ML20070S731
| ML20070S731 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 12/31/1990 |
| From: | Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML16341G030 | List: |
| References | |
| WCAP-12811, NUDOCS 9104030004 | |
| Download: ML20070S731 (166) | |
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ERRATA WCAP-12811 Page 4-8 Figure 4-2 is a simplified version of earlier drawings of the surveillance capsule design; e.g., figure 2-5 on pages 2-7/2-8 of WCAP-8783, Reference 1 to this report, in tl. ,
simplification process, the actual positions of the three thermal monitors and neutron dosimeters were inadvertently omitted from Figure 4-2. See the earlier drawings for the actual positions.
Page 6-26 The time averaged exposure rate for the lowest energy range of the neutron spectrum in Table 6-9 has an error in its exponent. The correct value for the flux is 8.17 x 10"'.
I o
. . WlST1NGh10VSECLASS3 WCAP-12811 ANALYSIS OF CAPSULE X FRur. THE PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM E. Terek S. L. Anderson L. Albertin
. December 1990 Work Performed Under Shop Order LSFP-106 Prepared by Westinghouse Electric Corporation for the Pacific Gas and Electric Company Approved by: -
/2M67 9-4 T. A. Meyer, Manhger Structural Reliability and Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 0 1990 Westinghouse Electric Corp.
[- ----- _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _____
PREFACE
. This report has been technically reviewed and verified.
i Reviewer l Sections 1 through 5, 7, and 8 J. M. Chicots be' L- I Section 6 E.P.Lippinco[_ f 1
i i
e I
l l 4 l
L
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l l
l TABLE OF CONTENTS 1
, Seetion Iltlg Eigg j l
l.0
SUMMARY
OF RESULTS 1-1
2.0 INTRODUCTION
2-1
3.0 BACKGROUND
3
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING 0F SPECIMENS FROM CAPSULE X 5-1 5.1 Overview 5-1 S.2 Charpy V-Notch Impact Test Results 5-4 S.3 Tension Test Results 5-6
- 5.4 Compact Tension Tests 5-7 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY- 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-7 7.0 SVRVEILLANCE CAPSVLE REMOVAL SCHEDULE 7-1
- 8. 0' REFERENCES 8 APPENDIX A - Load-Time Records for Charpy Specimen Tests D
11
)
a.. ,
LIST OF ILLVSTRATIONS
. Elaung Title Eagg 4-1 Arrangement of Surveillance Capsules in the Diablo Canyon 4-7 Unit 2 Reactor Vessel 4-2 Capsule X Diagram Showing Location of Specimens, Thermal 4-8 Monitors and Dosimeters 5-1 Charpy V-notch Impact Properties for Diablo Canyon Unit 2 5-15 Reactor Vessel Intermediate Shell Plate B5454-1 (Longitudinal Orientation) 5- 2 Charpy V-notch Impact Proper + tes for Diablo Canyon Unit 2 5-16 Reactor Vessel Intermediate Shell Plate B5454-1 (Transverse Orientation) 5-3 Charpy V-notch Impact Properties for Diablo Canyon Unit 2 5-17 Reactor Vessel Weld Metal 5-4 Charpy V-notch Imp. set Properties for Diablo Canyon Unit 2 5-18 Reactor Vessel Weld Heat Affected Zone Metal 5-5 Charpy impact Specimen Fracture Surfaces for Diablo Canyon 5-19 Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 (Longitudinal Oriee.ation) r 5-6 Charpy impact Specimen Fracture Surfaces for Diablo Canyon 5-20 Unit 2 Reactor Vessel Intermediat? Shell Plate B5454-1 (Transverse Orientation) iii
._ _ _ _ _ w
,u, s e e t t LIST OF ILLUSTRATIONS (Cont)
[.iourt Title PAgg .
5-7 Charpy impact Specimen Fracture Surfaces for Diablo Canyon 5-21 Unit 2 Reactor Vessel Weld Metal 5-8 Charpy impact Specimen Fracture Surfaces for Diablo Canyon 5-22 Unit 2 Reactor Vessel Weld Heat Affected Zone (HAZ) Metal 5-9 lensile Properties for Diablo Canyon Unit 2 Reactor Vessel 5-23 Intermediate She'l Plate B5454-1 (Longitudinal Orientation) 5-10 Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel 5-24 Intermediate Shell Plate B5454-1 (Transverse Orientation) 5-11 Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel 5-25 Weld Metal 5-12 Fractured Tensile Specimens from Diablo Canyon Unit 2 5-26 Reactor Vessel Intermediate Shell Plate B5454-1 (Longitudinal Orientation) 5-13 Fractured Tensile Specimens from Diablo Canyon Unit 2 5-27 Reactor Vessel Intermediate Shell Plate B5454-1
, (Transverse Orientation) 5-14 Fractured Tensile Specimens from Diablo Canyon Unit 2 5-28 Reactor Vessel Weld Metal iv C--.
LIST OF ILLUSTRATIONS (Cont)
. Fioure Title EAgg 5-15 Engineering Stress-Strain Curves for Diablo Canyon Unit 2 5-29 Reactor Vessel Intermediate Shell Plate 85454-1 Tension Specimens PL10 and PLll 5-16 Engineering Stress-Strain curves for Diablo Canyon Unit 2 5-30 l Reactor Vessel Intermediate Shell Plate B5454-1 Tension Specimens PL12 and PTIO 5-17 Engineering Stress-Strain Curves for Diablo Canyon Unit 2 5-31 l Reactor Vessel Intermediate Shell Plate B5454-1 Tension Specimens PTll and PT12 5-18 Engineering Stress-Strain Curves for Diablo Canyon Unit 2 5-32 Reactor Vessel Weld Metal Tension Specimens PW10 and PWil 5-19 Engineering Stress-Strain Curves for Diablo Canyon Unit 2 5-33 Reactor Vessel Weld Metal Tension Specimen PW12 5-20 True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor 5-34 j
Vessel Intermediate Shell Plate B5454-1 Tension Specimens
! PL10 and PTll 5-21 True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor 5-35 Vessel Intermediate Shell Plate B5454-1 Tension Specimens 1- PL12 and PTIO v
l
LIST OF ILLUSTRATIONS (Cont)
Fiaure T.tlle EiLqg .
5-22 True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor 5-36 Vessel Intermediate Shell Plate B5454-1 Tension Specimens PTll and PT12 5-23 True Stress-Striin Curves for Diablo Canyon Unit 2 Reactor 5-37 Vessel Weld Metal Tension Specimens PW10 and PWil 5-24 True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor 5-38 Vessel Weld Metal Tension Specimen Specimen PW12 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 Diablo Canyon Unit 2 1
l i
t O
v1 L
i LIST Of IABLES
. Table lith P_Lqst 4-1 Chemical Composition and Heat Treatment of the Diablo 4-4 Canyon Unit 2 Reactor Vessel Surveillance Materials i 4-2 Chemical Composition of Diablo Canyon Unit 2 Capsule X 4-5 Irradiated Charpy Impact Specimens 4-3 Chemistry Results from the NBS Certified Reference 4-6 Standards 5-1 Charpy V-Notch Impact Data for the Diablo Canyon Unit 2 5-8 Plate B5454-1 Irradiated at 550*F, Fluence 8.87 x 10 18 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Diablo Canyon Unit 2 5-9 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 8.87 x 10 18 n/cm2 (E > 1.0 MeV) 5-3 Instrumented Charpy Impact Test Results for Diablo Canyon 5-10 Unit 2 Shell Plate B5454-1 Irradiated at 550*F, Fluence 8.87 x 10 18 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results for Diablo Canyon 5-11 Unit 2 Weld Metal and HAZ Metal Irradiated at 550'F, Fluence 8.87 x 10 18 n/cm2 (E > 1.0 MeV) 5-5 Effect Irradiation of 550'F Irradiation to 8.87 x 10 18 5-12 (E > 1.0 MeV) on Notch Toughness Properties of Diablo Canyon Unit 2 Reactor Vessel Surveillance Materials v11
4 e a s LIST OF TAGLES (Cont)
Table Title Etqe .
5-6 Comparison of Diablo Canyon Unit 2 Surveillance Material 5-13 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-7 Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel 5-14 Surveillance Material Irradiated at 550'F to 8.87 x 1018 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 l Pressure Vessel Clad / Base Metal Interface 1
6-3 Relative Radial Distributions of Neutror Flux 6-17 (E > 1.0 MeV) within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-18 (E > 0.1 MeV) within the Pressure Vessel Wall I
j 6-5 Relative Radial Distributions of Iron Displacement Rate 6-19 (dpa) within the Pressure Vessel Wall i
6-6 Nuclear Parameters for Neutron Flux Monitors 6-20 6-7 Irradiation History of Neutron Sensors Contained in 6-21 -
Capsule X viii
LIST OF TABLES (Cont)
Table Title EAng 6-8 Measured Sensor Activities and Reactions Rates 6-24 6-9 Summary of Neutron Dosimetry Results 6-26 6-10 Comparison of Measured and Ferret Calculated Reaction 6-27 Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-28
$ Capsule Center 6-12 Comparison of Calculated and Measured Exposure Levels 6-29 for Capsule X 6-13 Neutron Exposure Projections at Key Locations on the 6-30 Pressure Vessel Clad / Base Metal Interface 6-14 Neutron Exposure Values for use in the Generation of 6-32 Heatup/Cooldown Curves 6-15 Updated Lead Factors for Diablo Canyon Unit 2 6-34 Surveillance Capsules e
e ix
SECTION 1.0
SUMMARY
Of RESULTS
, The analysis of the reactor vessel materials contained in surveillance Capsule X, the second capsule to be removed from the Pacific Gas and Electric Company
- Diablo Canyon Unit 2 reactor pressure vessel, led to the following conclusions:
o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 8.87 x 1018 n/cm2 af ter 3.11 EFPY of plant operation, o Irradiation of the reactor vessel intermediate shell plate 85454-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (Longitudinal Orientation),
to 8.87 x 1018 n/cm2 (E > 1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 101 and 121 'F, respectively.
o Irradiation of the reactor vessel intermediate shell plate B5454-1 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (Transverse Orientation), to 8.87 x 10 18 n/cm2 (E > 1.0 MeV) resulted in a 30 and 50 ft-lb transition temperature increase of 99 'f.
o Weld metal Charpy specimens irradiated to 8.87 x 1018 n/cm2 (E >
1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 204 and 214 'F, respectively, o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal
< Charpy specimens to 8.87 x 10 18 n/cm2 (E > 1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 252 and 247
'f, respectively.
o Irradiation of the intermediate shell plate to 8.87 x 1018 n/cm 2 (E > 1.0 MeV) resulted in a decrease of 30 ft-lbs (longitudinal orientation) and a decrease of 10 ft-lb (transverse or;entation) in upper shelf energy.
1-1
o The average upper shelf energy of the weld metal showed a decrease in energy of 47 f t-lbs af ter irradiation to 8.87 x 1018 n/cm2 (E >
1,0 MeV).
o Both the weld and plate metal Charpy test results are in close agreement with Regulatory Guide 1.99, Revision 2 predictions, e o
The calculated End-of-Life (EOL) (32 EFPY) maximum neutron fluence (E
> 1.0 MeV) for the Diablo Canyon Unit 2 reactor vessel is as follows:
Vessel inner radius * - 1.70 x 1019 n/cm 2 Vessel 1/4 thickness - 9.14 x 1018 n/cm 2 Vessel 3/4 thickness - 1.84 x 1018 n/cm 2
- clad / base metal interface o The surveillance capsule data presented in this report may be considered in the development of heatup and cooldown curves, however, in selecting the limiting material (s), the entire belt line '
components should be considered (Reference 5).
l l
l i
s G
l-2
o .
SECTION
2.0 INTRODUCTION
O This report presents the results of the examination of Capsule X, the second
- capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Pacific Gas and Electric Company Diablo Canyon Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Diablo Canyon Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties s' the reactor vessel materials is presented by Davidson and Yanichko in Refers .e 1. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based i 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 Hot Cell. The l postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed at the remote metallographic hot cell facility.
This report summarizes the testing of and the postirradiation data obtained from surveillance Capsule "X" removed from the Diablo Canyon Unit 2 reactor vessel and discusses the analysis of these data.
2-1
g=. t SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel contiining the reactor core and 1 - its primary coolant io resist fracture constitutes an important factor in ensuring safety 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 pmssure vessel steels such as SA533 Grade B Class 1 (base material of the -
Pacific Gas and Electric Company Diablo Canyon Unit 2 roactor pressure vessel intermediate shell plate B5454-1) 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 111-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)-
. RTNDT_is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-20Bl73) 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 cu'rve (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 Ki p curve, allowable stress intensity factors can-be obtained for this material as a function of- temperature. Allowable operating limits etn then be -
determined using these allowable stress intensity factors.
