ML20085C144
| ML20085C144 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/30/1991 |
| From: | Shaun Anderson, Madeyski A, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20085C142 | List: |
| References | |
| WCAP-12946, NUDOCS 9108300026 | |
| Download: ML20085C144 (177) | |
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BESTINGMOUSE CLASS 3 WCAP-12946 ANALYSIS OF CAPSULE Y FROM THE UNION ELECTRIC COMPANY CALLAWAY UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM E. Terek S. L. Anderson A. Madeyski June 1991 Work Performed Under Shop Order UMSP-106 Prepared by Westinghouse Electric Corporation for the Union Electric Company Approved by: (N T. A. Meyer, 'R~anape_ r/
Structural Reliabmty and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 e 1991 Westingbouse Elec.tric Corp.
PREFACE This report has been technically reviewed and verified.
Reviewer:
Sections 1 through 5, 7 and 8 J. M. Chicots /8 /fr d b Section 6 E. P. Lippincott Ih si O
9 9
e i
TABLE OF CONTENTS Section litle EA5Le 1.0
SUMMARY
OF RESULTS 1-1
2.0 INTRODUCTION
2-1 3,0 BACKGROUND 3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE 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 - PHOTOGRAPHS OF SPECIMENS BEFORE TESTING ii
L LIST OF TABLES i
lalle lille P.agn 4-1 Chemical Composition of the Callaway Unit 1 4-3 Reactor Vessel Surveillance Material 4-2 Heat Treatment History of the Callaway Unit No. 4-4 1 Reactor Vessel Surveillance Material 5-1 Charpy V-Notch Impact Data for the Callaway Unit 1 5-8 Lower Shell Plate R2708-1 Irradiated at 550'F, Fluence 1.29 x 1019 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Callaway Unit 1 5-9
. Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 1.29 x 1019 n/cm2 (E > 1.0 MeV) 5-3 Instrumented Charpy impact Test Results for the Callaway 5-10 Unit 1 Lower Shell Plate R2708-1 Irradiated at 550'F, Fluence 1.29 x 1019 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy impact Test Results for the Callaway 5-11 Unit 1 Weld Metal and Heat-Affected-Zone (HAZ) Metal, Irradiated at 550*F, Fluence 1.29 x 1019 n/cm 2 (E > 1.0 MeV) 5-5 2
~
Effect of 550*F Irradiation to 1.29 x 1019 n/cm 5-12 (E > 1.0 MeV) on the Notch Toughness Properties of the Callaway Unit 1 Reactor Vessel Surveillance Materials iii
LIST OF TABLES Iablg- Title Eagg 5-6 Comparison of the Callaway Unit 1 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 Callaway Unit 1 Reactor Vessel 5-14 Surveillance materials Irradiated at 550*F to 1.29 x 1019 n/cm2 (E > 1.0 MeV) 1 6-1 Calculated Fast Neutron Exposure Parameters at the 6-14 Surveillance Capsule Center 6-2 Calculated Fast Neutron Exposure Rates at the 6-15 Pressure Vessel Clad / Base Metal Interface .
6-3 Relative Radial Distributions of Neutron Flux 6-16 (E > 1.0 MeV) within the Pressure Vessel Wall
)
6-4 Relative Radial Distributions of Neutron Flux 6-17
- (E > 0.1 MeV) within the Pressure Vessel Wall 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 Four 6-20 Fuel Cycles of the Callaway Unit 1 Reactor iv
LIST OF TABLES lablq Tit 1e hat 6-8 Measured Sensor Activities and Reaction Rates 6-21 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 Coniparison 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 Upd .ed Lead Factors for Callaway Unit 1 Surveillance 6-29 Capsules t- ,
l l
LIST OF ILLUSTRATIONS l Fiaure Utle Eggt 4-1 Arrangement of Surveillance Capsules in the Callaway 4-5 Unit 1 Reactor Vessel 4-2 Capsule Y Diagram Showing Location of Specimens, Thermal 4-6 Monitors and Dosimeters 5-1 Charpy V-Notch Impact Properties for Callaway Unit 1 5-15 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) 5-2 Charpy V-Notch Impact Properties for Callavay Unit 1 5-16
. Reactor Vessel Lower She?1 Plate R2708-1 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for Callaway Unit 1 5-17 Reactor Vessel Core Region Weld Metal 5-4 Charpy V-Notch Impact Properties for Callaway Unit 1 5-18 Reactor Vessel Weld Heat Affected Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Callaway 5-19 Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) 5-6 Charpy. Impact Specimen Fracture Surfaces for Callaway 5-20 Unit 1 Reactor Vessel Lower Shell Plate R2708-1 .
(Transverse Orientation) vi
% -- r- - - - - - _------ - - - - - _ --_._--,-- -- -, -- - - - - - - - - - - - - - _ _ _ - - -
LIST OF ILLUSTRATIONS (Cont)
Fiaure lith hgg 5-7 Charpy Impact Specimen Fracture Surfaces for Callaway 5-21 Unit 1 Reactor Vessel Weld Metal 8 Charpy Impact Specimen Fracture Surfaces for Callaway 5-22 Unit 1 Reactor Vessel Weld Heat Affected Zone Metal 5-9 Tensile Properties for Callaway Unit 1 Reactor Vessel 5-23 Lower Shell Plate R2708-1 (Longitudinal Orientation) 5-10 Tensile Propertie: for Callaway Unit 1 Reactor Vessel 5-24 Lower Shell Plate 2708-1 (Transverse Orientation) 5-11 Tensile Properties for Callaway Unit 1 Reactor Vessal 5-25 Core Region Weld Metal .
5-12 Fractured Tensile Specimens from Callaway Unit 1 5-26 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)
L 5-13 Fractured Tensile Specimens from Callaway Unit 1 5-27 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens from Callaway Unit 1 5-28
~'
L Reactor Vessel Core Region Weld Metal
( -
15 Engineering Stress-Strain Curves for Plate R2708-1 5-29 l
Tensile Specimens CL12 and CLl4 (Longitudinal Orientation) vii
LIST OF ILLUSTRATIONS (Cont)
Fiaure lillt E3SS 5-16 Engineering Stress-Strain Curve for Plate R2708-1 5-30 Tensile Specimen CLIS (longitudinal Orientation) 5-17 Engineering Stress-Strain Curves for Plate R2708-1 5-31 Tensile Specimens CT13 and CT14 (Transverse Orientation) 5-18 Engineering Stress-Strain Curve for Plate R2708-1 5-32 Tensile Specimen CT15 (Transverse Orientation) 5-19 Engineering Stress-Strain Curves for Weld Metal 5-33
- Tensile Specimens CW13 and CW14 5-20 Engineering Stress-Strain Curve for Weld Metal 5-34 Tensile Specimen CW15 5-21 True Stress-Strain Curves for Plate R2708-1 Tensite 5-35 Specimens CL13 and CLl4 (longitudinal Orient. tion) 5-22 True Stress-Strain Curve for Plate R2708-1 Tensile 5-36 Specimen CL15 (Longitudinal Orientation) 5-23 True Stress-Strain Curves for Plate R2708-1 Tensile 5-37 Specimens CT13 and CT14 (Transverse Orientation) 5-24 True Stress-Strain Curve for Plate R2708-1 Tensile 5-38 Specimen CT15 (Transverse Orientation) viii
i LIST OF ILLUSTRATIONS (Cont) ;
l Fiaure Title hgg l 5-25 True Stress-Strain Curves for Weld Metal Tensile 5-39 Specimens CW13 and CW14 1
5-26 True Stress-Strain Curve for Weld Metal Tensile 5-40 l Specimen CW15 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 l
1 I
1 i
I e
ix
?
I SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule .!
Y, the second capsule to be removed from the Union Electric Company Callaway f Unit I reactor pressure vessel, led to the following conclusions: !
i i
o The capsule received an average fast neutron fluence (E > 1.0 MeV) of j 1.29 x 1019 n/cm2 after 4.60 EFPY of plant operation. l o Irradiation of the reactor vessel lower shell plate R2708-1 Charpy specimens, oriented with the longitudinal axis of the specimen .
parallel to the major rolling direction (longitudinal orientation), j to 1.29 x 10I9 n/cm2 (E > 1.0'HeV) resulted in a 30 ft-lb f transition temperature increase of 15'F and a 50 ft-lb transiti_on j temperature increase of 30'F. This results in a 30 ft-lb f
--transition temperature of 10*F and a 50 ft-lb transition !
t temperature of 55'F for longitudinally oriented specimens. j o Irradiation of the reactor vessel lower shell plate R2708-1 Charpy !
specimens, oriented with the_ longitudinal axis of the specimen normal j to the major rolling direction (transverse orientation), to 1.29 x )
1019 n/cm2 (E'> 1.0 MeV) resulted'in a 30 and 50 ft-lb transition temperature increase of 45'F. This results in a 30 ft-lb j transition temperature of 25'F and a 50 ft-lb transition f temperature of 75'F for transversely oriented specimens. !
19 2 o The weld metal Charpy specimens irradiated to 1,29 x 10 n/cm l (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 j 35'F. This results in a 30 ft-lb-transition temperature of i
-15'F and a 50 ft-lb transition temperature of 15*F for the !;
l weld metal. !
L i 1-1 t i
l
\
o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal ;
Charpy specimens to 1.29 x 10I9 n/cm2 (E > 1.0 MeV) resulted in a l 30 and 50 ft-lb transition temperature increase of 40*F. This
-[
results in a 30 f t-lb transition temperature of -50*F and a 50 !
ft-lb transition. temperature of -5'F for the weld HAZ metal.
- t o The average upper shelf energy of lower shell plate R2708-1
[
(longitudinal orientation) resulted in an energy decrease of 8 ft-lb I after irradiation to 1.29 x 1019 n/cm2 (E > 1.0 MeV). This [
results in an upper shelf energy of 118 ft-lb for longitudinally j oriented specimens.
k o The average upper shelf energy of lower shell plate R2708-1 [
(transverse orientation) resulted in a energy decrease of 13 ft-lb !
after irradiation to 1.29 x 1019 n/cm2 (E > 1.0 MeV). This f
results in an upper shelf energy of 91 ft-lb (transverse orientation); I o The average upper shelf energy of the weld metal decreased 15 ft-lb :
after irradiation to 1.29 x 1019 n/cm2 (E > 1.0 MeV). This ,f results in an upper shelf energy of 97 ft-lb for the weld metal. .
o The average upper shelf energy of the weld HAZ metal decreased 15 -j ft-lb after irradiation to 1.29 x 1019 n/cm2 (E > 1.0 MeV). This !
results in an upper shelf energy of 91 ft-lb for the weld HAZ metal, i t
o The surveillance capsule Y test results indicate that the !
surveillance material 30 ft-lb transition temperature changes and i upper shelf energy decreases are in close agreement with the
- Regulatory Guide 1.99 Revision 2 predictions. (
~!
I 1-2 i
o The surveillcnce capsule materials exhibit a more than adequate upper l
! 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.
o The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the Callaway Unit I reactor vessel is as follows:
Vessel inner radius * - 2.39 x 10 19 n/cm 2
19 2 Vessel 1/4 thickness - 1.31 x 10 n/cm 18 2 Vessel 3/4 thickness - 2.82 x 10 n/cm
- Clad / base metal interface W
G S
1-3 ;
SECTION
2.0 INTRODUCTION
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 neutrnn irradiation on the Union Electric Company Callaway Unit I reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Union Electric Company Callaway Unit I reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-9842 entitled " Union Electric Company Callaway Unit No.1 Reactor Vessel Radiation Surveillance Program" by L. R. Singer II}. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule "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 su' aillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance capsule "Y" removed from the Union Electric Company Callaway Unit I reactor vessel and discusses the analysis of these data.
2-1
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its prirl.ary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as A533 Grade B Class 1 (base material of the Union Electric Company Callaway Unit I reactor pressure vessel lower shell plate R2708-1) are well documented in the literature, Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation, A method for performing analyses to guard against fast fracture in reactor
- pressure vessels has been presented in " Protection Against Nonductile failure,"
Appendix G to Section 111 of the ASME Boiler and Pressure Vessel Code [5],
- The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)-
RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) 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 of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure
~
vessel steel, When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity f actors.
3-1
i ?
t
}
RTNDT and, in turn, the operating limits of nuclear power plants can be l adjusted to account for the effects of radiation on the reactor vessel material ,.
. t properties. The radiation embrittlement changes in mechanical properties of a !
given reactor pressure vessel steel can be monitored by a reactor surveillance ,
f program such as the Callaway Unit 1 Reactor Vessel Radiation Surveillance l Program,Ill in which a surveillance capsule is periodically removed from the l operating nuclear reactor and the encapsulated specimens are tested. The !
- increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due l to-irradiation is added to the original RTNDT to adjust the RTNDT for ;
radiation embrittlement. This adjusted RTNDT (RTNDT initial + {
i ARTNDT) is used to index the material to the KIR curve and, in turn, to i set operating limits for the nuclear pcwer plant which take into account the [
effects of irradiation on the reactor vessel materials. -I t
, i i
l
. t t
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l t
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?
SECTION
4.0 DESCRIPTION
OF PROGRAM
~'
Six surveillance capsules-for monitoring the effects of neutron exposure on the Callaway Unit I 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 pads and the vessel wall j as shown in Figure 4-1.- The vertical center of the capsules is opposite the-vertical center of the core, jl Capsule Y was removed after 4.60 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 R2708-1 and !
submerged arc weld metal representative of the beltline weld seams of the !
I rear tor vessel and Charpy V-notch specimens from weld HAZ material. l l
All test specimens were machined from the 1/4 thickness location of the plate. !
Plate test specimens represent material taken at least one plate thickness from l
, the quenched end-of the plate after performing a simulated post weld l stress-relieving treatment on the test plate. Weld and weld HAZ test specimens !
