ML20045H435
| ML20045H435 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/30/1993 |
| From: | Chicots J, Meyer T, Perock J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20045H433 | List: |
| References | |
| WCAP-13365, WCAP-13365-R01, WCAP-13365-R1, NUDOCS 9307200210 | |
| Download: ML20045H435 (300) | |
Text
.
k WESTINGHOUSE CLASS 3 (Non-Proprietary)
Revision 1 6
4 ANALYSIS OF CAPSULE Y FROM THE WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM J, M. Chicots J. Perock A. Madeyski April 1993 Work Performed Unde. Shop Order KZ0P-106 Prepared by Westinghouse Electric Corporation for the Wolf Creek Nuclear Operating Corporation b%M Approved by:
T.A.Meyer, Manager (
Structural Reliability and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION
~ -
Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
- 1993 Westinghouse Electric Corp.
All Rights Reserved 13 07200210 930713 PDA. ADOCK 05000482 P
i G
PREFACE This report has been technically reviewed and verified. This report was revised to correct a ca'. '41ational error in Appendix B.
Reviewer:
Sections 1 through 5, 7, 8 and E. Terek Appendix A Section 6 S. L. Anderson ODADL h
d 3-4/
Appendix B H. A. Ramirez o
i t
TABLE OF C0hTENTS
.I Section Title
.P_iLqq 1.0
SUMMARY
OF RESULTS 1-1 2.0 itiTRODUCTION 2-1
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0
. TESTING 0F SPECIMENS FROM CAPSULE Y 5-1 5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-4~
5.3 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 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1
8.0 REFERENCES
8-1 APPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS APPENDIX B - HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION APPENDIX C - CALIBRATION REPORTS ii
LIST OF TABLES Table Title P gg i
4-1 Chemical Composition and Heat Treatment of the Wolf Creek 4-3 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the Wolf Creek 5-8 Lower Shell Plate R2508-3 Irradiated at 550*F, Fluence 1.33 x 1019 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Wolf Creek 5-9 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 1.33 x 10 n/cm2 (E > 1.0 MeV) 19 5-3 Instrumented Charpy Impact Test Results for the Wolf Creek 5-10 Lower Shell Plate R2508-3 Irradiated at 550*F, Fluence 1.33 x 1019 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results for the Wolf Creek 5 Weld Metal and Heat-Affected-Zone (HAZ) Metal, Irradiated 19 2
at 550*F, Fluence 1.33 x 10 n/cm (E > 1.0 MeV) 5-5 Effect of 550*F Irradiation to 1.33 x 1019 2
n/cm 5-12 (E > 1.0 MeV) on the Notch Toughness Properties of the Wolf Creek Reactor Vessel Surveillance Materials iii
~
LIST OF TABLES (Continued)
Table Title Eagg 5-6 Comparison of the Wolf Creek 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 the Wolf Creek Reactor Vessel 5-14 19 Surveillance Materials Irradiated at 550*F to 1.33 x 10 n/cm2 (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Parameters at the 6-14 Surveillance Capsule Center 6-2 Calculated Fast Neutron Exposure Rates at the Pressure 6-15 Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux (E > 1.0 MeV) 6-16 t
within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux (E > 0.1 MeV) 6-17 within the Pressure Vessel Wall 6-5 Relative Radial Distributions of Iron Displacement Rate 6-18 (dpa) within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-19 6-7 Monthly Thermal Generation During the First Fuel Cycle of 6-20 j
the Wolf Creek Unit 1 Reactor 6-8 Measured Sensor Activities and Reactions Rates 6-21 iv
LIST OF TABLES (Continued) 1 Table Title
_P_iL41 6-9 Summary of Neutron Dosimetry Results 6-23 6-10 Comparison of Measured and FERRET Calculated Reaction 6-24 Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-25 Capsule Center 6-12 Comparison of Calculated and Measured Exposure Levels 6-26 for Capsule Y 6-13 Neutron Exposure Projections at Key Locations on the 6-27 Pressure Vessel Clad / Base Metal Interface 6-14 Neutron Exposure Values for Use in the Generation of 6-28 Heatup/Cooldown Curves 6-15 Updated Lead Factors for Wolf Creek Unit 1 Surveillance 6-29 Capsules v
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LIST OF ILLUSTRATIONS Fiaure Title EiLqe 4-1 Arrangement of Surveillance Capsules in the Wolf Creek 4-4 Reactor Vessel 4-2 Capsule Y Diagram Showing Location of Specimens, Thermal 4-5 Monitors and Dosimeters 5-1 Charpy V-Notch Impact Properties for Wolf Creek Reactor 5-15 Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-2 Charpy V-Notch Impact Properties for Wolf Creek Reactor 5-16 Vessel Lower Shell Plate R2508-3 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for Wolf Creek Reactor 5-17 Vessel Surveillance Weld Metal 5-4 Charpy V-Notch Impact Properties for Wolf Creek Reactor 5-18 Vessel Weld Heat-Affected-Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Wolf Creek 5-19 Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for Wolf Creek 5-20 Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) vi
~
LIST OF ILLUSTRATIONS (Continued) l Fiaure Title East 5-7 Charpy Impact Specimen Fracture Surfaces for Wolf Creek 5-21 Reactor Vessel Surveillance Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for Wolf Creek 5-22 Reactor Vessel Weld Heat-Affected-Zone Metal 5-9 Tensile Properties for Wolf Creek Reactor Vessel 5-23 Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-10 Tensile Properties for Wolf Creek Reactor Vessel 5-24 Lower Shell Plate R2508-3 (Transverse Orientation) 5-11 Tensile Properties for Wolf Creek Reactor Vessel 5-25 Surveillance Weld Metal 5-12 Fractured Tensile Specimens from Wolf Creek Reactor 5-26 Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-13 Fractured Tensile Specimens from Wolf Creek Reactor 5-27 Vessel-Lower Shell Plate R2508-3 (Transverse Orientation) 5-14 Fractured Tenstle Specimens from Wolf Creek Reactor 5-28 Vessel Surveillance Weld Metal 5-15 Engineering Stress-Strain Curves for Plate R2508-3 5-29 Tensile Specimens AL13 and All4 (Longitudinal Orientation) vii
~
LIST OF ILLUSTRATIONS (Continued)
Fiaure Titig Eggg 5-16 Engineering Stress-Strain Curve for Plate R2508-3 5-30 Tensile Specimens AL15 (Longitudinal Orientation) t 7
5-17 Engineering Stress-Strain Curves for Plate R2508-3 5-31 Tensile Specimens AT13 and AT14 (Transverse Orientation) 5-18 Engineering Stress-Strain Curve for Lower Shell Plate 5-32 R2508-3 Tensile Specimen ATIS (Transverse Orientation) 5-19 Engineering Stress-Strain Curves for Weld Metal 5-33 Tensile Specimens AW13 and AW14 i
5-20 Engineering Stress-Strain Curve for Weld Metal 5-34 Tensile Specimen AW15 i
6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 l
i i
f viii
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vesscl materials contained in surveillance Capsule Y, the second capsule to be removed from the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor pressure vessel, led to the following conclusions:
o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 1.33 x 1019 2
n/cm after 4.79 EFPY of plant operation.
o Irradiation of the reactor vessel lower shell plate R2508-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction (longitudinal orientation), to 19 1.33 x 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 30*F and a 50 ft-lb transition temperature-increase of 40*F.
This results in a 30 ft-lb transition temperature of 10'F and a 50 ft-lb transition temperature of 40*F for longitudinally oriented specimens.
o Irradiation of the reactor vessel lower shell plate R2508-3 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction (transverse orientation), to 1.33 x 19 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 40*F and a 50 ft-lb transition temperature increase of 45'F.
This results in a 30 ft-lb transition temperature of 40*F and a 50 ft-lb transition temperature of 85'F for transversely oriented specimens.
19 2
o The weld metal Charpy specimens irradiated to 1.33 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 50*F and a 50 ft-lb transition temperature increase of j
45'F.
This results in a 30 ft-lb transition temperature of 0*F and a 50 ft-lb transition temperature of 30*F for the weld metal.
i 1-1
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal o
Charpy specimens to 1.33 x 10 n/cm2 (E > 1.0 MeV) resulted in a 19 30 ft-lb transition temperature increase of 50*F and a 50 ft-lb transition temperature increase of 40*F.
This results in a 30 ft-lb transition temperature of -95'F and a 50 ft-lb transition temperature of -70*F for the weld HAZ metal.
Irradiation of lower shell plate R2508-3 (longitudinal orientation) o to 1.33 x 10 n/cm2 (E > 1.0 MeV) resulted in an average upper 19 shelf energy decrease of 27 ft-lbs, resulting in an upper shelf energy of 121 ft-lbs.
o Irradiation of lower shell plate R2508-3 (transverse orientation) to 1.33 x 10 n/cm2 (E > 1.0 MeV) resulted in an average upper 19 shelf energy increase of I ft-lb, resulting in an upper shelf energy of 94 ft-lbs.
The average upper shelf energy of the weld metal decreased 6 ft-lb o
after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV). This 19 results in an upper shelf energy of 94 ft-lb for the weld metal.
o The average upper shelf energy of the weld HAZ metal increased 19 ft-lb after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV). This 19 results in an upper shelf energy of 180 ft-lb for the weld HAZ metal.
o The surveillance capsule Y test results indicate that the lower shell plate R2508-3 30 ft-lb transition temperature shift is less than the Regulatory Guide 1.99 Revision 2 predictions. However, comparison of the 30 ft-lb transition temperature increase for the surveillance weld material is 15'F greater than the Regulatory Guide 1.99 Revision 2 predictions.
Regulatory Guide l'.99 Revision 2 requires a 2 sigma allowance of 56*F for weld metal be added to the predicted reference transition temperature to obtain a conservative l
upper bound value. Thus, the reference transition temperature l
1-2 i
i
l increase for the surveillance weld metal is bounded by the 2 sigma allowance for shift prediction.
o The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life (32 EFPY) of the vessel as required by 10CFR50, Appendix G.
The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 o
MeV) for the Wolf Creek reactor vessel is as follows:
l I9 2
Vessel inner radius * - 2.50 x 10 n/cm 19 E
Vessel 1/4 thickness - 1.36 x 10 n/cm 18 2
Vessel 3/4 thickness - 2.93 x 10 n/cm
- Clad / base metal interface 1-3
SECTION 2.0 INTRODUCTION i
This report presents the results of the examination of Capsule Y, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Wolf Creek reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-10015, entitled
" Kansas Gas and Electric Company Wolf Creek Generation Station Unit No.1 i
Reactor Vessel Radiation Surveillance Program" by L. R. Singerill. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-79, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels". Westinghouse Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule "Y" from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where, the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance capsule "Y" removed from the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor vessel and discusses the analysis of the data.
1 2-1 i
b 2t---
4--
6 4
w i-A 24 u-I SECTION 3.0 I
BACKGROUND The ability of the large steel pressure vessel containing the reactor core and -
its primary coolant te 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 un the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 533 Grade B Class 1 (base material of the Wolf Creek reactor pressure vessel lower shell plate R2508-3) 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 methud for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure,"
Appendix G to Section III of the ASME Boiler and Pressure Vessel Code [5],
The method uses fracture mechanics concepts and is based on the reference i
nil-ductility temperature (RTNDT)-
RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208)[6] 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 plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G to the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
i f
3-1
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, such as the Wolf Creek Reactor Vessel Radiation Surveillance ProgramIll, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase i
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 set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
I 3-2 2
6
~
Y SECTION
4.0 DESCRIPTION
OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Wolf Creek reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the neutron shielding 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 Y was removed after 4.79 Effective Full Power Years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact tension (CT) specimens (Figure 4-2) from the lower shell plate R2508-3 and submerged arc weld metal representative of the intermediate to lower shell beltline weld seam of the reactor vessel. Capsule Y also contained Charpy V-notch specimens from weld Heat-Affected-Zone (HAZ) material. All heat-affected zone specimens were obtained from within the HAZ of plate R2508-3 of the representative weld.
All test specimens were machined from the 1/4 thickness location of the plate.
Test specimens represent material taken at least one plate thickness from the quenched end of the plate.
Base metal Charpy V-notch impact specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation) and also normal to the major working direction (transverse orientation). Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction.
The CT specimens were machined such that the simulated crack in the specimen would propagate normal and parallel to the major working direction for the i
plate specimen and parallel to the weld direction.
4-1
The chemical composition and heat treatment of the surveillance material. is j
presented in Table 4-1.
4 l
Capsule Y contained dosimeter wires of pure copper,. iron, nickel, and -
aluminum-0.15 weight percent cobalt wire (cadmium-shielded and unshielded).
In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were placed in the capsule to measure the integrated flux at specific neutron energy levels.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two alloys and their melting points are as follows:
2.5% Ag, 97.5% Pb Helting Point:
579'F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Helting Point: 590*F (310*C)
The location of the surveillance capsules within the reactor vessel is shown in Figure 4-1.
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in capsule Y is shown in Figure 4-2.
e 4-2
TABLE 4-1 i
I CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE WOLF CREEK REACTOR VESSEL SURVEILLANCE MATERIALS Chemical Comoosition (wt%)
Lower Shell Element Plate R2508-3 Weld Metal
[a]
C 0.20 0.11 Mn 1.45 1.46 P
0.008 0.005 5
0.010 0.011 Si 0.20 0.48 Ni 0.62 0.09 Mo 0.55 0.56 Cr 0.05 0.09 Cu 0.07 0.04 Al 0.032 0.009 Co 0.014 0.010-Pb
<0.001
<0.001 W
<0.01
<0.01 Ti
<0.01
<0.01' Zr
<0.001
<0.001 V
0.003 0.005 Sn 0.002 0.003 As 0.007 0.004 Cb
<0.01
<0.01 N
0.007 0.006 B
<0.001
<0.001 Heat Treatment History Material Temoerature (*F)
Time (Hr)
Coolant Lower Shell Plate, Austenitizing 1575-1525 4
Water quenched t
R2508-3 Tempered 1200-1250 4
Air cooled Stress Relief 1100-1200 8.5 Furnace cooled Weld Metal Stress Relief 1100-1200 10.25 Furnace cooled a.