3-1 i
A-.- a - , , , , . , , ,,--.,c , i,,, .,e, , , , , . . , , . , . - -- ,__ _
i 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 1
given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Programill, in which a surveillance capsule is periodically removed from the I operating nuclear reactor and the encapsulated specimens 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 curve and, in turn, to I set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
l l
l l
l l .
l 3-2
SECTION
4.0 DESCRIPTION
OF PROGRAM Six rurveillance capsules for monitoring the effectr of neutm exposure on the Diablo Canyon Unit 2 reactor pressure vessel core region natorial were inserted in the reactor vessel prior to initial plant start-up. The six capsules were pos stioned in the reactor vessel between the neutron 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 after 3.11 effective full power years of piant operation. This capsule contained Charpy V-r.ctch, tensile, 1/2 T compact tension (CT), and bend bar specimens (see Figure 4-2) from intermediate shell plate B5454-1. Capsule X, also, .:entaine:1 Charpy V-notch, tensile, and 1/2 T compact tension (CT) specimens from weldment which was made from sections of intermediate shell plate B5454-1 and adjoining intermediate shel' ' late B5454-2 using weld wire representative of that used in the original fabrication and
~
Charpy V-notch specimens from weld Heat-Affected-Zone (HAZ) material. All weld HAZ specimens were obtained from within the weld HAZ of intermediate shell
~
l plate B5454-1.
The chemical composition and heat treatment of the surveillance material is presented in Table 4-1. The chemical analyses reported in Table 4-1 were obtained frz. unirradiated material used in the surveillance program. In addition, a chemical analysis using Inductively Co" 'ed Piasma Spectrometry (ICPS) was performed on one irradiated Charpy spec. men from intermediate shell plate B5454-1 and three weld metal Charpy specimens and is reported in Table j 4-2. The chemistry results from the NBS certified reference standards are l reported in Table 4-3.
4-1 l
J
l Test material was obtained from the intermediate shell course plate af ter thermal heat treatment and forming of the plate. All test specimens were machined from the 1/4 thickness location of the plate after performing a simulated postweld stress-relieving treatment on the test material. The test .
specimens represent material taken at least one plate thickness (9 5/8 inches) from the quenched end of the plate. -
Base metal Charpy V-notch impact specimens from intermediate shell plate B5454-1 were machined in both the longitudinal orientation (longitudinal axis of ti c! specimen parallel to the major working direction of plate B5454-1) and transverse orientation (longitudinal axis of the specimen perpendicular to the major working direction of plate B5454-1). The core region weld Charpy V-notch impact specimens were machined from the weldment such that the long dimension of the Charpy specimen was normal to the weld direction; the notch was machined such that the direction of crack propagation in the specimen will be in the weld direction, 1
Base metal tension specimens from the intermediate shell plate B5454-1 were machined so as to produce ;ome with the longitudinal axis of the specimen normal to and some perpendicular to the major rolling direction of the plate.
The core region weld tension specimens were machined from the weldment such that the long dimension of the specimen was oriented normal to the weld direction.
l The bend bar specimen, contained in Capsule X, was machined from plate B5454-1 with the longitudinal axis of the specimen oriented normal to the rolling direction of the plate such that the simulated crack would propagate in the l rolling direction of the plate. The bend bar specimen was fatigue precracked according to ASTM E399.
1 4-2
The 1/2T Compact Tension (CT) test specimens in Capsule X from intermediate shell plate B5454-1 were machined in both the longitudinal and transverse orientations. Thus, these CT specimens will generate fracture toughness data
- , both normal and parallel to the rolling direction of plate B5454-1. CT specimens from the weld metal were machined normal to the weld direction, with
. the notch oriented in the direction of the weld. All 1/2T CT specimens were fatigue precracked according to ASTM E399.
Capsule X contained dosimeters of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmiuni shielded dosimeters of neptunium (Np237) and uranium (U238) were included to measure the integrated flux at specific neutron energy levels.
The capsule contained two low-melting-point eutectic alloy thermal monitors.
These thermal monitors were used to more accurately define the maximum temperature attained by the test specimens during irradiation. The thermal monitors were sealed in Pyrex tubes and inserted in spacers located at three axial locations throughout the capsule The composition of the two low-melting-point eutectic 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.
4-3
TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE DIABLO CANYON UNIT 2 Rt' ACTOR VESSEL SURVEllLANCE MATERIALS Chemical Composition (wt%)*
Intermediate Shell Plate
- Element B5454-1 Weld Metal C 0.23 0,13 S 0,010 0,010 0,008 0,008 N2 to 0,002 0.012 Cu 0,15 0.22 Si 0.22 0,22 Mo 0.43 0,47 Ni 0.67 0,83 Mn 1.28 1,32 Cr 0.011 0.031 V 0.001 0,001 P 0,012 0.017 ,
Sn 0.010 0.010 Al 0,031 0.009
- Westinghouse Analysis HEAT TREATMENT Intermediate Shell Plate B5454-1 1550-1650*F 4 hr - Water Quench (Heat C5161-1) 1225i25 'F 4 hr - Air Cool 1150 25 'F 40 hr - Furnace Cool Weld 1150 25 'F 40 hrs '
4-4 l
l
_ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ )
t - 5 TABLE 4-2 CHEMICAL COMPOSITION OF DIABLO CANYON UNIT 2 CAPSULE X IRRADIATED CHARPY IMPACT SPECIMENS 0 Material ID. Diablo Canyon Unit 2: Irradiated Low Alloy Steel Specimen No. PT-56 PW-55 PW-56 PW-58 Metals Concentration in Weicht Percent Fe MATRIX ELEMENT: Remainder by Difference Co 0.003 0.018 0.004 0.010 Cu 0.137 0.225 0.213 0.225 Cr 0.070 0.043 0.071 0.043 Mn 1.226 1.432 1.355 1.391 Mo 0.453 0.545 0.484 0.506 Ni 0.656 0.875 0.856 0.877 P <0.0050 0.0131 0.0105 0.0118 V 0.002 0.006 0.009 0.006 C 0.237 0.152 0.134 0.066 S 0.0115' O.0054 0.0026 0.0025 Si 0.134 0.157 0.155 0.155 Analyses Method of Analysis Metals ICPS, Inductively Coupled Plasma Spectrometry Carbon EC-12, LEC0 Carbon Analyzer Sul fur Combustion / titration Silicon Dissolution / gravimetric Iron (Matrix Element: Remainder by Difference) 4-5
TA9tE 4-3
- CHEMISTRY RESULTS FROM THE NBS CERTlflED REFERENCE STANDARDS Material ID Low Alloy Steel: NBS Certified Reference Standards
__ NBS 361 NBS 362 Centified Measured (a) Certified Measured (a) ,
Metals Concentration in Weiaht Percent Fe
- 95.63 (matrix) 95.30 (matrix)
Co 0. 0.' 2 0.033 0.300 0.318 Cu 0.0 42 0.043 0.500 0.514 Cr 0.694 0.663 0.300 0.297 Mn 0.660 0.644 1.040 1.050 Mo 0.190 0.193 0.068 0.054 Ni 2.000 2.072 0.590 0.610 P 0.014 0.0144 0.041 0.0417 V 0.011 0.011 0.040 0.040 0 0.383 0.386 0.160 0.162/0.161 S 0.014 N/A 0.036 0.0354 Si 0.222 0.208 0.390 0.383 Material 10 Low Alloy Steel: NBS Certified Reference Standards NBS 363 NBS 364 Certified Measured (a) Certified Maasured (a)
Metals Concentration in Weiaht Percent Fe
- 94.40 (matrix) 96.70 (matrix)
Co 0.048 0.051 0.150 0.149 Cu 0.100 0.102 0.249 0.252 Cr 1.310 1.315 0.063 0.058 Ni 0.300 0.314 0.144 0.139 Mn 1.500 1.539 0.255 0.250 Ho 0.028 0.025 0.490 0.491 P 0.029 0.0285 0.010 0.0096 V 0.310 -----
0.105 0.100 C 0.620 -----
0.870 N/A S 0.0068 -----
0.0250 0.0247 Si 0.740 0.710 0.065 N/A
- Matrix element calculated as difference for material balance.
N/A - Not analyzed '
(a) Method of analysis -- Inductively Coupled Plasma Spectrometry (ICPS) for all elements except C, S and Si.
4-6 I
O. REACTOR VESSEL CORE BARREL 7 i NEUTRON PAD CAPSULE 3 1 56' l 56' J ' % e . * % ,<
270' l[f-
~"
\* _
'~
90' 7:.E \_
. 1 y i I I I X i W
180*
Figure 4-1. Arrangement of Surveillance Capsules in the Diablo Canyon Unit 2 Reactor Vessel 4-7 i
1
. ~ . .
t is 3
r.
.i
~1 1
4 .
M a
r s
l0 :
- i f
)--
i
.[;i 1
p -
a
,l lh . c.,na .i.. .o n.s ui im;;; im;;' c..., . c.e. . c.=,' imt;; imt;; <n..,
r: _ _.
' ' ..s , ..s , . .s . . .s .
..,, .. .... ,ri ,
. ,,, ..., . . . . .. . s . . , .
... ,e, ..s. ..s . ..s . ..s , . .s i ,0. ,o, .o. .o, ... ,
.. . .s . ..u em .m ..s r . .s r . ,
LEGEND: PH - HAZ PL - PLATE B5454-1 LONGITUDINAL PT'- PLATE B5454-1 TRANSVERSE PW - WELD METAL
= # %,
u... . _ _ _ _ . . _ _ . _ _ _ _ _
o
. . . , ..ni. un.a ,...., ....., ....., ....., c . . .. , );g;; ;;gg ,,,,,,,,
.. ,,,, ,,,, ,,,. ..., 1,.. . ... ,oi
....I ,,. , , , ,,,,- ,,,, .... l 4 .., .... ,,, ,s. ,,,, ,,,, , , . ,,,, .,,, ,,,, ,,,, ,,,, . . , ,,, ,,. .,,, ,,,, ,,,,- ....
,,,, ,,,. ,,,, ,,,, ,,,, .,,, ,, o ,,,. ,,,, .... ,,,, [ , , ,,
l S1 I Al'EITTURE CARD i'
\
Alio Ma;lable On
\pt r f ore Card a
f f'.gure 4-2 Capsule X Diagram Showing Location of Specimens, Thermal
% Honitors and Dosimeters TOA43m444-8
j SECTION 5.0 TESTING OF SPECIMENS FROM CAPSULE X
]
4 5.1 Overview 4
The port-trradd ation mechanical testing of the Charpy V-notch and tensile
. specir ns was performed at the Westinghouse Science and Technology Center with l consultation by Westinghouse Power Systems personnel. Testing was pert'crmed in accordance with 10CFR50, Appendices G and HI43, ASTM Specification 1 E185-82I83, and Westinghouse Procedure MHL 8402, Revision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.
L Upon receipt of the capsule at the laboratory, the specimens and spacer blocks l were carefully removed, inspected for identification number, and checked against the master list in WCAP-8783Ill. 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 1 monitor. Based on tt.is examination, the maximum temperature to which the test specisens1were exposed was less than 304'C (579'F). >
The Charpy impact tests were performed per ASTM Specification E23-88I93 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpv machine is instrumented with an Effects Technology i- Model 500 instr'umentation system. With this system, load-time and energy time-signals-can be recorded in addition to the standard measurement of Charpy
- i. energy (ED ). From the laid-time curva, the load'of_ general yielding _(Pay),
the time to genera 1' yielding (tcy), the maximum load (Pg), and the time to L
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 (Pr), and the L -load at which fast fracture terminated is identified as-the arrest load (PA)*
5 5-1
_ _ _ _ _ _ _ . _ _ _ . _ _ . . _ _ _ _ _ _ . _ . . _ . _ _ . . _ . _ _ _ _ ,_ _ _ _u _._,
_ - _. - ... ._. - -. ..__ _._-___-_ __-.=. _ ._.- .___ . _
The energy at maximum loa,1 (Eg) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen.
Therefore, the propagation energy for the crack ' ,) is the difference .
between the total energy to fracture (E )Dand the energy at maximum load.
4 The yield stress (oy) was calculated from the three-point bend formula having the following expression:
ay = Pay * (L / (B * (W-a)2 . c )) (i) where the constant C is dependent on the notch flank angle (4), 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 4 45' and 0 - 0.010", Equation 1 is valid with with C 1.21. Therefore (fo* L -
4W),
oy Pay * (L / (B (W-a)2
- 1.21)] - [3.3Pgy W)/[B(W-a)2] (2)
For the Charpy specimens. B 0.394 in., W = 0,394 in., and a - 0.079 in.
