ML19332D295
| ML19332D295 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 09/30/1989 |
| From: | Albertin L, Shaun Anderson, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19332D293 | List: |
| References | |
| WCAP-12396, NUDOCS 8911300295 | |
| Download: ML19332D295 (92) | |
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ANALYSIS OF CAPSULE Y FROM THE COMMONWEALTH EDISON COMPANY Il0N UNIT 2 REACTOR VESSEi.
RADIATION SVRVEILLANCE PROGRAM i
E. Terek S. L. Anderson I
i L. Albertin September 1989 l
Work performed under Shop Order No. CUAP-106 M
i APPROVED:
l T.A.Meyer,Mdager l:
Structural Materials and Reliability Technology Prepared by Westinghouse for the Commonwealth Edison Company
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WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 14230-2728 mr.noveno j
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PREFACE i
I o
This report has been technically reviewed and verified, l
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Reviewer Sections 1 through 5 and 7, 8 S. E. Yanichko. /r 4,[4h>
u Section 6 E. P. Lippincott W ie-C ff r
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TABLE OF CONTENTS o.
Section Title Page i
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SUMMARY
OF RESULTS 1-1 L
l 2
INTRODUCTION 2-1 3
BACKGROUND 3-1 4
DESCRIPTION OF PROGRAM 4-1 i
5 TESTING OF SPECIMENS FROM CAPSULE Y 5-1 5-1.
Overview 5-1 5-2.
Charpy V-Notch Impact Test Results 5-3 5-3.
Tension Test Results 5-5 5-4.
Wedge Opening Leading Tests 5-5 6
RADIATION ANALYSIS AND NEUIRON 00SIMETRY 6-1 6-1.
Introduction 6-1 6-2.
Discrete Ordinates. Analysis 6-2 6-3.
Neutron Dosimetry 6-7 l
7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 l
8 REFERENCES 8-1 L
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r LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the 4-8 i
Zion Unit 2 Reactor Vessel i
4-2 Capsule Y Diagram Showing Location of Specimens, 4-9 l
Thermal Monitors, and Dosimeters 5-1 Charpy V-Notch Impact Data for Zion Unit 2 5-15 I
Rear'.or Vessel Shell Plate C4007-1 (Transverse Orientation) 5-2 Charpy V-Notch Impact Data for Zion Unit 2 5-16 Reactor Vessel Shell Plate C4007-1 (Longitudinal Orientation) 5-3 Charpy V-Notch Impact Data for Ziun Unit 2 5-17 Reactor Vessel Weld Metal 5-4 Charpy V-Notch Impact Data for Zion Unit 2 5-18 Reactor Vessel Weld Heat Affected Zone Metal 5-5 Charpy V-Notch Impact Data for Zion Unit 2 A533 5-19 Grade B Class 1 Correlation Monitor Material (HSST Plate 02) 5-6 Charpy Impact Specimen Fracture Surfaces for Zion 5-20 i
Unit 2 Reactor Vessel Shell Pisto C4007-1 (Longitudinal Orientation) l l
5-7 Charpy I : pact Specimen Fracture Surfaces for Zion 5-21 l
Unit 2 Reactor Vessel Snell Plate C4007-1 1*
'(TransverseOrientation) 5-8 Charpy Impact Specimen Fracture Surfaces for 5-22 Zion Unit 2 Reactor Vessel Weld Metal 3947s/102784 10 jy
LIST OF ILLUSTRATIONS (Cont)
Figure Title Page 5-9 Charpy Impact Specimen Fracture Surfaces for 5-23 Zion Unit 2 Reactor Vessel HAZ Metal 5-10 Charpy Impact Specimen Fracture Surfaces for 5-24 Zion Unit 2 ASTM Correlation Monitor Mat ^. rial 5-11 Tensile Properties for Zion Unit 2 Reactor 5-25 Vessel Shell Plate C4007-1 (Longitudinal Orientation) 5-12 Tensile Properties for Zinn Unit 2 Reactor 5-26 Vessel Weld Metal 5-13 Fractured Tensile Specimens for Zion Unit 2 5-27 Reactor Vessel Shell Plate C4007-1 (Longitudinal Orientation) 5-14 Fractured Tensile Specimens for Zion Unit 2 5-28 Reactor Vessel Weld Metal 5-15 Typical Stress-Strain Curve for Tension Specimens 5-29 6-1 Plan View of a Reactor Vessel Surveii'.ance Capsule.
6-12 6-2 Fast Neutron (E > 1.0 MeV) Fluence at the 40 Degree 6-13 Surveillance Capsule Location as a Function of Full Power Operating Time
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e LIST OF TABLES Table Title Page 4-1 Chemical Composition of the Zion Unit 2 4-3 Reactor Vessel' Surveillance Materials 4-2 Chemical Composition foe Zion l, nit 2 Capsule Y 4-4 Irradiated Charpy Impact Specimenc 4-5 Chemistry Results from the NBS Certified Reference 4-5 Standards 4-4 Zion Unit 2 Reactor Vessel Toughness Data (Unirradiated) 4-6 4-5 Heat Treatment of the Zion Unit 2 4-7 t
Reactor Vessel Surveillance Materials 5-1 Cha'rpy V-Notch Impact Data for the Zion Unit 2 5-6 Reactor Vessel Shell Plate C4007-1 Irradiated s
at 550*F, tluence 1.48 x 10 n/cm2 (E > 1.0 MeV) 19 5-2 Charpy V-Notch Impact Data for the Zion Unit 2 5-7 Reactor Vessel Weld Metal and HAZ Iletal Irradiated at 550*F, Fluence 1.48 x 10 n/cm2 (E > 1.0 MeV) 19 5-3 Charpy V-Notch Impact Data for the Zion Unit 2 ASTM 5-8 Correlation Monitor Material (HSST Plate 02) Irradiated at 550'F, Fluence 1.48 x 10 n/cm2 (E > 10 MeV) 19 5-4 Instrumented Charpy Impact Test Results for Zion Unit 2 5-9 Reactor Vessel Shell Plate C4007-1 Irradiated at 550'F, Fluence 1.48 x 10 n/cm2 (E > 1.0 MeV) 19
+
5-5 Instrumented Charpy Impact Test Results for Zion Unit 2 5-10
?aaetor Vessel Weld Metal and HAZ Metal Irradiated at 550'F, Fluence 1.48 x 10 n/cm2 (E > 1.0 MeV) 19 so47,no2no to yj
LIST OF TABLES (Cont) l Table Title Page 5-6 Instrumented Charpy Impact Test Results for Zion Unit 2 5-11 ASTM Correlation Monitor Material (HSST Plate 02) 19 2
Irradiated at 550'F, Fluence 1.40 x 10 n/cm (E > 1.0 MeV) 19 2
5-7 The Effect of 550*F Irradiation at 1.48 x 10 n/cm 5-12 (E > 1.0 MeV) on the Notch Toughness Properties of the Zion Unit 2 Reactor Vessel Materials 5-8 Comparison of Zion Unit 2 Raactor Vessel Surveillance 5-13 Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions 5-9 Tensile Properties for Zion Unit 2 Reactor Vessel 5-14 Material Irradiated to 1.48 x 10 n/r.m2(E>1.0MeV) 19 6-1 Calculated Fast Neutron Exposure Parameters at the 6-14 Surveillanca Capsule Center 6-2 Calculated Fast Neutron Exposure Parameters 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 6-5 Relative Radial Distribution of Iron Di. placement 6-18 Rate (dpa) Within the Pressure Vessel ', fall no. nome io y$4 i
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LIST OF TABLES (Cont) 4 1
Table Title Page 6 Nuclear Parameters for Neutron Flux Monitors S-19 1
6-7 Irradiation History of Neutron Sensors Contained 6-20 in Capsule Y q
6-8 Measured Sensor-Activities and Reaction Rates 6-25
)
i 6-9 Summary of. Neutron Dosimetry Results 6-27 i
6-10 Comparison of Measured and FERRET Calculated 6-28 Reaction Rates at the Surveillance Ccpsule Center c
6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-29 Capsule Center 6-12 Comparison of Calculated and Measured Exposure 6-30 Levels f.or Capsules 6-13 Neutron Exposure Projections at Key locations on.the 6-31 Pressure Vessel Clad / Base Metal Interface 6-14 Neutror Exposure Values for Use in the Generation 6-32 of Heatup/Cooldown Curves 1
6-15 Updated Lead Factors for Zion Unit 2 Surveillance 6 33 l
Capsules O
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4 SECTION 1 StM4ARY OF RESULTS The analysis of the reactor vessel material contained in Capsule Y, the third surveillance capsule to be removed from the Commonwealth Edison Company Zion Unit 2 reactor pressure vessel, resulted in the following conclusions:
o The capsule received an average fast neutron fluence (E > 1.0 MaV) 19 2
of 1.48 x 10 n/cm,
o Irradiation of the reactor vessel lower shell Plate C4007-1, to 1.48 x 19 10 n/cmt resulted in 30 and 50 ft-lb transition temperature increases of 121 and 130'F, respectively, for specimens orientad normal to the major working direction (transverce orientation) and temperature increases of 88 and 103*F, respectively, for specimens orierted parallel to the major working direction (longitudinal orientation).
19 2
o Weld metal irradiated to 1.48 x 10 n/cm experienced increases
~
in the 30 and 50 ft-lb transition temperatures of 220 and 255'F, respectively.
This results in a 30 f t-lb transition temperature of 210*F and a 50 ft-lb transition temperature of 300*F.
19 2
o Irradiation to 1,48 x 10 n/cm resulted in no decrease in the average upper shelf energy of Plate C4007-1 (transverse orientation) and a decrease of 18 ft-lb for the weld metal.
This results in a weld metal average upper sh< elf energy of 51 f t-lb.
o Comparison of the 30 ft-lb transition temperature increases for the Zion Unit 2 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Re.ision 2, demonstrated that the Plate C4007-1 material and weld metal transition temperature increases were 31* and 22*F, respectively, greater than predicted.