- were taken from weld and weld HAZ metal _ of a stress-relieved weldment joining j lower shell plate R2708-1 and adjoining intermediate shell plate R2707-1. All {
HAZ specimens were obtained from within the weld HAZ of lower shell plate l
{
R2708-1. ;
f Base metal Charpy V-notch impact and tension specimens were oriented with the [
longitudinal axis of the specimen parallel to the major rolling direction of l I
the-plate (longitudinal orientation) and also normal to the major rolling j direction (transverse orientation). Charpy V-notch and tensile specimens from j the weld metal were oriented such that the long dimension of the specimen was [
normal to the welding direction, j i
i l
t 4-1 i h
t
p v
[ Capsule Y, also, contained 1/2T CT test-specimens from the lower shell plate
{ , R"08-1 and were machined in both the longitudinal and transverse
! orientations, Thus, the simulated crack in the specimen will propagate normal
! and parallel to the major rolling direction of plate R2708-1.
The 1/2T CT Test ,
j specimens from +.he weld metal were machined with the notch oriented in the
! direction of welding. Thus, the simulated crack,in the specimen will propagate
[ parallel to the weld direction. All CT specimens were fatigue precracked
} according to ASTM E399.
i The chemical composition and heat treatment of the surveillance material is
. presented in Tables 4-1 and 4-2. The chemical analyses reported in Table 4-1
} were obtained from unirradiated material used in the ,urveillance programill l and a chemical analysis performed on an irradiated Charpy specimen from the
.[ weld metal and base metal plate R2708-1 as reported in the capsule U l analysis (2),
i Capsule Y contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded). In -
addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were placed in the capsule to measure the integrated flux at specific -
neutron energy-levels.
Thermal monitors- made from the two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. The composition of the two alloys and their melting points are as follows:
2.5% Ag, 97.5% Pb Melting Point: 579'F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point: 590*F (310'C)
The arrangement of tne various-mechanical specimens, dosimeters and thermal monitors-contained in capsule Y are shown in Figure 4-2.
4-2
s j TABLE 4-1 CHEMICAL COMPOSITION OF THE CALLAWAY UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIAL Chemical Comoosition (wt%)
Element Lower Shell Plate R2708-1 Weld Metal C 0.22 3 0.23 3 0.15 3 0.11 3
! Mn 1.47l 1.32l 1.37 l 1.28l P 0.006l <0.005l 0.005l 0.005l S 0.014l 0.016l 0.008l 0.016l Si 0.25l 0.22l 0.44 l 0.47l Ni 0.59l 0,55l 0.07 l 0.06l Mo 0.57l 0.49l 0.54 l 0.50l Cr 0.05l(a) 0.06l(b) 0.04 l(a) 0.04l(b)
Cu 0.07l 0.06l 0.06 l 0.03l A1 0.025l l 0.003l --
l Co 0.013l 0.005] 0.011l 0.004l-Pb (c)l --
l <0.001l --
l
. W <0.01l --
l <0.01 l l
Ti <0.01l <0.001l <0.01 l <0.001l
- Zr <0.001l l <0.001l --
l V 0.003l 0.01 ) 0.004l <0.01l Sn 0.002l 0.003l --
l As 0,001l <0.001l --
l Cb <0.01l <0.01 l l
N 0.008l 0.007l --
l
-- )
Sb --
l 0.0015l B <0.001) 0.001 J
- a. Data from WCAP-9842 Ill
^
- b. Data from WCAP-ll374, Revision 1 (2)
- c. Not detected 4-3
TABLE 4-2 HEAT TREATMENT HISTORY OF THE CALLAWAY VNIT NO. 1 REACTOR VESSEL SURVEILLANCE MATERIAL Ill
- i i s
Heat Treatment History i
P Material Temperature ('F) Time (Hr) Coolant .
Lower Shell(a) Austenitizing 1600225 4 Water quenched l (Plate R2708-1) Tempered 1225 25 4 Air cooled Stre'is Relief 1150 50 13 Furnace cooled t
Weld Metal (b) Stress Relief Il50 ISO 7.75 Furnace cooled i
.~
I
- a. Plates were produced by Lukens Steel Company .
f
- b. Weldment was fabricated by Combustion Engineering, Inc. using 3/16 inch
[
Mil B-4 weld filler wire, heat number 90077 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.
- l T-4-4 t
I REACTOR VESSEL O'
CORE BARREL 530* [ NEUTRON PAD CAPSULE U l l V J s ,s. so,s. L
_Y l _
270* - -
-- 90
- 3- 2.s-
\
z SO' 3 79*
Y
/I I\
w X i W
210' 180' PLAN VIEW Figure 4-1. Arrangement of Surveillance Capsules in the Callaway Unit 1 Reactor Vessel 4-5
,y,, , w*
LAAEE cuanrys JAQ Tt ww 1 C?idr etti Cesaratti cuaarts (manegg (twacis towpacit Cmane CW11 CW71 CW 12 CH77 (W69 QEl (W66 CW14 Cw70 Cw19 CW18 CW11 CW7* CH74 CW71 CH71 Cwt 4 048 CL20 Clit Q 18 Q17 CW65 CW13 CW F3 003 CWF0 CH79 Cw67 Def (W64 Ce ; ; , f, l,,l l '
AI . i% Ce Ce r.
ggg r.
MON! TOR 0OO *> . m c. ten ,,
Mar 1 TOR ll11,;=,, si e
CENT TO TOP OF VESSEL SPECIMEN NUMBERING CODE:
CT - PLATE R2708- 1 (TRANSVERSE ORIENTATION)
CL - PLATE R2708- 1 (LONGITUDINAL ORIENTATION)
CW - CORE REGION WELD METAL CH - HEAT- AFFECTED-ZONE METAL i
traRPfl gasrrfat ity sa tt chacPT S twaarvt _. thaary g guaarts Emeracts Yt ata tt JiaRPt s touracti cwo osa ais cus an a s, a r,.
ent a r2 cTes cui cin lasi cTin cWL2 QE2 442 a14 0774 afe CITI ait Citt a 64 Cfe$ G 6) CTV a 62 dig cill CII4 Olli CI14 LW41 Ott C113 0173 3 73 c770 a F0 Citi utf CY64 a64 Cfit 2 61 CY13
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- c. - J jill h -.,..c.
Uuu '-
ouu nnn
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4%- n on FS - *i '=c ic)
.I 11, e - m .1 11 s- y: ~i REGION OF VESSEL TO BOTTOM OF VESSEL Si
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APERTURE CARD Also Available (b.
Apert ur.- 1:
Figure 4-2 Capsule Y Diagram Showint location of Specimens, Thermal Monitors and Dosimeters
%OTh000z g . m 4_e
~~
SEC110N 5.0 TESTING OF SPEClMENS 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 Power Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and HI33, ASTM Specification E185-82I71, and Westinghouse Prc:edure MHL 8402, Revision 1 as modH ed by RMF Proredures 8102, Revision 1 and 8103, Revision i.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9842(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'T (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 an Effects Technology Model 500 instrumentation systein. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED ). From the load-time curve (Appendix A), the load of general vielding (Pcy), the time to general yleiding (tcy), the maximum load (Pg), and the time to maximum load (tg) can be determined. Under some test
- .pditions, 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 (Pg).
5-1
i i
t The energy at maximum load (Eg) was determined by ccmparing the energy-time record and the load-time record. The energy at maximum load is roughly f
equivalent to the energy required to initiate a crack in the specimen. 'l Therefore, the propagation energy for the crack (E p ) is the difference ,l between the total energy to fracture (E )Dand the energy at maximum load. l The yield stress (ay) was calculated from the three-point bond formula !
having the following expression: 1 oy Pay * (L/(B*(W-a)2*C)) (1) l where L = distance betwee,4 tha specimen supports in tie impact testing machine; f
B - the width of the specimen measured parallel to the notch W height of the j specimen, measured perpendicularly to the notch; a - notch depth.
f The constant C is dependent on the notch flank angle (p), notch ront radius f (p), and the type of loading (i.e., pure bending or three-point bending). f in three-point bending a Charpy specimen in wnich 4 45' and p - .i 0.010", Equation 1 is valid with C 1.21. Tnerefore (for L - 4W), !
oy - Pgy * (L/(B*(W-a)2*'.01]) = (3.3Pgy W)/[B(W-a)2] (2) i for the Charpy specimens, B 0.394 it , W = 0.394 in., and a = 0.079 in, i Equation 2 then reduces to:
ay - 33.3 x PCY (3) i i
where oy'is in units of psi and Pgy is in units of lbs. The flow l stress was calculated from the average of the yield and maximum loads, also i using the three-potr', bend formula. !
h Parcent shear was determined from post-fracture photographs using the ,[
ratio-of-areas methods in compliance with ASTM Specification A370-89I93 f
l The lateral expansion was measured using a dial gage rig similar to that shown j in the same specification.
S-2 /
1 l -
i Tension tests were performed on a 20,000-pound Instron, split-console test l machine (Model 1115) per ASTM Specification E8-89b(103 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 pull rod was !
connected through a universal joint to improve axiality of loading. Thn tests l were conducted at a constant crosshead speed of 0.05 inches per minute I throughout the test, Deflection measurements were made with a linear variable displacemcht j transducer (LVDT) extensometer. The extensometer knife edges were !
spring-loaded to the specimen and operated through specimen failure. The f extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 !
per ASTH E8's-85(12),
Elevated test temperatures were obtained with a three-zone ele- *ic resistance split-tube furnace with a 9-inch hot zone All tests were conQ cted in air, j l
Because of the difficulty in remotely attaching a thermocouple directly to the I specimen, the following procedure was used to monitor specimen temperature, f
'. Chromel-alumel thermocouples were inserted in shallow ules in the center and (
each end of the gage section of a dummy specimen and in each grip. In the test i configuration, with a slight load on the specimen, a plot of specimen g temperature versus upper and lower grip and controller temperatures was l 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 j temperatures. Experiments indicated that this method is accurate to 12'F. l The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield j strength, ultimate strength, and-fracture strength were calc 91ated 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 l the fracture stress (true stress at fracture) and percent reduction in area _was j computed using the final diameter measurement. {
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5-3 [
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t i___._,____,____._._.,_ . . - . _ _ _ , . . _ - . _ _ _ _ . _ _ _ . . ,
f r
5.2 Charov V-Notch Imoact Test Results !
The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule Y irradiated to 1.29 x 1019 n/cm2 (E > 1.0 MeV) are presented in Tables 5-1 through 5-4 and are compared with unirradiated !
resultsOI as shown in Figures 5-1 through 5-4. The transition temperature I increases and upper shelf energy decreases for the Capsule Y materials are summarized in Table 5-5.
i irradiation ~of the reector vessel lower shell plate R2708-1 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major I rolling 1 ection of the plate (longitudinal orientation) to 1.29 x 10l9 !
n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-1) resulted in a 30 ft-lb i
transition temperature increase of 15'F and a 50 ft-lb transition temperature increase of 30'F. This resulted in a 30 ft-lb transition '
temperature of 10'F and a 50 ft-lb transition temperature of 55'F (longitudinal orientation).
~
The average upper shelf energy (USE) of the lower shell plate R2708-1 Charpy ,
specimens (longitudinal orientation) resulted in a energy decrease of 8 ft-lb .
after irradiation to 1.29 x 10I9 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average USE of 118 ft-lb (Figure 5-1).
Irradiation of the reactor vessel lower shell plate R2708-1 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation) to 1.29 x 1019 n/cm2 (E >
1.0 MeV) at 550*F (Figure 5-2) resulted in a 30 ft-lb transion temperature increase of 45'F and a 50 ft-lb transition temperature increase of 45'F. This resulted in a 30 ft-lb transition temperature of 25'F and a 50 ft-lb transition temperature of 75'F (transverse orientation).
5-4
i The average upper shelf energy (USE) of the lower shell plate R2708-1 Charpy specimens (transverse orientation) resulted in an energy decrease of 13 ft-lb after irradiation to 1.29 x 1019 n/cm2 (E > 1.0 MeV) at 550'F. This
{
resulted in an average USE of 91 ft-lb (Figure 5-2). !
~
1rradiation of the reactor vessel core region weld metal Charpy specimens to 1.29 x 1019 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-3) resulted in a 30 j ft-lb transition temperature increase of 40'F and a 50 ft-lb transition !
temperature increase of 35'F. This resulted in a 30 ft-lb transitinn i temperature of -15'F and a 50 ft-lb transition temperature of 15'r [
i r
The average upper shelf energy (USE) of the reactor vessel core region weld f metal resulted in an energy decrease of 15 ft-lb af ter irradiation to 1.29 x !
1019 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average USE of f 97 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens ;
to 1.29 x 1019 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-4) resulted in a {
30 f t-ib transition temperature increase of 40*F and a 50 ft-lb transition ;
. temperature increase of 40 'F. This resulted in a 30 ft-lb transition temperature of -50'F and a 50 ft-lb transition temperature of -5'r The average upper shelf energy (USE) of the reactor vessel weld HAZ metal resulted in an energy decrease of 15 ft-lb after irradiation to 1.29 x 10 19 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average USE of 91 f 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. -
b l
P 5-5 :
i
- . _ _ _ _ , , - . . _ . - . - . ~ - . . - , . - -
9 i
i A comparisnn of the 30 ft-lb transition temperature increases and upper shelf (
energy decreases for the various Callaway Unit I surveillance materials with ;
predicted values using the methods of NRC Regulatory Guide 1.99, Revision !
2l43 is presented in Table 5-6, This comparison indicates that the ,
transition temperature increases and the USE decreases resulting from '!
irradiation to 1.29 x 1019 n/cm2 (E > 1.0 MeV) are in good agreement with -
the Guide predictions.
i The load-time records for the individual instrumented Charpy specimen tests are shown in Appendix A. I Photographs of all Charpy impact specimens before testing are shown in Appendix <
B.
1 5.3 Tension Test Resulti The results of the tension tests performed on the various materials contained in Capsule Y_ irradiated to 1,29 x 1019 n/cm2 (E > 1.0 MeV) are presented in ;
Table 5-7 and are compared with unirradiated resultsfl3 as shown in Figures .
5-9 through 5-11.