Equation 2 then reduces to:
oy 33.3 x Pgy (3) where oy is in units of psi and P gy is in units of lbs. The flow stress was 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 using the ratio-of-areas methods in compliance with ASTM Specification AJ70-89(10),
The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
5-2
Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-89billl and E21-79(1988)(123, 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.
Extension measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded te 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 E83-85(133 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.
Chromel-alumel thermocouples sere 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 configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range of room temperature to 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 i2'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.
53
a 5.2 Charov V-Notch Impici Test Results The results of Charpy V-notch impact tests performed on the various materials contained in Capsulo X irradiated to 8.87 x 10I8 n/cm2 (E > 1.0 MeV) are presented in Tables 5-1 through 5-4 and are compared with unirradiated resultslll as ,hown in figures 5-1 through 5-4. The transition temperature -
increases and upper shelf energy decreases for the Capsule X surveillance materials are summarized in Table 5-5.
Irradiation of the reactor vessel intermediate snell plate B5454-1 Charpy specimens to 8.87 x 10 18 n/cm2 (E > 1.0 MeV) at 550F (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 101 'T and a 50 ft-lb transition temperature incruse of 121 'F for specimens oriented with the longitudinal axis parallel to the major working direction of the plate (longitudinal orientation). This resulted in a 30 ft-lb transition temperature of 105.7 'f and a 50 f t-lb transition temperature of 154.6 'I for specimens oriented with the longitudinal axis parallel to the major working direction of the plate (longitudinal orientation).
The average upper shelf energy (USE) of the intermediate shell plate B5454-1 Charpy specimens (longitudinal orientation) resulted in a average upper shelf energy decrease of 30 ft-lbs after irradiation to 8.87 x 1018 n/cm2 (E >
1.0 MeV) at 550'F. This results in a average USE of 115 f t-lb (figure 5-1).
Irradiation of tht reactor vessel intermediate shell plate B5454-1 Charpy specimens to 8.87 x 10 18 n/cm2 (E > 1.0 MeV) at 550*f (figure 5-2) resulted in a 30 ft-lb transition temperature increase of 99
- and 50 ft-lb l transition temperature increase of 99 'f for specimens oriented w; h the j longitudinal axis perpendicular to the major working direction of the plate (transverse orientation). This resulted in a 30 ft-lb transition temperature l of 126 'T and a 50 ft-lb transition temperature of 173 'T for specimens oriented with the longitudinal axis perpendicular to the major working -
direction of the plate (transverse orientation).
t .
5-4
- .. .--.- _-. . - - - . _. - - - = _ - - -
4 l 14 y;c;vge upper shelf energy (USE) of the intermediate shell plate 85454-1 harpy specimens (Transverse Orientation) resulted in a average upper shelf i energy decrease of 10 ft-lbs after irradiation to 8.87 x 1018 n/cm2 (E >
!, 1.0 MeV) at 550'f. This results in an average USE of 85 ft-lb (figure 5-2).
Irradiation of the reactor vessel core region weld metal Charpy specimens to 8.87 x 1018 n/cm2 (E > 1.0 Mei) at 550'f (figure 5-3) resulted in a 30 ft-lb transition temperature increase of 204 'T and 50 ft-lb transition temperature increase of 214 'f. This resulted in a 30 ft-lb transition temperature of 191 'f and a 50 f t-lb trans'ition temperature 'd 219 'f.
The average upper shelf energy (USE) of the reactor vessel core region weld metal resulted in a average upper shelf energy decrease of 47 ft-lbs af ter irradiation to 8.87 x 1018 n/cm2 (E > 1.0 MeV) at $50'f. This results in an average USE of 74 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld metal Heat-Affected-Zone (HAZ) specimens to 8.87 x IOM n/cm2 (E > 1.0 MeV) at 550'f (figure 5-4) resulted in a 30 f t-lb trar sition temperature increase of 252 'f and a 50 f t-lb transition temperature increase of 247 'F. This results in a 30 f t-lb transition temperature of 26 'F and a 50 ft-Ib transition temperature of 96
'f.
The average upper shelf energy (USE) of the reactor vessel weld HAZ metal resulted in a average upper shelf energy decrease of 46 ft-lbs after irradiation to 8,87 x 1018 n/cm2 (E > 1.0 MeV) at 550'f. This results in an average USE of 101 ft-lb (figure 5-4).
l 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 or tougher appearance with increasing test temperature.
l l
1 5-5 1 . - - - - -
A comparison of the 30 f t-lb transition temperature increases and the %
decrease in VSE for the various Diablo Canyon Unit 2 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2[5] is presented in Table 5-6. This comparison indicates that the actual .
transition temperature increases and VSE decreases resulting from irradiation to 8.87 x 1018 r/cm2 (E > 1.0 MeV) at 550 'f are in close agreement -
with the NRC Regulatory Guide 1.99, Revision 2 predictions.
5.3 Tension Test Results The results of tension tests performed on the intermediate shell plate B5454-1 (longitudinal and transverse Orientation) an:i the weld metal irradiated to 8.87 x 1018 n/cm2 (E > 1.0 MeV) are shown in Table 5-7 and are compared with unirradiated resultsill in figures 5-9, 5-10 and 5-11. l Irradiation of the reactor vessel intermediate shell plate B5454-1 tensile specimens to 8.87 x 1018 n/cm2 (E > 1.0 MeV) at 550'r (figure 5-9) resulted in an increase of 13 to 25 ksi in 0.2 percent offset yield strength -
and an increase of 13 to 14 ksi in ultimate tensile strength for specimens oriented with the longitudinal axis parallel to the major working direction of the plate (longitudinal orientation) 1rradiation of the reactor vessel intermediate shell plate B5454-1 tensile specimens to 8.87 x 10 18 n/cm2 (E > 1.0 MeV) at 550'f (figure 5-10) resulted in an increase of 13 to 18 ksi for the 0.2 percent offset yield strength and an increase of 12 to 15 ksi for the ultimate tensile strength for specimens oriented with the longitudinal axis perpendicular to the major working direction-of the plate (transverse orientation).
Irradiation of the reactor vessel weld metal tensile specimens to 8.87 x 1018 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-11) resulted in an iacrease of 20 tc 25 ksi for the 0.2 percent offset yield strength and an increase of 8 to 28 ksi for the ultimate tensile strength.
5-6 I
J
The fractured tension specimens for the plate material are shown in figures I
i 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.
True stress-strain curves for the tension specimens to the point of necking tre shown in figures 5-20 through 5-24.
5.4 Compact Tension Tet11 Per the surveillance capsule testing program with the Pacific Gas and Electric Company, 1/2 T-compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center Hot Cell.
+
9 5-7
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE DIABLO CANYON UNIT 2 PLATE B5454-1 1RRADIATED AT 550*F, FLUENCE 8.87 x 1018 n!cm2 (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shear Samele No. ('F) ('C) (ft-lb) ,Q1 gilsl M J)
Longitudinal Orientation PL54 7.0 PL47 25 80 21.0
( 9.5 9.0 0.23 5 (28.5 19.0 0.48 15 PL57 80 29.0 39.5 PL60 28.0 0.71 20 105 17.0 23.0 16.0 PL58 0.41 15 105 29.0 39.5 23.0 0.58 25 PL51 115 34.0 46.0 PL49 26.0 0.66 30 115 40.0 54.0 31.0 0.79 35 PL59 125 45.0 61.0 PL48 36.0 0.91 40 140 42.0 57.0 38.0 0.97 40 PL53 140 43.0 58.5 PL55 34.0 0.86 45 225 73.0 99.0 62.0 PL50 280 1.57 95 111.0 150.5 85.0 2.16 100 PL46 300 100.0 135.5 PL58 79.0 2.01 100 350 104.o 141.0 83.0 PL52 (2.11) 100 375 112.0 (152.0) 85.0 (2.16) 100 Transvers Orientation PT56 25 -4 8.0 11.0 PT48 6.0 0.15 5 80 27 21.0 28.5 20.0 PT59 80 0.51 15 27 17.0 23.0 18.0 0.46 15 PT54 120 49 27.0 38.5 23.0 0.58 20 PT52 125 52 26.0 35.5 PT46 25.0 0.64 25 135 57 38,0 51.5 31.0 PT60 0.79 35 140 60 34.0 46.0 30.0 0.76 30 PT53 100 82 47.0 63.5 42.0 1.07 45 PT51 200 93 55.0 74.5 PT58 47.0 1.19 50 225 107 82.0 111.0 68,0 PT49 1.73 100 250 121 76.0 103.0 64.0 1.63 100 PT55 280 138 76.0 103.0 70,0 PT47 320 1.78 100 160 85.0 115.0 68.0 1.73 100 PT57 330 166 79.0 107.0 PT50 68.0 100 350 177 82.0 111.0 69.0 (1.73
- 1. ;'5 100 5-8
- -. . - - . _ - . . _ = -
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE DIABLO CANYON UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL 1RRADIATED AT 550*f, FLUENCE 8.87 x 1018 n/cm2 (E > 1.0 MeV)
Temperature Impset Energy Lateral Expansion She ar
, Sample No. ('F) l'Ol (ft-lb) 12), (mils) imm1 (%),1dVetal PW58 100 15.0 20.5 14.0 (0.35 10 PW57 150 16.0 21.5 15.0 (0.38 15 PW55 185 19.0 26.0 17.0 0.43 35 PW60 185 25.0 34.0 25.0 0.04 25 PW49 195 32.0 43.5 32.0 0.81 60 PW50 195 33.0 44.5 28.0 0.71 60 PW59 210 37.0 50.0 30.0 0.76 60 PW47 210 50.0 88.0 41.0 1.04 95 PW56 250 1 53.0 72.0 45.0 1.14 95 PW46 250 63.0 85.5 50.0 1.42 95 PW53 300 74.0 100.5 62.0 1.57 100 PW51 350 71.0 96.5 62.0 1.57 100 PW52 375 76.0 103.0 67,0 1.70 100 PW48 400 72.0 97.5 62.0 1.57 100 PW54 420 73.0 99.0 85.0 2.16 100 RAZ Vetal PH59 0 - 18.0 15.0 0.38 15 PH51 50 36.0 24.5} 49.0 34.0 0.86 35 PH53 50 29.0 30.5 26.0 0.66 25 PH55 75 24.0 32.5 23.0 0.58 25 PH54 75 75.0 101.5 52.0 1.32 65 PH46 125 94.0 127.5 73.0 1.85 00 PH52 125 57,0 77.5 44.0 1,12 60 PH58 175 53.0 78.5 47.0 1.19 65 PH57 175 48.0 65.0 45.0 1.14 70 PH60 200 58.0 78.5 55.0 1.40 90 PH49 275 113.0 153.0 84.0 2.13 100 PH47 275 97.0 131.5 80.0 2.03 100 PH48 300 123.0 167.0 81.0 2.06 100 PH56 300 71.0 96.5 67.0 1.70 100 PH50 325 102.0 138.5 76.0 1.93 100 S 5-9
