O y
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L y r r o M c U e M r I X e m A E i M t M I d f T a p o P l
= [ i I t I I I I i d
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!> I-{
- e z
i l a e d-I s . g 1 A-e r u g i
, n F
_ W a m t M E D-N A - E O - G L
= D =
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i l l 8,,1 t e,t CD
+
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4
+ <
LAJ
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. t~. + Lf.3 Ca
+ ,,
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- + + gg .3 c.+.3 ,,
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+t:a.t<3 a rw. Lrm P9 i:sJ cm
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APPENDIX B ' HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION B-1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTND7 (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced ARTNDT-RTNDT is designated as the higher of either the drop weight nil-ductil ty transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.
, RTNDT increases as the material is exposed to fast-neutron radiation.
Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)(B-ll. Regulatory Guide 1.99, Revision 2 is used for the calculation of RTNDT values at 1/4T and 3/4T locations (T is the thickness of the vessel at the beltline region). B-2. FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory
~
Standard Review Plan [B-2] . The pre-irradiation fracture-toughness properties of V. C. Summer Unit 1 of the reactor vessels are presented in Table B-1. B-1
B-3, CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup
~
and cooldown rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or , cooldown cannot be greater than the reference stress intensity factor, KIR> for the metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code [B-3). The KIR curve is given by the following equation: KIR = 26.78 4 1.223 exp [0.0145 (T-RTNDT + 160)] (1) where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT , Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code [B-3) as follows: CKIM + KIT 5KIR (2) where KIM = stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RTNDT of the material C = 2.0 for level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical B-2
At any time during the heatup or cooldown transient, KIR is determined by the l metal temperature at the tip of the. postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed. From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of j the flaw is always at the inside of the wall because the thermal gradients l produce tensile stresses at the inside, which increase with increasing cooldown , rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, 1 composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. 1 During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K ip exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on
. temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various B-3
intervals along a cooldown ramp. The use of the composite curve eliminates i this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for , finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The inetal temperature at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4 T crack during heatup is lower than the KIR for the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR'S do not offset each other, and the pressure- temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, , both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the niculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady , state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the B-4
allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the 1983 Amendment to 10CFR50[B-4) has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RTNDT by at least 120'F for normal operation when the l pressure exceeds 20 percent of the preservice hydrostatic test pressure. Table B-1 indicates that the initial RTNDT of 10'F occurs in the vessel flange of V. C. Summer Unit 1, so the minimum allowable temperature of this j region is 130'F. These limits are shcan in figures B-1 and B-2 whenever
, applicable.
l B-4. HEATUP AND COOLDOWN LIMIT CURVES I Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in section B-3. Figure B-1 contains the heatup curves for 50 and 100'F/hr. Figure B-2 contains the cooldown curve up to 100*F/hr. Both Figures B-1 and B-2 are applicable for the first 14 EFPY of operation. Margins of 10' and 60 psig are included in these two figures to allow for possible instrumentation errors. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures B-1 and B-2. This is in addition to other criteria which must be met before the reactor is made critical. The leak limit curve shown in Figure B-1 represents minimum temperature requirements at the leak test pressure specified by applicable codes (B-2,B-3], B-5
The leak test limit curve was determined by methods of References B-2 and B-4. l The criticality limit curve shown in Figure B-1 , specifies pressure- ' temperature limits for core operation to provide additional margin during actual power production as specified in Reference B-4. The pressure- , temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum pressure-temperature curve for heatup and cooldown calculated as described in Section B-3. The maximum temperature for the inservice hydrostatic test for the V.C. Summer reactor vessel is 237'F. A vertical line at 237'F on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. Figures B-1 and B-2 define limits for ensuring prevention of nonductile f ailure for the V. C. Summer Unit I reactor vessel. , B-5. ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2 [B-1] the adjusted reference temperature (ART) for each material in the beltline is given by the following expression: ART = Initial RTNDT + ARTNDT + Margin (3) Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NS-2331 of Section 111 of the ASME Boiler and pressure Vessel Code. If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class. B-6
ARTNDT is tiis mean value of the adjustment in reference temperature l caused by irradistion and should be calculated as follows: ART NDT = [CI)f(0.28-0.10 log f) (4) To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth. f(depth X) " fsurface(e' ) (5) where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then put into equation (4) to calculate ARTNDT at the specific depth. CF (*F) is the chemistry factor, obtained from Reference B-1. All materials in the beltline region of V. C. Summer Unit I were considered for the limiting material . _ RTNDT at 1/4T and 3/4T are summarized in Table B-2. From l Table B-2, it can be seen that the limiting material is lower shell for heatup and cooldown curves applicable up to 14 EFPY, A sample calculation for RTNDT is shown in Table B-3. B-7
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-20 Vessel flange (b) - -
10 Intermediate Shell Plate, A9154-1 .10 .51 30 Intermediate Shell Plate, A9153-2 .09 .45 -20 Lower Shell Plate, C9923-1 .08 .41 10 Lower Shell Plate, C9923-2 .08 .41 10 Longitudinal and Circumferential Welds .06 .89 -4d 1
- a. The initial RTNOT (I) values for the plates and welds are measured values. .