NRC Regulatory Guide 1.99, Revision 2 requires a 2 sigma allowance, of 34*F for base metal and 56*F for weld metal, be added to the predicted no, nom.a.
g.3 i
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+
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reference transition torperature to obtain a conservative upper bound value. Tnus, the reference transition temperature increases for Plate C4007-1 material and the weld metal are bounded by the 2 sigma allowance for shift prediction.
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I SICTION 2 INTRODUCTION o
This report presents the results of the examination of Capsule Y, the third capsule to be removed from the reactor in the continuing surveillance program which monitors the eff acts of neutron irradiation on the Commonwealth Edison Company Zion Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Commonwealth Edison Corcany Zion Unit 2 reactor pressure vesse1 materials was designed and reccmmended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Yanichko and Lege.Ill The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based l
on ASTM E-185-70, " Recommended Practice for Survcillance Tests in Nuclear Reactors". Westinghouse Electric Corporation personnel performed the postirradiation mechan' cal testing of the Charpy V-notch impact and tensile surveillance specimens.
This report summarizes testing and the postirradiation data obtained frcm surveillance Capsule Y removed from the Commonwealth Edison Ccmpany Zion Unit l
J reactor vessel and discusses the analysis of the data. The data are also compared to results of the previously removed Zion Unit 2 Capsule U(2) and Capsule TI33 l
l s-3947s/102769 10 g.}
n SECTION 3 BACKGROUND 4
3 3
The ability of the large steel pressure vessel, which contains the reactor core and its primary coolant to resist fracture constitutes an im p tant 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 proporties of low alloy ferritic pressure vessel steels such as SA533 Grade B Class 1 (base material of the Zion Unit 2 reactor pressure vessel beltline) are well documented in industry literature. Generally, low alloy ferritic materials demonstrate 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 ia " Protection Against Non-ductile Failurs," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code.
The method utilizos fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)*
RT is defined as the greater of either the drop weight nil-ductility NDT transiilon temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 ft ib (and 35-mil laterci expansien) temperature as determined from Charpy specimens oriented normal (transverse) to the raajor working direction of the m:terial.
The RT f a given material is used to index that NDT material to a reference stress intensity factor curve (Kypcurve)which L
appears in Appendix G of the ASME Code.
The K curve is a lower bound of IR i
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 utilizing these allowable stress intensity factors.
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RTHDT and the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Commonwealth Edison Company Zion tinit 2 i
Reactor Vessel Radiation Surveillance Progrcm.I13 A surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated i
specimens are tested. The ince<ase in the average Charpy V-notch 30 ft Ib temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RT f r radiation embrittlement.
This adjusted RT NDT NDT (RT initial + ARTNDT) is used to index the material to the KIR NDT curve and to set operating limits for the nuclear power plant which reflect the effects of irradiation on the reactor vassel materials.
9
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4 4
SECTION 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for raonitoring the effects of neutron exposure on the Zion Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 4-1.
The vertical center of the capsulos is opposite tho vertical center of the core.
Capsule Y (Figure 4-2) was removed after 9.18 effective full power years of plant operation. This capsule contained Charpy V-notch impact and tentile specimens from the reactor vessel lower shell Plate C4007-1, and submerged are weld metal representative of that used for the beltline region intermediate shell longitudinal weld seams of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) material.
All heat-affected zone specimens were obtained from within the HAZ of Plate C4007-1 of the representative weld.
~
!a addition, the capsule also contained Charpy V-notch specimens of an ASTM
.orrelation monitor material (HSST Plate 02).
The chemical composition, toughness data and heat treatment of the surveillance material are presented in Tables 4-1 through Table 4-5.
The chemical analyses reported in Table 4-1 were obtained from unirradiated material used in the surveillance program.
In addition, a chemical analysis using Inductively Coupled Picsma Spectrometry (ICPS) was performed on irradiated Charpy specimens from the lower shell Plate C4007-1 and weld metal and is reported in Table 4-2.
The chemistry results from the NBS certified reference standards are reported in table 4-3.
All test specimens were machined from the 1/4 thickness location of the plate.
Test specimens represent material taken at least one plate thickness from the quenched end of the plate.
All base metal Charpy V-notch impact and tensile specimens were oriented with the longitudinal axis of the specimen both normal to (transverse orientation) and parallel to (longitudinal orientation) the principal working direction of the plate. Charpy V-notch specimens from the i
munem a 43
weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction.
Tensile specimens were oriented with the longitudinal axis of the specimens normal to the welding direction. Weld metal IT Wedge Opening Loading test specimens in Capsule Y were machined such that the simulated crack in the specimen would propagate parallel to the weld direction for weld specimens.
All specimens were fatigue precracked per ASTM E399-70T.
Capsule Y conta$ned dosineter wires of pure iron, copper, nickel, and unshielded aluminum-cobalt.
In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were contained in the capsule.
Thermal monitors made from two low-melting eutectic alloys end sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2.
The two eutectic alloys and their meeting points are:
2.5% Ag, 97.5% Pb Melting Point 579'F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590'T (310*C)
The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Cspsule Y are shown in Figure 4-2.
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TABLE 4-1 i
CHEMICAL COMPOSITION OF THE ZION UNIT 2 REACTOR VESSEL j
SURVEILLANCE MATERIALS i
ASTM Plate C4007-1(c)
Weld Metal {b)(c)
Correlation Element (Wt. %)
(Wt. %)
Monitor Material C
0.23 0.077 0.22 1
S 0.016 0.013 0.018 N-0.006 0.014 2
Co 0.001(a) 0.001 (a)
Cu 0.12 0.28 0.14 Si 0.22 0.47 0.25 Mo 0.54 0.39 0.52 Ni 0.53 0.55 0.68 Mn 1.39 1.51 1.48 Cr 0.065 0.064 V
0.001 0.003 P
0.010 0.017 0.012 Sn 0.011 0.004 Al 0.033 0.030 Ti 0.001 (a) 0.001(a)
Sb 0.001 0.001 Zr 0.001(a) 0.001(a)
As 0.008 0.003 B
0.003(a) 0.003 (a)
Zn 0.001(a) 0.003 (a) Not detected. The number indicates the minimum limit of detection.
L' (b) Surveillance weld specimens were made from same weld wire as the inter-mediate sheil longitudinal weld seams (Wire Heat 72105) and Linde 80 l-Flux Lot 8773.
p (c) Chemical composition data from WCAP-8132.
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Table 4-2 CHEMICAL COMPOSITION FOR ZION UNIT 2 CAPSULE Y IRRADIATED CHARPY IMPACT SPECIMENS Weld Metal Chemical Composition (wt.%)IE)
Specimen No.
Cu Ni C
Mn P
~ S Si Cr No V
Co W-50 0.26 0.57 0.09 1.68 0.016 0.008 0.49 0.08 0.34
<0.010 0.016 W-55 0.27 0.60 0.09 1.57 0.021 0.007 0.45 0.07 0.41
<0.010 0.015
~
W-49 0.26 0.59 1.59 0.015' 0.08 0.42
<0.010 0.015 W-51 0.28 0.60 1.58 0.020 0.08 0.40
<0.010 0.016 W-52 0.26 0.60 1.66 0.019 0.08 0.40
<0.010 0.015 w-53 0.27 0.60 1.67 0.018 0.08 0.39
<0.010 0.016 W-54 0.26 0.56 1.50 0.020 0.08 0.35
<0.010 0.015 W-56 0.23 0.59 1.60 0.020 0.08 0.40
<0.010 0.013 Plate C4007-1 Specircen No.
Cu Ni C
Mn P
S Si Cr Mo_.
V Co E-70 0.10 0.48 0.26 1.24 0.007 0.011 0.25 0.064 0.45
<0.010 0.010 (a) Method of analysis -- Inductively Coupled Plasma Spectrometry (ICPS) for all elements except C, 5 and Si.
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TABLE 4-3 CHEMISTRY RESULTS FROM THE NBS CERTIFIE0 REFERENCE STANDt.RDS ho Material ID Low Alloy Steel: NBS Certified Reference Standards NBS 361 NBS 362 Certified Measured (a)
Certified
. Measured (a)
Metals Concentration in Weight Percent Fe 95.6 (matrix) 95.3 (matrix)
Mn 0.66 0.678 1.04 1.053 Cr 0.694 0.732 0.30 0.307 Ni 2.00 above calib 0.59 0.602 Mo 0.19 0.211 0.068 0.056 Co 0.032 0.031 0.30 0.321 Cu 0.042 0.043 0.50 0.513 P
0.014 0.0147 0.041 0.0485 V
0.011 0.011 0.040 0.036 C
0.383 0.383 0.160 0.157 S
0.014 NA 0.036 0.0351 Material ID Low Alloy Steel:
NBS Certified Reference Standards NBS 363 NBS 364
~
Certified Measured (a)
Certified Measured (b)
Metals Concentration in Weight Percent Fe 94.4 (matrix) 96.7 (matrix)
Mn 1.50 1.543 0.255 0.261 Cr 1.31 1.396 0.063 0.064 Ni 0.30 0.319 0.144 0.147 Mo 0.028 0.032 0.49 0.520 Co 0.048 0.047 0.15 0.158 o
Cu 0.010 0.102 0.249 0.263 P
0.029 0.0283 0.01 0.0129 V
0.31 0.316 0.105 0.143 0.62 NA 0.87 NA S-0.0068 NA 0.0250 0.0249
- Matrix element calculated as difference for material balance.
Tentative value, certified + 100% of value.
NA - Not analyzed; NR, Not requested (a) Method of analysis -- Inductively Coupled Plasma Spectrometry (ICPS) for all elements except C, 3 and Si.
u n,nonesso 45
TABLE 4-4 ZION UNIT 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)
NMWO ')
I RT Material Cu Ni NOT NOT USE Component Heat No.-
Type
(%)
(%)
(*F)
(*F)
(FT-LB)
Closure Head Dome B9094-1 A5338, C1. 1
.14
.55
-20 11 72 Closure Head Seg.
C4787-1A
.13
.62 0
0 88 Closure Head Seg.