The results of the tension tests performed on the lower shell plate R2708-1 (longitudinal orientation) indicated that irradiation to 1,29 x 1019 n/cm 2
- (E > 1.0 MeV) at 550'F caused a less than 7 ksi increase in the 0.2 percent offset yield strength and a less than 6 ksi increase in the ultimate tensile strength when compared to unirradiated datall) (Figure 5-9).
The results of the tension tests performed on the lower shell plate R2708-1 (transverse orientation) indicated that irradiation to 1,29 x 10I9 n/cm2 (E
> 1.0 MeV) at 550'F caused a less than 6 ksi increase in the'0.2 percent offset yield strength and a less than 5 ksi increase in the ultimate tensile strength when compared to unirradiated dataIll (Figure 5-10).
5-6
The results of the tension tests performed on the reactor vessel core region weld metal indicated that irradiation to 1.29 x 10 l9 n/cm2 (E > 1.0 MeV) at 550'F caused a less than 8 ksi increase in the 0.2 percent offset yield
~
strength and a less than 7 ksi increase in the ultimate tensile strength when comparedtounirradiateddatall)(Figure 5-11).
The small increases in 0.2% yield strength and tensile strength exhibited by the lower shell' plate R2708-1 and the weld metal indicate that this material is not highly sensitive to irradiation to 1.29 x IC I9 n/cm2 (E > 1.0 Hey), as is also indicated by the Charpy impact test results.
The fractured tension specimens for the lower shell plate R2708-1 material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14. All scales shown in figures 5-12 through 5-14 and Appendix B are in 100ths of an inch.
The engineering stress-strain curves for the tension tests are shown in Figures
. 5-15 through 5-20.
- The true stress-strain curves for the tension tests are shown in figures 5-21 through 5-26.
Photographs of the untested tensile specimens are shown in Appendix D.
5.4 ggmpjtet Tension Tests Per the surveillance capsule testing program with the Union Electric Company, 1/2 T-compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center llot Cell.
5-7
r TABLE 5-1 i i
CHARPY V-NOTCH IMPACT DATA FOR THE CALLAWAY UNIT 1 l LOWER SHELL PLATE R2708-1 IRRADIATED AT 550'F. ;
FLUENCE 1,29 x 1019 n/cm2 (E > 1.0 MeV)
'l Temperature Impact Energy lateral Expansion l Samole No. F ,,1G_. it-lbi (Joules) mill (mm) % Shear l
l Lonoitudinal Orientation I CL75 -60 (-51) 7.0 ( 9.5) 5.0 (0.13) 5 !
CL68 -25 (-32) 16.0 ( 21.5) 11.0 0.28) 10 !
CL62 -5 -2 28.0 38.0 19.0 (0.48) 15 !
CL72 10 -1 15.0 20.5 13.0 (0.33) 10 !'
CL65 20 -
35.0 47.5 28.0 0.71) 20 CL61 CL64 30 40
-) 49.0 48.0 66.5 n.0 34.0 35.0 0.86) 35 !'
- 4) 0.89) 35 CL66 60 43.0 58.5 30.0 0.76) 35 .
CL74 80 33.0 44.5 31.0 0.79) 45 !
CL71 100 75.0 101.5) 52.0 1.32) 80 :
CL67 125 52) 89.0 120.5) 63.0 (1.60) 85 !
CL69 150 66) 103.0 (139.5) 74.0 (1.88) 100 CL73 200 117.0 (158.5) 75.0 (1.91 100 !
CL70 240 1 128.0 (173.5) 79.0 (2.01 100 -
CL63 275 1 124.0 (168.0) 79.0 (2.01 100
~
t Transverse Orientation f CT63 -60 -51) 11.0 ( 15.0) 5.0 (0.13 5 !
CT61 -25 -32) 9.0 ( 12.0 4.0 (0.10 5 CT64 -10 -23). 12.0 (16.5 7.0 (0.18 5 ;
CT62 0 (-18) 32.0 ( 43.5 22.0 (0.56) 15 '
CT68 20 -7) 28.0 ( 38.0) 19.0 (0.48) 15 !
CT65 40 4) 37.0 ( 50.0) 27.0 (0.69) 35 l CT67 50 10) 40.0 ( 54.0) 28.0 (0.71) 35 CT72 60 (16) 37.0 ( 50.0)- 30.0 (0.76) 35 :
CT66 75 24) 48.0 ( 65.0) 33.0 .(0.84) 45 CT70 100 38) 64.0 (87.0) 50.0 (1.27) 60 ,
CT74 125 52) 62.0 (84.0) 49.0 (1.24) 60 t CT73 150 ( 66) 66.0 89.5) 55.0 (1.40) 90 l CT75 175 ( 79) 93.0 126.0) 66.0 (1.68) 100 'l CT71 200 ( 93) 88.0 119.5) 62.0 (1.57) 100 i CT69 250 (121) 93.0 (126.0) 66.0 (1.68) 100 .
5-8 i
F TABLE 5-2 CHARPY V-NOTCH lHPACT DATA FOR THE CALLAWAY UNIT 1 REACTOR f VESSEL WELD METAL AND HAZ METAL 1RRADIAl[D AT $
550'F. FLUENCE 1.29 x 10 l9 n/cm2 (E > 1.0 MeV) i l
Temrerature Im)act Energy lateral Expansion Samole No. T ,_1Q. 1,L,ln (Joules) _mih ,_131)_ ' % She n (
Weld Metal f
CW72 -75 (-59) 5.0 ( 7.0) 4.0 (0.10) 5 CW61 -50 (-46) 13.0 17.5 14.0 (0.36 15 i CW65 -25 (-32) 29.0 39.5 22.0 (0.56 25 i CW69 -10 -2 39.0 53.0 35.0 (0.89 35 t CW75 0 -1 45.0 61.0) 38.0 (0.97) 50 ,
CW64 15 - 49.0 66.5) 40.0 (1.02 55 !
CW62 25 -4) 47.0 (63.5) 40.0 1.02 55 l CW63 50 (10) 83.0 li2.5) 63.0 1.60 80 t CW66 65 ( 18) 83.0 112.5) 59.0 1.50) 70 !
CW68 75 24) 100.0 135.5) 71.0 1.80) 90 CW70 100 38) 89.0 (120.5) 69.0 (1.75) 90 l
+
CW74 125 52) 63.0 ( 85.5 71.0 (1.80) 90 CW73 150 ( 66) 99.0 (134.0 82.0 (2.08) 100 CW67 175 ( 79) 98.0 (133.0 75.0 (1.91) 100 ,
CW71 200 ( 93) 94.0 (127.5) 70.0 (1.78) 100 ;
l MZ Metal CH67 -130 (-90) 4.0 ( 5.5 4.0 (0.10 5 P CH72 -100 (-73) 22.0 (30.0 14.0 0.36 15 :
CH66 -/5 (-59) 22.0 ( 30.0 16.0 0.41 15 :
CH71 -50 (-46) 33.0 44.5 24.0 0.61 20 :
CH64 -25 (-32) 27.0 36.5 26.0 (0.66 30 i CH65 -15 (-26) 53.0 72.0) 40.0 (1.02 45 i CH62 0 (-18) 48.0 ( 65.0) 36.0 (0.91 45 !
CH69 25 ( -4) 40.0 (54.0) 31.0 (0.79) 45 CH63 50 ( 10) 68.0 (92.0) 50.0 (1.27) 75 .
CH0c 75 24) 99.0 (134.0) 70.0 1.78) 100 :
CH75 100 38 86.0 (116.5) 63.0 1.60) 100 l CH61 125 52 85.0 (115.0) 60.0 1.52) 100 ;
CH74 150 66 80.0 (108.5) 62.0 (1.57) 100 -
CH73 175 -79) 94.0 (127.5) 65.0 (1.65) 100 !
+ CH70 200 93) 101.0 (137.0) 70.0 (1.98) 100 ;
i i
I 5-9 !
TABLE 5-3 10STRUMENTED DiARN 189ACT TEST RESULTS FOR THE CALLAWAY OIT 1 LOWER SHELL.
II PLATE R2708-1 IERADIATED AT 550*F. FLUENCE 1.29 X 10 n/cmI (E > 1.0 Mev)
Mornalired Enercies Test Charpy Cha my Mantmum Prop field Time Man teurs Time to Fracture Arrest Yleid Flow Sample . Temp Energy Ed/A Em/A Ep/A' Loed to Y6 eld Loed Mariatsa Loed Load Stress Stress Number (*F1 fft-1b) (ft-Tb/inE) (k +ps) (Usee) (k irs) fpsee) inirs) (bst) (bst). [isil, tenoitudinal Or
- tetten CL75 -60 7.0 56 36 20 2.95 85 3.25 125 3.25 0.00 97 1 02 CL68 -25 16.0 129 103 26 3.30 80 3.85 255 3.85 109 118 CL62 -5 28.0 225 169 56 2.90 85 3.80 430 3.80 s .. , 96 111 CL72 10 15.0 121 75 46 2.80 85 3.20 225 3.20 0.25 92 99 CLES 20 35.0 282 241 41 3.10 65 4.15 555 4.15 0.15 103 120 CL61 30 49.0 395 232 163 2.65 90 3.60 605 3.55 0.00 66 107 CL64 40 48.0 387 280 1 06 3.35 80 4.50 605 4.45 0.50 110 130 CL66 60 43.0 346 221 126 3.30 85 4.25 510 4.25 0.95 109 125 CL74 60 33.0 266 87 1 79 2.65 70 3.30 260 3.25 1.05 87 99 CL71 1 00 15.0 604 263 321 2.95 J5 4.10 650 3.90 1.55 97 116 Ct67 125 89.0 117 268 449 2.40 95 3.45 770 2.75 0.30 79 97 103.0 *
- CL69 150 829 210 619 2.55 90 3.65 570 64 1 02 y' CL73 20J 117.0 942 215 727 2.25 115 3.30 655 *
- 74 92 128.0 *
- Es -- CL70 240 1031 234 796 2.15 50 3.60 E25 71 35 CL63 275 124.0 996 259 140 1.95 70 3.35 765 54 87 Transverse C<icetetten CT63 -60 11.0 59 56 32 3.05 85 3.40 170 3.35 0.00 102 107 CT61 -25 9.0 72 41 31 3.40 90 3.60 130 3.50 0.00 112 115 CT64 -10 12.0 97 62 35 3.20 85 3.50 185 3.50 0.35 106 til CT62 0 32.0 258 198 60 2.85 85 3.75 500 3.75 0.15 94 109 CT68 20 28.0 225 180 45 3.05 85 4.05 425 4.05 0.00 101 117 CT65 40 37.0 298 226 72 3.20 85 4.25 520 4 15 0.35 106 123 CT67 50 40.0 322 205 117 2.55 95 3.55 585 3.50 0.15 85 101 CT12 60 37.0 298 208 90 3.05 85 4.25 495 4.20 0.50 101 121 CT66 75 48.0 387 217 169 2.60 40 3.90 520 3.70 0.95 45 107 CT70 100 64.0 $11 221 294 2.60 80 3.60 575 3.45 1.45 85 103 CT74 125 62.0 499 181 318 2.55 85 3.50 500 3.30 1.60 84 100 CT73 150 66.0 $31 2 04 328 2.80 60 3.90 500 3.65 2.45 92 111 CT75 175 93.0 749 204 545 2.15 55 3.40 580 71 92 200 88.0 *
- CT71 709 206 502 2.30 85 3.50 600 76 96 CT69 250 93.0 749 218 531 2.25 100 3.50 620 75 95
- Fully ductile fracture; no arrest load
%. - - , . - - - . . , ~. ,. . -, _ _ . -
~ __
TABLE 5-4 IN5TRUMENTED CHMPY lurACT TEST RESULT 5 FOR THE CALLAWAY tmIT I vet 0 METAL Aito HE AT- AF F ECTEC-ZOME (HAZ) MET AL.1: RADIATED AT 550*F. FLUEMCE 1.29 E 10 n/cm2 (E
- 1.0 m Norme1tred Enernses Y le M F low Prop Yteld Time Mastmum Time to Fracture Arrest Test Charpy Charpy Maninsu Stress Stress Manteue Loed Load Ed/4 Er/A Ep/A toad to Yield Lead Sample Temp Energy III?ec) (kir1} (ksil (6sil (6s')
(ft-15/6nE l (b *rs) 1/Isec) (k irs)
Number 1*J[1 (ft-ib}
Weld Eetel 3.35 0.15 100 106 3.05 120 3.50 130 5.0 40 26 14 CW72 -75 3.55 0.45 98 109 67 3.00 80 3.60 125
-50 13.0 105 38 122 CW61 435 4.05 0.60 110 45 3.30 85 4. 05
-25 29.0 234 !88 94 107 CV65 3.E0 560 3.60 0.55 219 95 2.85 85
-10 39.0 314 120 CW69 4.10 500 4.00 1.55 104 218 144 3.15 80 0 45.0 362 91 1 06 CW75 3.65 505 3.55 1.65 166 208 2.75 95 CV64 15 49.0 395
- 3. 75 1.65 97 112 2.95 60 3.25 500 47.0 378 203 176 CW62 25 585 2.90 2.05 86 1 02 206 462 2.60 90 3.55 CW63 50 B3.0 668 1.55 80 98 70 3.50 560 3.10 65 83.0 (68 206 4 63 2.45 109 CW66 2.83 2.00 92 250 555 2.80 40 3.80 EOS CV68 75 1 00.0 605 2.15 87 101 2.65 85 3.45 520 3.00 CW70 100 89.0 717 les 53 0 2.30 82 96 85 3.45 595 2.90 63.0 507 214 293 2.50
- CW78 125
- 92 110 9F 2.80 ES 3.90 610 248 550
- CW73 150 99.0 797 E55 70 90 228 561 2.10 60 3.35 CW67 175 93.0 789 *
- 52 80 1.60 60 3.25 765 94.0 157 244 513 .