This weldment was fabricated by Combustion Engineering, Inc., using 3/16 inch Mil B-4 weld filler wire,' heat number 90146 and Linde 124 flux, lot number 1061 and is identical to that used in the actual fabrication of the reactor vessel intermediate to lower shell girth weld.
4-3
REACTOR VESSEL O.
CORE BARREL 330 NEUTRON PAD (3.85) Z CAPSULE U-(3.85)
I I
V (3.65)
J se.s--
se.s. L
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270*
-- 90
- 35*
~
2.5*
i
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N i
5Q*_
48.7 *
(3.65) Y
/I I\\
(3.85) X i
V (3.85)
-l 210*
180*
i i
PLAN VIEW i
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Figure 4-1.
Arrangement of Surveillance Capsules in the Wolf Creek Reactor Vessel
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4-4
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Figste 4 2 Capsule Y Diagram Showing l
tocation of' 5pecimans.
Thermal Monitors and Dostmeters 45 l
SECTION 5.0 TESTING OF SPECIMENS FROM CAPSULE Y 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H[3], ASTM Specification E185-82l73, and Westinghouse Remote Metallographic Facility I
(RMF) Procedure MHL 8402, Revision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and 1
spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-10015[Il. No discrepancies were found.
Examination of the two low-melting point 579'F (304*C) and 590*F (310*C) eutectic alloys indicated no melting of either type of thermal monitor.
Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579'F (304*C).
The Charpy impact tests were performed per ASTM Specification E23-88[8] and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with a GRC 8301 instrumentation system, feeding information into an IBM XT computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E ).
From the load-time curve D
(Appendix A), the load of general yielding (Pgy), the time to general yielding (tgy), the maximum load (Pg), and the time to maximum load (tg) can be determined. Under some test conditions, a sharp drop in. load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture. load (Pp), and the load at which fast fracture terminated is identified as the arrest load (P )-
A 5-1 1
I The energy at maximum load (Eg) sas determined by comparing the energy-time record and the load-time record. The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen.
Therefore, the propagation energy for the crack (E ) is the difference p
between the total energy to fracture (E ) and the energy at maximum load.
D The yield stress (oy) was calculated from the three-point bend formula having the following expression:
ay - Pgy * {L/';B*(W-a)2*C])
(1) i where L - distance between the specimen supports in the impact testing machine; B - the width of the specimen measured parallel to the notch; W - height of the specimen, measured perpendicularly to the notch; a - notch depth.
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 ( - 45' and p -
0.010", Equation 1 is valid with C - 1.21.
Therefore (for L - 4W),
gy ]/[B(W-a)2]
(2) oy - Pgy * {L/[B*(W-a)2*1.21]} - [3.3P W
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 ay is in units of psi and Pgy 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 A370-89I93 The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
5-2
i Tension tests were performed on a 20,000-pound Instron Model 1115, split-console test machine, per ASTM Specification E8-89b(10] and E21-79 (1988)[Ill, and RMF Procedure 8102. Revision 1.
All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45. The upper pul? 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 to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-85[12),
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 dif'iculty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.
Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip.
In the test 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 obtain desired specimen temperatures.
Experiments indicated that this method is accurate to 2*F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5-3
5.2 Charov V-Notch Imoact Test Results t
The results of the Charpy V-notch impact tests performed on the various 19 2
materials contained in Capsule Y, which was irradiated to 1.33 x 10 n/cm (E > 1.0 MeV), are presented in Tables 5-1 through 5-4 and are compared with unirradiated resultsill as shown in Figures 5-1 through 5-4.
The transition temperature increases and upper shelf energy decreases for the Capsule Y materials are summarized in Table 5-5.
Irradiation of the reactor vessel lower shell plate R2508-3 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major 19 rolling direction of the plate (longitudinal orientation) to 1.33 x 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 30*F and in a 50 ft-lb transition temperature increase of 40*F.
This resulted in the 30 ft-lb transition temperature of 10*F and a 50 ft-lb transition temperature of 40*F (longitudinal orientation).
The average Upper Shelf Energy (USE) of the lower shell plate R2508-3 Charpy specimens (longitudinal orientriion) resulted in a energy decrease of 27 ft-lb a
after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV) at 550*F. This 19 results in an average USE of 121 ft-lb (Figure 5-1).
Irradiation of the reactor vessel lower shell plate R2508-3 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation) to 1.33 x 10 n/cm2 (E >
19 1.0 MeV) at 550*F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 40*F and in a 50 ft-lb transition temperature increase of 45'F.
This resulted in the 30 ft-lb transition temperature of 40*F and a 50 ft-lb transition temperature of 85'F (transverse orientation).
i 5-4 i
i
The average USE of the lower shell plate R2508-3 Ch:rpy specimens (transverse orientation) resulted in an energy increase of I ft..lb after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV) at 550*F. This res.nlted in an average I9 USE of 94 ft-lb (Figure 5-2).
l Irradiation of the reactor vessel core region weld metal Charpy specimens to 1.33 x 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-3) resulted in a 19 50*F increase in 30 ft-lb transition temperature and a 50 ft-lb transition temperature increase of 45'F.
This resulted in a 30 ft-lb transition I
temperature of 0*F and the 50 ft-lb transition temperature of 30*F.
The average USE of the reactor vessel core region weld metal resulted in an energy decrease of'6 ft-lb after irradiation to 1.33 x 10 n/cm2 (E > 1.0 19 MeV) at 550*F. This resulted in an average USE of 94 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 1.33 x 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-4) resulted in a 19 30 ft-lb transition temperature increase of 50*F and a 50 ft-lb transition temperature ii.. =ase of 40*F.
This resulted in a 30 ft-lb transition temperature of -95'F and the 50 ft-lb transition temperature of -70*F.
The average USE of the reactor vessel weld HAZ metal experienced an energy increase of 19 ft-lb after irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV) 19 at 550*F.
This resulted in an average USE of 180 ft-lb (Figure 5-4).
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.
5-5
4 A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the various Wolf Creek surveillance materials with predicted valt.es using the methods of NRC Regulatory Guide 1.99, Revision 2I43 is presented in Table 5-6.
This e.omparison indicates that the transition temperature increases and the upper shelf energy decreases of the 19 2
lower shell p' ate R2508-3 resulting from irradiation to 1.33 x 10 n/cm (E > 1.0 MeV) are less than the Regulatory Guide predictions. This comparison also indicates that the upper shelf energy decrease of the weld metal resulting from irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV) is less than the 19 Regulatory Guide prediction. Comparison of the 30 ft-lb transition temperature increase for the surveillance weld material is 15'F greater than the Regulatory Guide prediction. However, the NRC Regulatory Guide 1.99, Revision 2 requires a 2 sigma allowance of 56*F for weld metal be added to the predicted reference transition temperature to obtain a conservative upper bound value. Thus, the reference transition treperature increase for the surveillance weld metal is bounded by the 2 sigma allowance for shift prediction.
Table 5-6 also presents the E0L (32 EFPY) calculated upper shelf enecqy values for the surveillance materials.
The load ^ime records for the individual instrumented Charpy specimen tests are shown in Appendix A.
The calibration certification for all the equipment used in charpy and tensile tests are given in Appendix C.
5.3 Tension Test Results P
The results of the tension tests performed on the various materials contained in capsule Y irradiated to 1.33 x 10 n/cm2 (E > 1.0 MeV) are presented in 19 Table 5-7 and are compared with unirradiated resultsIll as shown in Figures 5-9 through 5-11.
The results of the tension tests performed on the lower shell plate R2508-3 I9 2
(longitudinal orientation) indicated that irradiation to 1.33 x 10 n/cm (E > 1.0 MeV) at 550'F caused less than a 6 ksi increase in the 0.2 percent 5-6
offset yield strength and less than a 6 ksi increase in the ultimate tensile strength when compared to unirradiated datall3 (Figure 5-9).
The results of the tension tests performed on the lower shell plate R2508-3 (transverse orientation) indicated that irradiation to 1.33 x 10 n/cm2 (E 19
> 1.0 MeV) at 550*F caused less than a 6 ksi increase in the 0.2 percent offset yield strength and less than a 9 ksi increase in the ultimate tensile strength when compared to unirradiated dataIll (Figure 5-10).
The results of the tension tests performed on the reactor vessel core region weld metal indicated that irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV) at 19 550*F caused less than a 5 ksi increase in the 0.2 percent offset yield strength and less than a 5 ksi increase in the ultimate tensile strength when compared to unirradiated dataIll (Figure 5-11).
The small increases in 0.2% yield strength and tensile strength exhibited by the lower shell plate R2508-3 and the weld metal indicate that this material is not highly sensitive to irradiation to 1.33 x 10 n/cm2 (E > 1.0 MeV), as 19 is also indicated by the Charpy impact test results.
The fractured tension specimens for the lower shell plate R2508-3 material are shown in Figures 5-12 r
>, while the fractured specimens for the weld metal are shown in Figu e 5-14 The engineering stress-strain cu. 'es for the tension tests are shown in Figures l
5-15 through 5-19.
5.4 Comoact Tension Tests Per the surveillance capsule testing program with the Wolf Creek Nuclear Operating Corporation, the 1/2-T compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center.
5-7 2
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE WOLF CREEK LOWER SHELL PLATE R2508-3 IRRADIATED AT 550*F, FLUENCE 1.33 x 10 n/cm2 (E > 1.0 MeV) 19 Cf f-lb) f mi$s fmm k
Sample No.
F Longitudinal Orientation AL63
-50
-46) 17 15 (0.38) 10 AL67
-35
-37) 22 17 (0.43) 15 AL68 0
-18) 23 15 (0.38) 20 AL61 10
-12) 60 45 (1.14) 40 AL72 25
- 4) 35 47) 25 (0.64) 25 AL75 35 2) 65 88) 45 (1.14) 45 AL62 50 10) 64
( 87) 45 (1.14) 45 AL64 75 24) 75 102) 44 (1.12 50 AL71 100 38) 53 72) 43 (1.09 60 AL74 125 52) 116 157) 67 1.70 100 AL70 150 66) 109 148) 72 1.83) 100 AL66 175 79 134 (182) 75 1.91) 100 AL69 250 (121 145 (197) 78 1.98) 100 AL65 275 135 88 97 78 1.98 100 AL73 300 149) 136 84 84 2.13) 100 Transverse Orientation AT71
-35
-37) 12 16 8
(0.20 10 AT63
-10
-23) 12 10 (0.25 10 AT69 10
-1 20 17 0.43 15 AT66 25 23 20 0.51 20 30 30 0.76 30 AT64 30 AT70 50 35 28 0.71 30 AT65 60 39 34 0.86 35 AT75 75 62 42 1.07 45 AT61 85 47 39 0.99 45 AT68 100 38) 51 43 1.09 50 AT67 125 64 51 1.30 65 AT62 150 74 56 1.42 80 AT73 175 89 69 1.75 100 AT72 225 1
100 78 1.98 100 AT74 275 1
94 68 (1.73 100 5-8
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE WOLF CREEK REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED I9 AT 550*F, FLUENCE 1.33 x 10 n/cm2 (E > 1.0 MeV) m$s)
)
)
)
Sample No.
C)
Weld Metal AW67
-125
- 87) 4 5
2 0.05) 10 AW72
-100
- 73 NA NA NA NA NA AW63
- 90
- 68 5
3 0.08 15 AW62
- 60
- 51 8
6 0.15 15 AW69
- 35
- 37) 23 14 0.36 45 AW71
- 10
- 23 29 22 0.56 50 AW68 0
- 18 47 35 0.89) 65 9
30 26 0.66) 75 AW65 15 is!
e n
- nj n
1.24 90 AW66 65 18 73 49 AW54 100 38) 82 60 1.52) 95 AW61 175 79) 88 64 1.63 100 AW74 250 121) 97 72 1.83 100 AW70 300 149) 97 74 1.88 100 HAZ Metal AH69
-200
-129 8
2 0.05 5
AH70
-160
-107 7
3 0.08 5
AH61
-125
- 87 18 9
0.23 10 AH74
-100
- 73 33 18 0.46 15 AH67
- 75
- 59 35 18 0.46 20 Et
- 1%!
1:'!
a AH68
- 25
- 32 05 31 0.79 40 AH65 0
- 18 95 49 1.24 70 9
114 56 1.42 75 AH72 15 AH62 50 10 153 80 (2.03 100 2N 9
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5-11
TABLE 5-5 I9 EFFECT OF 550*F IRRADIATION TO 1.33 x 10 n/cm2 (E > 1.0 MeV)
CM THE NOTCH T0uGHNESS PROPERilES OF THE WOLF CREEK REACTOR VESSEL SURVEILLANCE MATERIALS III III III III Average 30 ft-lb Average 35 mil Average 50 ft-lb Average Energy Transition lateral Expans ton Transition Absorption at Teaperat'ure ( *F)
Temperature ( *F)
Tebperature ( *F)
Full Shear (ft-lb)
Material unirradiated Irradiated AT unirradiated irradiated AT unstradiated irradiated AT untrradiated irradiated A(ft-ib)
Plate R2508-3
- 20 10 30
- 10 45 55 0
40 40 148 121
- 27 (longitudinal)
Plate R2503-3 0
to 40 25 45 20 40 85 45 93 94
+ 1 Y
g (Transverse)
Veld Metal
- 50 0
50
- 25 20 45
- 15 30 45 100 94
- 6
.