C5086-2
.09
.54 30 30 88 Closure Head Flange 124W609 A508, C1. 2
.08
.70 12(a) 12 105 Vessel Flange 2V-965
.12
.74 60(a) 60 79 4
Inlet Nozzle ZT4007-2
.11
.70 48(a) 48
>78 Inlet Nozzle ZT3885-1
.11
.58 60(a) 60 82 Inlet Nozzle ZT3885
.11
.56
'43 a) 43 78 Inlet Nozzle ZT3885
.11
.56 60(a) 60
>84 Outlet Nozzle ZV3930
.12
.66 58(a) 58 93 Outlet Nozzle ZV3930
.11
.65 48(a) 48
>80 Outlet Nozzle ZV3930
.12
.67 55(a) 55 84 Outlet Nozzle ZT3885-4
.11
.57 60(a) 60
>61 Upper Nozzle Shell Z03940 A508, C1. 2
.07
.62 10 10 106 Lower Nozzle Shell ZV3855
.09
.66 10 10
>80 2,
ln Lower Shell 88029-1 A5338, C1. 1
.12
.51
-10 22 81 Lower Shell C4007-1
.12
.53 10 22 94 (Actual)
Inter. Shell B8006-1
.12
.54 10 10 89 Inter. Shell B8040-1
.14
.52
-10 2
92 Bottom Head Trans. Ring 3V-433 A508, C1. 2
.09
.76 0
0 87 Bottom Head Dome C4007-2 A5338, C1. 1
.12
.53
-20 0
72 Inter. to Lower Shell Girth Weld Seam SA1769 (b)
SAW
.26
.60 0(a) 0 Lower Shell Long. Weld Seams WF29 (c)
SAW
.23
.63 0(a) 0 Inter. Shell Long. Weld Seams WF70 (d)
SAW
.32
.56 0(a) 0 (a) Estimated using method of U.S. NRC HUREG-0800 Branch Technical Position MIEB 5-2, July, 1981.
(b) Weld Wire Heat No. 71249 and Linde 80 Flux Lot No. 8738 (c) Weld Wire Heat No. 72102 and Linde 80 Flum Lot No. 8650 (d) Weld Wire Heat No. 72105 and Linde 80 Flux Lot No. 8669 (Chemical composition per WCAP-10962, December 1985)
(e) Normal to Major Working direction 3947s/102199 to e
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TABLE 4-5 1
HEAT TREATMENT OF THE ZION UNIT 2 REACTOR VESSEL ' SURVEILLANCE MATERIALS Material Temperature ('F)
Time (br)
Coolant Intermediate Shell 1600/1650 93/4 Brine quenched Plate C4007-1 1200/1225 9 3/4 Brine quenched 1100/1150 30 Furnace cooled Weld Metal 1100/1150 30 furnace cooled Correlation Monitor 1625 1 25 4
Air cooled (HSST Plate 02) 1600 1 25 4
Water quenched I
1225 1 25 4
Furnace cooled 1150 1 25 10 Furnace cooled f
9 L.
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47 1
h
.+
LOWER SHELL WELD REACTOR VESSEL (WF-29) y THERMAL SHIELD CORE SARREL X-CAPSULE (TYP)
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i Z
W
- INTERMEDIATE SHELL WELD
. INTERMEDIATE (WF-70)
SHELL WELD 4'
40*
(WF-70) f 4*
100' -
- o.
A 8
v 7
sa
- [
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T
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u LOWER SEELL WELD 90*
(WF-29) i Figure.4-1. Arrangement of Surveillance Capsules in the Zion Unit 2 Reactor Vessel no.,os =. io 4-8
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- 4. ; ;.L-C. -.;; ; b. --
=et -
lllL
.C-
- C.
- c. -
- e. -
.- C.
C..u.
m C.
ou uu uu uu ou 2..
n,
.n n
n.
y_
ua
- E.
ey-ly y
y
,0 70P OP VESSEL TO WOT,UE.0F WESEEt EPECaisE4 esUnset s.essG CODE
..............u..,
............,..u,.,
...u...,
Figure 4-2 Capsule Y Diagram Showing Locations of Specirr. ens, Thermal Monitors, and Dosimeters
..o
+
! ss.
b SECTION 5-TESTING OF SPECIMENS FROM CAPSULE Y
+
4 5-1.
OVERVIEW i.: gostirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. ' Testing was performed in accordance with 10CFR50, Appendices G and HI N, ASTM Specification E185-82 and Westinghoust Procedure MHL 8402,
. Revision 1 as modified by Westinghouse RMF Procecures 8102, Revision 1 and m
8103, Revision 1.
Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8132.l13 No discrepancies were found, l
i l'
Examination of the two low-melting 304'C (579'F) and 310'C (590'F) eutectic dlloys indicated no melting of either type of thermal monitor.
Based on thir l
examination, the maximum temperature to wnich the test specimens were exposed l
was less than 304'C (579'F).
l The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J machine.
The tup (atriker) of the Charpy machine is instrumented with an Effects Technology model. 500 in:;trumentation systeni. With this system, load-time and energy-time signals can be recorded in addition to the standard mea:,urement of Charpy energy (E ).
From the load-time curve, the load of general yielding D
(PGY), the ime to gs eral yielding (tgy), the maximum load (P ), and M
the time to maximum load (t ) can be determined. Under some test M
conditions, a sharp drop in load indicative of fast fracture was observed.
The load at which fast fracture was initiated is identified as the fast frecture load (P ), and the load at which fast fracture terminated is p
identified as the arrest load (P ).
The Charpy machine is maintained and
~*
A i
an.ncmuo 5-1
i?
- i p
v calibrated in accordance with ASTM Specification E-23.
The annual
. qualification of-the Charpy machine is accomplished with the Standardized
'Watertown test specimens. The' specimens and procedurea used by Westinghouse are recommended by the Army Materials Technology Laboratory and is identical to the a' proached used to certify machines that generate data for government-p contracts.
The energy at maximum load (E ) was determined by comparing-the energy-time.
g acord and the load-time record.
The energy at maximum load is approximately v.' valent to the energy required to initiate a crack in the specimen.-
Therefore, the propagation energy for the crack (E ) is the difference p
between the total energy to fracture (E ) and the energy at maximum load.
D The yield stress (oy) is calculated from the three point bend formula.
The flow stress is calculated from the average of the yield and maximum loads, also using' the three point bend formula.
Percentage shear was determined from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77.
The lateral expansion was measured using a dial gage rig similar to that shown in I
the same specification.
Tension tests were performed on a 20,000 pe.und Instron, split-console test machine (Model 1115) per ASTM Specificatior.s E8-83 and E21-/9, and RMF Procedure 8102, Revision 1.
All rull rods, grips, and pins were made of Inconel 718 hardened to Rc45.
The upper pull rod was connected through a universal joint to improve axiality of loading.
The tests were conducted at a
. constant. crossheed speed of 0.05 inch per rainute throughout the test.
l Deflection measurements were made with a linear variable displacement transducer (LVDT)ertensometer.
The extensometer knife edges were
?
spring-loaded to '.he specimen er,a operated through specimen failure.
The extensometer gage length is 1.00 inch.
The extensometer is rated as Class B-2
_per. ASTM E83-67.
w w,o w eso 5.g
w L
l L
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnue with a 9-inch hot zone. All tests were conducted in air.
Because of the difficulty in remotely attaching a thermocoupla directly to the specimen, the following procedure was used to monitor specimen temperature.
Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip.
In test L
~ config'uration,'with a slight' load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the. range room temperature to 550'F (288'C).
The upper' grip was used to control the furnace temperature.
During the actual testing the grip temperatures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to plus or minus 2'F.
The yield load, ultimate' load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension cu ve.
The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area.
The final diameter and final gage length were
, determined from postfracture photographs.
The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5.2.
CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact test
- performed on the various materials 19 2
contained _in Capsule Y irradiated to approximately 1.48 x 10 n/cm at SB0'F are presented in Tables 5-1 through 5-6 and Figures 5-1 through 5-5.
The transition temperature increases and upper shelf energy decreases for the Capsule Y material are shown in Table 5-7.
l Irradiation of the vessel lower shell Plate C4007-1 material (transverse orientation) specimens to 1.48 x 10 n/cm2 (Figure 5-1) resulted in a 30 19 l
and.50 ft-lb transition temperature increases of 121*F and 130*F, respectively, and an upper shelf energy increase of 4 ft-lb when compared to the unirradiated data.Ill It is probable that the in m ase in upper shelf energy is due to a random scatter in the data and it i: incre likely that the upper shelf energy is not changed.
l 3.o. nom.ao 5-3 l
k "1 -
Irradiation of the vessel lower shell Plate C4007-1 material (longitudinal orientation)' specimens'to 1.i8 x'10 m/cm2 (Figure 5-2) resulted in 30 19 and 50 ft-lb~ transition temperature increases of 88'F and 103'F, respectively, and'an upper shelf energy decrease of 17 ft-lb when compared to the unirradisied data.
Weld material irradiated to 1.48 x 10 n/cm2 (Figure 5-3) resulted in a 19
' ft-lb transition temperature increase of 220*F (from -10*F to 210*F) and a
- 50 ft-lb transition temp 2rature increase of 255'F (from 45'F to 300'F). Weld 19 2
metal irradiated'to 1.48 x 10 n/cm resulted in an upper shelf energy decrease of 18 ft-lb (from 69 ft-lb to 51 ft-lb).
19 Weld HAZ metal irradiated to 1.48 x 10 n/cm2 (Figure 5-4) resulted in 30 and 50 ft-lb transition temperature increases of 97*F and 99'F, respectively, and'an upper shelf energy decrease of 24 ft-lb.
'I A533 Grade B Class 1 correlation monitor material (HSST Plate 02) irradiated to 1.48 x 10 n/cm2 (Figure'5-5) resulted in 30 and 50 ft-lb transition 19 temperature increases of 135'F and 137'F, respectively, and an upper' shelf energy decrease of 14 ft-lb. The 30 ft-lb transition temperature increase is 23*F greater than predicted. 'The 23*F increase is bounded by the 2 sigma allowance of 34*F for shift predicu an.