CW71 200 MA? m e ts) 2.45 0.00 62 TI 1.85 10 2.45 35 4.0 32 10 23 CH67 -130 270 3.75 0.00 113 120 107 70 3.40 90 3.85 CH72 -100 22.0 177 4.20 0.20 123 132 3.75 90 4 25 270 22.0 177 116 61 CPE6 -75 505 4.35 0.00 120 132 235 30 3.50 90 4.35
-50 33.0 265 115 127 CH11 4.15 335 4.05 1.10 27,.0 217 145 72 3.50 85 CH64 -25 570 3.40 0.65 96 ill 231 195 2.90 90 3.80 CH65 -15 53 0 427 1,35 108 123 4.15 575 4.10 48.0 387 259 128 3.25 ES CH62 0 3.15 1.30 92 104 124 198 2.75 90 3.50 3 50 CH69 25 10.0 322 0.70 68 105 3.70 500 3.30 68.0 548 180 3 68 2.65 100 CH63 50 690
- 86 105 258 539 2.60 115 3.75 CH68 75 99.0 797 *
- 100 115 437 3.00 85 3.95 605 CH75 100 86.0 692 256 *
- 74 94 2.25 50 3.45 580 85.0 684 207 477 CH61 125 *
- 83 97 2.50 90 3.40 425 80.0 644 149 495 CH74 150 *
- 72 92 2.20 55 3.40 645 94.0 757 225 532 CH73 175 *
- 82 98 204 6 09 2.50 95 3.45 58 5 CH70 200 101.0 813
- Fully ductile f racture; no arrest load
. - _ _ - - b
i TABLE 5-5 EFFECT OF 550*F IRRADIATIOW TO 1.29 m 10II n/cm2 (E
- 1 0 MeV) 04 THE NOT31 TOUGimESS PROPERTIES OF THE CALLAWAY UNIT 1 REACTOR VESSEL SIRVEILLANCE MATERIALS i
Average 30 ft-Ib III Average 35 at1 III Average 50 ft-1b III Average Erergy III Transition Lateral Empens6cn Transitton Absorpt6on at Temperature (*F) Temperature (*F) Teeperatuee (*F) Full Shear (ft-lb)
Nterial llairradiated Irradiated AT un tradiated Irradiated AT untrradiated 'rrradi.ted Ar unieradiated Irradiated A(ft-1b)
Plate R2708-I - 5 10 15 25 50 25 25 55 30 126 118 - 8 (Longitudinal) y Plate R2706-1 - 20 25 45 25 70 45 30 75 45 104 91 - 13 O
(Transverse)
Weld Metal - 55 - 15 40 - 25 0 25 - 20 15 35 112 97 - 15 4
HAI Metal - 90 - 50 to - 45 0 45 - 45 - 5 80 106 51 - 15 I
(1) "A(ERAGE" is defined as the value read from t N curve fttted through the data points of the Charpy tests (Figs.res 5-1 througS 5-4).
i e
A
}.
s O l
TABLE S-6 COMPARISON OF THE CALLAWAY UNIT 1 SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS 30 ft-lb Transition Temp. Shift Upper Shelf Enerny Decrease R.G. I.99 Rev. 2 R.G. 1.99 Rev. 2 (Predicted) Measured Fluence (Predicted)(a) Measured l9 n/cm 2 (.7) (g) (y)
Material Capsule 10 (.7) 28 0 14.8 1.6 Plate R2708-1 0 0.327 44 15 20.3 6.3 (Longitudinal) Y 1.29 30 14.8 10.6 Plate R2708-1 U 0.327 28 44 45 20.3 12.5 b,
(Transverse) Y 1.29 70 14.8 9.8 L' eld Metal U 0.327 22 34 40 20.3 13.4 Y 1.29 50 -- 0 ;
MAZ Metal U 0.327 --
40 -- 14.2 Y 1.29 --
a)
Mean wt. % values of Cu and Ni from References 1 and 2 were used to calculate the chemistry factors for the surveillance material.
l TABLE 5-7 TENSILE PROPERTIES FOR THE CALLAWAY UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS IRRADIATED AT 550*F TO 1.29 N 10 ' n/cm (E > 1.0 Mev)
Ult imate Fractu*e Fracture Fracture Unifoem Total Reduct6cn Test 0.2% Yield Strength Load Stress Strenggh Elongat ica F. longat ice tn Area Sa cle Teap Strength (hie) (ksi) (b's t) (%r ft) f%)
Mateefst 4;mber 1*f_[ (ksil (bst) 194.0 60.5 12.9 26.8 64 Plate R27D8-1 CL13 70 73.5 95.7 2.97.
169.3 56.0. 11.1 24.3 62 (Longitudinal) CL14 225 67.7 87.6 2.75 3.00 160.6 61.1 10.5 21.6 64 CLIS 550 62.6 88.6
?. 3.30 134.1 67.2 12.9 23.1 54 Plate 22708-1 CT13 70 72.3 93.7 61.1 12.0 23.1 59 CT14 175 63.3 88.6 3.00 190.2 (Transverse) 54 88.6 3.35 170.2 68.2 9.6 18.6 CT15 550 63.7 l
208.8 63.2 10.5 23.4 67 Weld Metal CW13 70 80.5 95.7 3 10 i 92.7 3.00 184 7 61.7 9.0 21.3 E6 CW14 125 78.4 165.1 61.1 10.8 21.3 63 CV15 550 64.2 85.6 3.00
-150 -100 -50 0 50 100 l';0 c 00 250 l l _l l I I I I 2
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$ 60 - -~ 90 C 30'r zpcps-40 - 3 WKw
)$'r ;- # %- 0
, - 40 c0 - +
0 0
-200 -100 0 100 C00 300 400 500 TEMPERATURE Pf's rigure 5-1. Charpy V-Notch Impact Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) i I
i 5-15
- _ - - -- - .. .=- - -.- . .- - - .- - - . .-_.-. ..-- .. .
FC) : 150 -100 -50 0 50 100 150 200 250 !
~
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9 - e . , 8 80
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ofp'pc . i W s' a o 40 - 2 , 20 cv'ne L 0 M E*d I I I ! I ! i 100 25 h80 - 2g - 20 ; [ 60 - N - 15 - ! e 40 - 3" ofe p.43 r - 10 r E a 20 - 2- ,# 0' - 0.5 ! 0 I I' r ! I I I ! ! 0 o stAttAt[D o WADIA1[D (550't), rtut8tt 1;.9,10 0 vcd 160 : 200 140 120 - 160 : o
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- l I' l l 0 0 J
-200 -
100 0 100 ,?00 300 400 500 . TEMPERATURE ( n ( ! Figure 5-2. Charpy V-Notch Impact Properties for Callaway Unit 1 Reactor - L Vessel Lower Shell Plate R2708-1 (Transverse Orientation) j 5-16 i I i
,,u ,,,~----nsm-- ,a-,.w ,--
PC) :
-150 100 -50 0 50 100 150 200 250 '
I I I I I l l l l ! 100
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33r- g k E j40 - 1.0 2 20 - t, ,/ -*,.W - 05 I r*' " I I I I I I 0 0 1 o s tADIA'[D a :RRAliAf(D d50'rt TLLO([ 129 a 10b/cd j 160 ! . 200 140 120 0 160 n S 100 - f #n f O < 120 - I O [o'
': 80 - O *
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-200 -100 0 100 200 300 400 500 l TEMPE S E (*F) !
figure 5-3. .Charpy V-Notch Impact Properties for Callaway Unit 1 Reactor Vessel Core Region Weld Metal ; I 5-17 i t,
, _ . . _ . ,,. ._ __._,,__ , . _ . . _ _ . . _ . . . _. , _ ___ . - . _ . _. ..__._.._..__..__)
( C)
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- 3 3 -l 6: 2
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- 40 -
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g- 45'r 10 S-3 20 o - 0.5 0 - ' 0 o UNIGADIAf[D a !RRADIATED (550'r), rLutTC 12)i 10 N n/:4
~
140 r o ! 120 o - 160 l e 100 - to L hl n i 5 80 - a 1 ,$ G a
@ 60 -
R0 O -- 4c'r
~ '
40 - a S 20 - [ t g 40'r e 40 I r I 0 I l l l l - J-- 0
-200 -100 0 100 200 300 400 500 -
TEMPERATURE ( F) Figure 5-4. Charpy V-Notch Impact Properties for Callaway Unit 1 Reactor ; Vessel Weld Heat Affected Zone Metal j. 5-18 l t t
, n~- - , n - , - - - - . . - - - - - - - - _ _ . - - - - - _ _ - - _ . - - - - - - - - -
'.. s .
,i. '
g l JJ j_4 v FY3
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- - 1 3
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- 4. .
4 i- f. 3;g . g s s , s L s. , C. :~ /' ~ (
't i -.,.a CL75 CLOH CLG2 CL72 CLO5
. . _ . . . , . . - ,n ,,q. m
- 4,
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V . ,,. + s.,_ g. ;
., 3 i e
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-_ ;%SL%
CLG1 CLG4 CL60 CL74 CL71
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Yt . f 9%.{ % 54 , f.L, N4sf,
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. a. M p'4g ,~7,( '(,
l _ f = 8,,/ 5 ,6,1 ,. A s_ __ . + CLG7 C;09 CL73 ClJO CL63 Figure 5-5. Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) 5-19
~ ---
t, i e j s; .i
, 3 geg_~ v- C; gj *;$5$ frgy' + a- ,
.,_..m
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t.e . . y ,.,; , r
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i r i t CT03 CT61 CT64 CT62 CTG 8
~, - -~
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), 1 $
7
.._ . ; xmd1 .u_ . m m CT74 CT73 CT75 CT71 CT69 .
Figure 5-6. Charpy impact Specimen fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) 5-20
i l
?
._ ,, ;-e b
(}
, , , 1 f e t I
i
, i. . 1 ,
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i CW70 C%74 CW73 CWG7 CW71 ,
. i h
t Figure 5-7. Charpy impact Specimen fracture Surfaces for Callaway Unit 1 l Reactor Vessel Weld Metal ! 5-21 f
. ____ . . . _ ___ - _ _ . _ . _ _ -_ _ _ ~ _ _._ . _ ___._ ___ _ ___.. _ .__ _ .. _
+
h- ~_m; g 'h.e7~l~
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' ' 7' y , , j. . ul,
)
, , e g,sce u, 46; .a : a.,
e,y . o, # a
.y j
Cll75 C1161 C1174 Cil73 C1170 h i l Figure 5-8. Charpy impact Specimen Fracture Surfaces for Callaway Unit 1 ' Reactor Vessel Weld Heat Affected Zone Metal t 1 5-22
( C) 0 50 100 150 200 250 300 120 800 l l-
!!0 -
700 . 100 e ULilMAl[ 1[N!!LL $1R[t61H x o 90 y 0 N __ g- 600
" / s' g 80 - 2 W L -
500
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60 2 2' it - 400 02% W MP[NOTH 50 300 40 I I b- l l a o MRAmt[D A e IRRADIAt[D At 550 f, (LUENCE !?9 x1019n/cn2 80 2 RCDJCflDN IN AP[A
/ a 70 o
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20
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- = "
umr0RM [tasita I I I 0 0 100 200 300 400 500 TEMPERATURE (*D 9 Figure 5-9. Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) 5-23
(* C) 0 50 100 150 200 250 300 120 l l -l l j j j= 800 110 700 100 m r,, g0
-- "N VL11MATC T[Ni!LC 11RENGTH
-n -
600
\2 d5 1 o e
M 80 y y - 500 s. 60 -
" - 400 2
02% YlELD STRD4GTH 50
.i0 6 o UNIPRADIAILD AN 1RRADIAf[D Al $50 f, FLUENCE 129 x10I9 cn i 80 70 erI'UCil0N IN M1A o .
,[ j
@ 50 -
A 10TAL ELONGAll0N g
>- A y 40 -
$ 30 -
a 20 - 1h- A
----A s
n W % ,. 8 D 0 UNlrDRM [LQ4GAilDN 0 0 100 200 300 400 500 TEMPERATURE ('n Figure 5-10. Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) 5-24
('C) 0 50 100 150 200 250 300 120 j j l J l j j= 800 . 110 700 : 100
~
s (L11 mal [ ILNSit[ $TR[NG1H b U(
~
A " p 30 - A
%_ N d 95 W
G 70 - 6 N- %A- > w/ y ' c 500 5 NA 60 - 02% YlELD $1ECNGTH - 400 e0 - 40 , A 0 UNIERAD!A1(D Aa IREA0!At[D At 550 f, FLUENCC 129 =10 l9 n/&2 80 - 70 M110N IN M[ A 5 u- c- __, . 60 -
\ 2
\ 2 2
@ 50 -
3 40 -
=
- 2 2 30 tu1AL ELONGA110N
~_ g-- ~ ,g g 20 -
3 C __ a 10
*e uNirmw ttcNrAllDN 0
. 0 100 200 300 400 500 TEMPERATURE ("F) Figure 5-11. Tensile Properties for Callaway Unit 1 Reactor Vessel Core i Region Weld Metal 5-25
t t i i t n' 1 o q),ci;; 4 +
.w < . ,
,,in,M.pM._Nr -,
Specimen C1,13 70*F
~
. f 4, a; . . .
.+ !nlif M '
[- g ~$ .
~w 9 ' " N'
\'g ?
[$ .:l';.{ ~ Ss - ,
'% ~ >o._, ;
. - g 9e m .
av s (V Specimen CLl4 225'T ! i Specimen CL15 550*F 4 4 2 d ] Figure 5-12. Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel i Lower Shell Plate R2708-1 (Longitudinal Orientation) 5-26 1
^~ -
l, l l i ) Specimen CT13 70*f Specimen CT14 175'r i E Srecimen CT16 550*F l t l figure 5-13. Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel 3 Lower Shell Plate R2708-1 (Transverse Orientation) 5-27
. . - . . _ - - - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ - - _ _ _ _ . ~ _ _ _ _ . _ . _ _ . _ _ _ - . _ . . _ . - . _ _ . . _ . . _ _ _ - _ _ . _ - . _ = . _ _ . - _
- _ _ _ _ _ .i t
i i i I,
. Specimen CW13 70*F' i
=q , 4 -
- 3. g . i 2. . j .. - - re,, , n., ggs, b Specir.en CW14 i 125'F i
.i
.i r
I Specimen CW15 550*F , t i i Figure 5-14. Fractured Tensile Specimens from Caliaway Unit 1 Reactor Vessel ! Core Region Weld Metal i
?
y 5-28 i, i
100 9 90 . 80 s M 70-60-h50-h 40-30- , 20- SPEC CL13 t 10- 70 F , 0, . , i i i 0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN IN/lN 90 - i 80- ,
~
70d m 60 x !