- AC Metal
-145
- 95 50
- 90
- 25 65
-110
- 70 40 161 180
+ 19 4
(1) " AVERAGE" is defined as the value read f rom the curve fitted through the data points of the Charpy tests (Figures 5-1 through 5-4)-
g -i - 9. -e d 'R-o l
l 1
TABLE 5-6 COMPARISON OF THE WOLF CREEK SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SliELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS
-G %
30 ft-lb Transition Temp. Shift Upper Shelf Energy Decrease End-o f-Li fe Upper Shelf Fluence Predictsd (a) Measured Predicted (a)
Measured Eneray (a)
I9 2
Haterial Capsule 10 n/cm
(*F)
(*F)
(%)
(%)
(ft-lb)
Plate R2508-3 0
0.339 31 30 15 2
112.5 (Longitudinal)
Y 1.33 47 30 20 18 Plate R2508-3 U
0.339 31 25 15 0
112.5 h
(Transverse)
Y 1.33 47 40 20 0
Weld Metal U
0.339 23 20 15 8
76.0 Y
1.33 35 50 20 6
HAZ Metal 0
0.339 65 13 Y
1.33 50 0
Note:
(a) Based on Regulatory Guide 1.99, Revision 2
i TABLE 5-7 I
TENSILE PROPERTIES FOR THE WOLF CREEK REACTOR VESSEL SURVEILLANCE MATERIALS IRRADIATED AT 550*F TO 1.33 X 1019 n/cm2 (E > 1.0 MeV) l Test 0.25 Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp.
Strength Strength Load Streme Strensth Elonsstion Elonsation in Arem j
Material Number 1*D, (kei)
(kei)
(kio)
(kei) ikei)
(x)
(2)
(2)
Plate R250s-3 ALis so es.7 87.6 2.65 153.9 53.9 15.0 31.4 e6 (Longitudinal) AL14 115 83.2 88.3 2.43 180.9 50.4 12.0 20.8 72 AL18 550 57.8 34.5 2.75 157.s 56.0 11.7 27.7 e5 l
I Plate R250s-3 AT13 75 88.2 se.O a.10 138.8 63.2 13.5 28.1 54 (Traneveree)
AT14 140 64.2 83.0 2.80 147.s 57.o 12.0 23.0 61 AT15 550 5s.8 85.6 8.00 116.s 61.1 12.0 20.3 4e-Weld AW1s o
32.0 Os.s 3.20 17s.3 65.2 12.0 26.2 63 AW14 ISO 75.4 39.8 3.05 175.1 62.1 10.5 21.s 66 Y
AW15 Eso 71.s 91.7 3.0a 159.7 et.e 10.5 22.4 61
(
l L
4 D
,r-w -
n v
+
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+ -
--r s
ar, s
-e
+ -, --
..~
(*C)
-150
-100
-50 0
50 100 150 200 i
I.
I I
i 1
l l
4 100
_, ~
.G 80
% 60 w
G 40 P 'e j
20 0
100 2'5-
'M o.
-j80
?
V,"
2.0 t
g 60 o
1.5 7
9 40 o * *M
- 1.0 5 20 0.5 I
I "
I I
I I
O 0
O LNEADIATED U
e IRRADIATED 050*r1 FLtOCE 03 : In n/cn!
160 v
'N 200 140 5 "
e e
120 g
160 m
o
- 9 100 5
80 o
5 G
60 80 o
40 2
- r 20 E
I I
I I
I I
I I
0 O
-200
-100 0-100 200 300 4 00 TEMPERATURE (*f)
)
Figure 5-1.
Charpy V-Notch Impact Properties for Wolf Creek Reactor Vessel-Lower Shell Plate R2508-3 (Longitudinal Orientation) i 5-15 1
l
~
i
('C)
-150 -100
-50 0
50 100 150 200 l
I I
I 100 2%
I I
I g 80
!!f 60 W
2
" 40 3
20 0
100 2.5 3 80 e
2.0 6
n &
_mm
- 60 0/e "
1.5 m
ax c
9 40 2
1.0 6 i
W 2Ifr 3 20 0.5 I
I I
I I
I I
0 O
O lMRRAMATD e mAMATO c'ic'rt rtuon tn x #n/d 120 160 100 o o
- o o~
120 o
o 80 B
e
.h n
5 3
60 80 s
- F 6
2
$ 40 9
(r 40 20 I
0 O
Tuc
-200
-100 0
100 200 300 400 500 acasec TEMPERATURE (*D Figure 5-2.
Charpy V-Notch Impact Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) 5-16
l l
l l~
('C)
'i
-150
-100.
-50 0
50 100 150-200 1
I I
I I
I I
i 100
%gMt oo 8 80 60 8
o
" 40 20 I
0 100 2.5 h80
_8 2.0 1.5 m 60 0
2 k
o e
" 40 1.0 3
e dr 3 20 o
0.5 o
I' I
I I
0 0
o INDutAIDA B e suma es'rt ruso: os : IEn/cnt 120 160 100 7
e=
120 o 80 o
2 2
o b
n0 D
60 80 G
e Q
(r 40 m*r w
40 e
20 o
l i
i i
i i
0 0
-100 0
100 200 300 400 500
-200 wu Jmo TEMPERATURE ('F)
Figure 5-3.
Charpy V-Notch Impact Properties for Wolf Creek Reactor Vessel Surveillance Weld Metal 5-17
C'C)
-150 -100
-50 0
50 100 150 200 100
^^ -
~~
x__ et, 8 80 g60 4 40 20 0'
100
_ _ _ 33 2.5 e
a-g g 80 t
Lw, -
2.0 e 60 1.5 40 C*r 1.0 h20 0.5 I
- I I
I I
I 0
O o ualutato e nutmTo es% ruoa: L:n x d'a/d 200 m
180 240 160
,o
~
140 2
'.120 160 b
D 100 o
~
80 O 60 o
e 80 e*r 40 40 0
l l
l l
l 0
0
-200
-100 0
100 200 300 400 TEMPERATURE (*D m
acmu figure 5-4.
Charpy V-Notch Impact Properties for Wolf Creek Reactor Vessel Weld Heat-Affected-Zone Metal 5-18
l
~
W ![{~
ryWW;f
,, ~j,[,
A3 B-(*
1l$l j.?l&
. f:
E.
E
& '1 E
h__
N:
$N.--
y AL63 AL67 AL68 AL61 AL72
)
i'?
- 4}i.:..
.ff;,[.
AL75 AL62 AL64 AL71 AL74 REMRI AL70 AL66 AL69 AL65 AL73 figure 5-5.
Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 1 5-19 l
2LT I?Q&
rY[-T i%fh[
$144$
(
- ,'
- ;:l?5 Rf W *-
.}f&+f
$$5%.,
?07 T
fy m
m c-m M
N, "ff. -
.Q:l lQ. *': y d=jles
=- m r #_
i :!'.i 9e AT71 AT63 AT69 AT66 AT64 v
..L _,.-
T;3~sg
.a; ff l
,~,{.{
f.
- M is IO AT70 AT65 AT75 ATS1 AT68 l
AT67 AT62 AT73 AT72 AT74 i
i figure 5-6.
Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) i 5-20 4
C
t l,NA_ lhh!*
v.
i& ~~
__g_
.b
'M l
=~gtA
- +.
l f,
y-.q w AW67 AW72 AW63 AW62 AW69
,Tm
[~'
f?]
E, NA E
as AW71 AW68 AW65 Alt 75 AW73 M T*
$g hi?
~
r NY W:
~
AW66 AW64 AW61 AW74 AW70 l
l Figure 5-7.
Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Surveillance Weld Metal l
5-21
~.
I
. 's.
f'# ~ i
?
y-t';
& 4::
p.: ;p.
l
- D
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Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Weld Heat-Affected-Zone Metal 5-22 l
i
(*C) 0 50 100 150 200 250 300 120 l
l l
j l
l l-800 110 700 100 3 90 Ril*TE TEGLE STMTH 600 6
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Figure 5-9.
Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2503-3 (Longitudinal Orientation) 5-23
(* C) 0 50 100 150 200 250 300 I
l l
l l
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TEMPERATURE ('F) j Figure 5-10.
Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) i 5-24
i
(*C) 0 50 100 150 200 250 300 120 l
l l
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l-800
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Figure 5-11.
Tensile Properties for Wolf Creek Reactor Vessel Surveillance Weld Metal 5-25 l
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Fractured Tensile Specimens from Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) 5-26
d 4
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Figure 5-13.
Fractured Tensile Specimens from Wolf Creek Reactor Vessel Lower 1
Shell Plate R2508-3 (Transverse Orientation) i l
5-27 i
i I_,
e l
i I
t t
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figure 5-14.
Fractured Tensile Specimens from Wolf Creek Reactor Vessel Surveillance Weld Metal i
5-28
t 90.00 so.o&
70.oo 5
60.00-M (ff 50.00-0 C
40.00-
<n 30.00-AL13 20.00-50 F 1 o.oo-o.oC 0.0 0.10 0.20 0.30 STRAIN, IN/IN 1 oo.oC 90.o &
80.00-70.00-60.00-50.00-
=y 40.00-co.oe AL14 20.00-115 F 10.00-o.oc 0.0 0.10 0.20 0.30 STRAIN, IN/IN Figure 5-15.
Engineering Stress-Strain Curves for Plate R2508-3 Tensile Specimens All3 and All4 (Longitudinal Orientation) 5-29
100.00 90.00-80.00-70.00-O
\\
m, 60.00-50.00-cy 40.00-30.00-20.00-550*F 10.00-0.00 0.0 0.10 0.20 STRAIN, IN/IN Figure 5-16.
Engineering Stress-Strain Curve for Plate R2508-3 Tensile Specimens All5 (Longitudinal Orientation) 5-30
. ~.
4 400.00 90.00-
[
80.00-70.00-en, 60.00-50.00-Eg 40.00-30.00-20.00-AT13 10.00-75'F 0.00 0.0 0.10 0.20 O.30 STRAIN, IN/IN 100.00 90.00-80.00-70.00-E 60.00-ui 50.00-g[
40.00-30.00-20.00-AT14 10.00-140*F 0.00 0.0 0.10 0.20 0,30 STRAIN, IN/IN Figure 5-17.
Engineering Stress-Strain Curves for Plate R2508-3 Tensile Specimens AT13 and AT14 (Transverse Orientation) 5-31
t 100.00 90.00-70.00-
__m*
60.00-m 50.00-h 40.00-30.00-AT15 20.00-10.00 550 F 0.00 O.0 0.10 0.20 STRAIN, IN/lN Figure 5-18.
Engineering Stress-Strain Curve for Lower Shell Plate R2508-3 Tensile Specimen AT15 (Transverse Orientation) 5-32
11c 100-90-80-F i
70-vi 60-en w
CD 4&
AW13 20-O*F 10-0'0 0.10 052 0 O.30 STRAIN, IN/lN 100.00 90.00-80.00-70.00-en*
60.00-cn, 50.00-40.00-30.00-AW14 20.00-10.00-0'0U 0!10 0.20 O.0 STRAIN, IN/IN Figure 5-19.
Engineering Stress-Strain Curves for Weld Metal Tensile Specimens AW13 and AW14 5-33
i l
i
.)
100.00 90.00-30.00-70.00-i en*
60.00-vi 50.00-tcy 40.00-30.00-20.00-AW15 10.00-550 F o.00.
0.0 0.10 0.20 STRAIN, IN/IN I
Figure 5-20.
Engineering Stress-Strain Curve for Weld Metal Tensile Specimen AW15 5-34
d i
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 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 spectrum, flux, fluence) to which the test specimens were exposed must be known.
Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor 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 i
accurate evaluation of damage gradients through the pressure vessel wall.
l Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water Reactor l
l l
6-1
i Surveillance Results,"[25] recommends reporting displacements per iron atom (dpa) along with fluence (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..[23] 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 surveillance Capsule Y.
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 58.5, 61.0*, 121.5*, 238.5*,
241.0*, and 301.5* relative to the core cardinal axes as shown in Figure 4-1.
A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1.
The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core.
6-2
From a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pad and the reactor vessel.
In order to properly determine the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {p(E > 1.0 Mev), p(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/p(E > 1.0 MeV), within the pressure vessel geometry.
The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e.,
the 1/4T, 1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locai.iuns of interest for each cycle of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles.
It is important to note that the cycle specific neutron source distributions utilized in these analyses included.not only spatial variations of fission rates within the reactor core; but, also accounted for the effects of varying neutron yield 6-3
~
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 inner radius and through the thickness of the pressure vessel wall.
3.
Enable a direct comparison of analytical prediction with measurement.
4.
Establish a mechanism 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 DOT two-dimensional discrete ordinates code [13] and the SAILOR tross-section libraryII43. The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications.
In these analyses anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an S8 order of angular quadrature.
The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 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 periphery.
Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2a 6-4
level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.
All adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library.
Adjoint source locations were 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, O geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, p (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:
R(r,0)-[r[0[E I(r, 0, E) S (r, 0, E) r dr de dE where: R (r, 0)
- d (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 analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/d (E > 1.0 MeV) is insensitive to changing core source distributions.
In the application of these 4
adjoint importance functions to the Wolf Creek Unit I reactor, therefore, the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/d (E > 1.0 MeV) and p (E > 0.1 MeV)/d (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.
6-5
The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design reports for the first five operating cycles of Wolf Creek Unit 1[15 through 20],
Selected results from the neutron transport analyses 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),
l p(E > 0.1 MeV), and dpa] are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared.
Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycles 1 through 5 plant specific power distribution.
It is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself.
Radial gradient information for neutron flux (E > 1.0 MeV),
neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations.