The fracture appearance of each irradiated Charpy specimen from the various materials is shown_in Figures 5-6 through 5-10 and show an increasing ductile or tougher appearance with increasing test temperature.
Table 5-8 shows a comparison of the 30 ft-lb transition temperature
- (ARTNDT) increases for the various Zion Unit 2 survoillance meterials with predicted increases using the methods of NRC Regulatory Guioe 1.99, Revision 2.[5)' This comparison shows that the transition temperature increase 19 2
resulting from irradiation to 1.48 x 10 n/cm is greater than predicted by the Guide for Plate C4007-1.
The weld metal transition temperature 19 2
increase resulting from 1.48 x 10 n/cm is.also greater than the Regulatory Gui.de prediction.
m, nom,o 54
a-
~
15-3.. TENSION TEST RESULTS s,
The results of _ tens 4.on tests performed on Plate C4007-1 (longitudinal I9 2
orientatio'n).and weld metal irradiated to 1.48 x 10 n/cm are shown in 4<
Table 5-5'and Figures 5-11 and 5-12, respectively.
These results show that irradiation produced'a 0.2 percent yis1d strength increase of approximately 20 Ksi for. Plate C4007-1 and the weld metal.
Fractured tension specimens for each of the' materials are_shown in Figures 5-13 and 5-14.
A typical stress-strain curve for the tension specimens is shown in Figure 5-15.
~5-4. '. WEDGE OPENING LOADING It is common practice to store IT - Wedge Openirg loading (WOL) fracture
~
mechanics specimens at the Westinghouse Science and Technology Center Hot Cell..
W t
I' I
I '
h I
sw,isonne so 55
{
i e
TABLE 5-1 l
r y-CHARPY-V-NOTCH IMPACT DATA FOR THE ZION UNIT 2 REACTOR VESSEL SHELL PLATE C4007-1
)
IRRADIATED AT 550*F, FLUENCE 1.48 x 10 n/cm2 (E > 1.0 MeV) j 19 I
Temperature' Impact Energy.
Lateral Expansion Shear Samole No.,('Z),
(* C)
(ft-lb) g), '
(mils)-
,(gg,),
(%)
H l'
Lonstitudinal Orientation
)
li E62 75 (24) 12.0 ( 16.5) 10.0 (0.25) 10 4
l E61 125 (52) 32.0 ( 43.5) 27.0 (0.69) 20 l
i E70 125
( 52) 31.0 (42.0) 22.0 (0.56) 20 E63 150
( 66) 47.0 ( 63.5) 35.0 (0.89) 35 E66 150
( 66) 44.0 ( 59.5) 34.0 (0.86) 35 l
E67
.175
( 79) 53.0 (72.0) 41.0 (1.05) 45 E03 200 (93) 69.0
( 93.5) 45.0 (1.14) 70 E69 275 (135) 113.0 (153.0) 60.0 (1.52) 100 E64 350 (177) 108.0 (146.5) 73.0 (1.85) 100 E68 450' (232) 112.0 (152.0) 73.0 (1.85) 100 Transverse Orientation ET62 75
( 24) 14.0 ( 19 0) 14.0 (0.36) 10 L
ET61 150
( 66) 23.0 ( 31.0)
'22.0 (0.56) 20 ET67 175-(.79) 35.0 ( 47.5) 30.0 (0.76) 30 ET66 175
( 79) 35.0
( 47.5) 32.0 (0.81) 25 ET70 200
( 93) 43.0 (58.5) 35.0 (0.89) 45 1-ET65 210
( 99) 94.0 (127.5) 68.0 (1.73)'
100 L
ET68 240 (116)
F3.0 (112.5) 60.0 (1.52) 80
.ET64 275 (135) 78.0 (106.0) 59.0-(1.50) 80 l
ET69 350 (177) 102.0 (138.5) 67.0
- (1.70) 100 ET6.,
450 (232) 98.0 (133.0) 66.0 (1.68) 100 L
J an.ncme io 56
,,,.m,
.m
3;.
u v
I
',.[
TABLE 5-2 CHARPY V-NOTCH-IMPACT DATA FOR~ THE ZION UNIT 2 REACTOR-VESSEL ' WELD METAL AND HAZ METAL IRRADIATED - AT 550*F L*
FLUENCE 1.48 x 10 n/cm2 (E > 1.0 MeV) 1F L
V Temperature Impact Energy Lateral Expansion 'F'-sr Sample No. G
('C)
(ft-lb) M (sils)
',as),
Weld Metal W56 74 (23) 15.0 20.5) 10.0 c ~
W55 150 (66) 19.0 26.0) 15.0 3
W50 200 93) 20.0 27.0) 9 W53 225-107) 39.0 53.0}
- 0
?
i W54 225 107) 33.0 44.5
.0 W52-250 121) 44.0 59.5
'.0 (t
W49.
350 (177) 50.0
( 68.0',
0 (1.07)
W51l 450- (232) 52.0
( 70-5
.0 (1.R N
HAZ Metal H54
-50 (-46) 23.0
( 31.0) 19.0 (0.48) 10 H55
-10 (-23)
'10.0 (13.5) 14.0 (0.36) 10
-H56 0 (-18) 38.0 (51.5) 27.0 (0.69) 30-H49 25 -(- 4) 49.0
( 66.5) 33.0 (0.84) 45 H51 75 ( 24) 48.0
( 65.0) 29.0 (0.74).
40 H52 150 ( 66) 95.0 (129.0) 70.0 (1.78) 100 H50 250 (121) 118.0 (160.0) 76.0 (1.93) 100
.H53 350 (177) 83.0 (112.5) 53.0
-(1.35) 100
.-gi 3947s/102780 10 5-7
1 1
1 j
i TABLE 5-3 CHARPY V-NOTCH. IMPACT DATA FOR THE ZION UNIT 2 "j
- ASTMCORRELATIONMONITORMATERIAL(HSSTPLATE02) 19 IRRADIATED AT 550*F,' FLUENCE 1.48 x 10 n/cm2 (E > 1.0 MeV)
~
(
Temperature Impact Energy Lateral Expansion Shear geole No. Q
('C)
(ft-lb) 2),
(mils)
(aa)
(%L
.RSS
. 74 (23).
14.0 (19.0) 11.0 (0.28) 10 R51 150
( 66) 18.0 (24.5) 15.0 (0.38) 15 R54 175 (79) 23.0 ( 31.0) 21.0 (0.53) 20 R49 200 (93) 43.0 ( 58.5) 31.0 (0.79) 35 R50 200
( 93) 39.0 (53.0) 29.0 (0.74) 35 E56 250 (121) 75.0 (101.5) 53.0 (1.35 65' R56.
350 (177) 113.0 (153.0) 74.0 (1.88 100 152 450 (232) 108.0 (146.0) 70.0 (1.78 100
'4 l
a n o.~ne m io 5-8
=-
c 3
i N,
4 TABLE 5-4
' INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR ZION UNIT 2 REACTOR VESSEL SHELL PLATE C4007-1
-IRRADIATED AT 550*F, FLUENCE 1.48 x'10 n/cm2 (E > 1.0 MeV) lS f
f i
Normalised Energies Test Charpy Charpy -Maximum frop Yield-Time Maximum. Time to Fracture Arrest Yield Flow Saople Temp Energy Ed/A Ee/A Ep/A Load to Yield Load' Maximum Load Load
. Stress Stress 2
Number (*F)
(ft-lb)
(ft-lb/in )
(kips)
(ssec)
(kips) fasec).
(kipsl_ (kips)
(ksi)
(ksi)
Lonnitudinal Orientation E62 75 12.0 97 39 58 2.75 80 3.60 135 3.45 0.45 92 106 E70 125 31.0 250 151 99 2.95 85 4.25 360 4.25 0.65 98 119 E61 125 32.0 258 169 88 2.25 95 3.90 445 3.90 1.06 74..
102 E66 150 44.0 354 217 138 2.70 85 4.20 510 4.10 1.15 90 115 E63 150 47.0 378 206 173 2.60 90 4.00 510 3.95 1.45 86 109 "3
E67 175
~53.0 427' 253 174 3.00 60 3.85 610 3.85 1.45
'100 114 i
E65 200 69.0 556 266 290 2.55
- 60 4.40 575 4.15 2.05' 84 115 i
E69 275 113.0 910 298 612 2.55 145 4.15 750 84 til i
E64 350 108.0 870 266 604 3.00 100 4.20 610 99 fl9.
E65 450 112.0 902 259 642 2.G0
-110 4.05 625 97 115
'Iransverse Orientation ET62 75 14.0 113 55 58 3.45 45 3.95 130 3.95 0.15 114 122 ET61 150 23.0 185 94 91 2.45 65 3.65 260 3.65
'1.06 80 100.
ET66 175 35.0 282 142 139 2.30 80 3.80 375 3.75 1.55 76 101 ET67 175 35.0 282 159 123 2.80 65 4.15 370 4.10 1.2b 93 115.
I ET70 200 43.0 346
.176 171 3.30 95 4.20 410 4.15 1.95 109 124 ETG5 210 94.0 757 220 537 2.70 130 3.65 605 89 106 I
ET68 240 83.0 668 239 429 2.45 110 4.15 575 3.80 2.90 81 109 L
ETC4 275 78.0 628 173 455 2.20 80 3.65 480 72 96 ET69 350 102.0 821 258 563 2.90 105 4.15 605 96 117 ZT63 450 98.0 789 209 580 2.85 95 4.00 505 95 113 4
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r,n TABLE 5-6 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR ZION UNIT 2 ASTM CORRELATION MONITOR MATERIAL-(HSST PLATE 02).
u-
~ IRRADIATED AT 550*F, FLUENCE 1.'48 x 10 n/cm2 (E.> 1.0 MeV).
,[
19 t
Normalised Energies-l Test Charpy
~Charpy Maximum Prop Tield' Time
-Maximum. Tine to ' Fracture Arrest Yield Flow Sample Temp' Energy.