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. M -
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= pA ra r r/
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t p) l emf 5Oos6ss50050005 2Sso0ll24428057 5o0055000500000 i ee* 2s822348025s235 t TT( l1II11122333 111112222333 c u d er y l e pb l mm 477es199s350e82 545e55454555455 69s428o31895770 l au LLLLLLLLLLLLLLL 554554e55545455 u
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TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE DIABLO CANYON UNIT 2 18 WELD METAL AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 8.87 x 10 n/cm2 (E > 1.0 MeV) Normalised Enerstice Time Maximum T1== to Fracture Arrest Yield Flo* Test Charpy Charpy Maximum Prop Yield Load Load Streme Stree. Ep/A Loe.d to Yield Load Maximum Sample Temp Energy Ed/A Em/A 2 (kipe) (ue c) _ (kipe) (usec) (kipe) (kipe) (kei) (bei) Number (*F) (ft-lb) (ft-lb/in ) Weld Metal 3.60 270 3.60 0.10 109 114 100 15.0 121 89 32 8.80 125 121 PW58 145 3.80 220 3.75 0.85 117 PW57 150 16.0 129 65 64 3.55 3.15 1.25 97 103 38 115 2.95 125 8.30 170 PW55 185 19.0 153 4.25 375 4.25 1.05 117 129 185 25.0 201 150 52 S.55 725 1.85 107 Ils PW60 125 S.90 870 3.90 PW49 195 32.0 758 136 121 3.25 4.8 1.00 115 129 68 3.45 115 4.30 460 PW50 195 33.0 266 198 535 3.80 1.30 99 113 207 91 3.00 125 3.80 PW59 210 37.0 298 4.35 480 4.30 3.20 112 1sa 210 50.0 403 198 304 s.40 185 104 119 PW47 145 4.05 485 4.00 3.60 PWLS 250 53.0 427 186 241 8.15 3.65 3.20 96 172 304 2.90 125 3.83 535 P'J46 250 63.0 507 203 -* -- 122 131 218 378 3.70 220 4.20 535 PW53 300 74.0 596 545 -* -- 118 129 [ 225 847 8.55 210 4.25 PW51 350 71.0 572 4.20 550 -* -- 106 123 375 76.0 612 224 388 8.20 190 96 113 PW52 145 3.95 535 -- -* 420 73.0 588 205 383 2.90 81 95 PW54 140 3.30 605 -* -* PW13 300 63.0 507 205 802 2.45 HAZ Me.tal 4.45 310 4.45 0.10 132 140 0 18.0 145 124 21 4.00 135 128 PC59 3.60 125 4.15 425 4.15 0.85 118 PC53 50 29.0 234 174 60 4.75 1.20 125 141 43 3.75 175 4.75 545 PZ4 50 36.0 290 247 - - - - - - P155 75 24.0 194 COMPUTER MALFUNCTION 605 3.70 1.55 112 128 eO4 259 345 8.40 125 4.05 PZ54 75 75.0 4.40 550 4.05 1.85 117 131 57.0 450 231 228 3.55 140 113 PH52 125 4.10 530 2.60 1.50 90 125 94.0 757 214 543 2.70 105 125 P46 125 4.30 535 4.20 2.9 109 175 48.0 387 224 los 3.30 1.50 97 114 P!57 262 2.95 125 4.00 535 3.30 PH58 175 58.0 467 205 3.75 2.05 los 123 27 3.2 ISO 4.25 555 PZ60 200 58.0 467 220 -* -- 98 118 602 2.95 130 4.20 555 P247 275 97.0 781 219 -- -* 94 118 e12 2.85 125 4.30 705 PE49 275 113.0 910 298 -* -- 92 175 359 2.80 115 4.15 530 PESS 300 71.0 572 213 715 -* -- 98 L2 282 708 2.95 135 4.00 PB48 300 123.0 990 4.20 895 -* -- 90 115 325 102.0 821 291 531 2.75 IIS P'f50
- Fully ductile fractures no arrest load 5-11 l
u - __
, g
. e e . . , e TABLE 5-5 EFFECT OF 550*F 1RRADIATION TO 8.87 x 1018 ,jc ,2 (E > 1.0 MeV)
OM NOTCH TOUGHNESS PROPERTIES OF DIABLO CA4 TOM tmIT 2 REACTOR VESSEL stRVEILLANCE MATERIALS Average 35 et t Average Energy Average 30 ft-lb Lateral Expansion Average 50 ft-ib Abs e tton at Temperature (*F) Temperature (*F) Temperature (*F) Full Shear (f t-Ib) Material : (% irradiated Irradiated AT unirradiated Irradiated AT unieradiated Irradiated AT unirradiated Irradiated Atit-ib) 4.6 105.7 101.1 18.4 135.6 117.2 33.3 154.6 121.4 144.7 114.9 29.8 Plate B5454-1 (longitudinal) 125.5 98.9 54.4 149.1 94.7 74.6 173.1 98.5 94.9 85 9.9 Plate 85454-1 26.6
-(transverse)-
204.2 -0.7 212.3 213.0 14.9 228.7 213.8 120.9 74.4 45.5 Weld N tal -13.6 190.6 26.1 -169.9 69.9 239.8 -151.2 96.2 247.4 147.6 101.2 46.4 HAZ Metal -226.0 252.1 5-12 m b ir
TABLE 5-6 COMPARISON OF DIABLO CANYON UNIT 2 SURVEIELANCE MATERIAL 30 FT-tB TRANSITION TEMPERATURE SHIFT 5 AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE I 99 REVISION 2 PREDICTIONS 30 ft-lb Transition Teco. Shift Upper Shelf Eneroy Decrease R.G. 1.99 Rev. 2 R.G. I.99 Rev. 2 Measured Fluence (Predicted) Measured (Predicted) 18 2 (*F) (%) (%) Material Capsule 10 n/cm (*F) N01E I NOTE 2 72.1 71.7 65.0 18 14.3 Plate B5454-1 0 3.51 8.87 98.4 98.0 101.1 24 20.6 (longitudinal) X 3.51 72.1 71.7 73.0 18 0.9 Plate 85454-1 U 24 10.4 8.87 98.4 98.0 98.9 (transverse) X 3.51 149.3 158.0 174.0 28 29.7 Weld Metal U 38 38.8 X 8.87 204.0 215.9 204.2 232.7 -- 40.4 HAZ Metal U 3.51 --- --- 252.1 -- 31.4 X 8.87 --- --- NOTES: These CF's are based on the
- 1. ART NDT calculated with a CF - 101.5 (plate) and CF - 210.3 (weld).
mean wt. % values of Cu and Ni from the unirradiated data {Il and capsuler UI2} and X analyses. Tnese CF's are based on the
- 2. ART NDT calculated with a CF = 101.0 (plate) and CF - 222.6 (weld).
measured _ shifts from surveillance capsules U{2] and X per Regulatory Guide 1.99 Revision 2 using longitudinal and transverse data. 5-13 m
TABLE 5-7 TENSILE PROPERTIES FOR DIABLO CANYON UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIAL IRRADIATED AT 550*F TO 8.87 x 1018 n/cm2 (E > 1.0 MeV) Test O.25 Tield Ultimate Fracture Fracture Fracture Unifore Total Reduction sample Temp. Strength Strength Load Streme Strength Elongation Elongation in Arem Material Number I'F) (kei) (kel) (klo) (hel) (hel) (%) (%) (%) t Plate B5454-1 PL10 74 77.9 98.s 3.20 109.s e5.2 13.5 26.1 62 (Longitudinal) PL11 300 71.3 90.7 3.30 20s.s e3.2 11.1 22.2 70 PL12 550 s1.0 93.7 3.90 200.2 79.5 9.0 16.2 60 Plate B5454-1 PTIO 74 79.5 200.7 3.70 396.1 75.4 12.0 22.1 62 (Tranverse) PT11 300 71.s 90.7 3.30 159.1 87.2 10.5 19.5 58 PT12 550 66.2 91.7 3.50 ses.s 71.3 10.5 19.1 Es Geld PW10 74 90.7 105.9 3.70 196.1 75.4 12.5 24.3 62 PW11 300 si.5 96.8 3.00 197.1 73.3 10.5 21.3 63 PW12 550 77.4 35.8 3.75 151.5 7s.4 10.2 1s.0 SO l 4 9 a f 4 5-14
127 10& @ @a , - _m
-- m.,__
@ 80-6 60- .
(2
- n. 40 27 e l o
l 2 ,I 0 , , , , , 120
$ 100-
. o 80- ,
60- " " ui 117 p de "
=
E 20-0 V , , , , , , 160
, 144.7 ~
140-Unirradiated 12&
~ 100- ,
=
f 80- = a w g 6& " 121 1rradiated at 550 'F h 40-
/=
[ 101/ [ 8.87 x 10 18 n/cm t 27 g / o P200 100 b ibo 2b0 3b0 4b0 Sbo 600 Temperature (F) Figure 5-1. Charpy V-Notch Impact Properties for Diablo Canyon Unit 2 i Reactor Vessel Intermediate Shell Plate B5454-1 l (Longitudinal Orientation) 5-15
1201 - - - - . , l 107 @ n a n @ -. s QL _ y BCr 6 T3 GO' 0 o a 4& (3 , 0
~-
120
, 100-
'g 80 ,
I 3 a ,4-en u -
- 8. G& -
m - o s 40- g 95 b e, b 20-0- T*
.) . , . . . .
120 s 100- Unirradiated 94.9 go kw 60- -[g a Irradiated at 550 'f 40- /g- 8,87 x 1018 n/cm 2 0 / 20-se, _ biOO 100 b 1b0 2bo 3b0 4b0 Sb0 600 Temperature (F) figure 5-2. Charpy V-Notch Impact Properties f r Di blo Canyon Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 (Transverse Orientation) 5-16 t
120 - - 100- - - = ==*. o , ! 80-6& ,
, 44 20-
_-)", O O . . , . . . 4 120 , g10e . e , 87 o 60- , l 40- 213 o
. 24 , o O < . . . . , .
- 140
. a 120.9 120-
_ Unirradiated .
-u M10&
80- , m 74.4 a ww 60- -
. 214 =,
4} 204 hirradiatedat550*r
. - 8.87 x'10I0 n/cm 2 2& o o .
A _a 300 2'00 100 b ib0 2bo 3d0 4b0 500
. Temperature (F) t figure 5-3. Charpy V-Notch Impact Properties for Diablo Canyon Unit 2-Reactor Vessel Weld Metal 1
( i 5-17 \ 1 eww- , , - - , 1<s-v en ----,,.,,-o.,m-a sm me e ,e,-w---e- w -e s p - www ~r= w s. nv , r-<,-v---- < r w g ww m v e e,- r .r.--w-. -- -,e*
120 . 100- @ '.D LSE ,
, , F8 l- o E
(&
- c. 40- y '
27 0 j . , k 120
$107 E e
.{ 80-E -~6e
. o <-
W - a g . o 8 w 41 240 = g 3 20-
/- -
o o 0 200 ' 180- = 160-7 Unirradiated 147.6 j140- g) 5 m_ h 127 , e a 101.2
$ 100- e o w
80- = =
. o a i 60- 247 ,.e oa 4
20-52 * [o trradiated at 650 'r t.87 x iole njc,2 300 200 100 0 1b3 2b0 300 4b0 500 77 pi Temperaturo (F) Figure 5-4 Charpy V-Notch Impact Properties for Diablo Canyon Unit 2 , Reactor Weld Heat Affected Zone Metal 5-18
i i
! ) 1 PL54 PL47 PL57 PL60 PL58 p'.4 % m ,8"***-
f; '!
.$ f_'
h
/<
\
&y::,,a i: ~ m.m o ,
' ?. ! ---
,. 27 Ff r V j.l;i' [;;
, ',, ,, .e , i PL51 PL40 PL59 PL48 PL53 l 9 s .. tYA . W,r. . . .
, s PL55 PL50 PL40 PL50 PL52 figure 5-5. Charpy Impact Specimen fracture Surfaces for Diablo Canyon Unit 2 Reactor Vessel Intermediate Shell Plate 85454-1
, (Longitudinal Orientation) 5-19
- i. ] l pure, . , w PT50 PT48 PT59 PT54 PT52
- e W :. F'N ,M' PT46 PT60 PT53 PT51 PT58 asse- .
- w We '9,.
M b, M # 'v,d y \!Ng Q f: ; PT49 PT55 PT47 PT57 PT50 ( Figure 5-6. Charpy impact Specimen fracture Surfaces for Diablo Canyon Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 (Transverse Orientation) - 5-20
* ;* -v , 7&i mesm w .. . . _ _
PW58 M57 PW55 PWSU PW49
,m .
% .f .. ' ma I ,*"'
) i . ;* ,
s9). Y .- , IO , e.g n ,f
- l. {
- g. -
h _, y ( ,_ f ,*.:% p% PW50 PW59 PW47 PW56 PW40 pr- & t 5- Ps 4-I
} -[ l I '
i .. r - . Il yQ
} t.
PW53 PW51 PW52 PW48 PW54 Figure 5-7. Charpy Impact Specimen fracture Surfaces for Diablo Canyon Unit 2 Reactor Vessel Weld Metal 5-21
i l -
==mmy> w. m gM".-.
A;d i
' 11
! (,
j *~ - f qjyrj. " Ti
'f
[- . PH59 Pl!51 Pil53 Pil55 Pil54 N iT. .. ." 7 - 7* s- , , . l! !?"
- cui uw .,
_, y; !> .