Exposure parameter distributions within the wall may be obtained by normalizing the caiculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.
6-6
a For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by:
41/4T(45*)
- d(220.27, 45*) F (225.75, 45*)
where:
(1/4T(45*)
- Projected neutron flux at the 1/4T position on the 45' azimuth 4 (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 ( (E > 0.1 MeV) and dpa/sec.
The DOT calculations were carried out for a typical octant of the reactor.
However, for the neutron pad arrangement in Wolf Creek Unit 1, the pad extent for all octants is not the same.
For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron pad extending from 32.5 - 45.0 degrees which produces the maximum flux. Other octants have neutron pads spanning larger azimuthal sectors which provide more shielding. For'the octant with the 12.5 degree pad, the maximum flux to the vessel occurs near 25 degrees and the values in the tables for the 25 degree angle are vessel maximum values.
Exposure values for 0,15, and 45 degrees can be used for all octants; values in the tables for 25 and 35 degrees are maximum values and only apply to octants with a 12.5 degree neutron pad.
6.3 Neutron Dosimetry The passive neutron sensors included in the Wolf Creek Unit I surveillance program are listed in Table 6-6.
Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the 6-7
evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest [d (E > 1.0 Mev), p (E > 0.1 MeV), 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 form, were placed in holes drilled in spacers at ;everal axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest.
Rather, the activation or fission process is a measure of the integrated effect 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 well known.
In particular, the following variables are of interest:
o The specific activity of each monitor.
o The operating history of the reactor.
o The energy response of the monitor.
o The neutron energy spectrum at the monitor location.
o The physical characteristics of the monitor.
The specific activity of each of the neutron monitors was determined using established ASTM procedures (21 through 34]. 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 Wolf Creek Unit I reactor during cycles 1 through 5 was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the i
applicable period.
The irradiation history applicable to Capsule Y is given in Table 6-7.
Measured and saturated reaction product specific activities as well as measured 6-8
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 [35),
~
TheFERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra.
In the FERRET evaluations, a log-normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations.
In general, the measured values f are linearly related to the flux 4 by some response matrix A:
(s,a)
(s)
(a) f
-I A
d 9
19 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 a
p i
9 19 g
relates a set of measured reaction rates R$ to a single spectrum p by g
the multigroup cross section ajg.
(In this case, FERRET also adjusts the cross-sections.) The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.
6-9
In the FERRET analysis of the dosimetry data, the continuous quantities (i.e.,
fluxes and voss-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [36]. -This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed to the group scheme used in FERRET.
The cross-sections were also collapsed into the 53 energy-group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/B-V dosimetry file.
Uncertainty estimates and 53 x 53 s.ovariance matrices were constructed for each cross section.
Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.
For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight.
In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is used:
gg,-Rf+R N
R,P g g gg, where RN specifies an overall fractional normalization uncertainty (i.e.,
complete correlation) for the corresponding set of values. The fractional uncertainties R specify additional random uncertainties for group g that are g
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 a (0 specifies the strength of the latter term).
6-10 1
For the a priori calculated fluxes, a chort-range correlation of a - 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1.
Strang long-range correlations (or anticorrelations) were justified cased on information presented by R.E.
MaerkerI373 Haerker's results are closely duplicated when a - 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 Capsule Y dosimetry are given in Table 6-9.
The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1.33 x 10 n/cm2 (E > 1.0 MeV) with an associated 19 uncertainty of i 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 achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.
A summary of the measured and calculated neutron exposure of Capsule Y is -
presented in Table 6-12.
The agreement between calculation and measurement falls within 14% for all fast neutron exposure parameters listed. The thermal neutron exposure calculated for the exr,osure period under predicted the measured value by 56 percent.
Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13.
Along with the current (4.79 EFPY) exposure derived from the Capsule Y measurements, projections are also provided for an exposure period of 15 EFPY and to end of vessel design life (32 EFPY).
In the evaluation of the future exposure of the reactor pressure vessel the average exposure rates derived from cycles 1 through 5 were employed.
In computing these average exposure rates, the calculated averages were also scaled by the average measurement / calculation ratios observed from evaluations of dosimetry from Capsules Y and U.
This procedure resulted in the following 6-11
i bias factors being applied to the analytical results:
Flux (E > 1.0 MeV)
Bias - 1.182 Flux (E > 0.1 MeV)
Bias - 1.142 dpa/sec Bias - 1.149 In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Wolf Creek Unit I reactor coolant system, exposure projections to 15 EFPY and 32 EFPY were also employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14.
In order to access RTNDT vs. fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations:
p' (1/4T)
- p (Surface) {dpa (Su face)}
p' (3/4T)
- p (Surface) (dpa (Surface))
Using this approach results in the dpa equivalent fluence values listed 'in Table 6-14.
In Table 6-15 updated lead factors are listed for each of the Wolf Creek Unit I surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules.
i l
6-12
i l
i q
(TYPICAL)
Co
- 58.50
- 61.0' Fe Cu y
Y r
{
h:I c
3
- 81.625 IN.
j I
L h
v gg I
NEUTRON PAD l
Figure 6-1.
Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 i
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER p(E > 1.0MeV)
((E > 0.lMev)
Iron Olsplacement Rate 2
2 In/cm -sec1 In/cm -sec1 Idoa/sec1 29.0*
31.5*
29.0*
31.5*
29.0*
31.5*
DESIGN BASIS 1.13 X 10ll Il 11 1.21 X 10 5.07 X 10 5.42 X 10ll 2.21 X 10-10 2.36 X 10-10 CYCLE 1 8.31 X 1010 8.86 X 1010 3.73 X 10ll 11 1.63 X 10-10 1.74 X 10-10 3.98 X 10 es CYCLE 2 8.84 X 1010 9.76 X 1010 3.96 X 10ll Il 1.73 X 10-10 1.91 X 10-10 4.39 X 10 h:
10 CYCLE 3 7.18 X 10 7.78 X 1010 3.22 X 1011 ll 1.40 X 10-10 1.52 X 10-10' 3.50 X 10 CYCLE 4 6.99 X 1010 7.68 X 1010 3.14 X 1011 ll 1.37 X 10-10 1.50 X 10-10 3.45 X 10 CYCLE 5 7.05 X 1010 7.50 X 1010 11 3.16 X 10 3.37 X 10ll 1.38 X 10-10 1.47 X 10-10 l
e 4
wa e,
TABLE 6-2 l
CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 2
In/cm -seci d(E > 1.0MeV) 0.0*
15.0*
25,0*
35,0*
15E.
10 2.81 X 1010 10 2.45 X 10 10 3.01 X 10 10 2.66 X 10 DESIGN BASIS 1.78 X 10 10 10 10 2.05 X 1010 10 2.93 X 10 2.21 X 10 1.79 X 10 CYCLE 1 1.31 X 10 10 2.51 X 1010 10 10 2.31 X 1010 2.06 X 10 CYCLE 2 1.44 X 10 2.06 X 10 10 10 10 10 1.69 X 1010 1.89 X 10 1.55 X 10 1.71 X 10 CYCLE 3 1.18 X 10 10 10 10 10 10 1.78 X 10 1.87 X 10 1.58 X 10 1.82 X 10 CYCLE 4 1.31 X 10 10 1.51 X 10 1.69 X 1010 10 10 10 1.93 X 10 CYCLE 5 1.27 X 10 1.82 X 10 2
In/cm -seci 6(E > 0.1MeV) 0.0*
15.0*
2L,_0
- 35.0*
45.0*
10 10 7.05 X 10 10 10 6.95 X 10 DESIGN BASIS 3.71 X 1010 5.61 X 10 8.22 X 10 10 10 10 10 6.04 X 1010 5.10 X 10 5.14 X 10 CYCLE 1 2.72 X 10 4.06 X 10 10 10 10 4.34 X 1010 6.31 X 1010 5.84 X 10 6.28 X 10 CYCLE 2 3.00 X 10 10 10 4.29 X 10 10 10 10 4.41 X 10 CYCLE 3 2.45 X 10 3.55 X 10 5.15 X 10 10 10 10 10 5.09 x 1010 4.49 X 10 4.56 X 10 CYCLE 4 2.71 X 10 3.74 X 10 10 10 4.23 X 10 CYCLE 5 2.64 X 1010 3.84 X 1010 5.26 X 1010 4.29 X 10 Iron Atom Displacement Rate Idoa/secl 0.0*
15,0*
25,0*
35,0*
45.0*
DESIGN BASIS 2.78 X 10-11 4.13 X 10-Il 5.04 X 10-11 4.14 X 10-Il 4.48 X 10-11 CYCLE 1 2.04 X 10-Il 2.99 X 10-Il 3.70 X 10-11 3.04 X 10-11 3.27 X 10-11 CYCLE 2 2.24 X 10-II.3.19 X 10-Il 3.87 X 10-Il 3.48 X 10-II 4.00 X 10-Il CYCLE 3 1.84 X 10-II 2.61 X 10-Il 3.16 X 10-11 2.63 X 10-11 2.73 X 10-Il CYCLE 4 2.03 X 10-Il 2.75 X 10-Il 3.12 X 10-Il 2.68 X 10-Il 2.90 X 10-11 CYCLE 5 1.98 X 10-11 2.82 X 10-Il 3.23 X 10-13 2.56 X 10-11 2.69 X 10-Il 6-15
TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0*
15*
25' 35'
-45' 220.27(1) 1.00 1.00 1.00 1.00 1.00 220.64 0.976 0.979 0.980 0.977 0.979 221.66 0.888 0.891 0.893 0.891 0.889 222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 0.554 0.552 0.543 226.95 0.462 0.460 0.465 0.463 0.452 228.28 0.386 0.384 0.388 0.386 0.37I 229.60 0.321 0.319 0.324 0.321 0.311 230.92 0.267 0.263 0.275 0.267 0.257 232.25 0.221 0.219 0.225 0.221 0.211 233.57 0.183 0.181 0.185 0.183 0.174 234.89 0.151 0.149 0.153 0.151 0.142 236.22 0.124 0.122 0.126 0.124
-0.116 237.54 0.102 0.100 0.104 0.102 0.0945 238.86 0.0828 0.0817 0.0846 0.0835 0.0762 240.19 0.0671 0.0660 0.0689 0.0679 0.0608 241.51 0.0538 0.0522 0.0550 0.0545 0.0471 242.17(2) 0.0506 0.0488 0.0518 0.0521 0.0438 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-16 i
I
TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0*
15*
25' 35' 45'
+
220.27(l) 1.00 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 1.00 221.66 1.00 1.00 1.00 0.999 0.995 222.99 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901 225.63 0.874 0.865 0.874 0.850 0.842 226.95 0.818 0.808 0.818 0.792 0.782 228.28 0.761 0.750 0.716 0.734 0.721 229.60 0.705 0.693 0.704 0.677 0.662 230.92 0.649 0.637 0.649 0.621 0.605 232.25 0.594 0.582 0.594 0.567 0.549 233.57 0.540 0.529 0.542 0.515 0.495 234.89 0.487 0.478 0.490 0.465 0.443 236.22 0.436 0.428 0.440 0.416 0.392 237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201 242.17(2) 0.233 0.226 0.237 0.223 0.188 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-17
y
-.~.
j:
TABLE 6-5 RELATIVE. RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT. RATE-(dpa)
WITHIN THE PRESSURE VESSEL WALL -
Radius (cm).
0*
15*'
-25* -
'35*
45*'
i 220.27(l)'
1.00 1.00
'1.00 -
1.00 1.00' 220.64 0.984 0.981 0.984 0.983 0.984-221.66 0.912 0.909 0.917 0.921-0.915 222.99-0.815 0.812 0.826-0.833 0.821 224.31 0.722 0.719 0.737.
0.747
.0.730' 225.63
.0.638 0.634 0.656 0.668 0.647.
226.95 0.563 0.559 0.584 0.597 0.572 228.28 0.497.
0.493 0.519' O.533 0.506 229.60 0.439 0.435 0.462 0.475-0.447 l
230.92 0.387 0.383 0.410 0.423 0.394.
j 232.25 0.341 0.338 0.364 0.376 0.347-233.57 0.300 0.297 0.322 0.334 0.305' 234.89 0.263 0.261-0.285 0.295 0.266 236.22 0.230 0.228 0.250 0.260 0.231:
237.54 0.199 0.198 0.218 0.227 0.199 238.86 0.171 0.170 0.189-0.196 0.169 l
240.19 0.145 0.144 0.161 0.167 0.140 241.51 0.121 0.119 0.135 0.139-0.113' 242.17(2) 0.116
-0.113 0.128-
.0.134-0.106 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius i
J 6-18
TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Weight
Response
Product Yield Material Interest Fraction Ranae Hal f-Li fe
(%)
Copper Cu63(n,a)Co60 0.6917 E > 4.7 MeV 5.272 yrs Iron Fe54(n,p)Mn54 0.0582 E > 1.0 MeV 312.2 days Nickel NiS8(n,p)CoS8 0.6830 E > 1.0 MeV 70.90 days
?