Ed/A Em/A Ep/A Load to Yield Load Maxioma Load Load Stress Stress.
2 Namber M (ft-lb)
'(ft-lb/in )'
(kips)'
(msec)
(kips)
(asec)
_Lkips) _ C ins)
(ksi)
(ksi)
R55 74 14.0
'113
'46
~
67 2.55 100-3.90 155~
3.95' O.30 84 106' R51 150 18.0 145-83 62 2.40 60 3.45 235 3.45 0.55'
'79' 97.
R54 175 23.0 185 90 95 2.10
'90 3.40 285 3.40 1.25 69 91' R50 200 39.0 314 216-98 2.75 85 4.25 505 4.25 1.60 91
-116 R49 200 43.0 346 246 100 2.30 60 4.05 580 4.05 1.20 77
-10E T
K56 250 75.0 604 220 384 3.15 100 4.15 510 106 121
.[
R53 350 113.0 910 284 626 2.50 65 4.05-665 82 108 i
R52 450 108.0 870 242 628 2.75 95 3.80 605 90
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TABLE 5-8 COMPARISON OF ZION UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS WITH REGULATORY GUIDE 1.99 REVISION 1 PREDICTIONS j
ARTNDT ( F)
USE DECREASE (%)
Fluence "3 I
19 2
- Material Capsule 10 n/cm Meas.
Pred.
Meas.
Pred.
Plate C4007-1 0
0.27 38 53.3 5.5 16 (Longitudinsi)
T 0.78 75 76.3 18.8 19 Y
1.48 88 91.0 13.3 22.5 Plate C4007-1
.U 0.27 49 53.3 0
16
~
(Transverse)
T 0.78
.90 76.3 14.9 19 Y
l'48 121 91.0 0
22.5 Weld Metal V
0.27 128 117 27.5 32 T
0.78 175 167 36.2 40 Y
1.48 220 200 26.1 46 Correlation 0
0.27 50 66.3 10.5 17.5 Monitor T
0.78 100 94.9 28.2-22 Y
1.48 135 113 11.3 25.5
[a]
These have been recalculated using the methods of Section E.
t.
mwimee io '
5-13
, ~.
7, g;;
i,h g
TABLE 5-9
_t j
TENSILE PROPERTIES FOR ZION UNIT 2' i
REACTOR VESSEL MATERIAL IRRADIATED TO 1.48 x-'10 n/cm2 (E > 1.0 MeV)'
I9 i.
o Test 0.25 Yield Ultimate Fracture Fracture ' Fracture ' Uniform Total-Reduction-Sample Temp. Strength Strength Load Stress
. Strength Blnegation Blongation in Aren-Material Number,[*F1 (ksi)
(ksil (k b l (ksi) '
(ksi)
(%)
(5)~
(5)
Plate C4007-1 E6 79 89.1 114.1 3.CO
-2"%.6-73.3 71.2
'23.2
'68 l
(1,ong.
ES 550 75.4~
99.8~
'3.40 189.1~
69.3 9.6 20.3 63 i
Orient.)
m 1
O Weld W14 79 91.7 114.1 4.27 254.1 86.9 9.1 19.5' 66 W13 550 89.6 105.9 4.60 210.8 93.7 8.3 13.5 5F I
t l
4 l
i i
I b47sm9208910 3
9
.,g.
(
- C)
-150 -100
- 50 0
50 100 150 200 250 c
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FIGURE 5-1. CHARPY V-NOTCH IMPACT DATA FOR ZION UNIT 2 REACTOR VESSEL SHELL PLATE C4007-1 (TRANSVERSE ORIENTATION)
cu rv e-757563-A -
(
- C) 1
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FIGURE 5-2 CHARFY V-NOTCH IMPACT DATA FOR ZION UNIT 2 REACTOR VESSEL SHELL PLATE C4007-1 (LONGITUDINAL ORIENTATION) l l
i 7347s/092089 10 5-16 o
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100 200 300 20 500 Temperature ("F) i FIGURE 5-3 CHARPY V-NGTCH IMPACT DATA FOR Z?DN UNIT 2 REACTOR Vf.5SE f
METAL 1
L 5-17
... l
P Curve 757562-A
('C)
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FIGURE 5-4 CHARPY V-NOTCH IMPACT DATA FOR ZION UNIT 2 REACTOR VESSEL WELD HEAT AFFECTED ZONE METAL H
no.mme io 5-18
curve 757 %0-A
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(
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FIGl.lRE 5-5 CHARPY V-NOTCH IMPACT DATA FOR ZION UNIT 2 A533 GRADE B CLASS 1 CORRELATI0t; MONITOR MATERIAL (HSST Pl. ATE 02) 9 am. mms io 5-19
u l
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FIGURE 5-6 CHARPY IMPACT SPECIMEN FRACTilRE SURFACES FOR ZION UNIT 2 REACTOR VESSEL SHELL PLATE C4007-1 (LONGITUDINAL ORIENTATION) sw. mms ia 5-20 RM-21085
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~
3947s/092089.10 5-N RM-21089
__.=;.,
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FIGURE 5-11 TENSILE PROPERTIES FOR ZION UNIT 2 REACTOR VESSEL SHELL PLATE C4007-1 (LONGITUDINAL ORIENTATION) an. ***
- 5-25
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curve 757565-A
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FIGURE 5-12 TENSILE PROPERTIES FOR ZION UNIT 2 REACTOR VESSEL WELD METAL 3947s/092089 10
_ _... _ _,... _ _ _.. _,. _., _ _ _ _ _ _. _ _ _. _... _. _.. _., _ _ _.. _, _. _.. _. _ _ _ _.. _.. _. ~ _ _...
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t FIGURE 5-13 FRACTURED TENSILE SPECIMENS FOR ZION UNIT 2 REACTOR VESSEL SHELL PLATE C4007-1 (LONGITUDINAL ORIENTATION) 3947s/092089 10 5-27 9
RM-21090
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FIGURE 5-14 FRACTURED TENSILE SPECIMENS FOR ZION UNIT 2 REACTOR /ESSEL l
WELD METAL l
M47 s.'092089 10 5-28 f
RM-21091
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FIGURE 5-15 TYPICAL STRESS-STRAIN CbRVE FOR TENSION SPECIMENS l
r e.
3s47 /ostoss-io 5-29 l
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'l 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 1
6.1 ~ INTRODUCTION Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LHR 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 presernt and future condition of the reactor vessel, a relationship autt be established between L
the neutron environment at'various positions within the reactor vessel and e
that experienced by the test specimens. The former requirement is normally
- met by employing a combination of rigorous analytical techniques and measurements obtaineo with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from anclysis.
The use of fast. neutron fluence (E > 1.0 MeV) to correlate measured materials
- properties changes to the neutro's exposure of the material for light water reactor applications has traditionally been accepted for dewelopment of damage trend curves as well as for the implementation of trend curve data to assess vessel condition.
In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853,
" Analysis and Interpretation of Light Hater Reactor Surveillance Results,"
recommends reporting displacements per iron atom (dpa) along with fluence l
l l
1 9225e:1d/102089 6-1 1!.
-m.
--,._,m
.m..-._.,,..-._m,,,,,
,,., _.. ~. -
,,,. ~ _ -
i
}
f (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is gicified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Aton." The_ application of the dpa parameter to the assessment of embrittlement gradients through the thickness of. the pressure vessel wall has already been promulgated in. Revision 2'to the i
Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials."
7 This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in 4
surveillance Capsule Y.
Updated evaluations of dosimetry from prior surveillance capsule withdrawals are also presented to provide a complete data i
L base for use in establishing the integrated exposure of the pressure vessel L
wall. These re-evaluations include data from Capsules U and T withdrawn at the conclusion of the first and fourth fuel cycles, respectively.
Fast neutron exposure parameters in terms of fast neutrons fluence (E > 1.0 Mov), fast neutron fluence (E > 0.1 Mov), and iron atos displacements (dpa)
- are established for the irradiation history. The analytical formalism relating the measured capsule exposures 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 are provided.
6.2 DISCRETE ORDINATES ANALYSIS l
A plan view of the reactor geometry at the core midplane is shown in Figure Eight irradiation capsules attached to the thermal shield are included l
4-1.
in the reactor design to constitute the reactor vessel surveillance program.
Four capsules are located synetrically at azimuthal angles of 4' and 40' relative to the core cardinal axis as shown in Figure 4-1.
A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 6-1.
The stainless steel specimen containers are 1-inch square and approximately 38 inches in height. The containers are positioned 9225e:1d/102089 6-2 i
'A L,-
l axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core.
.7 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 thermal shield 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 l
carried out. The first, a single computation in the conventional forward L
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 {$(E > 1.0 Mev.) $(E > 0.1 Mov), and dpa) through the vessel wall. The neutron sportral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratics; i.e.,
dpa/$(E > 1.0 MeV), within the pressure vessel geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel' wall; i.e.,
the 1/4T, 1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core.
The importance l-functions generated from these adjoint analyses provided the basis for all l
absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for the first 10 cycles of irradiation; and established l
l 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 9225e:1d/102089 6-3
'1
)
spatial variations of fission rstes within the reactor core; but, also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased.
)
The absolute cycle specific data from the adjoint evaluations together with i
relative neutron energy spectra and radial distribution information from the forward calculation provided the means to:
1.
Evaluate neutron dosimetry obtained from surveillance capsule i
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.
l L
The forward transport calculation.for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R, 9 geometry using the DOT two-dimensional j
discrete ordinates code [6] and the SAILOR cross-section library [7]. The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications. In these analyses anisotopic scattering was treated with a P expansion of the cross-sections and the angular 3
discretization was modeled with_an S order of angular quadrature.
8 1
The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Nestinghouse 4-loop I
plants.
Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core
- t periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant 'nd a
cycle to cycle variations in peripheral power was used.
Since it is unlikely t
l l
9225e:1d/102089 6-4 L
E-
that a single reactor would have a power distribution at the nominal +2e level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.