*,A~*
f Ti' fi4
<- y sc' Q
s , , - PH46 Fil52 Pil58 Pl!57 Pl!00
,me BKL. ,ff gy- &1
. g __e PIl49 Pil47 PII48 Pil56 Pil50 Figure 5-8. Charpy Impact Specimen fracture Surfaces for Diablo Canyon Unit 2 Reactor Vessel Weld Heat Affected Zone (HAZ) Metal 5-22
l ('C) i 0 50 100 150 200 250 300 120 i i ' ' ' i '- 800 110 - Utilmate Strength 700 100 - q
~
b ,2 A~ 600 g
$g70 _
I 500 2
.6 1
^ ;
60 2" a
/2 -
400 n 50 - 0,2 % Yield Strength 40 I t i I i 300 Code: Open Points - Unirradiated I8 2 Closed Points - Irradiated at 550 F ( 8. 87 x 10 n/cm ) 80 i i i i i , , 70 - o d2 g_ l . 60 - - l
- Reduction in Area
$M - -
- ~
h40 Total Elongation l ,8 30 - b '
,f 2 ,
3 - ! E 2% Y P-10 - 0 , pniforn) Elonggtlon , 0 100 200 300 400 .500 600 L Temperature ('F) Figure 5-9. Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 (Longitudinal Orientation) 5-23
1 ('C) 0 50 100 150 200 3g 250 300 i i i i i i 110 -
'- 800
-Ultimate Strength -
100 - - 700
=
? ,*o 0
- 2% ;- 6*
g 70 - 5 - 500 3 M -2 4@ 50 -
- 0. 2 % Yield Strength 2 40 i ' ' i i M0 Code:
Open Points - Unirradiated L 2 Closed Points - Irradiated at 550 F ( 8. 87 x 10 n/cm ) 80 i - i i i i i i 70 - Reduction in Area 60 -
$- ^ -
3 50 -
.I 40
]s 30 ,
Total Elongation - 3 - a W
~
A 10 - 6 e - 0 i Mniforty Elongation i 0 100 200 300 400 500 600 Temperature ('F) Figure 5-10. Tensile Properties for Diablo Canyon Unit 2 Reactor Vessei Intermediate shell Plate B5454-1 (Transverse Orientation) 5-24
( C) 0 50 100 150 200 250 300 120 ' '
' i ' I i-800 110 -
- Ultimate Strength 100 -
700 j 90 - N - 600 -
! 80 -
L N , -2 E
^ -
500 3 E 70 5 2 j 60 _ n 400 50 - 0.2 % Yleid Strength 40 I i i t i
%0 Code:
Open Points - Unirradiated 8 Closed Points - Irradiated at 550 F ( 8.87 x 10 n/cm ) 80 l 10 i 2* i i
~-
i i i i 60 - 3 50 - Reduction in Area - E' 40 - - j 30 - A Total Elongation _ S ~#2 20 10 -
*" a-r2 0 , Unliorm Elgngatior) ,
0 100 200 300 400 500 600 Temperature ( *F) Figure 5-11. Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel Weld Metal 5-25
i i f i l 4
\ -
l 1 g
- 1 #
'.g ; at v h
- n s .& <
n .c
,~# , s 4
1 ,. j Specimen PL10 i 74*P i l l i ! i l f- ' 1 1, , . Specimen PL11 300'P I ' I wy
. h, It1
.y.
l% , ;
<jn ,, e..
,-., p, ;j.' .. ,
3 .i
})
t* '
-j i :' e
. fe
' i . t,
]O 4f q l
Uc: 8. i r Specimen PL12 5S0*F Figure 5-12. Fractured Tensile Specimens from Diablo Canyon Unit 2 Reactor - Vessel Intermediate Shell Plate B5454-1 (Longitudinal Orientation) 5-26 L._..__.__._,._.__.___....-_-----------------------
m,,y q ,m. .~ y >; 6 4
.h
- 2 3 4 .c I. q
,1et m
. e q!
4 Jtotm. J' : - i r Specimen PT10 74'F
.,~ R '
'?T*
t '},p) . - y",7;[ ' cT
..; g Specimen PT11 300'F i
4 i ;1 R- 1 l,cl . i j, :!o aa4 -- 9- -(,,
.gim 7 -
O. Nil % . d#, 'l' . ., ,,g 8 i-f A- .j Specimen PT12 550'F Figure 5-13. Fractured Tensile Specimens from Diablo Canyon Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 (Transverse Orientation) 5-27
E g q q.,oye %, gp p rvy p.- m ., e t p ,.
, py g 6
M,. s x 3. f j . .t. . . Q'ta;41j-fr,' i-.) i ;-j [ , S t *10 f x; D sW1;d 3 j' f3,no,ns;;s.m2x a
+
- .{
,s..I_. (-
Specimen PW10 74*p 97p my
~ [.fi ff; * ,y , ,
h pf!? ' h;ff"IfA>$@M ff wm ~ e _ . [_.-;, 3
- a. . a .
yev . 1;.3
, pto' ,
Specimen PW11 3Co* F
't a
e 5 e J ,, i }cj - l-t l; ! .j
-3 1
- v m ;' e <
- p e fl0 f tri N,W.h 4. : , ',- h . )' r .~ j I ,
Specimen F 2 550'F Figure 5-14. Fractured Tensile Specimens from Otablo Canyon Unit 2 Reactor Vessel Weld Metal - 5-28
120 100-g 80 x s N g 60-u
$ 40 -
20- SPEC PL10 74 F 0 0 0.b5 0.' 1 0.i 5 0.'2 0.25 0.3 STRAIN, IN/IN 100 90-80-70-tn x 60-
'l 50-E 40.-
M 30-20-SPEC PL11 10-300 F 0 0.b5 0.' 1 0.I 5 0.'2 0.25 STRAIN, IN/IN Figure 5-15. Engineering Stress-Strain Curves for Diabic Canyon Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 Tension Specimens PL10 and PLll 5-29 l
.4 -g
- 100;
^
90-80- (( , 70-m x so-vi :m- . . E 40-w 30-p;; l_ 20- _ SPEC PL12 l _ 550F c 0 ', . . 0- 0.02 0.04 0.06 0.08. 0.1 0.12' O.14 '0.16 0.18 0.2 STRAN, N/N i i l-I 120 i 100- _ -
- g. 80 - -
.g ;
4 Vi m 60-w . 1 m 401 20- SPEC PT10 74 F 0 0.b 5'. 0.' ) 0,'15 0.'2 0.25-STRAN, N/N l: i i Fig ~ure-5-16; Engineering-Stross-Strain Curves for Diablo Car. yon Unit 2- ' ' Reactor Vessel Intermediate Shell Plate B545',-1 Tension - Specimens PL12<and PT10 5-30 .
i 100 90- e 80-70-m x 60-l 50- I E 40 30-20-SPEC PT11 10-300 F 0 . . , , 0 0.05 0.1 0.15 0.2 0.25 STRAlN, IN/lN 100 90-80-70-i?i M 60-p 50-E 40-th 30-20-SPEC PT12 10-550F 0 , . , , , 0 0.02 0.04 0.06 0.08 0.1 0. I 2 0. I 4 0.I 6 0.I 8 0.2 STRAIN, IN/IN Figure 5-17. Engineering Stress-Strain Curves for Diablo Canyon Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 Tension Specimens PT11 and PT12 5-31 ; i
120 100-g 80-x
$ 60-u m 40-20- SPEC PW10 0 , , ,
74 F - 0 0.G 5 0.1 0.15 0.2 0.25 0.3 STRAIN, iN/IN l 100 - l 90-80-70- ' ! m l- x 60-vi 50-d
$ 40 -
! M 30-20-SPEC PW11 10-0 , 300 F 0 0.05 0.1 0.15 0.2 0.25 STRAIN, IN/IN Figure 5-18. Engineering Stress-Strain Curves for Diablo Canyon Unit 2 Reactor Vessel Weld Metal Tension Specimens PW10 and PWil 5-32
--.= ... l 100 90-80.-
g 70-
.x 60-VI u) 50- <
W i El40-m 20-SPEC PW12 - 550F l1 0 , .. , , 0.02 0.04 0.06 0.08 0 - 0.!1 0.I 2 0.I 4 0.I 6 0.I8-0.2 : STRAN, N/N
-i Figure 5.19. Engineering. Stress-Strain Curves for Diablo Canyon Unit 2- _
Reactor Vessel Weld Metal Tension Specimen PW12 l 5-33 l
1 150 12E 100- - a R 7& m
& S0-25-PL10 b.00 0.'02 0.'04 0.'06 0.'08 0.'10 0.'12 0.'14 0.16 True Strain (IMn) 150 12& ^ 100-k n
37& W e f 50-25-PL11 b.00 0.'02 0.'04 0.'06 0.'08 0,'10 0.'12 0.'14 0.16 Tnje Strain (iMn) Figure 5-20. True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor
- Vessel Intermediate Shell Plate B5454-1 Tension Specimens PL10 and PLll 5-34
__ __ _ - _ _ ~
, 150 -- i 12E
^ 100-b [ . 7&
m H S& 29 PL12 b.00 0.'02 0.'04 0.'06 0.'08 0.'10 0.'12 0.'14 0.16 ' True Strain (IMn) i 150 - 12E p 100-7m m E 7&
& 50-2&
PT10 b.00 0.'02 0.'04 0.'06 0.'08 0.'10 ' 0.'12 0.'14 0.16 True Strain (IMn) 1 Figure 5-21. True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor Vessel Intermediate Shell Plate B5454-1 Tension Specimens PL12 and PTIO 5-35
-150 .
12& g 100- , 7& ~ . i H 50-29 PT11 b.00 0.'02 0.'04 0.'06 0.'08 0.'10 0.'12 0.'14 0.16 True Strain (iMn) . 150 12& g 100- ,-- M i 7E e H 50- i 29 PT12-- b.00 - 0.'02 0.'04 0.'06 0.'08 : 0.'10 0.'12 0.'14 0.16 Tnje Strain (iMn)- Fig'ure 5-22. . True. Stress-Strain Curves for-Diablo Canyon. Unit 2 Reactor -* Vessel Intermediate Shell Plate B5454-1 Tension Specimens PTil and PT12 5-36 1 __ -J
i 1
,. l 4
150 i
.12&
^ 106 1
7& 3 H 50-
= 25-PW10 8.00 0.'02 0.04 0.'06 0.'08 0.'10 - 0.'12 0.'14 0.16 True Strain (IMn) 150 12&
i. g 100-w
$ 7&
50-t 25-PW11 [_ l.00- : 0.'02 ~ 0.'04 - 0.'06 0.'08 0.'10 : 0.'12 0.'14 - 0.16 True Strain (lMn). l E Figure 5-23. -True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor l ,,
- Vessel-Weld Metal Tension Specimens PW10 and PWil 5-37
160 12E
~
^ 100-k m
f 7E 50-RV PW12 , b.00 0.'02 0.'04 0.'06 0'08 0'10 0.'12 0.'14 0.16 True Strain (irdn) Figure 5-24. True Stress-Strain Curves for Diablo Canyon Unit 2 Reactor Vessel Weld Metal Tension Specimen PW12 5-38
l . . SECTION 6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY
, 6.1 Introduction . 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 programs for two reasons, first, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy 'pectrum, 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 si various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing r. combination of rigorous analytical techniques and measurements obtained with passive neutron flw ionitors contained in each of the surveillance capsules. The latter information is derived solely from snalysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials prcperties changes to the neutron exposure of the material for lignt water reactor applications 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 from a thresho'd fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853,
" Analysis and Interpretation of Light Water Reactor Surveillance Results,"
recommends reporting displacements per iron atom (dpa) along with fluence 6-1
(E > 1.0 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, " 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, " 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 surveillar,ce 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 surveillance 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 56.0', 58.5', 124.0', 236.0', 238.5', and 304.0' relative to the core cardinal axis 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. - 6-2 l __ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - -- 0
from a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on bot l the distribution of neutron flux and the neutron energy spectrum in ". water annulus between the neutron
. pad and the reactor vess91, 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 ev A ations 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), d(E > 0.1 MeV), and dpa] through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/4(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 impt rtance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for the first 3 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 wittiin the reactor core; but, also accounted for the effects I 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. Evaluate neutron dosimetry obtained from surveillance capsule locations.
- 2. Extrapolate dosimetry results to key locations at the ir.ner radius and through the thickness of the pressure vessel wall.
- 3. Enable a direct comparison of analytical prediction with measurement.
l 4. Establish a mechanism for 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, 0 geometry using the 00T 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. 1 The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westingbouse 4-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 j 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 cut using an Sg order of angular
. quadrature and the P3 cross-section approximation from the SAILOR library.