Uranium-238*
U238(n,f)CsI37 1.0 E > 0.4 HeV 30.12 yrs 5.99 G
Neptuntum-237*
Np237(n,f)CsI37 1.0 E > 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum
- CoS9(n,0)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobalt-Aluminum 0059(n,8)Co60 0.0015 E > 0.015 MeV 5.272 yrs
- Denotes that monitor is cadmium shielded.
w g
g,vw w
.-ww
TABLE 6-7 MONTHLY THERMAL GENERATION DURING THE FIRST FIVE FUEL CYCLES OF THE WOLF CREEK UNIT 1 REACTOR I
THERMAL THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION GENERATION MONTH (MW-hr) MONTH (MW-br) MONTH (MW-hr) MONTH (MW-hr) 5/85 0
1/87 1533313 9/88 2450165 5/90 1003923 6/85 356676 2/87 2192444 10/88 492163 6/90 2442569 7/85 1025780 3/87 2471746 11/88 0
7/90 2515109 8/85 1643803 4/87 2247475 12/88 0
8/90 2534494 9/85 2053023 5/87 2436662 1/89 2095086 9/90 2453417 10/85 2086772 6/87 2250313 2/89 2113705 10/90 2533710 11/85 2366472 7/87 2066874 3/89 253b552 11/90 2421081 12/85 2368666 8/87 2527262 4/89 2454150 12/90 2531359 1/86 2480479 9/87 1954923 5/89 2498149 1/91 2363291 2/86 2005668 10/87 0
6/89 2448863 2/91 1840498 3/86 2513225 11/87 0
7/89 2493515 3/91 1969185 4/86 933250 12/87 0
8/89 2534633 4/91 1506284 5/86 2341310 1/88 1216547 9/89 2453774 5/91 1692964 6/86 1670026 2/88 956585 10/89 2516573 6/91 2434282 7/86 2210358 3/88 2526972 11/89 2450503 7/91 2534580 8/86 2439547 4/88 2452604 12/89 2536033 8/91 2466385 9/86 2406802 5/88 2533966 1/90 2534772 9/91 1221097 10/86 1219774 6/88 2451743 2/90 2017613 l
11/86 0
7/88 2531412 3/90 509723 12/86 650000 8/88 2533606 4/90 0
l l
l 6-20
~
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec-om)
(dis /sec-om)
(RPS/ NUCLEUS)
Cu-63 (n,a) Co-60 5
5 3.41 x 10 Top 1.37 x 10 5
5 2.99 x 10 Middle 1.20 x 10 5
5 Bottom 1.21 x 10 3.01 x 10 5
5 4.78 x 10'17 Average 1.26 x 10 3.14 x 10 Fe-54(n.p) Mn-54 6
6 Top 1.66 x 10 3.03 x 10 6
6 Middle 1.49 x 10 2.72 x 10 6
6 Bottom 1.48 x 10 2.71 x 10 6
6 4.49 x 10-15 Average 1.54 x 10 2.82 x 10 Ni-58 (n p) Co-58 6
7 Top 8.04 x 10 4.45 x 10 6
7 Middle 7.38 x 10 4.08 x 10 Bottom 7.33 x 10 4.06 x 107 6
6 4.20 x 107 5.99 x 10-15 Average 7.58 x 10 U-238 (n,f) Cs-137 (Cd) 6 3.00 x 10-14 5
5.40 x 10 Middle 5.43 x 10 6-21
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec-am)
(dis /sec-am)
(RPS/ NUCLEUS)
Np-237(n,f) Cs-137 (Cd) 6 7
2.65 x 10-13 Middle 4.40 x 10 4.38 x 10 r
Co-59 (n,8) Co-60 I
Top 2.59 x 10 6.45 x 107 7
7 7
Bottom 2.57 x 10 6.40 x 10 7
7 4.19 x 10-12 Average 2.58 x 10 6.42 x 10 Co-59 (n,B) Co-60 (Cd) 7 7
Top 1.30 x 10 3.24 x 10 7
7 Middle 1.36 x 10 3.39 x.10 7
7 Bottom 1,39 x 10 3.46 x 10 7
7 2.19 x 10-12 Average 1.35 x 10 3.36 x 10 6-22
TABLE 6-9
SUMMARY
OF NEUTRON 00SIMETRY RESULTS TIME AVERAGED EXPOSURE RATES 2
10
( (E > 1.0 MeV) {n/cm -sec}
8.77 x 10 8%
2 ll p (E > 0.1 MeV) {n/cm -sec}
3.90 x 10 i 15%
dpa/sec 1.69 x 10-10 11%
2 10 p (E < 0.414 eV) (n/cm -sec) 8.29 x 10 21%
INTEGRATED CAPSULE EXPOSURE 2
19 8%
4 (E > 1.0 MeV) {n/cm )
1.33 x 10 2
20 4 (E > 0.1 MeV) {n/cm )
5.91 x 10 15%
dpa 2.56 x 10-2 11%
2 I9
+ (E < 0.414 eV) {n/cm )
1.25 x 10 21%
NOTE: Total Irradiation Time - 4.79 EFPY 6-23
I
. TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER 1
Adjusted Reaction Measured Calculation
[M Cu-63 (n,a) C0-60 5.50x10-17 5.64x10 1.03 Fe-54 (n,p) Mn-54 5.92x10-15 5.80x10-15 0.98 Ni-58 (n,p) Co-58 7.88x10-15 7.87x10-15 1,00 U-238 (n,f) Cs-137 (Cd) 3.41x10-I4 3.30x10-14 0.97 Np-237 (n,f) Cs-137 (Cd) 3.11x10-13 3.23x10-13 1.04 Co-59 (n,8) Co-60 (Cd) 2.66x10-12 2.67x10-12 1.00 Co-59 (n,0) Co-60 5.30x10-12 5.26x10-12 o,gg 6-24
TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy AdjusgedFlux Energy AdjusgedFlux (n/cm sec)
Group (Mev)
(n/cm -sec)
Group (Hev) 1 6
28 9.12x10 1.72x1010 1
1.73x10 6.95x10 2
1.49x10 1.56x10 29 5.53x10-3 2.22x1010 1
7 1
7 9
3 1.35x10 5.98x10 30 3.35x10-3 6.94x10 j
4 1.16x10 1.32x10 31 2.84x10-3 6.62x109 1
8 5
1.00x10 2.89x10 32 2.40x10-3 6.35x109 1
8 8
10 0
4.89x10 33 2.03x10-3 1.78x10 6
8.61x10 7
7.41x10 1.11x10 34 1.23x10-3 1.62x1010
]
0 9
0 9
10 8
6.07x10 1.57x10 35 7.49x10-4 1.49x10 9
4.97x10 3.31x10 36 4.54x10-4 1.41x1010 0
9 0
9 10 10 3.68x10 4.42x10 37 2.75x10-4 1.50x10 0
9 10 11 2.87x10 9.44x10 38 1.67x10-4 1.55x10 i
12 2.23x10 1.33x10 39 1.01x10-4 1.62x1010 0
10 0
10 40 6.14x10-5 1.61x1010 13 1.74x10 1.90x10 0
10 10 14 1.35x10 2.12x10 41 3.73x10-5 1.59x10 10 0
10 42 2.26x10-5 1.56x10 15 1.11x10 3.90x10 10 10 43 1.37x10-5 1.52x10 16 8.21x10-1 4.48x10 10 10 44 8.31x10-6 1.46x10 17 6.39x10-1 4.66x10 10 45 5.04x10-6 1.35x1010 18 4.98x10-1 3.38x10 10 19 3.88x10-1 4.74x1010 46 3.06x10-6 1.27x10 10 10 47 1.86x10-6 1.17x10 20 3.02x10-I 4.86x10 9
21 1.83x10-1 4.79x1010 48 1.13x10-6 8.68x10 10 49 6.83x10-7 1.09x1010 22 1.11x10-1 3.81x10 23 6.74x10-2 2.63x1010 50 4.14x10-7 1.44x1010 10 51 2.51x10-7 1.41x1010 24 4.09x10-2 1.48x10 10 52 1.52x10-7 1.32x1010 25 2.55x10-2 1.94x10 9
10 26 1.99x10-2 9.52x10 53 9.24x10-8 3.74x10 10 27 1.50x10-2 1.20x10 NOTE: Tabulated energy levels represent the upper energy of each group.
6-25
TABLE-6-12 COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE Y Calculated Measured Q3 f(E > 1.0 MeV) {n/cm )
1.15 x 1019 1.33 x 1019 2
0,87 2
19 f(E > 0.1 MeV) (n/cm }
5.16 x 1019 5.91 x 10 0.97 dpa 2.25 x 10-2 2.56 x 10-2 0.88 2
18 19 f(E < 0.414 eV) (n/cm }
5.47 x 10
.1.25 x 10 0.44 6-26
TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 4.79 EFPY O'
15*
25' 30' 45'
+
(E > 1.0 Mev) 2.32 X 1018 18 18 18 3.31 X 10 3.62 X 10 3.62 X 10 3,44 x 1018
[n/cm2]
+
(E > 0.1 MeV) 4.67 7 1018 18 6.73 X 10 9.55 X 1018 9.56 X 1018 8.31 X 1018
[n/cm2]
Iron Atom Displacements 3.52 X 10-3 4.98 X 10-3 5.89 X 10-3 5.90 X 10-3 5.32 X 10-3
[dpa]
15.0 EFPY I
0*
15*
25*
30' 45*
+
(E > 1.0 Mev) 7.47 X 1018 1.06 X 1019 19 1.17 X 10 1.17 X 1019 1.12 X 1019 l
[n/cm2]
i m
A f
(E > 0.1 MeV) 1.50 X 1019 2.17 X 1019 19 3.09 X 10 3.09 X 1019 2.72 X 1019 (n/cm2]
Iron Atom Displacements 1.13 X 10-2 1.60 X 10-2 1.91 X 10-2 1.91 X 10-2 1.74 X 10-2
[dpa]
32.0 EFPY O'
15*
25*
30' 45' t
(E > 1.0 Mev) 1.59 X 1019 I9 I9 19 19 2.27 X 10 2.50 X 10 2.50 X 10 2.39 X 10
[n/cm2]
+
(E > 0.1 MeV) 3.20 X 1019 19 4.62 X 10 6.59 X 1019 6.60 X 10l9 5.80 X 1019
[n/cm2]
Iron Atom Displacements 2.41 X 10-2 3.42 X 10-2 4.07 X 10-2 4.07 X 10-2 3.71 X 10-2
[dpa]
l l
TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES 7
15 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2
2 (n/cm )
(equivalent n/cm )
j Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T I7 18 18 1.63 x 1018 0*
7.47 x 1018 18 8.67 x 10 7.47 x 10 4.72 x 10 4.06 x 10 18 19 18 18 15' l.06 x 10I9 5.78 x 10 1.24 x 1018 1.06 x 10 6.72 x 10 2.33 x 10 7.39 x 10 2.56 x 1018 18 19 I9 18 18 1.17 x 10 25*(a) 1.17 x 10 6.36 x 10 1.36 x 10 19 18 18 19 18 18 30*(a) 1.17 x 10 6.39 x 10 1.37 x 10 1.17 x 10 7.68 x 10 2.85 x 10 45*
1,12 x 10 6.09 x 10 1.30 x 1018 1.12 x 10 7.08 x 10 2.46 x 1018 l9 18 19 18 32 EFPY g,
NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2
2 (n/cm )
(equivalent n/cm )
Surface 1/4 T 3/4_I Surface ll4_T 3/4 T l9 3.49 x 1018 18 1.85 x 10 1.59 x 10I9 18 19 8.66 x 10 1.01 x 10 O'
l.59 x 10 19 18 18 15*
2.27 x 10 1.23 x 1019 2.64 x 10 2.27 x 10I9 19 4.97 x 10 1.43 x 10 18 18 19 5.47 x 10 19 1.36 x 1019 2.90 x 10 2.50 x 1019 1.58 x 10 25*(a) 2.50 x 10 18 19 19 6.08 x 10 19 I9 2.93 x 1018 2.50 x 10 1.64 x 10 30*(a) 2.50 x 10 1.36 x 10 18 19 5.24 x 10 19 19 2.78 x 10 2.39 x 1019 18 1.51 x 10 45*
2.39 x 10 1.30 x 10 (a) Maximum point on the pressure vessel
.. ~.
TABLE 6-15 UPDATED LEAD FACTORS FOR WOLF CREEK UNIT 1 SURVEILLANCE CAPSULES Caosule Lead Factor U
Withdrawn Y
3.75(a)
V 3.76(b)
W 4.08(b)
X 4.08(b)
Z 4.08(b)
(a) Plant specific evaluation based on end of cycle 5 calculated fluence.
(b) Projection based on average flux through cycle 5.
9 6-29
~
SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE
~
The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Wolf Creek reactor vessel:
Capsule Estimated Location Lead Fluence Capsule (deg.)
Factor Retaoval Time (a)
(n/cm )
2 U
58.5 3.85 1.08 (Removed) 3.39 x 10I0 (Actual)
Y 241.0 3.75 4.79 (Removed) 1.33 x 1019 (Actual)
V 61.0 3.76 8.5 2.50 x 1019 (b)
X 238.5 4.08 14.0 4.46 X 10I9 W
121.5 4.08 Standby Z
301.5 4.08 Standby (a) Effective Full Power Years (EFPY) from plant startup.
(b) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY).
l 7-1 l
i 0
- -m -
m-
=
SECTION
8.0 REFERENCES
1.
L.R. Singer, et. al., " Kansas Gas and Electric Company Wolf Creek Generating Station Unit No.1, Reactor Vessel Radiation Surveillance Program," WCAP-10015, June 1982.
2.
S.E. Yanichko, et. al., " Analysis of r.apsule U from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program," WCAP-11553, August 1987.
3.
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.
4.
Regulatory Guide 1.99, Proposed Revision 2, " Radiation Damage to Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, February,1986.
5.
Section III of the ASME Boiler and Pressure Vessel Code, Appendix G,
" Protection Against Nonductile Failure."
6.
ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels."
7.
ASTM E185-82, " Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)."
l 8.
ASTM E23-88, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials."
9.
ASTM A370-89, " Standard Test Methods and Definitions for Mechanical Testing of Steel Products."
- 10. ASTM E8-89b, " Standard Test Methods of Tension Testing of Metallic Materials."
8-1 l
+
4 1
- 11. ASTM E21-79(1988), " Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."