L order of angular All adjoint analyses were also carried out using an S8 cross-section approximation from the SAIL.0R library.
quadrature and the P3 Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, e geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, ( (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:
I(r, 9. E) S (r..e, E) r dr de dE R (r, e) f I IE r6
$ (E > 1.0 MeV) at radius r and azimuthal angle e where: R(r. e)
=
L'.
Adjoint.importance function at radius, r, azimuthal I (r, 9. E) angle e, and neutron source energy E.
j 1
Woutron source strength at core location r, e and S (r. 8 E) energy E.
Although the adjoint importance functions used in the Zion Unit 2 analysis were based on a response function defined by the threshold neutron flux F
(E > l'.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the l
spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/$ (E > 1.0 MeV) is insensitive to changing core source distributions.
In the application of these adjoint important functions to the Zion Unit 2 reactor, therefore, calculation of the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/$ (E > 1.0 MeV) and $ (E > 0.1 MeV)/$ (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific $ (E > 1.0 MeV) solutions from the individual adjoint evaluations, 9225e:1d/102089 6-5
The reactor core power distributions used in the plant specific adjoint calculations were _taken from the fuel cycle design reports for the first ten operating cycles of Zion Unit 2 [8 thru 17].
Selected results from the neutron transport analyses performed for the Zion Unit 2-reactor are provided in Tables 6-1 through 6-5.
The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the l
pressure vessel wall.
In Table 6-1, the calculated exposure parameters {$ -(E > 1.0 MeV),
( (E > 0.1 MeV), and dpa} are given at the geometric center of the two surveilitnce 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 l
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 thred pertinent exposure parameters are listed for both the design basis and the cycle 1 through 10
- average plant specific power distributions.
It is important to note that the l
l 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 atos displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations.
Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.
9225e:1d/102089 6-6
- - * ' * ' + '
e e
a a
w.-%
,,.--,,a
,,w'.-e
,g.
e----m,,-,%9.g
-4_mg--.,,yy.qytg,--
-r ww ym eeg yw g we yqypw-* wet.-w'WF 9. ieur > y re-t39v t-w
I
]
L i
i For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by:
e
$(220.27, 45') F (225.75, 45')
$1/4T(45')
=
Projected neutron flux at the 1/4T position on where
$1/4T(45')
=
the 45' azimuth Projected or calculated neutron flux at the 4 (220.27, 45')
=
vessel inner radius on the 45' azimuth.
Relative radial distribution function from Table F (225.75, 45')
6-3.
Similar expressions apply for exposure parameters in terms of $(E > 0.1 MeV) and dpa/sec.
6.3 NEUTRON DOSIMETRY Tt.a pessive neutron sensors included in the Zion Unit 2 surveillance program are listed in Table 6-6.
Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determina-tion of the various exposure parameters of interest ($ (E > 1.0 Mev), $
(E > 0.1 NeV), dpa}.
The relative locations of the neutron sensors within the capsules are shown in Figure 4-2.
The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest.
l l
9225e:Id/102089 6-7
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'ac.tivation measurements only if the irradia' ion parameters are well t
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.
The neutron energy spectrum at the monitor location.
o The physical characteristics of the monitor.
o The specific activity of each of the neutron monitors was determined using L
established ASTM procedures [18 through.313. Following sample preparation and weighing, the activity of each monitor was determined by means of a i
lithium-drifted germanium, Ge(Li), ganna spectrometer. The irradiation history of'the Zion Unit 2 reactor during cycles 1 through 10 was obtained from NUREG-0020. " Licensed Operating Reactors Status Sunatary 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.
l Values of key fast neutron exposure parameters were derived from the measured l
reaction rates using the FERRET least squares adjustment code [323. The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra.
9225e:1d/102089 6-8
e k
Y
- 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) 4(a) f (s,a), g g i
ig g
g where i indexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be simultaneously adjusted.
For example, R _= I -ogg (g g
g relates a set of measured reaction rates Rg to a single spectrum (g by the multigroup cross section ogg.
(In this case, FERRET also adjusts the cross-sections.) The lognormal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.
In the FERRET. analysis of the dosimetry data, the continuous quantities (i.e.,
fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [33]. This precedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed to the group scheme used in FERRET.
The cross-sections were also collapsed into the 53 energy-group structure i
using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/B-V dosimetry file.
Uncertainty estimates and 53 x 53 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.
9225e:1d/102089 6-9
]
w e
4I For each set of data or a priori values, the inverse'of the corresponding g
relative covariance matrix M is used as a statistical weight.-
In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is used:
gg,-Rf+R R, P,g,
.M g
g where R specifies an overall fractional normalization uncertainty (i.e.,
y complete correlation) for the corresponding set o'f values. The fractional uncertainties R specify additional random uncertainties for group g that g
are correlated with a correlation matrix:
')23 Pgg, - (1 - e) sgg, + e exp (- (a-2y The firstxtern specifies purely random uncertainties while the second term describes short-range correlations over a range y (9 specii'ies the strength of the latter ters.)
For the a priori-calculated fluxes, a short-range correlation of y = 6 groups was used. This choice implies that neighboring groups are strongly correlated when e is close to 1.
Strong long-range correlations (or' anticorrelations) were justified based on information presented by R. E.
Maerker-[34). Maerker's results are closely duplicated when y - 6.
For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.
Results of the FERRET evaluation of the capsule Y dosimetry are given in Table 6-9 The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1.48 x 10 n/cm2 (E > 1.0 MeV) with an associated I9 uncertainty of 2 8%. Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atos 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.
9225e:1d/102089 6-10
p A summary cf the measured and calculated neutron exposure of capsule Y is j
presented in Table 6-12 along.with the updated exposure evaluations for capsules T and U.
This information is also shown graphically as a function of reactor full power operating time in Figure 6-2.
The agreement between calculation and measurement is good for all exposure parameters in each of the capsules withdrawn to date. The measured data from the early capsules (T and i
U) exceed prediction by approximately 14%, whereas, the data from capsule Y falls roughly 7% below prediction.
In general, the measured data tracks the predicted exposure at the capsule locations quite well and, therefore, th.e calculated values will be used in the assessment of the exposure of the vessel itself.
Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (9.18 EFPY) exposure projections are also provided for an exposure period of 20 EFPY and to end of vessel design life (32 EFPY). The time averaged exposure rates for low leakage fuel cycles were used to perfohn projections beyond the end of the cycle 1 through 10 exposure period.
[.
l-In the calculation of exposure gradients for use in the development of heatup L,
and cooldown curves for the Zion Unit 2 reactor coolant system, exposure projections to 20 EFPY and 32 EFPY were employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14.
In order to access RT vs. fluence trend NDT curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T l
positions were defined by the relations 6' (1/4T) = 6 (Surface) { dpa (1/4T)
}
dpa (Surface) 4' (3/4T) = 4 (Surface) { dpa (3/4T)
}
dpa (Surface)
Using this approach results in the dpa equivalent fluence values listed in Table 6-14.
In Table 6-15 updated lead factors are listed for each of the Zion Unit 2 surveillance capsules. These data may be used as a guide in establishing i
future withdrawal schedules for the remaining capsules.
l l
9225e:1d/102089 6-11
~ -. --
e" i
NOTE: ALL DIMENSIONS ARE IN CENTIMETERS
.j i
3.49 2.63 2.48 1.60.
I M
RADIUS (cm) 0.27
- W 214.56 213.
y 212.85 C
1212.35 FLUX WIRES-N $
C y
V CAPSULE CENTER -
\\ p
+
212.12 FLUX WIRES-
\\ p y
211.85.
C-FLUX WIRES -
C y
211.35 y
r,,,,,,,,,,)
210.85 g,(xxxxxx 210.52-g 209.68 THERMAL SHIELD
+
Figure 6-1 Plan View of a Reactor Vessel Surveillance Capsule.
t 6-12
,__2 s
...m..
_m t.
r
[.
' '." g=Ff(=b i si 4-pH p=hytyeips Q j'l g l Q l l l l l f [
[E llE t
L,3 s.. =
f 7.. -
6.. ;
5.. -
4..
3.. ;
1 I
I I
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I I 2.. l,
f f '
'{,
ll
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l i l
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si i
i si i i i i
e Am i i l l
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ad" i
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s II I
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- 7 I
i
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i it ill I
riiti i
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Calculation
- 3-2.
. I I
I I
1 I
I I
E E
l l I
E I
i
)
.I I
I I I
I I
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i I i
l i
i I
t I
a l
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ii i i t J
l I
I il I
i 10 5 0
1 2
3 4
5 6
7 8
9 10 11 12 OPERATINGTIME(EFPY) l Figure 6-2.
Fast Neutron (E >1.0 Mev) Fluence at the 40 Degree I
Surveillance Capsule Location as a Function of Full j
Power Operating Time i
L 9225e:1d/102089 6-13
_.,. _. _ ~ _ _..
4 I
TABLE 6-1
~
d CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLaurT CAPSULE CENTER I,
$ (E-> 1.0 Mev)
((E > 0.1 Mev) dpa/sec Design Cycle Design Cycle Design Cycle.
Basis 1-10 Basis 1-10 Basis 1-10 40* Capsule 7.54 x 10 5.45 x IO 2.22.x IO 1.63 x IO 1.24 x 10-10 8.80 x 10-U 10' N
N N
4* Capsule 2.49 x IO 1.86 x.10 6.30 x IO 4.71 x 10 3.92 x 10-U 2.92 x 10-"
N D
M N
'[
i F
)
9225e:1d/102089
-8 e
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l 5
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o i
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8 S.