Adjoint source locations were chosen 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, 0 geometry to provide neutron source distribution importance functions for the exposure pa ameter 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) = [7 [0 [E 1(r, 0, E) S (r, 0, E) r dr de dE where: R (r, 0) - p (E > 1.0 MeV) at radius r and azimuthal angle 0 I (r, 0, E) - Adjoint importance function at radius, 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 Diablo Canyon Unit 2 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. l location the ratio of dpa/d (F > 1.0 MeV) is-insensitive to changing core l source distributions. In the application of these adjoint important functions i to the Diablo Canyon Unit 2 reactor, therefore, calculation of the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/p (E > 1.0 MeV) and ( (E > 0.1 MeV)/p (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific p (E > 1.0 MeV) solutions from the individual adjoint evaluations. l 6-5
. . l l
The reactor core power distributions, used in the plant specific adjoint calculations, were taken from the fuel cycle design report for the first three operating cycles of Diablo Canyon Unit 2 (16 through 18]. The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of the pressure 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 Diablo Canyon Unit 2 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 irradiation 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), p(E
> 0.1 HeV), 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 cycles 1 through 3 plant specific power distribuMons. It is important to note that the data for the vessel inner radius were taken at the clad / base metal interfate; 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 > l 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 - l calculation, are presented on a relative basis for each exposure parameter at 6-6 l
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. For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45'
. azimuth is given by: l 4
p}f4T(45') - d(220.27, 45') F-(225.75, 45') where: dl/4T(45') - Projected neutron' flux at the 1/4T position on the 45' azimuth p (220.27,45') Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth. F (225.75, 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. 6.3 -Neutron Dosimetry The passive neutron sensors included in the Diablo Canyon Unit 2 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-evaluation of the neutron energy spectrum within the -capsule and-the subsequent determination'of the various exposure parameters of interest [p (E > 1.0-MeV),pi(E>0.1MeV),dpa]. 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 t
l'- form, were placed in holes drilled in spacers at several axial levels within ji the capsules. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule. 6-7 i
- i. . . ~ . ;-._.. -- . . - _. . - _ _ _ , . . _ . _ , -
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 that the time- and 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 weil 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 [19 through 32). 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 Diablo Canyon Unit 2 reactor during cycles 1 through 3 was obtained from NUREG-0020, " Licensed Operating Reactors Status Sammary 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. Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code [33]. The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the center of the surveillance caosule as input and 6-8
proceeded to ' adjust a priori (calculated) group fluxes to produce a best fit (in- a-least squsres. sense) to- the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted
, spectra, J
4
.. In the FERRET evaluations, a log nonnal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertai nti es and correlations. In general, the measured values f are linearly related to the flux ( by some response matrix A:
(s,a) (s) (a) f- =I A ( g ig 9 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, R-I o p i g 'ig g relates.a set of measured reaction rates Rt to a single spectrum pg by the multigroup cross section ogg . (In this case, FERRET also adjusts- the cross-sections.) The lognormal approach automatically accounts for the
- physical constraint' of positive fluxes,- even with the large assigned uncertainties.
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 FERRET group
-structure using the 5AND-II code (34]. This procedure was-carried out.by first-
, expanding-th'e a priori _ spectrum into the SAND-II 620' group structure e 6-9 l
- - 2,._.. . . _ ,;. . . _ - . , _-
using a SPLINE l 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 i 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 $3 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, L as for the crost sections, a multigroup covariance matrix is used. More often, a simple parameterized form is used: t M gg,-Rf+Rg R,P g gg, where NR specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional L uncertainties Rg specify additional random uncertainties for group g that are L correlated with a correlation matrix: Pgg, -_(1 - 0) 6gg, + 0 exp ( ) The first- term -specifies purely random uncertainties while the second term '
' describes short-range correlations over a range 7 (0 specifies the strength of :the latter term).
6-10
For the a priori calculated fluxes, a short-range correlation of 7 - 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlations (or
, anticorrelations) were justified based on information presented by R.E.
Maerker(35). Maerker's results are closely duplicated when 7 6. For
, the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.
Results of the FERRET evaluation of the car.sule X dosimetry are given in Table 6-9. The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 8.87 x 10 18 n/cm2 (E > 1.0 MeV) with an associated uncertainty of 8%. 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 schieved in the fits of the adjusted spectrum to the individual experimenul reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure. A summary of the measered and calculated neutron exposure of capsule X is presented in Table 6-12. The agreement between calculation and measurement falls within i fr10% 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 (3.11 EFPY) exposure derived from the capsule X measurements, projections are also provided for exposure periods of 8 EFPY, 16 EFPY and to end of vessel design life (32 EFPY). The time averaged exposure rates for the first 3.11 EFPY of operation were used to perform projections beyond the end of the Cycle 1 through 3 exposure period. (e 9 48 EFPY = 8 EFPY (4'3.11 EFPY / 3.11 EFPY)]
. In the calculation of exposure gradients for use in the development of heatup j and cooldown curves for the Diablo Canyon Unit 2 reactor coolant s tstem, 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 6-11
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 t' (1/4T) = + (Surface) [dpa (Surface)l 4' (3/4T) = 4 (Surface) (dpa (Surface)) Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Tabla 6-15 updated lead factors are listed for each of the Diablo Canyon Unit 2 surveillance capsules. These data may be used as a guide in establishing f sture withdrawal schedules for the remaining capsules. 4 2 l a i e 6-12
CHARPY SPECIMEN A
/
> ' e
% %,,////KA
. /,/,/ / -
~
NNNNNNNNNNNNNNNNNNN NEUTRON PAD \
~
NNNNNNNNNNNNNNNNNN }. l l' Figure 6-1. -Plan View of a Dual Reactor Vessel Surveillance Capsule l 6-13 i-
i 1.01 1.04 0.96 0.77 DCSICN 1% SIS 0.73 0.77 0.68 0.56 CYCLE 1 0.59 0.50 0.54 0.36 CYC1E 2 0.51 0.53 0.43 0.33 CYCLE 3 1.02 1.10 1.00 1.05 1.10 0.71 0.99 1.04 0.96 0.98 0.87 0.51 0.94 1.19 0.94 1.10 0.88 0.40 1.19 1.18 1.14 1.05 0.83 0.33 i L 1.05 0.87 0.87 1.07 1.00 1.05 L 1.13 1.10 1.11 1.06 0.99 0.97 e 1.10 0.99 1.30 1.13 1.09 1.01 1.11 1.17 1.17 1.15 1.18 0.74 1 \ ' 1.09 1.06 0.88 1.10 1.04 1.14 1.17 1.13 1.12 1.18 1.12 1.29 1.04 1.30 1.10 1.13 1.30 1.00 1.30 1.10 l 0.90 1.04 1.12 0.92 1.18 1.15 1.18 1.14 1.27 1.03 1.26 -1.04 1 1.30 1.11 1.30 1.14 I Figure 6-2. Core Power Distributions Used in Transport Calculations for Diablo Canyon Unit 2 6-14
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER t e IRRADIATION 4 (E > 1.0 MeV) d (E > 0.1 MeV) dpa/sec 2 2 CYCLE TIME [n/cm -sec] [n/cm -sec] (EFPS) 31.5* 34.0* 31.5* 34.0* 31.5* 34.0* 1 DESIGN BASIS 1.11 X 10 II 1.29 X 1011 4.88 X 10 ll 5.93 X 10 ll 2.21 X 10-10 2.62 X 10-10 , CYCLE 1 3.13 X 10 7 8.35 X 10 10 9.52 X 10 10 3.67 X 10 ll 4.37 X 10 11 1.66 X 10-10 1.93 X 10-10 CYCLE 2 3.15 X 10 7 7.00 X 10 10 8.01 X 10 10 3.08 X 10 11 3.68 X 10ll 1.40 X 10-10 1.62 7 10-10 CYCLE 3 3.52 X 10 7 6.70 X 10 10 7.61 X 10 10 2.95 X 10 ll 3.49 X 1011 1.34 X 10-10 1.54 1 10-10 . 6-15
1 TABLE 6-2 I l CALCULATED TAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE
.-3 p(E > 1.0 MeV) [n/cm2.3ec) 0' 15' 30' 45' _
DESIGN BASIS 1.45 x 1010 2.21 x 10 10 1,69 x 1010 2.44 r 1010 Cycle 1 1.08 x 1010 1.62 x 1010 1.27 x 1010 1.81 x 1010 Cycle 2 7.99 x 109 1.20 x 1010 1.02 x 1010 1.49 x 1010 Cycle 3 8.19 x 10 9 1.20 x 1010 9.84 x 109 1.37 x 1010 2 9(E > 0.1 MeV) (n/cm -sec) O' 15' 30' ___ 45' DESIGN BASIS 3.02 x 1010 4.66 x 1010 4.25 x 1010 6.11 x 1010 4.53 x 10 10 Cycle 1- 2.25 x 1010 3.41 x 1010 3.20 x 1010 Cycle 2 1.67'x 1010 2.53 x 1010 2.5/ x 1010 3.73 x 1010 Cycle .3 1,71 x 1010 2.53 x 1010 2.48 x 1010 3.43 x 10 10 dpa/sec 0' 15' 30' 45' DESIGN BASIS 2.25 x 10-11 3.41 x 10-11 2.73 x 10-11 '3.88 x 10-11 Cycle 1 1.68 x 10-11 2.50 x 10-11 2.05 x 10-11 2.88 x 10-11 Cycle 2 1.24 x 10-11 1.85 x 10-11 1.65 x 10-11 2.37 x 10-11 Cycle 3 1.24 x 10-11 1.85 x 10-11 1,59 x 10-1I 2.18 x 10-11 . 6-16 i
TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0' 15' 30' 45' 220.27(l) 1.00 1.00 1.00 1.00 220.64 0,979 0.979 0.980 0.979 221.66 0.891 0.891 0.893 U.889 222.99 0.771 0.769 0.773 0.765 224.31 0.655 0.652 0.658 0.648 225.63 0.552 0.549 0.555 0.543 226.95 0.463 0.459 0.467 0.452 228.28 0.387 0.383 0.390 0.376 229.60 0.322 0.318 0.326 0.311 230.92 0.268 0.253 0.271 0.257 232.25 0.222 0.218 0.225 0.211 233.57 0.183 0.180 0.187 0.174 234.89 0.151 0.148 0.155 0.142 236.22 0.125 0.121 0.128 0.116 237.54 0.102 0.0992 0.105 0.0945 238.86 0.0831 0.0807 0.0862 0.0762 240.19 0.0673 0.0650 0.0703 0.0608 241.51 0.0539 0.0512 0.0567 0.047? 242.17(2) 0.0508 0.0477 0.0536 J.0438 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius t
6-17
TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (L > 0.1 HeV) , WITHIN THE PRlSSURE VESSEL WALL Radius . __f em) 0' 15' _ 30' 45' l 220.27(I) 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 221.66 1.00 1.00 1.00 0.995 222.99 0.974 0.966 0.982 0.956 224.31 0.928 0.915 0.938 0.902 225.63 0.875 0.859 0.886 0.843 226.95 0.819 0.802 0.832 0.782 228.28 0.762 0.743 0.777 0.722 229.60 0.705 0.686 0.721 0.663 230.92 0.649 0.629 0.665 0.605 232.25 0.594 0.575 0.611 0.549 233.57 0.540 0.522 0.550 0.495 234.89 0.489 0.470 0.506 0.443 236.22 0.436 0.421 0.455 0.392 237.54 0.306 0.373 0.406 0.343 238.86 0.337 0.326 0.358 0.296 240.19 0.290 0.280 0.310 0.248 241.51 0.244 0.232 0.261 0.201 242.17(2) 0.233 0.219 0.249 0.188 i NOTES: 1) Base Matal Inner Radius
- 2) Base Metal Outer Radius e
6-18
, TABLE 6-5 - RELATIVE RADIAL DISTRIBUTIONS Of IRON DISPLACEMENT RATE (dpa)
'.! THIN THE PRESSURE VESSEL WALL . Radius (cm) _
O' __ 15' 30' 45' 220.27(l) 1.00 1.00 1.00 1.00 220.64 0.982 0.982 0.986 0.984 221.66 0.911 0.910 0.923 0.915 222.99 0.813 0.812 0.837 0.821 224.31 0.721 0.718 0.751 0.730 225.63 0.637 0.633 0.073 0.646 226.95 0.562 0.558 0.602 0.572 228.28 0.496 0.491 0.539 0.505 229.60 0.438 0.433 0.481 0.447 230.92 0.387 0.381 0.430 0.394 232.25 0.341 0.335 0.383 0.347 233.57 0,300 0.295 C.341 0.305 234.89 0.263 0.258 9.?02 0.266 236.22 0.230 0.225 0.207 0.231 237.54 0.199 0.195 0.234 0.199 238.86 0.171 0.168 0.203 0.169 240.19 0.145 0.142 0.174 0.140 241.51 0.121 0.117 0.146 0.113 242.17(2) 0.116 0.110 0.140 0 *406 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 9
6-19
TABLE 6-6 NUCLEAR PA!!AMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission , Monitor of Weight Response Product Yield Material Interest Fraction Rance Hal f-Li fe (%) Copper Cu63(n e)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)Csl37 1.0 E > 0.7, deV 30.12 yrs 5.99 Neptunium-237* Np237(n,f)Csl37 1.0 E > 0.08 MeV 30.12 yrs 6.50 i l Cobalt-Aluminum
- CoS9(n,7)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobalt-Aluminum
- CoS9(n,7)Co60 0.0015 E > 0.015 MeV 5.272 yrs I
- Denotes that monitor is cadmium shielded.