- 12. ASTM E83-85, " Standard Practice for Verification and Classification of Extensometers."
13.
R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data I
Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970.
14.
"0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".
15.
P.C. Cook, et. al., "The Nuclear Design and Core Physics Characteristics of the Wolf Creek Generating Station Unit 1 - Cycle 1",
WCAP-10483, February 1984. (Proprietary) i 16.
D.S. Leach, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 2", WCAP-11251, Revision 1, December 1986 (Proprietary) 17.
D.S. Leach, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 3", WCAP-11543, September 1987 (Proprietary) 18.
D.S. Leach, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 4", WCAP-11956, October 1988 (Proprietary) 19.
M.M. Baker, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 5",
WCAP-12530, April 1990 (Proprietary) 20.
H.Q. Lam, et. al., "The Nuclear Parameters and Operations Package for Wolf Creek Unit 1 - Cycle 6", WCAP-13079, November 1991 (Proprietary) 8-2
- 21. ASTM Designation E482-89, " 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, 1991.
l
- 22. ASTM Designation E560-84, " Standard Recommended Practice for Extrapolating l
Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section l
12, American Society for Testing and Materials, Philadelphia, PA,1991.
/
- 23. 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, 1991.
- 24. ASTM Designation E706-87, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991,
- 25. ASTM Designation E853-87, " 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,1991,
- 26. ASTM Designation E261-90, " 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, 1991.
27.
ASTM Designation E262-86, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 28. ASTM Designation E263-88, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
8-3
- 29. ASTM Designation E264-87, " 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,1991.
- 30. ASTM Designation E481-86, " 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, PA, 1991.
- 31. ASTM Designation E523-87, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, 1
American Society for Testing and Materials, Philadelphia, PA,1991.
- 32. ASTM Designation E704-90, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
33.
ASTM Designation E705-90, " Standard Method for Measuring Fa.d-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 34. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
35.
F. A. Schmittroth, FERRET Data Analysis Core, HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
i 36.
W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
37.
EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981.
8-4
6 f
a APPENDIX A Load-Time Records for Charpy Specimen Tests The load-time records for the individual instrumented Charpy specimen tests are
- hown in Figures A-1 through A-31.
As shown in Figures A-14, A-19, A-24, A-26 and'A-31, some instrumented test records exhibit a sharp drop in load during the early portions of the test.
This is caused by resonances that are excited in either the specimen or the test machine by the initial shock of the tup striking the specimen. This phenomenon is generally referred to as " ringing".
It does not adversely affect the Charpy results.
It is possible to apply smoothing technique's to'the data to remove some of these effects, but resolution in other portions of the curve would be sacrificed. The most pronounced effects of ringing usually occur before general yielding occurrs and the analysis of the data is not seriously affected. Therefore, the data has been presented without significant smoothing.
k A-0
D O
A A
O O
L L
E T
R S
U E
T R
C R
A A
R
=
F p
A w
=,
P P
/
i I D
A O
i l i i i L
r d
M r
U oc M
e I
r X
A e
t m
M M
i
=
I t
T y
da P
=
o ig I
I 1 1 I I
I I I I I I i I i l
de z
i lae d
I g
t 1
A erug a
W, a
i m
F t
L A
R D
EN A E O G L
= O L
I I i l l g g Y
y E
Gl g
P Y t
04 3
>A
m ipa.M eat 4.sa w
g i
E 1
I e
-e-S e
_ '{
L
- A
^
q
- 8. 9 2,6
- 3. 8
- 4. s te
,=
..=c.
e
.a =
w g
h e-:
l
.s..
I S
II k
A, 5.
TM 4 suMC 8 Figure A-2.
Load-time records for Specimens AL63 and AL67 A-2
i I
I 1
i i
a -.
n.
i I
8 s
In.
W M A m a _..
=
pnew t.L M8 68 s
I 5
g :-
' U A.
u u
=
Figure A-3, 1,oad-time records for Specimens AL68 and AL61 A-3
.w a
1 g
t I
8 s
~
l ;_.1 tit 4 vestt 3
.u, i
s a
g 3
I I
s 8
s l.
}
=
_-nn~
Figure A-4.
Load-time records for Specimens AL72 and AL75 A-4
= u.
u.
I, 2
s" I8 i
i.
t.
m
. it.
= =.
=
I i
8 3
I s'
Ei i
VM 4 IEEE D Figure A-5.
Load-time records for Specimens AL62 and AL64 A-5
u.
c.
3 I
S 18 S
~ ~_
YtfE 4.EEE 9
. c.
e.
4 i
e g
g.
I t
8 S
l3 11st a sIEE 3 Figure A-6.
Load-time records for Specimens AL71 and AL74 A-8 y-i-
~
~.
..v.
c.
i i
i s
g I
i s
1.:.
TI4 i 815Z l 3
I s"
I :.
i 1
1IE 6 slutE '>
1 i
1 Figure A-7.
Load-time records for Specimens AL70 and AL66 4
A-7
F
. m.
m.
i i
S I
s' I :..
rw
.=>
nn a u, u.
g S
I 8
s II r=
.=>
I Figure A-8.
Load-time records for Specimens AL69 and AL65 A-s
=
EEL dLFB aft MU i
i i
g l
J
.r..
2 s
w l
I i
\\
e
.9 1.9
- 8. 9
- 8. 0 4.9
,, e i
718E 463
..m
.m g
f g
-e-s' l i-1 m.
. a n_
5.
Tlst t net 3 Figure A-9.
Load-time records for Specimens AL73 and AT71 A-9
e m hr
.sL etu etu i
s i
g 3
IE..
S I ;_
1 h
.9 B. 9
- 8. 0
- 3. 0
- 4. 6
.. e Ytst 4 sEEE $
i i
s g
i I:
8 s
1 E8 t
i Y %,w m m, _
i
=
c.
I Figure A-10.
Lond-time records for Specimens AT63 and AT69 A-lo
NM est eves else 8
3 i
a g
I I
E 3,
5
- =
l l
b -
f A A e a m.
49
- 4. 0
- 4. 0
- 3. 0 e, p
- 3. 3 TisE t sett >
t
.e e,ee f
3 3
I q
i I
r._
.i lt f'
g
^"^
r-i..
- e..
=
.a, Figure A-11.
Load-time records for Specimens AT66 a.nd AT64 A-11
..n.
.n.
.x I8 ft E 8M3 $
I S
I8
\\
..-3 Figure A-12.
Load-time records for Specimens AT70 and AT65 A-12
s*
I :_
- i..
.e, peggsty edL AT&B 4f48 i
s i
3 3
I
,i li i
fig t etKC D Figure A-13.
Load-time records for Specimens AT75 and ATS1 A-13
i
+
~
=~
~
3 i
i s
g 3
I..
3 S
h
~
l iA. A A A _ _
B. O
- 8. 9
- 3. 9
- 4. 9 Le TM i sugg 3 i
~
s a
i I
g 3
I3..
8 s
I:
~ __ _
s.
,=
.c, t
Figure A-14.
Load-time records for Specimens AT68 and AT67 j
A-14
I i
1 1
4 1
'i l
..q 1
i i
i s
g Is..
.a.
[ a.
1 N
f tpE
( suEC 7
.=
3 I
s" l.
TIME e suK 1 Figure A-15.
Load-time records for Specimens AT62 and AT73 A-15
AT72 No record.
Computer malfunction.
8 s
88
=
Figure A-16.
Load-time records for Specimens AT72 and AT74 i
A-16
aw
, ~ ~
3 I
s" R.
~
s.
n.
..a.
AW72 No record.
Computer malfunction.
Figure A-17.
Load-time records for Specimens AW67 and AW72 A-17
e
- EIL.43 esa g
i s
s j-I I-j-5 e
i l
% P-l
.9 L. 8
- 4. 0
- 3. 6
- 4. 0 3.e
=
c.
t W
s I8 l
.m. _ _ _ __
YIIC 4 E-E 3 Figure A-18.
Load-time records for Specimens AW63 and AW62 A-18
4 i
i s
g I
E_
- _
S w
h _. _ _
e i
i
- 1. 8
- f. 0
- 3. 0 4,0 m
.a.
. -i I
g-I g
s
[_
E_
i m
.a.
Figure A-19.
Load-time records for Specimens AW89 and AF71 A-19
e eview -
mit sans aime i
4 g
I
.E e.
S e
e.
R.,
I i
l a
e
.3
- 1. 0
- 8. 0
- 3. s
- 4. s S.c -
TIst i sen 3 a
g
- 7..
I
. g.
5 l8 f
T3st 4 am 3 Figure A-20.
Load-time records for Specimens AY68 and AW65 i-A-20
AW75 No record.
Computer malfunction.
g 3
I E
8 s
f i
If n
u TIE i fMt 3
(
l Figure A-21.
Load-time records for Specimens AY75 and AY73 i
l l
A-21 I
a a
i SnenF esk mas mas 3
3 i
g b
I E_ ;_
e" I :-
e a
I I
g
.9 a.0
- 8. 8 3e te
- 5. 9 TisK 4 seK t 4
5 4
g'
=
I s ;_
58 L
i 1
1 e_
e N.
- a...
..e
=
,s.=,
i i
Figure A-22.
Load-time records for Specimens AW66 and AW64 A-23 i
i
ENtew IEL mal maa 4
3 I
a g
I a *_
~
5" E
.4 8.6
.8
- 3. 0
- 4. 0 S8 THE t HKC D 6
3 3
I g
P I
.i l
lt i
m
.. = >
Figure A-23.
Load-time records for Specimens AWS1 and AW74 A-sa
4 e
I 1
i i
g i
I
~-
e w
I I ~-
i
\\
l l
1 e
r i
.9
- 9. 8
- e. 9 S,.
TM fM3 I
2 I
S
\\
l 18 i-f l
l l
k.
Figure A-24.
Load-time records for Specimens AT70 and AH69 A-24
k l
MM 4EL met ep.
g s
s t-3 I
t e
- I=
=
3 3
w r
2 L.
a e
- 1. 0
- 8. 0
- 3. 8
- 4. 9
- 3. s itM i surc a DoubM 4EL ett met 6
i e
I g
I 8
s I :.
i
- - t L_
_t_
_^..
- e. _, m.. -
- 4..
Figure A-25.
Load-time records for Specimens AH70 a.nd AH61 A-25
, ~.
3 J,
I E
5 Ii I
t ru e.__ _ _ _
.=>
l i
n mer mar
, onow l
l I
8 s
Ii 1
u~--_
TM 4 EuRE D Figure A-26.
Load-time records for Specimens AH74 and AH67 A-26
~~
i i
I i
.z. _
.s If 4
3
..4 6
g..
m
.a.
888'I hEL 8DEP8 yvsew V
g r
?
g I
8 s
If
.a.
Figure A-27.
Load-time records for Specimens AH64 and AH71 A-27 a
g I
s s
I e* * ~
8 3
g }.
- ~
\\
m
.9 8.8 4,8
- 3. 8 4, 0 S. 9 n
4 i
i I
g
- 3. _
I 8
I :.
r=
.a, Figure A-28.
Load-time records for Specimens AH68 and AHB5 I
I l
A-28
^
~~
=-
i i
g k
f I
i 5 ;.
lt
~
=
.m.
AH62 No record.
Computer malfunction.
f Figure A-29, Load-time records for Specimens AH72 and AH62 A-29
I
~~
= ~n
-n I
i i
g I
s s*
l I" l
f 3_
71st 4 #EEE 3 AH63 No record.
Computer malfunction.
S Figure A-30.
Load-time records for Specimens AH73 and AH63 A-30
4 4
1 i
r
. i i
1 I
I f
1 WM tEL ene een l
i i
i i
g 1
- =
3 I
s"
)
i I
.3
- 1. 6
- 4. 4
- 3. 0
- 4. 0
.. e ifE t susc 3
}
g I
I 3
g ;._
i Tiet i figEE 3 Figure A-31.
Lond-time records for Specimens AH66 and AH75 9
A-31
APPENDIX B Heatup and Cooldown Limit Curves for Normal Operation i
s 4
i s
i B-0
s TABLE OF CONTENTS
~
Section Title E191 LIST OF ILLUSTRATIONS B-2 LIST OF TABLES B 1 INTRODUCTION B-4 2
FRACTURE TOUGHNESS PROPERTIES B-4 3
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS B-5 P
4 HEATUP AND C00LDOWN PRESSURE-TEMPERATURE LIMIT CURVES B-8 5
CALCULATION OF ADJUSTED REFERENCE TEMPERATURE B-9 6
REFERENCES B-20 ATTACHMENT 1: DATA POINTS FOR HEATUP AND C00LDOWN CURVES' B-21 e
B-1
LIST OF ILLUSTRATIONS.
Fiaure-Title Eagg B-1 Wolf Creek Reactor Coolant System Heatup Limitations
. B-14'
'I (Heat up rate up to 60*F/hr and 100*F/hr) Applicable for the First 10 EFPY~(With No Margins For Instrumentation Errors)
B-2 Wolf Creek Reactor Coolant System Cooldown Limitations-B-15 (Cooldown Rates up to 100*F/hr) Limitations Applicable for the First 10 EFPY (With No Margins For Instrumentation Errors) 1 B-3 Wolf Creek Reactor Coolant System Heatup Limitations B-16~
(Heat up rate up to 60*F/hr and 100*F/hr) Applicable for the First 12.3 EFPY (With No Margins'For Instrumentation Errors).
j B-4 Wolf Creek Reactor Coolant System Cooldown Limitations
' B-17 i
(Cooldown Rates up to 100*F/hr) Limitations Applicable for; the First 12.3 EFPY (With No Margins For. Instrumentation. Errors)
B-5 Wolf Creek Reactor Coolant System Heatup Limitations B (Heat up rate up to 60*F/hr and 100*F/hr). Applicable -
for the First.16 EFPY (With No Margins For Instrumentation Errors)
B-6 Wolf Creek Reactor Coolant System Cooldown Limitations B-19 (Cooldown Rates up to 100*F/hr) Limitations Applicable for '
the First 16 EFPY (With No Margins For Instrumentation Errors) l B-2 e.,
- ~
m..
r w
,-...y
LIST OF TABLES Table Title Eaqe B-1 Wolf Creek Reactor Vessel Toughness Table B-11 (Unirradiated)
B-2 Summary of Adjusted Reference Temperatures (ART's) at 1/4T B-12 and 3/4T Locations B-3 Calculation of Adjusted Reference Temperatures for the B-13 Limiting Wolf Creek Reactor Vessel Material - Lower Shell P1 ate, R2508-3 4
B-3
- 1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced ARTNDT-is designated as the higher of either the drop weight nil-ductility RTNDT transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.
increases as the material is exposed to fast-neutron radiation.