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=
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d
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+
m m
h
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~
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=m o
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+
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9 N
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1 6-15
'1 1
w v
-,e
. -. _...+..,.., _,. _.,,,
[7 7
Y'
.l TA8LE 6-3 1
RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV)
- l MITHIN THE PRESSURE VERSEL NALL
.!j Radius' (cm) o'
_r;i'_,
X X
220.27(I) 1.00 1.00 1.00 1.00 220.64 0.977 0.978 0.979 0.977 221.66 0.884 0.887 0.889 0.885 222.99 0.758 0.762 0.765 0.756 224.31 0.641 0.644 0.648 0.637 225.63 0.537 0.540 0.545 0.534 226.95 0.448 0.451 0.455 0.443 228.28 0.372 0.373 0.379 0.367 229.60 0.309 0.310 0.315 0.303~
230.92 0.255 0.257 0.261 0.250 232.25 0.211 0.212 0.216 0.206 i
233.57 0.174 0.175 0.178 0.169 234.89 0.143 0.144 0.147 0.138 236.22 0.117 0.118 0.121 0.113
-237.54 0.0961 0.0963 0.0989 0.0912 h
238.86 0.0783 0.0783 0.0807-0.0736 240.19 0.0635 0.0632 0.0656 0.0584 241.51 0.0511 0.0501 0.0519 0.0454 242.17(2) 0.0483 0.0469 0.0487 0.0422 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 9225e:1d/102089 6-16
-+
t s.
3
?
i u
t TABLE 6-4 e
RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius (ce) o*
1s' X
X
' 220.27(I}
1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 t
221.66 1.00 0.996 1.00' O.994 222.99 0.965 0.958 0.968
-0.953 224.31 0.916 0.906.
0.91'9 0.898
- 225.63 0.861 0.849 0.865 0.838 226.95 0.803 0.790 0.809 0.777 228.28 0.746 0.732-0.752 0.717 229.60 0.689 0.675 0.695 0.657
'230.92-0.633 0.619 0.640 0.600
~232.25 0.578 0.565 0.586-0.544 I
233.57 -
0.525 0.513 0.534 0.490 234.89 0.474 0.463 0.483 0.437 236.22 0.424 0.414 0.433 0.387 237.54 0.375 0.367 0.385 0.338 238.86 0.328 0.322 0.338 0.291
)
L
-240.19 0.283 0.277 0.292 0.244 241.51 0.239 0.232 0.245 0.196 242.17(2) 0.229 0.220 0.232 0.183 t
[.. --
NOTES:
- 1) Base Metal Inner Radius j
- 2) Base Metal Outer Radius s.
\\I*
I 9225e:1d/102089 6-17 n
L ;.
i/
1 TABLE 6-5
- 1 RELATIVE RADIAL DISTRI8UTIONS OF IRON DISPLACEMENT RATE (dpa) el c.
i:
h.
WITHIN THE PRESSURE VESSEL MALL p
)
L l
a E
Radius fen) c' E
X X
l 220.27II) 1.00 1.00 1.00 1.00 220.64-0.983 0.983 0.984 0.983 221.66 0.913 0.914 0.918 0.915 222.99 0.818 0.819 0.827 0.820 o
224.31 0.728 0.728 0.739 0.730
. 225.63 0.647 0.646 0.659 0.647' 226.95 0.574 0.573
. 0.587 0.573-L 228.28 0.510-0.507 0.523 0.507 229.60:
0.453 0.450 0.466 0.449 f
{
230.92
.0.402 0.399
- 0.414 0.397
-232.25 0.356 0.353 0.368 0.349 233.57 0.315 0.312 0.327 0.307 234.89 0.277 0.275 0.289 0.269 i
236.22 0.243 0.241 0.254 0.233 237.54 0.212 0.210 0.222 0.201 238.86 0.182 0.181 0.192 0.170-l 240.19 0.155 0.154 0.164 0.141 241.51 0.131 0.128 0.137 0.113 242.17(2) 0.125 0.122 0.130 0.106 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius v
t i
4 4
9225e:1d/102089 6-18 s
s
,me.e-ew.
y
..ervn-,
- w., -,..
,,,e-,-
ww-.,r-r.,,
s.-.
q-
h 1
i v.
TABLE 6-6 T
NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Height
Response
Product Yield Material Interest Fraction Ranoo Half-Life (O
]
60 63(n.a)Co 0.6917 E> 4.7 MeV 5.272 yrs Copper Cu
.- Iron Fe (n.p)Mn 0.0582 E> 1.0 MeV 312.2 days 58 58(n.p)Co 0.6830 E> 1.0 MeV 70.90 days Nickel Ni 137 238(n,f)Cs 1.0 E> 0.4 MeV 30.12 yrs 5.99 q
Uranium-238*
U 137 Neptunium-237*
Np237(n.f)Cs 1.0 E) 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum
- Co (n,y)Co 0.0015 0.4ev <E< 0.015 MeV 5.272 yrs Cobalt-Aluminum
- Co '(n,y)Co 0.0015 E'< 0.015 MeV 5.272 yrs l-
- Denotes that monitor is cadmium shielded.
L 1
l-9225e:1d/102089-6-19
y
(
L TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS e
CONTAINED IN,MPSULE Y Irradiation Decay PJ Time Time HQnih XAE DtQ PJ/PMAX (davn)
(days) 12 1973 305.5-
.0.0940
.8 5616 1
1974 394.2 0.1213 31 5585 2
1974 436.4-0.1343 28 5557 3
1974 394.2 0.1213 31 5526 4
1974 219.0 0.0674 30 5496 5-1974 0.0 0.0000 31 5465' 6
1974 0.0 0.0000 30 5435 7
1974 0.0 0.0000 31 5404 8
1974 76.9 0.0237-31.
5373 9
1974 998.0 0.3071 30 5343 10 1974 849.6 0.2614 31 5312 11 1974 2243.2 0.6902 30 5282 12 1974 1073.5 0.3303 31 5251 1-1975 1637.2 0.5037 31 5220 2
1975 1765.6 0.5433 28 5192 3
1975.
1776.8 0.5467 31 5161 4
1975 1550.6 0.4771 30 5131 5
1975 1950.6 0.6002 31 5100 6
1975 452.0 0.1391-30 5070 7
1975 2631.2 0.8096 31 5039 8
1975 2434.3 0.7490 31 5008 9
1975 635.6 0.1956 30 4978 10 1975 2366.4 0.7281 31 4947 11 1975 2499.3 0.7690 30 4917 12 1975 2321.1 0.7142 31 4886
,i 1
1976 925.4 0.2847 31 4855
.)
2 1976 1269.8 0.3907 29 4826
)
3 1976 2764.6 0.8506 31 4795 4
1976 31.4 0.0097 30 4765 5
1976 1754.8 0.5399 31 4734 l
6 1976 1612.0 0.4960 30 4704 7
1976 3093.1 0.9517 31 4673 8
1976 2351.3 0.7235 31 4642 9
1976 1510.4 0.4647 30 4612 10 1976 122.2 0.0376 31 4581 11 1976 3131.5 0.9635 30 4551 12 1976 2522.3 0.7761 31 4520 9225e:1d/102089 6-20
7 1
j TA8LE 6 cont'd
.o IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE Y
]
Irradiation Decay PJ Time Time g
g g
PJ/PMAX (days)
(days) 11
'1977 524.3 0.1613 31 4489 2-1977 0.0 0.0000 28 4461 3
1977 162.8 0.0501 31 4430 4
1977 2645.5 0.8140 30 4400 5
1977 3153.0 0.9702 31 4369 6-1977 2882.8 0.8870 30 4339-7 1977 2762.2 0.8499 31 4308
- 8 1977 3190.4 0.9817 31 4277 9
1977 3227.7 0.9932 30 4247 10 1977 3084.2 0.9490 31
~4216
-11 1977 3062.7 0.9424 30 4186 6
12.
1977 2848.2 0.8764 31 4155 1
1978 2635.6 0.8110 31 4124-2 1978-220.9 0.0680 28 4096 3
1978 0.0 0.0000 31 4065 4
1978
.1248.9 0.3843 30 4035 5-1978 3180.9 0.9787 31 4004 6
1978 3127.9 0.9624 30 3974 7
1978 3122.1.
0.9606 31 3943 8
1978 3234.1 0.9951 31 3912 9
1978 3147.3 0.9684 30 3882 t -
10 1978 3090.1-0.9508 31 3851 l
11 1978 3239.6 0.9968 30 3821 12 1978-3158.3 0.9718 31 3790
'l 1979 2785.2 0.8570 31 3759 2
1979 1362.4 0.4192 28 3731 o
3 1979 312.8 0.0962 31 3700 L
4 1979 947.9 0.2917 30 3670 l'
5 1979 2725.3 0.8385 31 3639 6
1979 2731.8 0.8406 30 3609 7
1979 2780.2 0.8554 31 3578 8
1979 2786.3 0.8573 31 3547 9
1979 2843.1 0.8748 30 3517
'10 1979 1798.7 0.5534 31 3486 l.
11 1979 0.0 0.0000 30 3456 12 1979 0.0 0.0000 31 3425 l
9225e:1d/102089 6-21
~.
~
-TABLE 6 cont'd IRRADIATION HISTORY OF NEUTRON SENSORS ej 9
J CONTAINED IN CAPSULE Y T
I i
Irradiation Decay PJ Time
. Time Month
- y. gar DM1 PJIPMAX (davs)
(davs) 1 1980 1069.0 0.3289 31 3394.
2.
1980 3131.4 0.9635 29 3365 3
3 1980 3175.6 0.9771 31 3334 4
1980 2745.4 0.8448 30 3304 5
1980 91.2 0.0281 31 3273 6
1980 0.0 0.0000 30 3243 l
7 1980 302.1 0.0929 31 3212 l
8 1980 2633.7 0.8104 31 3181 9
1980 2720.5 0.8371 30 3151 10 1980 2997.7 0.9224 31 3120 11 1980.