1 l 6-20
TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE X Irradiation Pg Pg/ Irradiation Decay
.- Period (MW) PMax Time (days) Time (days) 10/85 734 .215 13 1750 11/85 1272 .373 30 1720 12/85 1305 .383 31 1689 1/86 750 .220 31 1658 2/86 341 .100 28 1630 3/86 1845 .541 31 1599 4/86 2859 .838 30 1569 5/86 3088 .905 31 1538 6/86 2789 .8: ti 30 1508 7/86 2396 .702 31 1477 8/86 32 6 .951 31 1446 9/86 2779 .815 30 1416 *~
10/86 3206 .940 31 1385 11/86- 3279 .961 30 1355 12/86 3128 .917 31 1324 1/87 3103 .910 31 1293 2/87 2647 .776. 28 1265_' I 3/87 2029 .595 31 1234 4/87 254 .075 30- 1204 5/87 0 .000 31 1173 6/87 0 .000 30 1143 7/87 980 .287 31 1112
. NOTE: Reference Power - 3411 MWt 6-21
TABLE 6-7 (Cont'd) 1RRADIAT10H HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSVLE X Irradiation Pa Pa/ Irradiation Decay Period (MW) PHax Time (days) Time (days) . 8/87 3116 .914 31 1081 9/87 3405 .998 30 1051 10/87 3317 .973 31 1020 11/87 2587 .758 30 990 12/87 3190 .935 31 959 1/88 3330 .976 31 928 2/88 2336 .685 29 899 3/88 3034 .890 31 868 4/88- 3336 .978 30 838 5/88 3402 .997 31 807 6/88 3274 .960 30 777 7/88 1619 .475 31 746 8/88 2281 .669 31 715 9/88 1541 .452 30 685 10/88 0 .000 31 654 11/88 0 .000 30 624 12/88 1954 .573 31 593 1/89 3400 .997 31 562 2/89 3379 .991 28 534 3/89 3403 .998 31 503 4/89 2017 .591 30 473 5/89 3309 .970 31 442 6-22 I
TABLE 6-7 (Cont'd) 1RRADIAT10N HISTORY Of NEUTRON SLNSORS CONTAINED IN CAPSULE X Irradiation Pg P/ J Irradiation Decay
. Period (MW) PHax Time (days) Time (days) 6/89 3401 .997 30 412 7/89 2662 .781 31 381 8/89 3052 .895 31 350 9/89 3337 .978 30 320 10/89 2782 .816 31 289 11/89 2767 .811 30 259 12/89 3300 .967 31 228 1/90 3405 .998 31 197 2/90 3365 .987 28 -169 3/90 2487 .729 4 165 es i
9 I 6-23 l
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated Reaction Monitor and Activity Activity Rate - Axial location ,(dts/sec-om) (dis /see-am) (RPS/ NUCLEUS) Cu-63 (n.o) Co-60 Top 8.85 x 10 4 3.00 x 106 Bottom 9.23 x 104 3.13 x IOS Average 9.04 x 104 3.06 x 105 4.67 x 10-I7 i Fe-54(n.p) Mn-54 Top 1.41 x 106 2.79 x 106 l
- Middle 1.49 x 106 2.95 x 106 i Bottom 1,48 x 106 -
2.93 x 106 Average 1.46 x 106 2.89 x 106 4.61 x 10-15 Ni-58 (n.p) C0-58 l l Top 7.23 x 106 4.32 x 107 Middle 7.64 x 106 4.57 x 107 Bottom 7.57 x 106 4.53 x 107 Average 7.48 x 106 4,47 x 307 6.38 x 10-15 U-238 (n,f) Cs-137 (Cd) Middle 2.99 x 105 4.44 x 106 2.93 x 10'I4 - 6-24
TADLE 6-8 (Cont'd) HEASURED SENSOR ACTIVITIES AND REACTION RATES Heasured Saturated Reaction
, Monitor and Activity Activity Rate Axial Location (dis /sec-9m). idis/sec-am) (RPS/ NUCLEUS)
Np-237(n,f) Cs-137 (Cd) Middle 3.13 x 106 4.65 x 10 7 2.82 x 10~l3 C0-59 (n,1) Co-60 Top 2.17 x 10 7 7.35 x 10 7 Middle 1.95 x 107 6.59 x 107 Bottom 2.12 x 107 7.17 x 107 Average 2.08 x 10 7 7.04 x 10 7 4.59 x 10'I2 C0-59 (n,1) Co-60 (Cd) Top 1.21 x 107 4.10 x 107 Middle 1.12 x 10 7 3.80 x 107 Bottom 1.20 x 10 7 4.07 x 107 Average 1.18 x 10 7 3.99 x 107 2.60 x 10-12 l 6-25
TABLE 6-9
SUMMARY
OF NEUTRON DOSlHETRY RESULTS TIME AVERAGED EXPOSURE RATES . p (E > 1.0 MeV) (n/cm2-sec) 9.05 x 1010 8% . p (E > 0.1 MeV) (n/cm2-sec) 4.24 x 1011 15% dpa/sec 1.80 x 10-10 i 11% p (E < 0.414 eV) (n/cm2 -sec) 8.17 x 1011 1 22% INTEGRATED CAPSULE EXPOSURE 2 4 (E > 1.0 MeV) (n/cm ] 8.87 x 1018 3 gg 2 4 (E > 0.1 HeV) (n/cm ) 4.16 x 1019 15% dpa 1.77 x 10-2 g 3)g t (E < 0.414 eV) (n/cm2 ] 8.02 x 1018 22% NOTE: Total Irradiation Time - 3.11 EFPY I 9 6-26
=--m uni
TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER
. Adjusted Reaction Measured {alculation (Bj Cu-63 (n.o) Co-60 4.67x10-17 4.66x10-17 1.00 Fe-54 (n.p) Mn-54 4.61x10-15 4.67x10-15 3,o3 Ni-58 (n.p) C0-58 6.38x10-15 6.40x10-15 3,og U-238 (n,f) Cs-137 (Cd) 2.93x10-14 2.74x10-14 0.94 Np-237 (n,f) Cs-137 (Cd) 2.82x10-13 2.90x10-13 1.03 Co-59 (n.1) Co-60 (Cd) 2.60x10-lE 2.61x10-12 0.99 Co-59 (n,1) 00-60 4.59x10-12 4.55x10-12 1.00 6
{ 6-27 o i' 1 - . - - - .-
i' TABLE 6-11 l ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER l Energy Adjusjedflux Energy Group Group Adjusgedflux . (Mov) (n/cm -sec) (Mov) (n/cm-sec)
! 1 1.73x101 3.97x106 28 9.12x10'3 1.98x1010
- 2 1.49x10I 9.69x106 29 5.53x10~3 2.57x1010 3 1.35x10I 4.45x107 30 3.36x10~3 8.06x109 4 1.16x101 1.13x108 31 2.84x10~3 7.74x109
- 5 1.00x10I 2.71x108 32 2.40x10'3 7.48x109 6 8.61x100 4.86x108 33 2.04x10'3 2.llx1010 7 7.41x100 1.15x109 34 1.23x10~3 1.93x1010 8 6.07x100 1.67x109 35 7.49x10-4 1,78x1010 9 4.97x100 3.52x109 36 4.54x10'4 1.69x1010 10 3.68x100 4.68x109 37 2.75x10'4 1.82x1010 11 2.87x100 9.79x109 38 1.67x10~4 1.96x1010 12 2.23x100 1.36x1010 39 1.0lx10~4 1.96x1010 0 -
13 1.74x10 1.93x1010 40 6.14x10-5 1.95x1010 14 1.35x100 2.19x1010 41 3.73x10-5 1.89x1010 15 1.lix100 4.10x1010 42 2,26x10-5 1.83x1010 16 8.21x10'I 4.77x1010 43 3,37xio-5 1.77x1010 17 6.39x10'I 5.04x1010 44 8.3tx10-6 1.68x1010 18 4.98x10"I 3.70x1010 45 5.04x10-6 1.53x1010 19 3.8Bx10'I b.37x1010 46 3.06x10-6 1.42x1010 20 3.02x10'I 5.38x1010 47 1.86x10-6 1.29x1010 21 1.83x10-1 5.42x1010 4g 3,33xio-6 9.72x109 22 1.11x10*1 4.35x1010 49 6.83x10-7 1.19x1010 23 6.74x10'2 3.00x1010 50 4.14x10'7 1.48x1010 24 4.09x10'2 1.69x1010 51 2.51x10-7 1.45x1,10 25 2.55x10-2 2.30x1010 52 1.52x10'7 1.35x1010 26 1.99x10'2 1,09x1010 53 9.24x10~0 3.88x1010 27 1.50x10-2 1.36x1010 , NOTE: Tabulated energy levels represent the upper energy of each group. 6-28
TABLE 6-12 COMPARISON Of CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE X
, Calculated Measured [Id 2
f(E > 1.0 MeV) (n/cm ] 8.18 x 1018 8.87 x 1018 0.92 f(E > 0.1 MeV) [n/cm2 ] 3.76 x 1019 4.16 x 1019 0.90 dpa 1.66 x 10-2 1.77 x 10-2 o,94 2 f(E<0.414eV)(n/cm) 3.22 x 1018 8.02 x 1018 0.40 e ~ 6-29
- _ - . - , . _._ .- _. . - - _ - .. = = - - . . . - - - . . - - . - -
TABLE 6-13 4 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE AZIMUTHAL ANGLE 0' 15' 30' 45' . 3.11 EFPY 4(E>l.0 MeV) 9.55 x 1017 1.42 x 1018 1.16 x 1018 1.65 x 1018 2 (n/cm ) 4>(E>0.1 MeV) 2.03 x 1018 3.07 x 10 18 2.99 x 10I8 4.22 x 1018 2 , (n/cm ) dpa 1.45 x 10-3 2.15 x 10-3 1.84 x 10-3 2.57 x 10-3 8.0 EFPJ f(E>l.0 MeV) 2.46 x 1018 3.65 x 1018 2.98 x 1018 4.24 x 1018 2 (n/cm ) 4(E>0.1 MeV) 5.22 x 10I8 7.90 x 1018 7.69 x 1018 1.09 x 1019 2 (n/cm ) dpa 3.73 x 10-3 5.53 x 10-3 4.73 x 10-3 6.61 x 10-3 16.0 EFPY 4'(E>l.0 MeV) 4.91 x 1018 7.31 x 1018 5.97 x 1018 8.49 x 1018 2 (n/cm ) f(E>0.1 MeV) 1.04 x 10I9 1.58 x 1019 1.54 x 10I9 2.17 x 1019 2 (n/cm ) dpa 7.46 x 10-3 1.11 x 10~2 9.47 x 10-3 1.32 x 10-2 , _ 6-30
TABLE 6-13 (continued) i NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VfSSEL CLAD / CASE METAL INTERFACE AZIMUTHAL ANGLI O' 15' _ _3 0 ' 45' 32.0 EFPY f(E>1.0 MeV) 9.83 x 1018 1.46 x 1019 1.19 x 10 19 1.70 x 1019 2 (n/cm ) f(E>0.1 MeV) 2.09 x 1019 3.16 x 10l9 3.08 x 1019 4.34 x 1019 2 (n/cm ) dpa 1.49 x 10-2 2.21 x 10-2 1.89 x 10-2 2.64 x 10*2 i b 4 1 6-31
4 r j TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES l 8 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dea SLOPE 2 (n/cm ) (equivalent n/c:r2) Surface I/4 T 3/4 T Surface I/4 T 3/4 T 0* 2.46 x 10 18 1.34 x 10 18 2.86 x 10 I7 2.46 x 10 l8 1.55 x 10 18 5.38 x 10 II 15* 3.65 x 10 18 1.98 x 10 18 4.12 x 10 17 3.65 x 10 l8 2.29 x 10 18 7.81 x 10 I7 30* 2.98 x 10 18 1.63 x 10 18 3.56 x 10 I7 2.98 x 10 18 1.99 x 10 18 7.60 x 10 I7 i 45* 4.24 x 10 18 2.27 x 30 18 4.59 x 10 I7 4.24 x 10 18 2.71 x 10 I8 9.31 x 10 I7 16 EFPY NEUTRON FLUENCE (E > I.0 MeV) SLOPE doa st0PE (cfcm2 ) (equivalent n/cm2) Surface 1/4 T 3/4 T Surface I/4 T 3/4 T 0* 4.91 x 10 18 2.67 x 10 l8 5.72 x 10 I7 4.91 x 10 18 3.09 x 10 18 1.07 x 10 18 15* 7.31 x 10 I8 3.96 x 10 I8 8.25 x 10 I7 7.31 x 10 I8 4.58 x 10 18 1.56 x 10 18 30* 5.97 x 10 18 3.26 x 10 18 7.14 x 10 17 5.97 x 10 18 3.98 x 10 18 1.52 x 10 I8 45* 8.49 x 10 18 4.54 x 10 18 9.21 x 10 17 8.49 x 10 18 5.42 x 10 18 1.87 x 10 l8 6-32
TABLE 6-14 (continued)
- NEUTRON EXP05UPE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES i
j 32 EFPY NEU1RON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 (n/cm ) (equivalent n/cm2 ) l Surface 1/4 T 3/4 T Surface. 1/4 T 3/4 T l 0* 9.83 x 10 18 5.35 x 10 18 1.14 x 10 18 9.83 x 10 I8 6.17 x 1018 2.15 x 10 18
! 15* 1.46 x 10 19 7.% x 10 18 1.64 x 10 18 1.46 x 10 19 9.14 x 10 18 3.13 x 10 18 30* 1.19 x 10 I9 6.53 x 1018 1.42 x 1018 1.19 x 10 I9 7.93 x 1018 3.04 x 10 18
- f. 45* 1.70 x 10l9 9.14 x 10 I8 1.84 x 1078 1.70 x 10 I9 1.09 x 1019 3.73 x 10 18 n
1 l I 1 J e 4
- 6-33
TABLE 6-15 UPDATED LEAD FACTORS FOR DIABLO CANYON UNIT 2 3URVEILLANCL CAPSULES hpult Lead factor (a) U 5.28 X 5.28 W 5.28 Z 5.28 V 4.62 Y 4.62 (a) Plant specific evaluation s F ) 6-34 __ i .