RTNDT Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)Ill. Regulatory Guide 1.99, Revision 2 is used for the calculation of RTNDT values at 1/4T and 3/4T locations (T is the thickness of the vessel at the beltline region).
2.
FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan [2]. The pre-irradiation fracture-toughness properties of the Wolf Creek reactor vessel are presented in Table 1.
B-4
3.
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Ky, for l
the combined thermal and pressure stresses at any time during heatup or 1
cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time. KIR is obtained from the reference-fracture toughness curve, defined in Appendix G to the ASME Code [3]. The i
1 KIR curve is given by the following equation:
KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)]
(1) i where KIR - reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature l
RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code [3] as follows:
C*Kg+ KIT 5KIR (2) i where KIM - stress intensity factor caused by membrane (pressure) stress KIT - stress intensity factor caused by the thermal gradients KIR - function of temperature relative to the RTNDT f the material C
- 2.0 for Level A and Level B service limits C
- 1.5 for hydrostatic and. leak test conditions during which the reactor core is not critical t
I B-5 i-
i is determined by the At any time during the heatup or cooldown transient, KIR metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature g'radients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed.
From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the referen e fb w of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.
From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.
During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation.
It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4' T location for finite cooldown rates than for steady-state operation.
Furthermore, if conditions exist so that the increase in KIR exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various B-6
intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4 T crack during heatup is lower than the KIR for the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves fcr finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the B-7
q allowable pressure is taken.to be the lesser of the three values taken from the
~
curves under consideration. The use of the composite' curve is necessary to set conservative heatup limitations-because'it is possible for conditions to exist 1
wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the 1983 Amendment to 10CFR50I43 has a rule which ' addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RTNDT by at least 120*F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Wolf Creek).
Table B-1 indicates that the initial RTNDT of 20*F occurs in both the closure head flange and the vessel flange of Wolf Creek, so the minimum allowable temperature of this region is 140*F. These limits are shown in-Figures B-1 through B-6 whenever applicable.
4.
HEATUP AND C00LDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor pressure vessel have been calculated using the methods discussed in-Section 3.
Since pressure readings are measured at other locations than the limiting beltline region, the pressure differences between the pressure transmitter and the limiting beltline region must be accounted for when using the pressure-temperature limit curves herein. The indicated pressure and temperature labels provided on the curves relate to the limiting beltline region of the reactor vessel.
Figures B-1, B-3 and B-5 contain the heatup curves for 60'F/hr and 100'F/hr for 10, 12.3 and 16 EFPY, respectively.
Figures B-2, B-4 and B-6 contain the cooldown curves up to 100*F/hr applicable for the first 10, 12.3 and 16 EFPY of operation, respectively. No margins for possible instrumentation errors are included in the development of heatup and cooldown curves.
i B-8 u
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures B-1 through B-6.
This is in addition to other criteria which must be met before the reactor is made critical.
The leak limit curve shown in Figure B-1, B-3 and B-5 represents minimum temperature requirements at the leak test pressure specified by applicable codes [2,3). The leak test limit curve was determined by methods of References 2 and 4.
The criticality limit curves shown in Figures B-1, B-3, and B-5, specify pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 4.
The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum pressure-temperature curve for heatup and cooldown calculated as described in Section 3.
The maximum temperature for the inservice hydrostatic test for the Wolf Creek reactor vessel at 10,12.3 and 16 EFPY is 217'F, 219'F and 222'F, respectively. A vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures B-1 through B-6 define limits for ensuring prevention of nonductile failure for the Wolf Creek reactor vessel.
5.
CALCL'LATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2Ill the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
ART - Initial RTNDT + ARTNDT + Margin (3)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure B-9
for the material in Vessel Code.
If measured values of initial RTNDT question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
NDT - (CF]f(0.28-0.10 log f)
(4)
ART To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
f(depth X) " fsurface(e.24x)
(5) where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then put into equation (4) to calculate ARTNDT at the specific depth.
CF (*F) is the chemistry factor, obtained from Reference 1.
All materials in the beltline region of Wolf Creek were considered in determining the limiting material. The results of the RTNDT at 1/4T and 3/4T are summarized in Table B-2.
From Table B-2, it can be seen that the limiting material is the lower shell plate, R2508-3, for heatup and cooldown curves applicable up to 10, 12.3 and 16 EFPY. Sample calculations to determine the RTNDT values for 10 EFPY are shown in Table B-3.
B-10
L TABLE B-1 WOLF CREEK REACTOR VESSEL TOUGHNESS TABLE (Unirradiated)[5]
CU NI I-RTNDT (a)
Material Description
(%)
(%)
('F)
Closure Head Flange 20 (b)
Vessel Flange 20 (b)
Intermediate Shell, R2005-1 0.04 0.66
-20 Intermediate Shell, R2005-2 0.04 0.64
-20 Intermediate Shell, R2005-3 0.05 0.63
-20 Lower Shell, R2508-1 0.09 0.67 0
Lower Shell, R2508-2 0.06 0.64 10 Lower Shell, R2508-3 0.07 0.62 40 Intermediate and Lower Shell 0.04 0.04
-50 Longitudinal Welds Circumferential Weld 0.05 0.05
-50 (a.)
The initial RTNDT (I) values for the plates and welds are measured values.
(b.)
To be used for considering flange requirements for heatup/cooldown curvesI43 B-11
TABLE B-2
SUMMARY
OF ADJUSTED REFERENCE TEMPERATURES (ART's) AT 1/4T and 3/4T LOCATION (*F) 10 EFPY 12.3 EFPY 16 EFPY MATERIAL DESCRIPTION 1/4-T 3/4-T 1/4-T 3/4-T 1/4-T 3/4-T Intermediate Shell Plate, R2005-1 21 7
24 10 28 13 Intermediate Shell Plate, R2005-2 21 7
24 10 28 13 Intermediate Shell Plate, R2005-3 29 12 33 16 37 20 Lower Shell Plate, R2508-1 79 61 83 67 87 71 Lower Shell Plate, R2508-2 68 49 73 53 78 57 Lower Shell Plate, R2508-3 (84)* (75)* (86)* (77)* (89)* (79)*
Longitudinal Welds
-6
-21
-3
-18 1
-14 Circumferential Weld (10)
(-7)
(13)
(-3)
(16)
(2)
RTNDT numbers within ( ) are based on chemistry factors calculated using surveillance capsule data.
These RTNDT numbers were used to generate heatup and cooldown curves.
B-12
TABLE B-3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR THE LIMITING WOLF CREEK REACTOR VESSEL MATERIAL AT 10 EFPY - LOWER SHELL PLATE, R2508-3 l
Reculatory Guide 1.99 - Revision 2 10 EFPY Parameter 1/4 T 3/4 T Chemistry Factor, CF (*F) 44 (34) 44 (34) n/cm )(a)-
.459
.163 2
19 Fluence, f (10 Fluence Factor, ff
.783
.522 ARTNDT - CF x ff (*F) 34 (27) 23 (18)
Initial RTNDT, I (*F) 40 40 Margin, M (*F) (b) 34 (17) 23 (17)
Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 108 (84) 86 (75)
ART - Initial RTHDT + ARTNDT + Margin 19 2
(a) Fluence, f, is based upon fsurf (10 n/cm, E>l Mev) = 0.77 at 10 EFPY. The fluence was based on projections for rerated conditions. The Wolf Creek reactor vessel wall thickness is 8.625 inches at the beltline region.
2 0.5 The standard deviation (b) Margin is calculated as, M - 2 [ oj2,y 3
for the initial RTNDT margin term, og, is assumed to be 0*F since the initial RTNDT is a measured value. The standard deviation for ARTNDT term, I
aA, is 17'F for the plate, except that aA need not exceed 0.5 times the 8,
is 8.5*F for the plate (half the value) when mean value of ARTNDT-A surveillance data is used.
The numbers within ( ) are calculated using surveillance capsule data.
F B-13
MATERIAL PROPERTY BASIS LIMITING ART AT 10 EFPY:
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B-14
MATERIAL PROPERTY BASIS LIMITING ART AT 10 EFPY:
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Figure B-3 Wolf Creek Reactor Coolant System Heatup Limitations (Heat up rates up to 50*F/hr and 100*F/hr) Applicable for the First 12.3 EFPY (With No Margins For Instrumentation Errors)
B-16
MATERIAL PROPERTY BASIS LIMITING ART AT 16 EFPY:
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Figure B-5 Wolf Creek Reactor Coolant System Heatup Limitations (Heat up rates up to 60*F/hr and 100*Ffhr) Applicable for the First 16 EFPY (With No Margins For Instri. mentation Errors) 1 4
1 a
B-18
MATERIAL PROPERTY BASIS j
i LIMITING ART AT 16 EFPY:
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lNDICATED TEWPERATURE (DEG.F)
Figure B-6 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100*F/hr) Applicable for the First 16 EFPY (With No Margins For Instrumentation Errors)
B-19 i
l I
6.
REFERENCES 1
Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988.
2
" Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981.
3 ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Appendixes,
" Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure", pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986.
4 Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Washington, D.C.,
Federal Register, Vol. 48 No. 104, May 27, 1983.
5 WCAP-ll553, " Analysis of Capsule U from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program", S.
E. Yanichko, et al., August 1987.
t B-20
I I
ATTACHMENT 1 DATA POINTS FOR HEATUP AND C00LDOWN CURVES (With No Margins for Instrumentation Errors)
L i
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SAP 60 & 100 DEG-F/HR HEATUPS REG. GUIDE 1.99,REV.2 WITHOUT MARGIN 05/27/92 THE FOLLOWING DATA WERE CALCULATEDFOR THE INSERVICE HYDROSTATIC LEAK TEST.
MINIMUM INSERVICE LEAK TEST TEMPERATURE t
'E 000 EFPY) 10.0 PRESSURE (PSI)
TEMPERATURE (DEG.F) 2000 195 2485 217 l
i PRESSURE PRESSURE STRESS 1.5 K1M (PSI)
(PSI)
2000 21130 89059 2485 26254 111616 l
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SAP 60 & 100 DEG-F/HR HEATUPS REG. GUIDE 1.99.REv.2 w1THOUT MARGIN 05/27/92 COMPO5tTE CURVE PLOTTED FOR HEATUP PROFILE 3 HEATUP RATE (5) (OEG.F/HR) 100.0
=
1RRADIATION PERIOD =
12 000 EFP YEARS FLAW DEPTH * ( 1-AOWIN)T 10.0 INDICATED IND! RATED INDICATED INDICATED INDICATEO INDICATED TEMPERATURE PRE 35URE TEMPERATURE PRES 5URE TEMPERATURE PRESSURE (DEG.F)
(PSI)
(DEG.F)
(PSI)
(DEG.F)
(PSI) 1 85.000 G4+-96 15 155.000 660.05 29 225.000 1160.90 2
90.000 6e?-65 16 160.000 678.94 30 230.000 1221.71 3
95.000 "J99-SS,@l.N 17 165.000 699.82 31 235,000 1287.50 4
100.000 589-99 18 170.000 723.09 32 240.000 1357.89 5
105.000 Set-99 19 175.000 748.50 33 245.000 1433.44 6
110.000 581.19 20 180.000 776.51 34 250.000 1514.28 7
115.000 581.83 21 185.000 806.91 35 255.000 1601.18 8
120.000 584.56 22 190.000 839.94 36 2G0.000 1694.08 9
125.000 589.43 23 195.000 876.02 37 265.000 1793.30 10 130.000 596.14 24 200.000 914.91 38 270.000 1899.81 11 135.000 605.00 25 205.000 956.99 39 275.000 2013.45 12 140.000 615.79 26 210.000 1002.38 40 280.000 2134.88 13 145.000 628.60 27 215.000 1051.37 41 285.000 2264.20 14 150.000 643.22 28 220.000 1104.13 42 290.000 2402.16 03 i
N 10 5
3!