- b. To be used for considering flange requirements for heatup/cooldown curves [B-4), the values are from Table A-1 in Reference (B-5),
l-9 ( l I I B-8
TABLE B-2
SUMMARY
OF ADJUSTED REFERENCE TEMPERATURE (ART) AT 1/4T and 3/4T LOCATION 14 FFPY RTNDT AT Component 1/4T ('F) 3/4T (*F) Intermediate Shell, A9154-1 87 77
. Intermediate Shell, A91$3-2 73 SB Lower Shell, C9923-1 96* 83*
l Lower Shell, C9923-2 96* 83* Intermediate to lower Shell 8 1
. Longitudinal Welds ,
Circumferential Weld 19 10 , *- These RTNDT numbers used to generate heatup and cooldown curves applicable up to 14 EFPY. 0 B-9
TABLE B-3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING V.C. SUMMER UNIT 1 REACTOR VESSEL MATERIAL - LOWER SHELL Reaulatory Guide 1.99 - Revision 2 14 EFPl P3r3m_e_in 1/4 T 3/4 T Chemistry Factor, CF ('F) 51 51 Fluence, f (10 19 n/cm2 )(a) 1.08 0.43 , Fluence Factor, ff 1.023 0.765 ARTNDT - CF x ff (*F) 52 39 Initial RTNDT, I ('F) 10 10 Margin, M ('F) (b) 34 34 . Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 96 83 ART = Initial RTNDT + ARTNDT + Margin (a) Fluence, f, is based upon fsurf (10 19 n/cm2 , E>l Mev) = 1.73 at 14 EFPY. The V. C. Summer Unit I reactor vessel wall thickness is 7.75 inches at the beltline region. (b) Margin is calculated as, M - 2 (oj 2 s y 23 0.5 The standard deviation for the initial RTNDT margin term (oi) is as. aed to be O'F since the inititl RT NDT is a measured value. The , standard deviation for ART NDT' (06 ) is 17'F for base metal, except that oA need not exceed 0.50 times the mean value of ARTNDT* { B-10 1 1
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(, up rates up to 50 and 100'F/hr) Applicable for the First 14 i EFPY (With Margins 10'F and 60 psig for Instrumentation Errors) B-11
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B-6. REFERENCES B-l' Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May,1988. B-2 " Fracture Toughness Requirements," Branch Technical Position MTED 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981. B-3 A1ME Boiler and Pressure Vessel Code, Section Ill, Division 1 - Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure," pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York,1986. B-4 Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements," V.S. Nuclear Regult. tory Commission, Washington, D.C., Federal Register, Vol . 48 No.104, May 27,1983. B-5 WCAP-10814, " Analysis of Capsule U from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program", R. S. Boggs, et al, June 1985. B-13
ATTACHMENT 1 DATA P0lHTS FOR HEATUP AND C00LDOWN CURVES (With Margins 10'F and 60 psig for Instrumentation Errors) . W e B-14 l
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