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M i y 40-i -
+ !
- 30 J i
20" i SPEC CL14 J 10 : 225 F i i 0 . i i 0 0.05 0.1 0.15 0. 2 0.25 STRAIN IN/IN Figure 5-15. Engineering Stress-Strain Curves for Plate R2708-1 Tensile Specimens CL12 and CL14 (longitudinal Orientation) 5-29 I
10 0 ~ --- - --- --- -- 90i N :' 802 i _. 70 d N 60-h50-h 40-30 2 20- SPEC CL15 10- 550 F ; O. . , , , ! 0 0.05 0.1 0.15 0.2 0.25 STRAIN IN/IN Figure 5-16. Engineering Stress-Strain Curve for Plate R2708-1 Tensile Specimen CL15 (Longitudinal Orientation) 5-30
100 90- ' . 80 ' l i
-701 I i @ 60-i !
g 50-i m i ! M 40 , m ' 30-20-SPEC CT13 ; 10-70 F {
- 0. i i i '
O 0.05 0.1 0.15 0.'2 0.25 STRAIN IN/lN 100] 90 J l
. 80 ,
70 J a t x 60- : 8 50-t.a l m 40- ' 30- ! 20-SPEC CT14 . 10-175 F ' 0. . i e > ' O 0.05 0.1 0.15 0.2 0.25 . STRAIN IN/lN Flgure 5-17. Engineering Stress-Strain Curves for Plate R2708-1 Tensile Specimens CT13 and CT14 (Transverse Orientatica) l 5-31 ______-_-_____A
f 100=-- - 90- _ l 50 d f a 7~/ ! l x 601 . l s 50 -l I 5 40- . m . 3 0 ,, - i - SPEC CT15 20-101 550 F ' 0' , , , , , d 0.b2 0.04 0.06 0.08 0.1 0.12 0.I 4 0.i 6 0.i 8 0.'2 STRAIN IN/IN Figure 5-18. Engineering Stress-Strain Curve for Plate R2708-1 Tensile Specinen CT15 (Transverse Orientation) 5-32 l
. ~.
100] 90i 80-i( ' 70- 1 i a 60 J l l x
. 50 , <
m 1 0 x-40- ! t
$ 30- !
1 ; 20- SPEC CW13 l l 10- 70 F ' Oi . , , 0 0.05 0.1 0.16 0. ,. 0.25 STRAIN lN/lN 100 q 90- . 1 80-i i : i 70-m
- 60, !
N 50- ! u s 40- i M i 30- i i 20- SPEC CW14
.- 10- 125 F l
i 0 > > > . 0 0.05 0.1 0.15 0.2 0.25 STRAIN IN/lN Figure 5-19. Engineering Stress-Strain Curves for Weld Metal Tensile Specimens CW13 and CW14 5-33
100 - m 90- _ 80 i I 70- i rn
/ i 601.#
E. 50 J ! E i y .40- l 30 ' !, 20 SPEC CW15 i 10- 550 F i
- 0. . . . .
0- 0.05 0.1 0.15 0.2 0.25 STRAIN IN/IN Figure 5-20. Engineering Stress-Strain Curve for Weld Metal Tensile Specimen CW15 5-34
15 0 ----- - - - - - - - - - - - - - 125- CL13 -; Maximy Loud _. i T100-5 N - t E 75-
- 1 M !
o , H 2 50- ! l I 25-I 0 0.00 0.05 0.10 0.15 0.20 0.25 - True Strain (in/in) 150-F ! 125- CL14-t i T100- M ximum Loa _d_ _ 6 i M M
- 75- .
m e ! 2 50-H I ! I ! 25 ! ! l l r i : 0! > i . i 0.25 0.00 0.05 0.10 0.15 0.20 True Strain (in/in) I Figure 5-21. True Stress-Strain Curves for Plate R2708-1 Tensile Specimens l . CL13 and CLl4 (Longitudinal Orientation) i 5-35
g t I
-150 -- . . - - - _.
i 125- CL15 , i
! f y100- p L.M_d. _ i v .
m m
- 75- 4 m ,
e ! a 4 H' 50- i
, L I
25 J - l r i 0! - I 0.00 0.05 0.10 0.15 0.$0 0.$5 , True Strain (in/In) i Figure 5-22. True Stress-Strain Curve for Plate R2708-1 Tensile Specimen CL15 (Longitudinal Orientation) . l i 5-36 [ a __ . _ _ _ _ _ . _ . _ . _ _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . . _ . _ - _ ______ m-___ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . . _ _ _ _ . _ _ - . . _ _ _ _ _ . _ . - _ _ - _ _ _ _ _ _ __
150- --- 125- CT13 ; i i ; Maximum L_od._ ! T 1001 6 l l f i E' I , S 75 4 ; , e tn e - I s2 50 ! - i 25-
- 0. . . . . -i 0.00 0.05 0.10 0.15 0.20 0.25 True Strein (in/in) 150-- --- ---
1 ! i Y
. 125q CT14 )
- i, 100- kd--
n- , n- ! > S 75 ; ; l e r2 50- : 25- ! I
+- 0 . . . . i 0.00 0.05. 0.10 0.15 0.20 0.25 ,
I True Strain (in/in) i Figure 5-23. True Stress-Strain Curves for Plate R2708-1 Tensile Specimens I CT13 and CT14 (Transverse Orientation) , 5-37
. =. -- . . . - _. -
150 i i 4 ' I 125- CT15 ! T100 1 Maximum Loc _d _ U , E- I o 75- i m ! E i i C H 50-4i :
! I i ,
4 25 l . O' O.00. 0.05 0.10 0.15 0.20 0.25 True Strein (in/in) Figure 5-24. True Stress-Strain Curve for Plate R2708-1 Tensile Specimen CT15 (Transverse Orientation) 5-38
150- -- i t l 125 d CW13 ! Maximum Load _, j T1004 i 0 h m 0 75-li t o 2 50 - H i i 25- ! l 0 . i i i
- 0.00 0.05 0.10 0.15 0.20 0.25 True Strein (in/in) 150q .
I i 125d CW14 , Maxim L_oc_d _
,100d p; 5 '
8 75 j tn o H 2 50 - 25-
\
i 0.10 0.k 5 0.b0 0.N5 h.b0 0.b5 True Strain (in/in) Figure 5-25. True Stress-Strain Curves for Weld Metal Tensile f aecimens CW13 and CW14 5-39
150 7 -- -- - - - --'-- -~ 7 I 125- CW:5 ; i n 1 j 100j
.v Maximu Load._
E ! > 2 75 i { Di i.
,o n- i s2 50 ' ;
{ i 25- : , , 0 . . . . > 0.00 0.05 0.10 0.15 0.2 t, 0.25 True Strain (in/In) Figure 5-26. True Stress-Strain Curve for Weld Metal Tensile Specimen CW15 , 5-40
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY l
- 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 contaired 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 r.s for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall. Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853,[23] " Analysis and Interpretation of Light Water Reactor 6-1
Surveillance Results," 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,l213 " Characterizing Neutron Exposures ;n Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the tk.;ckr. ass of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99,I43 " 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 Qrdinates 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), d(E > 0.1 Mev), and dpa} through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/d(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 inisrior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations. The second set of calculations consisteo of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal loutions on the pressure vessel inner radius to neutron source distributions within the reactor core. The imp:*tance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for 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
- - - - - - - - - - - _ _ _ - - _ _ _ _ ___ _ _ __ __ J
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:
- l. 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 calculatioi. 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 codell33 and the SAILOR cross-section libraryIl43 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 Sg order of angular quadrature. The reference core power distribution utilized in the forwaru analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in tne 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 j quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source locations were chosen at several azimuthal locations along the l pressure vessel inner radius as well,as the geometric center of each I surveillance capsule. Again, these calculations were run in R, 0 geometry to provide neutron source distribution importance functions for the exposure j parameter of interest; in this case, d (E > 1.0 MeV). Having the j importance functions and ag opriate core source distributions, the response of interest could be calculated as: R(r,0)-fr[0[E I(r, 0, E) S (r, 0, E) r dr de dE where: R (r, 0) - 4 (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. j 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 icplementation of low leakage loading j patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relathe neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/d (E > 1,0 MeV) is insensitive to changing core source distributions. In the application of these ; adjoint importance functions to the Callaway 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 4 (E > 0.1 MeV)/d (E > 1.0 MeV) ratios from the forward analysis in f conjunction with the cycle specific d (E > 1.0 MeV) solutions from the ! t individual adjoint evaluations. , 6-5 t [
The reactor core power distribution used in the plant specific adjoint calculations.was taken from the fuel cycle design reports for the first four operating cycles of Callaway Unit 1[15 through 18], 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 [d (E > 1.0 MeV), d(E > 0.1 MeV), and dpa] are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure , parameters are listed for both the design basis and the cycles 1 through 4 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 azituthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5. ' 6-6
For example, the neutron flux (E > 1.0 MeV) at 'the 1/4T position on the 45' azimuth is given by: ) i
= p(220.27, 45*) F (225.75, 45') j 41/4T(45*)
where: = Projected neutron flux at the 1/4T position on 41/4T(45') the 45* azimuth l p (220.27,45') = Projected or calculated neutron flux at the I 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 00T calculations were carried out for a typical octant of the reactor. However,- for the neutron pad arrangement in Callaway Unit 1, the pad extent for j all octants is not the same. For the analysis of the flux to the pressure l 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 j with 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. f 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 Callaway Unit I surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary i nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent > determination of the various exposure parameters of interest [p (E > 1.0 Mev), p (E > 0.1 MeV), dpa]. : 6-7 f l
The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptuntua 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 [19 through 32] . Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the Callaway Unit I reactor during cycles 1 through 4 was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the 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 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 (333 The 6-8
l FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at 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 & by some response matrix A: (s,a) (s) (a) f -I A 4 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 o d i 9 19 9 relates a set of measured reaction rates Rg to a single spectrum pg by the multigroup cross section ajg . (In this case, FERRET also adjusts the cross-sections.) Tne log-normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties. In the FERRET analysis of the dosimetry data, the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53 groups. The calculated
. fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-Il code [34]. This procedure was carried out by
- first, expanding the a priori spectrum into the SAND-Il 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed 6-9
L to the group scheme used in FERRET. The cross-sections were also collapsed into the 53 energy-group structure using SAND !! 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 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant. For each set of data or a priori values,- the inverse of the corresponding relative covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is used: M gg,=Rh+R g R,Pg gg, where RN specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties Rg specify additional random uncertainties for group g that are correlated with a correlation matrix: Pgg, = (1 - 0) 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). For the a priori calculated fluxes, a short-range correlation of 8 - 6 groups was used. This choice implies that neighboring groups are strongly . correlated when 0 is close to 1. Strong long-range correlations (or -anticorrelations) were justified based on information presented by R.E. - Maerker[35l. Maerker'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. 6-10
Results of the FERRET evaluation of the Capsule Y dosimetry bre given in Table , 6-9. The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1.29 x 10 I9 n/cm2 (E > 1.0 MeV) with an associated uncertainty of 8%, Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the adjusted spectrum are provided in Table 6-10. In general, excellent results were 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 exposure period underpredicted the measured value by 53 percent. Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (4.60 EFPY) exposure derived from the Capsule Y measurements, projections are also provided for an exposure period of 17 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 4 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 V. This procedure resulted in the following bias factors being applied to the analytical results: Flux (E > 1.0 MeV) Bias - 1.147 Flux (E > 0.1 MeV) Bias = 1.059 dpa/sec Bias = 1.087 6-11
In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Callaway Unit I reactor coolant system, exposure projections to 17 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 RTNOT vs. fluence trend
~
curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations d' (1/4T) = $ (Surface) (dp ( rf e)} d' (3/4T) - p (Surface) (dp ( urf e}} Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Table 6-15 updated lead factors are listed for each of the Callaway Unit I surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules. O e e 6-12
(TYPICAL) Co - 58.5 0 - 61.0 0 Fe
, g N L NEUTRON PAD 1 As xxxxxxx cN.
Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVE!LLANCE CAPSULE CENTER p(E > 1.0MeV) ((E > 0.lMev) Iron Displacement Rate In/cm2-sec1 In/cm2-sec1 Idoa/seci ; 29.0* 31.5" 29.0* 31.5* 29.0* 31.5* DESIGN BASIS 1.18 X 10 Il 1.26 X 1011 5.31 X 1011 5.66 X 10 ll 2.31 X 10-10 2.47 X 10-10 CYCLE 1 8.88 X 1010 9.46 X 10 10 3.99 X 10 ll 4.25 X 10 ll 1.74 X 10-10 1.85 X 10-10 i CYCLE 2 7.69 X 10 10 7.99 X 10 10 3.46 X 10ll 3.59 X 10ll 1.51 X 10-10 1.56 X 10-10 CYCLE 3 7.93 X 10 10 9.04 X 10 10 3.57 X 1011 4.06 X 10 11 1.55 X 10-10 1.77 X 10-10 CYCLE 4 6.82 X 10 10 7.51 X 10 10 3.07 X 10ll 3.37 X 10 ll 1.34 X 10-10 1.47 X 10-10 l
TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE d(E > 1.0MeV) In/cm2 -sec)
. 0 . 0 '. M M M M DESIGN BASIS 1.87 X 1010 2.78 X 10 10 3.15 X 1010 2.56 X 1010 2.94 X 1010 10 CYCLE 1 1.40 X 1010 2.06 X 10 10 2.36 X 10 10 1.92 X 10 10 2.19 X 10 10 CYCLE 2 1.35 X 10 10 1.97 X 10 10 2.13 X 1010 1.59 X 1010 1,72 X 10 CYCLE 3 1.32 X 1010 1.92 X 10 10 2.08 X 1010 1.98 X 10 10 2.52 X 10 10 CYCLE 4 1.14 X 10 10 1.56 X 1010 1.77 X 1010 1.54 X 1010 1.78 X 1010 (ILy.J.lMeV) In/cm2 _ igd
. 0.0* M 21d' 15 0* 45.0*
DESIGN 8 ASIS 3.87 X 1010 5.85 X 1010 8.59 X 10 10 7.27 X 10 10 7.36 X 10 10 , CYCLE 1 2.90 X 1010 4.34 X 1010 6.45 X 1010 5.45 X 10 10 5.49 X 1010 CYCLE 2 2.81 X 1010 4.16 X 1010 5.81 X 1010 4.51 X 1010 4.32 X 10 10 ' CYCLE 3 2.74 X 1010 4.04 X 1010 5.69 X 1010 5.63 X 1010 6.30 X 10 10 CYCLE 4 2.38 X 1010 3.28 X 1010 4.84 X 1010 4.38 X 10 10 4.46 X 10 10 Iron Atom Disniacement Rate Idoa/sec1 0.0* 15,0* 21&* 35.0* 45.0* DESIGN BASIS 2.90 X 10-11 4.31 X.10-II 5.27 X 10-II 4.35 X 10-11 4.68 X 10-11 .