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SAP COOLDOWN CURVES REG. GUIDE 1.99.REV.2 WITHOUT MARGIN 05/19/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 1 1 STEADV-STATE COOLDOWN )
IRRADIATION PERIOD
- M-400 EFP YEARS FLAW DEPTH = ADWIN T 12.3 INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERAIURE PRESSURE (DEG.F)
(PSI)
(DEG.F)
(PSI)
(DEG F )
(PSI) i 1
85.000 669-991 12 140.000 4 %,44 CII 23 195.000 1470.60 2
90.000 Gee-64 13 145 000 951.27 24 200.000 1545.70 3
95.000 69?-95 14 150.000 988.11 25 205.000 1626.15 4
100.000 44b40 15 155.000
- 1027.89 26 210.000 1792.28 5
105.000 N
16 160.000 1070.49 27 215.000 1804.71
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110.000 TS*H!M!b 7
115.000
?ff-79 18 170.000 t165.37 29 225.000 2009.36 8
120.000 40444 19 175.000 1218.08 30 230.000 2122 56 9
125.000 et?-94 20 180.000 1274.58 31 235.000 2243.67 10 130.000 959--+9 21 185.000 1335.46 32 240.000 2372.76 11 135.000 964-91 22 190.000 1400.66 y
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( 20 DEG-F / HR COOLDOWN )
IRRADIATION PERIOD W EFP YEARS
=
FLAW DEPTH = ADWIN T l2.3 INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG. F )
(PSI)
(DEG.F)
(PSI)
(DEG. F )
115.000 M'i 13 145,000 936.97 2
90.000 649-69 8
120.000 fte-?O 4 2 I + 14 150.000 976.71 3
95.000 665-59 ' (3l i 9 125.000 804 M 15 155.000 1019.30 4
100.000 4G+-94 10 130.000 SW 16 160.000 1065.03 5
105.000 N
j 11 135.000 N l 17 165.000 1114.22 2
6 110.000 M'
12 140.000 000.;C k E\\
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( 40 DEG-F / HR COOLDOWN I l
[
IRRADIATION PERIOD
- t9-099 EFP VEARS FLAW DEPTH = ADWIN 1 12 3 l
l INDICATED INDICAIED INDICATED INDICATED INDICATED INDICATED l
TEMPERATURE PRESSUDE TEMPERATURE PRESSURE T E MPE R A T URE PRESSURE (DEG.F )
(PSI)
(DEG.F)
(PSI)
(DEG.F)
(PSI) 1 85.000 597.38 7
115.000 346-4M '
13 145.000 925.03 2
90.000 615.02 8
120.000 95t-99 14 150.000 967.61 l
3 95.000 N
9 125.000 343-64
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100.000 664-43 Qlg 10 130.000 0+*-99 16 160.000 1062.66 l
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105.000 636-46 11 135 000 Ste-St t 17 165.000 1115.59 6
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12 140.000 ee9-t?)
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SAP COOLDOWN CURVES REG. GUIDE 1.99,REV.2 WITHOUT MARGIN 05/19/92 THE FOLLOWING DATA WERE PLDifED FOR COOLDOWN PROFILE 4
( 60 DEG-F / HR COOLDOWN )
IRRADIATION PERIOD 15 090 EFP VEARS
=
FLAW DEPTH = ADWIN T 82.3 INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRES 5URE (DEG.F)
(PSI)
(DEG. F )
(PSI)-
(DEG.F)
(PSI) 7fM H HD 12 140.000 434,46 h2I h 790-48]
1 85.000 563.71 7
115.000 2
90.000 582.54 8
120.000 13 145.000 915.41 3
95.000 602.75 9
125.000 7Gib-40 (all 4 14 150.000 961.28 4
100.000 fre4-68p 10 130.000 994HND 15 155.000 1080.69 43.1 4 11 135.000 N
16 160.000 1063.76 5
105.000 6*e.-M r
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SAP COOLDOWN CURVES REG. GU10E 1.99,REV.2 WITHOUT MARGIN 05/19/92 THE FOLLOWING DATA WERE PLDITED FOR COOLDOWN PROFILE 2
( 20 DEG-F / HR COOLDOWN I IRRADIATION PERIDO = N EFP YEARS FLAW DEPTH a ADWIN T S,o INDICATED INDICATED INDICATED INolCATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG. F )
(PSI)
(DEG. F )
(PSI)
(DEG.F )
(PSI) 1 85.000 499-48' 7
115.000 799-ft ' j 13 145.000 916.99 2
90.000 939-69 8
120.000 M
14 150.000 955.06 3
95.000 BMi 9
125.000 tes-t4 N
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100.000 67+-+e ' NI4 10 130.000 SM 16 160.000 1039.97 5
105.000 499-99 11 135.000 e49-44<
17 165.000 1087.29 6
110.000 ttS I 12 140.000 e& W 18 170.000 1138.14 k Gl y e regtdre m sk to e
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SAP COOLDOWN CURVES REG. GUIDE t.99.REV.2 WITHOUT MARGIN 05/19/92 THE FOLLOWING OATA WERE PLOTTED FOR COOLDOWN PROFILE 3
( 40 DEG-F / HR COOLDOWN )
IRRADIATION PERIOD =
+o-see EFP VEARS FLAW DEPTH = ADWIN T
!&,o INDICATED INDICATED INDICATED INDICATED I ND IC A T E D INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERAIURE PRESSURE (DEG.F )
IPSI)
(DEG.F)
(PSI)
(DEG.F )
(PSI) 1 85.000 588.40 7
115.000 5'T 13 145.000 903.31 N [' @l I 2
90.000 605.22 8
120.000 N
14 150.000 944.22 3
95.000 699-5+'
9 125.000 15 155,000 988.26 4
100.000 649-04 10 130.000 Tse-es' 16 160.000 1035.62 5
105.000 6eW
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110.000 k Ft4. a p: o c^Y b
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1RRA01ATION PERIOD = N ETP YEARS FLAW DEPTH = AOWIN T 4.0 INDICATED INDICATED INDICATED IHOICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F)
(PSI)
(DEG.F)
(PSI)
(DE G. F )
(PSI) 1 85.000 553.92 7
115.000 666-66')
13 145.000 891.92 2
90.000 572.01 8
120.000 M +-+9{
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95.000 591.54 9
125.000 W
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SAP COOLDOWN CURVES REG. GUIDE 1.99.REV.2 WITHOUT MARGIN 05/19/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 5
( 100 DEG-F/HR COOLDOWN )
IRRADIATION PERIOD =
44-049 EFP YEARS FLAW DEPTH = ADWIN T g g,o i
INDICATED INDICATED INDICATED INDICATED INDICATED IN0!CATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F)
(PSI)
(DEG. F )
(PSI)
(DEG.F )
(PSI) 1 85.000 485.49 7
115.000 G8FP-46 12 140.000 "20.02 (o51 W 2
90.000 506.17 8
120.000 N
13 145.000 877.82 3
95.000 528.66 9
125.000 49G-OO ' (p3.I t 14 150.000 929.37 4
100.000 552.81 10 130.000 944-t+
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105.000 579.04 11 135.000 h
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110.000 607.20
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TEMPERATURE (DEG.F )
2000 200 2485 222 PRESSURE PRESSURE STRESS 1.5 K1M (PSI)
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l APPENDIX C Calibration Reports f
C-0
=
WESTINGHOUSE SCIENCE & TECHNOLOGY CENTER CALIBRATION RECORD DATE: 11/16/91 INSTRUMENT BEING CALIBRATED:Tinius-Olsen Mod. 74, Ser. # 123159 Impact machine LOCATION: Bldg. 302A low level cell OPERATOR: L. M. Thomas CALIBRATION METHOD: NIST standards per ASTM LM-91 RESULTS:
Series SRM No. Average Value Expected Range Acceptable Not Acceptable ft-lb ft-lb LL23 2092 12.4 11-15 X
HH24 2096 71.8 66-78 X
CALIBRATION ACCEPTABLE X
NOT ACCEPTABLE COMMENTS: Machine dial values only.
APPROVAL:
Jb
'T ti 12 L f 91 R. P. Shogan Date Manager, Nuclear Services & Materials Testing I
C-1
TReport & Gertificate of Berification C&M COLLINS CAllBRATION SERVICE THIS IS TO CERTIFY that the following described testing machine has been calibrated by C&M Call-BRATION and the loading range (s) shown below found to be within a maximum tolerance of l
MACHINE ZN SII[Ub/
7 V8/
TYPE d ced b// -Iceb l
LOCATION W P5 N ne.lwa.>r fltelN r levs.
CAPACITY 2 0 CCC lbr A f D (en 4 rn C h v.-e l. / /
)#A SERIAL NO.
IdA DATE OF VERIFICATION Iune V' M9/
Method of verification and below recorded data is in accordance with A.S.T.M. E-4 F9
& Mll. STD. 45662-A. The testing device (s) used for this calibration have certifications traceable to the NationalInstitute of Standards Technology.
Machine Readings 13 Pounds O Newtons D Kilograms Readings temperature corrected for 12
- C.
MACHINE C AUSRAISON DEVICE MACMINE ERROR C.C MACMANE CAuBRATION DiveCE M ACMdNE ERROR CD L
READING READING fg No R EADINO READING
/Q No i
l'O Co. ? ?
-t7 w
I 200 2 0 I si
/.1/
.SL I
/00
/ ocs 3 9
.39
.19 i
'tO O v o ?. Js
-p 13
.sg 2
200 Pol ?]
l.23
.GE I
SCO 80+ G ?
. ale
.5E
?
]OO 30139
- 1. 3 Cr
. +5 L
./200 1707?
- 2. ?
.bo
?
'f00 y o 3. i t
~3 I?
7%
I lbOO
/4 0/. 3
~l. I
.oE E
500 507 50
. T. 50
.so E
2000
?006 5-
~ 4. S
.31 3
t
/OO Joo,73
- ES
.El f
S*o O 5 0? 31
~ r. 3 %
.4C 2
200 20 L 66
- [ bi
.81 I
/000 1007.?
~7E 7?
400
<+c t 7 3
- 7. ~15
.br E
2000 7015.1
-13.8
.bt 3
b00 (03 r?
-3??
53 E
3000 J o ? ?. 7
- r ?. ?
.75 3
9'00 80 5.31
- S.1/
.66 I
'f 0 oD Uo tt. /
~ ?S' /
.70
)
l000 to o 7. L
- r. G
.rG G~D00 5004.5
-45
,o 9 CAllBRATION APPARATUS - Morehouse Proving Rings. Stainsense load cells & Troomner Dead Weights. Verifications traceable to the Nationalinstitute of Standards Technolooy, in accordance with A.S.T.M. latest speelfications.
C.D.
SERI AL NO.
LOADING RANGl C ALIB.
NATIONAL INSTITUTE OF STANDARDS TECHNOLOGY LAB NO.
TEMP.
CODE CLASS AVALUE DATE O F C.D.
1
't'7 51 1 & - 3 oo alt 5-ts-so
( C/tst007
%3 *C 3
4957 777-Pfr llt s'-ES 90 7 If ? ?? 75 y g2 *C 3
47Fy
/7tfr. Po /r 2-?F-9/
7 7 7 tr9 7s9 / 712 07 / /s.5 r97 PJ *C 4
- C
- C W ACHINE RANGE "'
LCADING RANGE Primary Leadindicating Devies i
It '
t v el c4.,4 0
roo ro - soo C & M COLLINS CALlBRATION SERVICE J
O f ono fn o. fog n.
230 Haymont Dr., Box 149, Gibsonia, PA 15044 (412)443 7631 0
roon eco - reno gy O,
7o00 FOO - 2000
' ' (SERVICE ENGINEER) i Witnessed By:
i C-2 r
e
-~
-a.w e-
? Report & Gertificate of Berification C&M COLLINS CALIBRATION SERVICE THIS IS TO CERTIFY that the following described testing machine has been calibrated by C&M CALI.
BRATION and the loading range (s) shown below found to be within a maximum tolerance of
/.
IA/ 5780 N 3 Vi/
TYPE bd C' // - Tr.m/o.-
MACHINE LX/e3 Nny 6 w e 8 /ec k e (c" A CAPACITY P o ocice / 6e LOCATION 89 d fro J u
( Le 4.'/ /
0/A SERIAL NO.
/88 DATE OF VERIFICATION 1 eor 4 /99 /
Method of verification and below recorded data is in accordance with A.S.T.M. E-4 f9
& Mll. STD. 45662 A. The testing device (s) used for this calldration have cert!!! cations traceable to the NationalInstitute of Standards Technology.
Machine Readings G Pounds O Newtons O Kilograms Readings temperature corrected for PT
'C.
M A CHINE CAUSRATjON DIVlCL M ACMINE ERROR C.0 MACHINE CAllBRAfloN DEVICE M ACHINE ERROR C.D RE A Dateo READINo fg _
No R EADINo READINo No i
_ /DOO
/003 6
-36 36 2000 70/4,3
-N 3 77 2
Le o o O Lect 1 t
>7 ?
M3 2
(CCO f s3 h b
~%6 si E
s
@ O00 Bogs o
. L c. o
.F/
e
/000o
/oo3F
- 38 3r 2
2000 2 c os.s
- S. s
.?2
?
9000 3 9 0s. fr 4 v. 2
.</
?
- 000 7999.o 4/0
.o /
r 17000
>>nt/
-si
.0 'f E
/ /> 000
/60?3
~23
.sv
?
i 20000 20 c off
- I E'
.09 CALIBRATION APPARATUS - Morehouse Proving Rings. Sta5 sense load cells & Troomner Dead Weights. Verifications 7
traceable to the National Institute of Standards Technolooy. In accordance with A.S.T.M. latest speelfications.
C.D.
sE RI AL No.
LOADING RANGE CALIB.
NAT10NAI,1NSTITUTE OF STANDARDS TECHNOLOGY LAB NO.
TEMP.
CODE CLASS AVALUE DATE OF C.D.
1 4i57 2 77- ?k 14 C47-90 927
??9139
?)
- C 2
a,7 3.y p qt.fo y 7 7 3,,,,
7 7 7_ r 2 9 7 f 9 7 7 2. 0 7/ ?v 3-f 4r rj
'C 3
C 4
.C 5
- C W ACHINE RANGE LDADING RANCE Prunary Loadindicating Device
[
j '
Lrn decado d -
/o oco /&i
/e00 - /oooo C & M COLLINS CAllBRATION SERVICE d-70 mo /61 700o - 30 voo 230 Haymont Dr., Box 149, Gibsonia, PA 15044 (412)4434 631 By:
h fCA
' ' (SERVICE ENGINEER)
Witnessed Br*
i C-3 n
I