1 1 SECTION 7.0 SVRVEILLANCE CAPSVLE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Diablo Canyon Unit 2 reactor vessel: Capsule Estimated Location Lead Removal fluence 2 Capsule (deg.) Factor Time (b) (n/cm ) U 56.0 5.28 0.99 (a) 3.51 x 1018 (a) X 236.0 5.28 3.11 (a) 8.87 x 1018 (a) Y 238.5 4.62 7.0 1.7 x 1019 (c) W 124.0 5.28 10.0 2.8 x 1019 V 58.5 4.62 Standby --- Z 304.0 5.28 Standby --- e l (a) Plant specific evaluation actual fluence and EfPY (b) Effective full power years (EfPY) from plant startup. (c) Approximate fluence at the reactor vessel clad / base metal interface at end of . life (32 EFPY). l l s 7-1 {- _ . - _ . . . _ . -- __- .
SECTION
8.0 REFERENCES
- 1. Davidson,J.A. and Yanichko,S.E., " Pacific Gas and Electric Company Diablo Canyon Unit No. 2 Reactor Vessel Radiation Surveillance Program,"
WCAP-8783, December 1976.
- 2. Yanichko, S.E., Anderson, S.L, and Albertin, L., " Analysis of Capsule U from the Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel- Radiation Surveillance Program". WCAP-ll851, May,1988.
- 3. "Charpy Curve Fits for DCPP Unit 2 Survelliance Capsule X", Letter from M.
D. Sullivan of Pacific Gas and Electric Company to E. Terek of Westinghouse, Dated November 2, 1990
- 4. Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.
- 5. Regulatory Guide 1.99, Revision 2, " Radiation Errbrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988.
- 6. Section III 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
- 10. ASTM 4370-89,
- Standard Test Methods and Definitions for Mechanical Testing of Steel Products."
- 11. ASTM E8-89b,
- Standard Test Methods of Tension Testing of Metallic .
Materi al s . " l
- 12. ASTM E21-79 (1988), ' Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."
l l
- 13. ASTM E83-85, ' Standard Practice for Verification and Classification of Extensometers."
- 14. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation. Vol. S--Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970.
- 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. Vestovich, J. A., Lancaster, D.B , Signorella, T.L. and Radcliffe, R.E.,
"The Nuclear Design and Core Phsics Characteristics of the Diablo Canyon Unit 2 cycle 1" WCAP-10593, July 1984, (Westinghouse Proprietary)
- 17. Fecteau, M.W., Piplica, A.N., Radcliff, R.E. and Zimmermann, M. W., "The Nuclear Design and Core Phsics Characteristics of the Diablo Canyon Power Plant Unit 2 Cycle 2" WCAP-ll450, May 1987, (Westinghouse Proprietary)
- 18. Fecteau, M.W., Piplica, A.N. and Radcliff, R.E. "The Nuclear Design and Core Phsics Characteristics of the Diablo Canyon Power Plant Unit 2 Cycle 3" WCAP-11962, October, 1988 (Westinghouse Proprietary) 8-2
- 19. 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, l
. 1984.
1 i
. 20. 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.
- 21. ASTM Designation E693-79, " Standard 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.
- 22. 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.
- 23. 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,
- 24. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 25. 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.
l. 8-3
l
- 26. ASTM Designation E263-82,
- Standard Method for Determining fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, i American Society for Testing and Materials, Philadelphia, PA,1984.
- 27. ASTM Designation E264-82, " Standard Method for Determining fast-Neutron
- flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, .
American Society for Testing and Materials, Philadelphia, PA,1984. , l
- 28. ASTM Designation E481-78,
- Standard Method for Measuring Neutron-flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, FA, .
1984. j
- 29. ASTM Designation E$23-82, " Standard Method for Determining fast-Neutron j flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 30. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American I
Society for Testing and Materials, Philadelphia, PA,1984. l
- 31. ASTM Designation E705-79, " Standard Method for Measuring fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM Standards, Section
! 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 32. ASTM Designation E10(5-84, " Standard Method for Application and Analysis of Radiometric Monitcrs for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, l Philadelphia, PA,1984.
- 33. F. A. Schmittroth, FERRET Date_ Analysis Core, HEDL-TME 79-40, Hanford L Engineering Development Laboratory, Richland, WA, Septembtr 1979.
e 1 j 8-4 l u (
- 34. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Soectra Determined by Foil Activation, AFWL-TR-7-41, Vol.1-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
, 35. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al., 1981.
9 4 4 8-5 l
. . _ _ . . . . . . ...- - - . . . - _ . . . . . - . - - - _ . . _ . . . _ = . _ . . _ = . _ - . _ .
i 4 8 4 l l APPENDIX A Load-Time Records for Charpy Specimen Tests i S. l l l A-0
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- - ._ _- . = .
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l t
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w_ 8.0000E402 1.6000E+03 ' 2.4000E+03 Time, microseconds Figure A-11. Load-time record for Specimen PL53.
__ 0909 c. E+03 6.0000 E+03 I 4 . es O- ,.a.
. 4.009u ' '-
>a 1O E+03 _'-~~-. __. g a . u w' S
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e O e A a
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E.4000E+03 E+00 O.000:E+00 B.0000E+02 1.6000E+03 Time, microseconds Figure A-13. Load-time record for Specimen PL50. g
8.Uu00, E+03 6.0000 E403
.a m
A.
- 4.0000 ~ -~ - '+ v s .
,7 3a E+03 , d 'NN x -_
s_ 2.0000 N.s m E+03 .
'm
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0.0000l E+00 0,0000E+00 3.0000E+02 1.600C{+03 2.40C0E+03 Time, microseconds Figure A-14. Load-time record for Specimen PL46. 6 1
e
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~~ ~ ~ ~ ~ ~ " ' ~
8.0000
~~~~~-~~~~~-"~~~~~~~---~~~^~~~~~~~~~~~~~~~~
E403 1 6.0000~ E+03 1 M to I a.
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~~~
0.0000 _
~
E+00 0.0000E+00 3.0000E+02 ~~~"i.~6500E+05~"~'~530555755~~"~ Time, microseconds Figure A-15. Load-time record for Specimen PL56.
t-l ( . . . . _ .... . ... . . ... .... .. __... ..... . . .. .. .... .. .. . . . . . . . . . . . . . . _ . . . . . . . . . . . . . _ _ . ! E+03 i 6.0000 E+03 a. _ 4.0000 _-_ _m 8 1o E+03 c+' ~' s.-s a ' x.%
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2.0000 .
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N~%~ . 0.0000 1 l E+00 'O.000CD 00 8.0000E+02 1.600C9 03 2. 40C4E+03 Time, microseconds Figure A-16. Load-time record for Specimen PL52. D O U S . O &
E+03 ' 6.0000 E+03 M a.
- 4.0000 t
1O E403 m
- 2
-4 2.0000 E+03
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E+00 0.000H +00 a.0000E402 i . 600CE+03
- 2.4006E+03 Time, microseconds Figure A-17. Load-time record for Specimen PT56.
' i oen ,
f *"> . r O j f+ LAJ l
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2.0900 E+03 _s b .,~* ' . s v ,.' , O.0000 " ^ v% z -e n e.e .=m ,._ . - _ _ _ ___- E+00 0.0000E+00 9.0000E402 1.6000E+03 2.4000E403 Time, microseconds Figure A-20. Load-time record for Specimen PT54. 9 6
. . & Q 9
l1 1: Ii 1l1lIjlll41lll1ii1i1I i - 1l1!;1 J u u m m
=
=_4 c.
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.+ . 4 + .+ .+
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~~~~~~~-~~~~-~~~~~~~~~~~~~~~~~~~~~
8 . 0 0 0 0 '~ ~ ~ ~ ~~ ~ " ~~ ~ ~ ~ ~ E+03 6.0000 E+03
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2.0000 E+03 s A, zgO T 0.0000 -
- A. r. .w. . . . . . . ... . . . . . -
. .. . dwa .. .. .Cred- CCwur.C W E+00 0.000 M +00 9.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-22. Load-time record for Specia..a PT46.
S e O O $ 9 9
. . . .- a
'm
~ ~~ "-~ ~" ~"""~~ ~~ ~ ~ - ~ ~ ~ ~ " ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' - ~ ~ -
8.0000 E403 6.0000 E403 M On 4.0000 -
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V ** ^., f. n O.0u00 '.c.crewo -- ._ew.mc._ , . _e___._ E+00 0.000 M +00 3.0000E402 1.6000E+03 2. 40C@E+03 Time, microseconds Figure A-23. Load-time record for Specimen PT60. r
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ " ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ " - ~ ~ ~ " ~ ~ ~ ~ "
8.0000 E+03 6.0000 E+03
.a
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~ ~ ^ ~ ~+ _.
0.0000 3. , .._. .. . ~ E+00 0.0000E400 3.0000E+02 1.600Cf+03 2.4000E403 Time, microseconds Figure A-24. Load-time record for Specimen PT53. W I
l I
~ " ~ ~ - ~" ~ ~ ~ ~ ~~ "" ~ ~~~ ~~~~ ~
8.0000 ~~~~~""-"~~~--"'""~~ E+03 6.0000 E+03 91
", 4.0000 j_,,
> ^
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,, ) 8 E+00 0.000CE+00 .____:_K - - -
B.0000E+02 :_ : _ : .=-f 2. 4000E +03 - . 1.6000E+03 Time, microseconds Figure A-25. Load-time record for Specimen PT51.
1
>s e
CD No record - computer malfunction Figure A-26. Load-time record for Specimen PT58. Pb-O S 6 S S p
\
~ ~ ~ " - ' ~ ~ " " ' " ~ ~ ' " ~ ~ ~ - ~ ~ " " ~ " ~ ~ ~ " " ~ ~ - - " " ~ ~ " " ~ ~ ~ " " ~ ~ ~ " ~ " " " " -
8.0000 E+03 6.0000 E+03 o. i 4.0000 ~___
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.~~~~-
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..........,........____........._....[.._......zr:r........,
E+00 0.0000E+00 B.0000E+0) 4.6000E+03 ' 2. 4 00-3-E+03 1 Time, microseconds Figure A-27. Load-time record for Specimen PT49. y
E+03 6.0000 t E+03
.a m
o. 4.0000 _-~_. Y fo E+03
~
as Q .,.i- ' CD ;. , 2.0000 's~~ E+03 s ._
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0.0000 !. ...........g..........___....... ..... .. ....... . E+00 0.0000E+00 0.0000E+02 'i.600E+03
- 2.400=3E+03 Time, microseconds Figure A-28. Load-time record for Specimen PT55.
. e =
1 . . . . . . l-l 1 l E+03 6.0000 E+03 40 o. g 4.0,000
- r. o.
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o
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=._
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^m 2.0000 ~'
E+03 k 0.0000 . E+00 0.0000E+00 3.0000E+02 ........... 1.6000E+03 2.4000E+03 Time, microseconds Figure A-29. Load-time record for Specimen PT47.
8.0,0u0 . . . . . . . . . _ . . . . . . . . . ... ............. _ _ ......... .. ........... _ .._............ _ ......... ..... . . .. ..... E+03 6.0000 E+03 4.0000 t 'jo E+03 -
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- - - - - , - , , - , - - - . - , - - - - - - , - - - - -.-- - - - , - - - - - - --.---.-- - - - _ _ - . - - -n- - - - _ _ - - - , . , - - , - - _ - - - - , - . - - e t l i ' i
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