~
CYCLE 1 2.17 X 10-11 3.19 X 10-11 3.95 X 10-11 3.25 X 10-11 3.49 X 10-Il CYCLE 2 2.11.X 10-11 3.06 X 10-11 3.56 X 10-Il 2.69 X 10-11 2.75 X 10-Il CYCLE 3 2.05 X 10-11 2.97 X 10-II 3.49 X 10-Il 3.35 X 10-Il 4.01 X 10-Il , CYCLE 4 1.78 X 10-11 2.41 X 10-II 2.97 X 10-11 2.61 X 10-Il 2.84 X 10-11 . 6-15 ,
TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL Radius _ (cm). _ O' 15' 25' 35' 45' . 220.27(l) 1.00 l'.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 0.375 229.60 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 I
L 6-16
- f TABLE 6-4
?.CLATIVE RADIAL DISTRIBUTIONS Of NEUTRON FLUX (E > 0.1 MeV)
W11HIN THE PRESSURE VESSEL WALL Radius Lqal, 0' 15' L ., ' ,,_Ji'_ ,_,15 ' 220.27(I) 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 9.567 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 E36.22 0.436 0.428 0.440 0.416 0.392 237.54 0.386 0.380 0.392 0.369 0.343 236.86 0.337 0.333 0.344 0.324 0.29: 240.19 0.289 0.287 0.298 0.279 0.248 241.5i 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
TABLJ 6-5 l RELATIVE RADIAL DISTRIBUTIONS Of IRON DISPLACEMENT RATE (dpa) I W11HIN THE PRESSURE VESSEL WALL ; Radius
'\ ;
.(raL O' __.n *__. 25' ,__n ' 45' -l !
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 0.033 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 0.533 0.506 ; i 229.60 0.439 0.435 0.46% 0.475 0.447 ! 230.92 0.387 C.383 0.410 0.423 0.394 232.25 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 0.169 240.19 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 6-18
TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Weight Response Product Yield Material Interest Fraction Rance Hal f-Life (%1 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 l Nickel NiS8(n,p)CoS8 0.6830 E > 1.0 MeV 70.90 days 4 p, Uranium-238* U238(n.f)CsI37 1.0 E > 0.4 MeV 30.12 yrs 5.99
.e Neptunium-237* Np237(n,f)Cs137 1.0 E > 0.08 MeV 20.12 yrs 6.50 Cobal t-Aluminum
- CoS9(n,3)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs
. Cobalt-Aluminum CoS9(n,d)Co60 0.0015 E > 0.015 MeV 5.272 yrs
- Denotes that monitor is cadmium shielded.
I
TABLE 6-7 ; HONTHLY THERMAL GENERATION DURING THE FIRST FOUR IVEL CYCLES : Of THE CALLAWAY UNIT 1 REACTOR I THERMAL THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION GENERATION
- M0111] (MW-hr) M0 Hill (MW-hrt. MQL(Ilj _ (MW-hr) MQRIll (MW-hr) I 10/84 73090 4/86 558152 10/87 0 .4/89 0 j 11/84 878649 5/86 1628231 11/87 776234 5/89 459251 ,
1001525 ' 12/84 6/86 1891846 12/87 2498378 6/89 2230340 1/85 1825824 7/86 2279207 1/88 2400761 7/89 2646374 -l 2/85 1997026 8/86 2236572 2/88 1944633 8/89 2567249 : 3/85 2093839 9/86 2395748 3/88 2534237 9/89 2346488 l 4/85 1476721 10/86 2478466 4/88 140149S 10/89 2607932 ! 5/85 2351261 11/06 2000521 5/88 2444952 11/89 2529924 . 6/85 1990891 12/86 2476793 6/38 2547644 12/89 2644439 , 7/85 2204437 1/87 2310037 7/88 2624336 1/90 2595499 . 8/85 2226770 2/87 2278365 8/88 2590033 2/90 2367162 . ; 9/85 2399444 3/87 2310093 9/88 2295718 3/90 2604203 10/85 2206396 4/87 99068- 10/88 2526306 4/90 2557687 < 11/85 1940798 5/87 1527462 11/88 2355533 5/90 2394320 i 12/85 2212562 6/87 2442877 12/88 1913183 6/90 2193115 j 1/86 2445738 7/87 2426722 1/89 2640944 7/90 2587861 j 2/86 2161005 8/87 2466901 2/89 2391524 8/90 2614245 3/86 0 9/87 673770 3/89 2483347 9/90 1500131 I 4
)
6-20 . L i l
. _ _ _ .. -. . - _ . - _ - _ . . - . - . . . ~ . , ~ . . . . - - - . _ _ -.. ..
r TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REAC110N RATES Measured Saturated Reaction Monitor and Activity Activity Rate I Axial location (dis /see-ami (dis /sec-am) (RPS/ NUCLEUS) Cu-63 (n a) Co-60 ; i Top 1.50 x 10 5 3.66 x 105 Middle 1.34 x 10 5 3.27 x 105 Bottom 1.31 x 10 5 3.19 x 105 , Average 1.38 x 10 5 3.37 x 10 5 5.14 x 10-17 l Fe-54(n.p) Mn-54 i i 6 Top 2.12 x 10 6 3.20 x 10 Middle 1.90 x 106 2.86 x 10 6 j
., Bottom 1.88 x 10 6 2.83 x 106 Average 1.97 x 10 6 2.96 x 10 6 4.72 x 10-15 l l
Ni-58 (n.p) Co-58 i Top 1.89 x 107 5.16 x 10 7 ; Middle 1.73 x 10 7 4.73 x 107 Bottom 1.70 x 10 7 4.64 x 10 7 '! Average 1.77 x 10 7 4.84 x 107 6.91 x 10-15 U-238 (n,f) Cs-137 (Cd) I Middle 4.52 x 10 5 4.59 x 10 6 3.03 x 10-14 f f 6-21
~_ _
TABLE 6-8 NEASURED SENSOR ACTIVITIES AND REACT 10*l RATES - cont'd Heasured Saturated Reaction - Monitor and Activity Activity Rate Axial location (dis /sec-om) (dis /sec-om) 1RPS/ NUCLEUS) Np-237(n.f) Cs-137 (Cd) Middle 4.07 x 106 4.14 x 107 2.51 x 10-13 Co-59 (n.a) Co-60 Top 2.53 x 107 6.16 x 107 Hiddle 2.45 x 107 5.97 x 107 . Bottom 2.54 x 107 6.19 x 107 Average 2.51 x 107 6.11 x 107 3.99 x 10-12 , Co-59 (n a) Co-60 (Cd) . Top 1,32 x 107 3.22 x 107 Middle 1.30 x 107 3.17 x 107 Bottom 1.34 x 107 3.27 x 107 Average 1.32 x 107 3.22 x 107 2.10 x 10-12
=
l- 6-22 l u l
TABLE 6-9
SUMMARY
OF NEUTRON DOSIMETRY RESULTS TIME AVERAGED EXPOSURE RATES 2 4 (E > 1.0 MeV) (n/cm -sec) 8.89 x 1010 8% 2 3.76 x 1011 4 (E > 0.1 HeV) {n/cm -sec) 1 15% dpa/sec 1.67 x 10-10 3 33g 2 7.90 x 1010 ( (E < 0.414 eV) {n/cm -sec) 1 22% INTEGRATED CAPSULE EXPOSURE 2 f (E > 1.0 MeV) (n/cm ) 1.29 x 1019 1 8% 2 4 (E > 0.1 HeV) {n/cm ) 5.46 x 10 20 i 15%
. dpa 2.42 x 10-2 i 11%
2 1.15 x 1019
+ (E < 0.414 eV) (n/cm ) 1 22%
NOTE: Total Irradiation Time - 4.60 EfPY 4 l f 6-23 f l
TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER - Adjusted Reaction tietsured Calculation (13 Cu-63 (n,a) Co-60 5.14x10*I7 5.19x10~17 1.01 Fe-54 (n.p) Mn-54 4.72x10-15 4.81x10-15 1.02 Ni-58 (n.p) Co-58 6.91x10-15 6.80x10-15 0.98 U-238 (n.f) Cs-137 (Cd) 3.03x10'I4 2.7613-I4 0.92 Np-237 (n.f) Cs-137 (Cd) 2.51x10-33 2.65x10*I3 1.05 Co-59 (n.a) Co-60 (Cd) 3.99x10*I2 3.96x10-12 0.99 Co-59 (n,0) Co-60 2.10x10-12 2.11x10-12 3,og 4 e 6-24
i i TABLE 6-11 ADJUSTED NEV1RON ENERGY SPECTRUM AT ) THE SURVEILLANCE CAPSULE CENTER [ Energy Energy Adjusgedflux I AdjusjedFlux '
- Group (Mev) (n/cm-sec) Group (Mev) (n/cm -sec)
I 1.73x101 7.27x106 28 9.12x10~3 1.72x10 10 ( 2 1.49x10I 1.65x107 29 5.53x10~3 2.22x1010 l 7 6.94x109 I 3 1.35x101 6.37x10 30 3.36x10~3 4 1.16x10I 1.42x10 8 31 2.84x10-3 6.62x109 f 8 2 2.0x10~3 6.35x109 5 1.00x101 3.13x10 32 l 0 8 2.04x10~3 1.78x1010 } 6 8.61x10 5.32x10 33 7 7.41x10 0 1.21x109 34 1.23x10~3 1.62x1010 l 10 8 6.07x10 0 1.72x10 9 35 7.49x10'4 1.49x10 l 0 3.59x109 36 l.41x1010 9 4.97x10 4. 54 x 10' 10 3.68x10 0 4.74x10 9 37 2.75x10~4 1.50x1010 l 11 2.87x10 0 9.94x109 38 1.67x10~4 1.55x1010 t . 12 2.23x10 0 1.37x1010 39 1.0lx10~4 1.62x1010 13 1.74x10 0 j,93xio10 40 6.14x10-5 1.61x1010 j
. 14 1.35x100 2.09x1010 41 3.73x104 % 10I9 15 1.11x10 0 3.79x1010 42 2.26x10-5 1.5bx1010 !
10 j 16 8.21x10~I 4.29x10 10 43 1.37x10-5 1.52x10 17- 6.39x10-I 4.42x1010 44 8.32x10-6 1.46x1010 18 4.98x10*I 3.19x10 10 45 5.04x10-6 1.35x1010 19 3.88x10"I 4.46x1010 46 3.06x10-6 1.27x1010 l i 20 3.02x10-1 4.58x1010 47 1.86x10-6 1.17x1010 21 .l.83x10~l- 4.53x1010 48 1.13x10-6 8.68x10 9 , i 22 1.llx10-1 3.63x1010 49 6.83x10~7 1.09x1010 23 6.74x10-2 2.53x1010 50 4.14x10-7 1.44x1010 j 24 4.09x10-2 g,44xiol0 51 2.51x10~7 1.41x1010 } 25 2.55x10-2 1.89x1010 52 -1.52x10-7 1.32x1010 l 26 1.99x10-2 9.35x10 9 53 9.24x10-8 3.74x10 10 27 l',50x10-2 3,igxiol0 ; 1 NOTE; Tabulated energy levels represent the upper energy of each group. 6-25 ;
1ABLE 6-12 l COMPARISON Of CALCULAlED AND MEASURED EXPOSURE LEVELS FOR CAPSULE Y - [1]sylaigd Measured CM 2 f(E > 1.0 MeV) {n/cm ) 1.13 x 1019 1.29 x 1019 0.88 2 f(E > 0.1 MeV) {n/cm ) 5.08 x 1019 5.46 x 1019 0.93 dpa 2.21 x 10-2 2.42 x 10-2 o,93 2 f(E < 0.414 eV) (n/cm ) 5.39 x 1018 1.15 x 1019 0.47 6-26
I TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS
- ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE
, 4.60 EFPY i 0* 15* 25* 35' 45* t (E > 1.0 Mev) 2.16 X 10 18 3.11 X 10 I8 3.44 X 10 18 2.91 X 10 18 3.42 X 10 18 [n/cm2]
+ (E > 0.1 MeV) 4.14 X 10 18 6.03 X 10 18 8.69 X 10 18 7.66 X 10 18 7.90 X 10 18
, [n/cm2] Iron Atom Displacements 3.18 X 10-3 4.55 X 10-3 5.47 X 10-3 4.68 X 10-3 5.16 X 10-3 [dpa] 17.0 EFPY 0* 15* 25* 35' 45*
+ (E > 1.0 Mev) 7.99 X 10 18 1.15 X 10 l9 1.27 X 10 l9 1.07 X 10 I9 1.27 X 10 I9 l [n/cm2]
[ + (E > 0.1 MeV) 1.53 X 10 19 2.23 X 10 19 3.21 X 10 19 2.83 X 10 l9 2.92 X 10 19 [n/cm2] Iron Atom Displacements 1.17 X 10-2 1.68 X 10-2 2.02 X 10-2 1.73 X 10-2 1.91 X 10-2
, [dpa]
32.0 EFPY f 0* 15* 25* 35* 45*
+ (E > 1.0 Mev) 1.50 X 10 19 2.16 X 10 19 2.39 x 10 I9 2.02 X 10 19 2.38 X 10 19
- [n/cm2]
4 + (E > 0.1 MeV) 2.88 X 10 19 4.19 X 10 19 6.04 X 10 l9 5.33 X 10 19 5.49 X 10 19 [n/cm2] Iron Atom Displacements 2.21 X 10-2 3.16.X 10-2 3.81 X 10-2 3.25 X 10-2 3.59 X 10-2 f- [dpa]
, - . . , , . . , , . . - _ _ , . _ - , _ . , _ _ . _ , _ _ . . _ . . - . . _ . . . . _ . _ . . , _ . . . , - - - - - . . , _ _ , - . _ . , _ _ ~ , , _ . . _ , _ _ , , _ _ _ _ . _ _ _ _ _ . ~ -.
TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES 17 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 (n/cm ) (equivalent n/cm2) Surface 1/4 T 3/4 T Surface 1/.4_I 3/4 T 0* 7.99 x 10 18 4.34.x 10 18 9.27 x 1017 7.99 x 10 18 5.04 x 10 18 1.75 x 1018 15" 1.15 x 10 I9 6.22 x 10 18 1.31 x 10 18 1.15 x 10 19 7.20 x 10 18 2.49 x 10 18 25*(a) 1.27 X 10 I9 6.94 X 10 I8 1.50 X 10 18 1.27 X 10 19 8.25 X 10I8 3.03 X 10 I8 35* 1.07 x 10 I9 5.84 x 10 18 1.25 x 10 18 1.07 x 10 I9 7.11 x 10 18 2.66 x 10 18 45' l.27 x 10 I9 6 77 x 10 18 1.37 x 10 18 1.27 x 10 I9 8.10 x 10 I8 2.77 x 10I8
? 32 TFPY E NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dea SLOPE 2 (equivalent n/cm2)
(n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 4 0* 1.50 x 10 I9 8.17 x 10 18 1.75 x 10 18 1.50 x 10 19 9.49 x 10 I8 3.29 x 10 I8 15* 2.16'x 10I9 1.17 x 10 I9 2.46 x 10 18 2.16 x 10 I9 1.36 x 10 I9 4.69 x 10 I8 25*(a) 2.39 X 10 I9 1.31 X 10 I9 2.82 X 10 18 2.39 X 10 I9 1.55 X 10 I9 5.70 X 10 18 35* 2.02 x 10 I9 1.10 x 10 I9 2.34 x 10 18 2.02 x 10 I9 1.34 x 10 I9 5.01 x 10 18 i 45* 2.38 x 10 I9 1.27 x 10 I9 2.57 x 10 I8 2.38 x 10 I9 1.52 x 10 I9 5.22 x 10 18 (a) Maximum point on the pressure vessel
1ABLE 6 15 UPDATED LEAD FACTORS FOR CALLAWAY Ullii i SURVEltLAtiCC CAPSUlls OR1912 LtiLDfdOI U 3.85(*) (withdrawn) Y 3.77(b) (withdrawn) V 3.77(U) W 4.10(CI X 4.10(c) Z 4.10(C) (a) Plant specific evaluation based on operation at 3411 Mwt for 1.05 EfPY * (b) Plant specific evaluation based on end of cycle 4 calculated fluence. (c) Projection based on average flux through cycle 4. 9 e 4 6-29 I
SECTION 7.0 SURVEILLAllCE CAPSULE R[f40 VAL SCliEDULE 4 The following removal schedule meets ASTM E185-82 and is reconmended for future capsules to be removed from the Callaway Unit I reactor vessel:
- Capsule Estimated i Location Lead fluence Capsule (deg.) Factor Removal Time (b) (n/cm )
2 V 58.5 3.85 1.05 (Removed)(a) 3.27x1018(Actual) , Y 241.0 3.77 4.6 (Removed)(a) 1.29x1019(Actual)(c) V 61.0 3.77 9.0 2.5 x10I9 (d) X 238.5 4.1 15.0 4.6 x1019 W 121.5 4.1 Standby --- Z 301.5 4.1 Standby --- (a) Plant Specific Evaluation T (b) Effective full Power Years (EFPY) from plant startup. (c) Approximate fluence at 1/4 thickness of reactor vessel wall at end of life (32 EFPY). (d) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY). f t , 7-1
SECTION 8.0 l REFERENCES !
- 1. L.R. Singer, et. al., " Union Electric company Callaway Unit No. 1, Reactor [
Vessel Radiation Survol11ance Program," WCAP-9842, May 1981. {
- 2. S.E. Yanichko, et. al., " Analysis of Capsule U from the Union Electric f Company Callaway Unit 1 Rtactor Vessel Radiation Surveillance Program," f WCAP-ll374, November 1986.
j
- 3. Code of Federal Regulations,10CFR50, Appendix G, "Fractt+e Toughness ;
Requirements", and Appendix H. " Reactor Vessel Material Surveillance ! Program Requirements " U.S. Nuclear Regulatory Commission, Washington, f D.C. l
- 4. Regulatory Guide 1.99, Revision 2 " Radiation Embrittlement of Reactor !
Vessel Materials", U.S. Nuclear Regulatory Commission, May, 1988. {
- 5. Section 111 of the ASME Boiler and Pressure Vessel Code, Appendix G,
" Protection Against Nonductile failure." i i . i
- 6. AS1H E208, " Standard Test Method for Conducting Drop-Weight Test to l Determine Nil-Ductility Transition Temperature of Ferr"ic Steels."
[
- 7. ASTM E185-82, "".tandard Practice for Conducting Surveillance Tests for h Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF).'
- 3. 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." ! I 8-1 ! i
I
- 10. ASTM E8-89b, " Standard Test Methods of Tension Testing of Metallic ,
Materials." i
- 11. AS1H 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. Soltest, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data 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 Coup'ad Self-Shielded, 47 Neutron, 20 Gamma-R3y, P3, Cross Section Library for Light Water Reactors". !
i
- 15. P.C. Cook, et. al., "The Nuclear Design and Core Physics Characteristics ,
of the Callaway Unit 1 Nuclear Powtr Plant - Cycle 1", WCAP-10249, Revision 1, October 1984. (Proprietary) ,
- 16. K.P. Mock, et. al., "The Nuclear design of the Callaway Unit 1 Nuclear Power Plant - Cycle 2", WCAP-lll20 Hay 1986 (Proprietary)
- 17. J.R. Secker, et. al., "The Nuclear design of the Callaway Unit 1 Nuclear Power Plant - Cycle 3", WCAP-ll552, August 1987 (Proprietary)
- 18. J.R. Socker, et. al., "The Nucioar design of the Callaway Unit 1 Nuclear Power Plant - Cycle 4", WCAP-12134, May 1989 (Proprietary)
- 19. ASTM Designation E482-82, " Standard Guide for Application of Neutron
- Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
B-2
. = - _ - . _ - . _ _ - _ _ - - . _- - - . - -
- 20. ASTM Designation ES60-77, " Standard Reconaended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 21. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron j Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12 American Society for Testing and Materials, l Philadelphia, PA, 1984.
- 22. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor !
Pressure Vessel-Surveillance Standard", in ASTM Standards Section 12 American Society for Testing and Materials, Philadelphia, PA,1984.
- 23. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM i Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
s
- 24. ASTM Deatgration E261-77, " Standard Method for Determining Neutron Flux, Fluenet, and Spectra by Radioactivation Techniques", in ASTM Standards,
. Section 22 American Society for Testing and Materials, Philadelphia, PA, ; 1984. t
- 25. AS1H Designation E262-77, " Standard Method for Measuring Thermal Neutron !
Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984. ;
- 26. ASTM Designation EP.53-82, " Standard Method for Determining fast-Noutron !
Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA,1984.
- 21. ASTM Designation E264-82, " Standard Method for Determining fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984, i 8-3 t
. . . _ _ . , . _ _ , _ _ _., . .. ,-r ,. .,r ._. __ ,
- 28. ASTM Designation E481-78, ' Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12. American Society for Testing and Haterials, Philadelphia, PA, 1984. -
- 29. ASTM Designation 2523-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in AS1H Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 30. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American I Society for Testing and Materials, Philadelphia, PA, 1984.
- 31. ASTM Designation E705-79, " Standard Method for Measuring fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
I
- 32. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitirs for Reactor Vessel Surveillance", in ASTM ,
Standards, Section 12, American Society for Testing and Materials, t Philadelphia, PA, 1984. .
- 33. f. A. Scl.mittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
! 34. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spectta_ Determined by Foil Activation, AFWL-TR-7-41, Vol. 1-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967, 1
- 35. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Aoplications", R. E. Maerker, et al., 1981.
4 8-4
APPENDIX A Load-Time Records for Charpy Specimen Tests
~
h 9 A-0
1 I
=
V/g = -
./p
- Pg = M A xtMUM LO AD P, = F R ACTU A E LO A D e
Pgy GENERAL g YlELO LOAD I . t
- a I O
a 1 i
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l Te M E Figure A-1. Idealized Load-Time Record for an Instranted Charpy Impact Test i f I i
, e '
4
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=s .e *.
(J.9999 i .........J.m. a.h_safumnaherr un.s ~e.e~ -- t E+00 0.0000E+00 3.0000E+02 1.6000E+03 ' 2.4000E+03 , i l i Time, microeeconds b t' l Figure A-3. Charpy Impact Test load-Time Record for Specimen CL68. ; i
, i 1 !
i I i t 8 e p k
. - - _ - - - .- _ _ _ _ . _ ________.._______.___..__.___._._._..__._.m_ _ . _
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"~~~~
l 0.0000"I.'iOO0E+00"~'~~~i5563f+02 E+00 . E.~id50E+03 '2 400iE+03 a Time, microseconds i, i Figure A-4. Charpy Impact Test Load-Time Record for Specimen CL62. 4
- l .
4
--# .o.,,--,,w-.,,y..,,y7,,.,,,..y. ..v-,,,-.-.-,~,.,w.._,7m ,._,,,,,.7,,..,.-,,r
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,s.U,v0V ............... .................................... ............................ ... ........ ....... ........................
t+0- ,3 3.7500 E+03
,-?
e C O ' y a E. ' e - 2.5000 7 0 E+03 2 1.2500 . E+03 ,
. 7 , _ ,
00000j............:....~.u..i.%.-l.-lh.h.:w';u.-a &a~~.ex , .t,w .uw.:r . l E+00 0.0000E400 3.0000E+02. 1.6000E+03 2.4000E+03 l l Time, microseconds 1 t Figure A-5. Charpy Impact Test Load-Time Record for Specimen CL72. l
if t a f t 5.00,9,9.....- . E+03
~
, .^ l 3.7500 g'2 '
E+03 ~./ .
. C .
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- $ . I Time, microseconds i
Figure A-6. Charpy Impact Test load-Time Record for Specimen CL65. [ I )
- - - - , _ - . , _ _ , _ , . - - . . - . - . - . _ . , ~ , _ , - - , _ , - - , . . . , _ . . , . . _ . . . , , , . . , - . . , , . _ . . -,.-- .....-_ .__ ._, - ,.,..,.. . _. ._.,._ -..,- .-. _ - -.-,_.. ...-- - .- ~ , _... --
i
'-~~~~~~~~~~"~~ ~~' ~~~~~ - '~~' ~~~~ ~ ~ ~ ~ ~ ' ~ " " " " ~ ' ~
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. i 4
E= 49.0 ft-lbs 1.2500 - 3 E+0,a . 4 1 - 1 *. f h~ 0.0000 5 ' :.\E=f'. v -. m. p i.
~
E+00 "'E3550E+00 ~~~E0000E+02'W
- m 2.4000E+03 1.60YdE r[%^-#'
W. i i l Time, microseconds i l Figure A-7. Charpy Impact Test load-Time Record for Specimen CL61. i _ _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _ _ . _ _ . _ _ _ _ _ _ _ . - , - - , - . .. - _ - . . ~- .- . . . . , . ..-, , .
I i 1 1 1 t t t v.? i s. . j . ..
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-- n ? 1. t n. n. nE+ 0 3
- 0.4000E+y^'a
) i Time, sicroseconds i 1 1 i i Figure A-8. Charpy Impact Test load-Tir.s iacord for Specimen CL64. i {
i Y i I 1 ' lt Y .h .
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- E+00 ' O.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03
! Ti e, microseconds i Figure A-9. Charpy Irpact Test Load-Time Record for Specimen CL66. i i i t-
< - -e .- - , , . - , , , - . - , . . , - - . . - - . - .
t E+03 3.7500 E+03
.- C
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0.0000 - 1 E+00 0.0000E+00 8.0000E+02 1.6000E+03 2.4000E+03 Time, microseconc; Figure A-10. Charpy Impact Test Load-Time Record for Specimen CL74.
. . a.. . - -.~ _. ~ . . _ . . . _ . . . _ . . . - . .u . _. _. . . . . . _ . .
8 E+03
,-w,s_
d./ ,500 - E+03
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. r i l M.fw"-%-
0.0000;i~.6666353~~~~[".~66665562 E+00 B "~~~~i.'UUUUEidi"~~~~2.4000E+03 Time, microseconds Figure A-11. Charpy Impact Test Load-Time Record for Specimen CL71. t
.i.n.vv E+03 :
i 3.7500 i E+03 . , . . . .
- A. ; ~- *
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..... . . ..... .- i . :-1: :.c n E+00 0.0000E+00 8.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds ,
s a Figure A-12. Charpy Impact Test load-Time Record for Specimen CL67. __ ._ . .. ._. . . . . _ _ , . - - . . . . . . _ . - . . . . _ . _ - . . - - _ . _ - - .~ . ._.
l t i O J G S) i t
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' O.0000E+00 8.0000EiE5'~~~~~I.'UUUU5'+'03 53EEE5id5~~~~
h+hh Time, microseconds Figure A-14. Charpy Impact Test Lead-Time Record for Specimen CL73.
i i i 5.0000 l E+03 3.7500 E+03 ..
, .J-Qvmh
%., C e
o v .
. T.~ L
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.L 2.5000 ','
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1.2500 . v.% ! E+03 wag N. - O.OfJ00- ............. ........ ..........
._. .............. ....... . ............ ....=-
E+00 0.0000E+00 1.2050E+03 2.4100E+03 3.6150E+03 Time, microseconds Figure A-15. Charpy Impact Test Load-Time Record for Specimen CL70. 8 # 0 . . .
i t h . M' '. . E+03
, ,/
- e. gsv ,,u E+03 4
r-
'-m J r e r.- .
s
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- s. 6
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+.
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- o. n ,.
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t n fiQfiQ t k+bb ' O.0000E+00 3,0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-16. Charpy Impact Test Load-Time Record for Specimen CL63 ;
l E+03 3.7500 E+03 v; C e . O ! u 5 - 2.5000 6 e E+03 3 1.2500 ~ E+03 . 1 0*000 p.........Ossda.JA .s2.Laa ? '4 -
. --%_- *- - ' m.e n.,ts.n.n =~ ucec:o E+00 ' O.0000E400- 8.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-17. Charpy Impact Test Load-Time Record for Specimen CT63
103 3.7500 F+0.3 C T e 6 o 2.5000
?
E - t+0,a i m ba C* 1.2500 E+03 O.0000 ' n " : N. A A_m. a;3 rue. ss.L ~ a, ---- 2,4000E+03 E+00 0.0000E+00 3.0000E+02 1.6000E+03 Time, microseconds i Figure A-18. Charpy Impact Test Load-Time Record for Specieren CT61
5.UVVV
- r. ,
t.+V3 1 75A.A iJv E+03 I C e o p I
>8 -
n.
.,.e E s At f .i i
l W b& b*VVV D s* e E+0,a 4 : l i 1.2500 , l E+03
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r.
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E+bb 'O.'U5UdEIUU~ ~ ~E.'EEEEEiEE ~~ ~""dEUUEI55~
- 1. "E.IUUEIU5 Time, microseconds Figure A-19. Charpy Impact Test Load-Time Record for Specimen CT64 l
l
.=.....
t+03
"'S T . .
E407 - s""
.a y e T o ? 8. 2.5000 -
6 3 Lo- E+03 2 1.2500 E+03 . i i V- w'i+-~ew.%w,.,. c, m.e_e_ 0.0000 . _w_e.-e: =,, cm ; E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds i Figure A-20. Charpy Impact Test Load-Time Record for Specimen CT62 r c
? [ 1
. . . . . . . . . . . . . . . . . = =.
I 5. C)v,. ,.v0, E403
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O.0000 ,.... =
*-?'N 'Y/'WR^vwveMa~~c E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 :
i i i Time, microseconds r Figure A-21. Charpy Impact Test Load-Time Record for Specimen CT68 O . O g I
4 L I L t ij j
.i. .
- .m00
- _. m ,
./
C e T o y . 8. 2.5000 6
~
5
% Le E+03 .
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E+03 . .'-. . s, /
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~.~..~."..'.^...'....._
U.0000 i -
... - =.....-- g. .
E+00 0.0000E+00 8.0000E+02- 1.6000E+03 2.4000E+03 Time, microseconds i. Figure A-22. Charpy Impact Test Load-Time Record for Specimen CT65 i I
, . . . . . , , , - . , . . . , . . ~ , _ , . --_...- ,,. _..... . _ ... m ,. . _ . _ ~ _ - . . . ..-.-._,- . - . - , , . . . _ _ . - . , .
E+03 3.7500 E+03 *
. ,:!' . . -- l c
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. ' ^! \~"^^-'".. ...N. w.'1 &. .fw- m....
0.00.u^0 3.0000E+02 1.6000E+03 2.4000E+03 E+00 0.0000E+00 Time, microseconds Figure A-23. Charpy Impact Test I.oad-Time Record for Specimen CT67
. e
5 . 0 0 0 0 ............................................................................ ......... r,.0,s c l.F :. 3.7500 - c+
- m. rs,o ,
4g t C e - T O . .
? K. 2.5000 : 7 m . ,
- E+03 2
~ ; j.
e . A 1.2500 t .-! ~ ,
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-:. n ,.
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U.,* c." R.R , h =xm h()0.0 . \. -=..... . E+00 '0.0000E+00 B.0000E+02 i 6000E+03
. 2.4000E+03 Time, microseconds Figure A-24. Charpy Impact Test Load-Time Record for Specimen CT72
.~: - . . - . . . ,,e ., ,...,,,.-,,.,,'m., ,.n .,.. . - . , y e ic-...,-.i.- e-.. , , . + ,,,,-y-, , ,,--, --,g., ,,.n..,n- ,c-.- e.r . ,-,, 9 . , , ,w. ., ,.%,,e ,w,-- # ..,~,3- , ,m-,- , - vv.,, .-
a 0.00,VU e
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- - - u ---- --'-- " --- "
0, . U O U rJ .-...-..... .------..--------;.....-------}- ---.----------------
}......... 1.6000E+03 2.4000E+03 E+00 0.0000E+00 3.0000E+02 Time, microseconds Figure A-25. Charpy Impact Test Load-Time Record for Specimen CT66
3 P 1 ( !
.s .41.1111, ,. ,.,l, P
c_+n;
-s v
,-r 3.
v !DO(J . E+03 -^~,,
- 1 i
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- 0. 0 0J0 n. ......... = -y -^ ~ ~ - N ~-- - - - - - - -
-l-' .-
,0,0000E+00 B.0000E+02 1.6000E+03 2.4000E+03 E+00 Time, microseconds Figure A-26. Charpy Impact Test Load-Time Record for Specimen CT70
. . _ _ _ . ~ . _ _ . . . . - _ . _ . . _ . . . . - - . - . - - - . _ . . - .
i h 0.0000 E+03 3.7500 E+01 ' /-,.% s s - C l - e .e T o 7 T
~ .E. 2.5000 .
f' 1 4
;;: E+03 a\
~
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1.2500 s E+03 _- t
+
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.~.v 0.0000 ...........................l..........._ -
w ~ ~ __ ,.4 v v e 1 c, B.0000E+02 1.6000E+03 2.4000E+03 l E+00 0.0000E+00 Time, microseconds Figure A-27. Charpy Impact Test Load-Time Record for Specimen CT74 I i 1 . . . . + .
. ........................................................ ...=
- c. 00 E+03 3.7500 JP-r -
E+03 ,r e
- s C'
e s ." o ..- - T
> a .
~
7' -
$ ; 2.5000 : ,..
e 3 E+0a - i s i 's i .. i.2500 ' E+03 - i i
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/ "
- - ~~
o 0000 ~ ~ '~
+0b~ ' O.0000E+00 B.0000E+02 1.6000E+63 ~Y.II00E+I3
' Time, microseconds Figure A-28. Charpy Impact Test Load-Time Record for Specimen CT73 ( l L
~ ~"~~~~~"" ~~
~~"""~"~~~~~~~~~"~~~~~~~~~~
5.0000 E+03 3.7500 E+03 ' s
.- s:E5 7'. .
/ -; .- C e : .
8 o s
.-A
- ' t 7-y 5- 2.5000 ..;
~~ ~
^ 5 3 -
e E+03 _ N,
. N 1.c,.o 00 % t E+03
- N s
'~~
w _s
~N 0.0000 fi................._............
E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 i Time, microseconds Figure A-29. Charpy Impact Test Load-Time Record for Specimen CT75 t m , , m _ _ _ _ _
E+03 i.'/5nn a av E+03 _ p, l .f f /. *;
.?.
e
.,e
/: ..
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?.5000 :.:
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r s. 1.2500 ' r. E+03 . i .~ e .
- -. - [%, u a s
+s .ar. i.
t) (.usn o ~ E+bbO.0000E+00 B.0000E+02 ~~I.iOdOE~+~63 l34000iiT~~~ Time, microseconds , Figure A-30. Charpy Impact Test Load-Time Record for Specimen CT71 _ _ - . _ . - _ - _ . - , _ . . . - . . - _. --_-~ _ , . . , _ - . . . - . _ . . . . . . . . . . , _ _ . . . , _ _ , . - . _ - , - - . , _ - - . _ . - - - - . _ - , . . . . , _ - - . . - . , . - . _ _ , . . -
-~~~~~ ~ ~ ~ ~ ' ~ ~ " " " " ~ ~ ~ ~ ~~~~~~~~~~~~~~~~~~~~
5.0000 E+03 3.7500 E+03 . .- 1. ~
,'s .',1,:; ~.
. * .: y ...
3 - . . C e .-
- o. .:
T 6
?
f 2.5000 N
-' 4- - ' 9 E+03 N .
m; %
. M.
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v n . r.,. A
%A - :'v i
0.0000 ..... ..... ....................--
:.e<. .
E+00 0.0000E+00 3.0000E+62 1.6000E+03 2.4000E+03 - Time, microseconds Figure A-31. Charpy Impact Test t.oad-Time Record for Specimen CT69
. O h
, . . , , , . . , , , . , , , - .,a . , - - , , - - - , - . , m.-.,..y.., ... y ,. ,- .,,n-, , . . , . , , , , , , - , - - - , , .-,...,.-,_,-,.-,-,,,,-..---y .
,-,n, ,,,,-,v-- -
-a,-,,,,.,v .n.,,-~~,-., .
**e.eeeeemeeeee.eesee..e.....................................e..................................
E+03 3.7500 E+03 Rm PI. Pg3 C e N o y E- 2.5000 7 M - E+03 ': 2 e i.2500 - E+03 .-' E= E.0 ft-lbs 0.0000 ,.cEn.kaitaw'~i.wahuma.o: yeanr:rnuw=-- E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-32. Charpy Impact Test Load-Time Record for Specimen CW72
._. _ . _ - . . .. _ , - _ _ _e . . --_-_e. . - _ . . . _ . . - _- .. . , . . _ _ . _ _ . _ . . , . . _ . . _ .-
- 0. 0000 E+03 3.7500 E+03 -.
C o,
>
- 6 a, -
2.5000 . 1 e E+03
. 1 1.2500 .-'
E+03
? ; *. - : ,, . ~ . >
0.0000 !
- l : .- !~1 <.c s.c w z uw-w_.a_-_ _ .= . _ _ _ = _ .
' O.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 E+00 Time, microseconds figure A-33. Charpy Impact Test Load-Time Record for Specimen CW61 f
L_______
E+03 nr 3.7500 j' E+03 sj.?" C g 5 e . L, P 2.5000 - 6
- 5
;;, E+03 .
1.2500 ~ E+03 -
+.
, i j: .-
0.0000)..........................h.C.[2AiC2 E+00 ~ 0.0000E+00 B.0000E+02 1 6000E+03 e wh. a e _ & a w s 2.4000E+03 Time, microseconds l'igure A-34 Charpy Impact Test Load-Time Record for Specimen CW65
~
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Time, microseconds i Figure A-35. Charpy Impact Test load-Time Record for Specimen CW69 i
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1 l Figure A-38. Ch ,rpy Irnpact Test load-Time Record for Specien CW62 l i l
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; Figure A-41. Charpy Impact Test Load-Time Record for Specimen CW68 -
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4 APPEt4 DIX B Photographs of specimens Before Testing 4
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- 12.7 mm figure B-1. Charpy Impact Specimens CL75, CL68, CL62, CL72 and CL65 from Plate 2708-1 (Longitudinal Orientation) Before Testing.
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CL74 d; , a,, [ 05" } . 12,7 mm Figure B-2. Charpy impact Specimens CL61, CL64, CL66, CL74 and CL71 from Plate 2708-1.(Longitudinal Orientation) Before Testing. l B-2 .
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r CL63- L "" J : 12.7 mm i Figure B-3. Charpy impact Specimens.CL67, CL69, CL73, CL70 and CL63 from l Plate 2708-1-(Longitudinal Orientation) Before Testing. ! I i t B ' RM-25899 i I
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. , ..,.,...,m n lib 8: L 0.5" l 12.7 mm 4 Figure B-4. Charpy Impact Specimens CT63, CT61, CT64, CT62 and CT68 from Plate 2708-1 (Transvers? Orientation) Before Testing.
B-4 RW-25900
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- NCT69i l 0.5" l 12.7 mm Figure B-6. Charpy Impact Specimens CT74, CT73, CT75, CT71 and CT69 from Plate 2708-1 (Transverse Orientation) Before Testing, t
8-6 RM-25902 ,
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CW75 [ 0 S" j 12.7 mm Figure B-7. Charpy Impact Specimens CW72, CW61, Ct!65, CW69 and CW75 from the Weld Metal, Before Testing. B-7 RM-25903
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- 12.7 mm Figure B-ll, Charpy Impact Specimens CH65, CHE2, CH69, CH63 and CH68 from the Heat-Affected Zoia (HAZ), Before Testing.
B-11 RM.25907 1
mm . # g; y .p n ,, - z a
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- Figure B-12. Charpy Impact Specimens CH75, CH61, CH74, CH73 and CH70 from the Heat-Affected Zone (HAZ), Before Testing.
B-12 RM-25908
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,g CT15 Figure B-14. Tensile Specimens CT13, CT14 and CT15 from Plate 2708-1 '
(Transverse Orientation), Before Testing. B-14 RM-25910 l
":y{J , ,l
+
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CW15 Figure B-15. Tensile Specimens CW13, CW14 and CW15 from the Weld Metal,
, Before Testing.
B-lc RM-25911 i _ - - _ _ _ _ _ _ - _ - - - - _ _ _ _ _ _ _ _ - _ - _ _ _ - - _ _ _ _ . - _ - _ ____ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _}}