2572.4 0.7915 30 3090 12
'1980 2509.3 0.7721 31 3059 1
1981 2812.4 0.8653 31 3028 2
1981 2966.1 0.9126 28 3000
- ~
3 1981 2854.6 0.8783 31 2969 4
1981 2906.9 0.8944 30 2939 l
5 1981-2810.3 0.8647 31 2908 6
1981 2962.6 0.9116 30 2878 L
7 1981 2777.4 0.8546 31 2847
.8 1981 2164.0 0.6658 31 2816 9
1981 566.5 0.1743~
30 2786 10 1981 0.0 0.0000 31 2755 L
11-1981 6.3 0.0019 30 2725 12 1981 1485.9-0.4572 31-2694 1
1982 1533.6 0.4719 31 2663-2 1982 342.9 0.1055 28 2635 3-1982 543.7 0.1673 31 2604 4
1982 2120.3 0.6524 30 2574 5
1982 3098.8 0.9535 31 2543
(
6 1982 2071.6 0.6374 30 2513 7
1982 2730.2 0.8401 31 2482
.8 1982 3122.9 0.9609 31 2451 i
9 1982 699.2 0.2151 30 2421 10 1982 795.4 0.2447 31 2390 11 1982 3229.7 0.9937 30 2360 L
12 1982 2945.8 0.9064 31 2329 L
9225e:1d/102089 6-22
i
-TABLE 6 cont'd-
/
'o IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE Y l
Irradiation Decay PJ Time Time Moath Y. tat Dia PJ/PMAX (davn)
(davt)
\\
1 1983 3080.4 0.9478 31 2298 1
2 1983 2557.5 0.7869 28 2270 3
1983 0.0 0.0000 31~
2239 4
1983 0.0-0.0000 30 2209 5
1983 207.5 0.0638 31 2178 6
1983 3001.5 0.9235 30
-2148 7
1983 3066.5 0.9435 31 2117 8
1983 3021.4 0.9297 31 2086 9
1983 3222.9 0.9917 30 2056 10-1983 3227.7 0.9932 31 2025 11 1983 2835.5 0.8725 30 1995 12
-1983 3226.5 0.9928 31 1964 1
1984 2928.0 0.9009 31 1933 o~
2 1984 3226.5 0.9928 29 1904 3
1984 2393.6 0.7365 31 1873 i
4-
-1984 0.0 0.0000 30 1843 5
1984 0.0 0.0000 31 1812 6
.1984 0.0 0.0000 30 1782 7
1984 1793.9 0.5520 31 1751 8
1984 3080.5 0.9479 31 1720 9
1984.
3165.4 0.9740-30 1690 L
10' 1984 3214.7-0.9891 31 1659 L
11 1984 3218.7 0.9904 30 1629 L
12 1984 3174.1 0.9766 31 1598 1
1985 3163.2 0.9733 31 1567 2
1985 3194.7 0.9830 28 1539 I
L
-3 1985 3055.3 0.9401 31 1508 4
1985 3206.3 0.9866 30 1478 5
1985 3065.5-0.9432 31 1447 6
1985 2983.3 0.9179 30 1417 7
1985 2300.5 0.7078 31 1386 8
1985 1637.9 0.5040 31 1355 L
9 1985 182.4 0.0561 30 1325 L
10 1985 0.0 0.0000 31 1294 11 1985 0.0 0.0000 30 1264 12 1985 0.0 0.0000 31 1233 l..
922Ea:1d/102089 6-23 4.
1 TA8LE 6 cont'd IRRADIATION HISTORY Or NEUTRON SENSORS CONTAINED IN CAPSULE Y i
Irradiation Decay P.1 Time Time.
IlDaib 1A&E Mid fall.l3lg (davt) fdavs) 1 1986-0.0 0.0000 31 1202 2
1986' 2169.0 0.6674 28 1174 3
1986 2813.0-0.8655 31 1143 4
1986 3188.0 0.9809 30 1113 i
5 1986 3043.7 0.9365 31 1082 6
1986 2788.8 0.8581 30 1052 7
1986 1737.8 0.5347 31 1021 8
1986 3216.8 0.9898 31 990 9
1986 2838.3 0.6733 30 960 J
10 1986 3217.9 0.9901 31 929 11 1986 3227.3 0.9930 30 899 i
12 1986 3226.5 0.9928 31 868 1
1987 3226.7 0.9928 31 837 2
1987 2870.9 0.8833 28 809 3
1987 2007.2 0.6176 31 778 4
1987 0.0 0.0000 30 748 5
1987 0.0 0.0000 31 717 6
1987 0.0 0.0000 30 687 3
7 1987 0.0 0.0000 31 656 4
8 1987 1937.8 0.59ft 31 625 9
1987 3168.4 0.SD 3 30 595 10 1987 2653.4 0.8164 31 564 11 1987 3151.4 0.9697 30 534 3
12 1987 3049.2 0.9382 31 503 1
1988 3142.8 0.9670 31 472 2
1988 3121.3 0.9604 29 443 3
1988 2920.1 0.8985 31 412 4
1988 2906.9 0.8944 30 382 l
l 5
1988 3082.5 0.9484 31 351 l
6 1988 3176.0 0.9772 30 321 l
7 1988 3214.1 0.9889 31 290 8-1988 2694.2 0.8290 31 259 9
1988 3166.3 0.9742 30 229 l
10 1988 1950.3 0.6001 13 216 L
l L
l:
9225e:1d/102089 6-24 L
i
(
]
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES l'
r
~
Adjusted Measured Saturated Reaction Monitor and Activity Activity Rate Arial Location (dis /see-an)
(dis /see-en)
(RPS/ NUCLEUS)
Cu-63 (n.a) Co-60 5
5 Middle 1.32 x 10 2.78 x 10 5
5 Bottom-Middle 1.36 x 10 2.86 x 10 Average 1.34 x 10 2.82 x 10 4.30 x 10~I7 5
5 Fe-54(n.p) Mn-54 6
6
- O Top 1.10 x 10 2.52 x 10 0
Top-Middle 1.12 x 10 2.56 x 106 6
6 i
Middle 1.08 x 10 2.47 x 10 6
0 l
Bottom-Middle 1.14 x 10 2.61 x 10 6
6 Botton 1.10 x 10 2.52 x 10 Average 1.11 x 10 2.54 x 10 4.04 x 10-15 0
0 Ni-58 (n.p) Co-58 6
7 Top-Middle 4.22 x 10 3.94 x 10 0
7 Middle 4.21 x 10 3.93 x 10 6
7 Botton-Middle 4.31 x 10 4.02 x 10 Average 4.25 x 10 3.96 x 10 5.65 x 10-15 6
7 i
Middle 4.09 x 10 2.31 x 10 1.52 x 10'I4 5
6 L
9225e:Id/102089 6-25
t i
TABLE 6-8 i
NEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd Men'sured Saturated Reaction Monitor and Activity Activity Rate Antal Location (dis /tec-am)
(dit/sec-ami (RPS/NirlfuS)
Np-237(n.f) Cs-137 (Cd)
Middle 3.88 x 10 2.19 x 10 1.33 x 10~I3 0
7 Co-59 (n,y) Co-60 (Cd) 7 7
Top 1.23 x 10 2.47 x 10 Average 1.23 x 10 2.47 x 10 1.61 x 10-12 7
7 5
1 9225e:1d/102089 6-26
TABLE 6-9 M RY OF NEUTRON 00EIMETRY RESULTS TIME AVERACLD EXPotuRE RATES 2
10
$ (E) 1.0 MeV) {n/cm -sec) 5.10 x 10 i gg 2
II
$ (E) 0.1 MeV) {n/cm -sec) 1.64 x 10 i IST.
8.47 x 10~II i 10%
dpa/sec
~
INTEGRATED CAPSULE EXPOSURE 2
I 4 (E) 1.0 MeV) (n/cm )
1.48 x 10 '
i 81 2
I 4 (E) 0.1 MeV) {n/cm )
4.76 x 10 '
i 15%
2.46 x 10-2 i 10%
dpa l
l NOTE: Total Irradiation Time. 9.18 EFPY l
r l
l l.
l.
l-i l
l 9225e:1d/102089 6-27
TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALC 1JLATED
)
REACTION RATES AT THE tuRVEILLANCE CAPtULE CENTER
_i i
Adjusted Reaction Measured calculation Clg 1
4.30x10'I7 4.34x10'II 1.01 Cu-63 (n.s) Co-60 4.04x10-15 3.98x10-15 0.98 Fe-54 (n.p) Mn-54 5.65x10-15 5.49x10-15 0.97 N1-58 (n.p) Co-58 U-238 (n.f) Cs-137 (Cd) 1.52x10*I4 1.77x10-I4 1.16 Np-237 (n f) Cs-137 (Cd) 1.33x10'I3 1.36x10~I3 1.02 Co-59 (n,y) Co-60 (Cd) 1.61x10-12 1.59x10-12 0.99 e
~
i i
6-28 9225e:1d/102089
l TA8LE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE EURVEILLANCE CAPSULE CENTER i
Energy Adjusted Flux Energy AdjustedFlux iglg3 M E
fdGgg (Nav) ind;gZ,lagl fgggg (Nav) 1 1.73x10 6.45x10 28 9.12x10'3 7.63x10' I
6 2
1.49x10 1.49x10 29 5.53x10~3 9.16 10' I
7 3
1.35x10 5.49x10 30 3.36x10~3 2.89x10' I
7 4
1.16x10 1.23x10' 31 2.84x10~3 2.79x10' I
5 1.00x10 2.71x10' 32 2.40x10-3 2.75x10' I
6 8.61x10 4.63x10' 33 2.04x10~3 8.24x10' 0
7 7.41x10 1.07x10 34 1.23x10-3 8.45x10' 0
9 8
6.07x10 1.54x10' 35 7.49x10-4 8.71x10' 0
9 4.97x10 3.10x10 36 4.54x10 8.98x10' 0
I d
10 3.68x10 3.68x10' 37 2.75x10 9.70x10' 0
4 10 O
11 2.87x10 6.92x10' 38 1.67x10'4 1.23x10 0
0 9
10 12 2.23x10 7,9g,jg 39 1.01x10 1.03x10 l
13 1.74x10 9.78x10' 40 6.14x10-5 9.97x10' f
0 14 1.35x10 9.21x10' 41 3.73x10-5 9.50x10' 0
15 1.11x10 1.49x10 42 2.26x10-5 8.99x10' 0
10 16 8.21x10'I 1.52x10 43 1.37x10-5 8.54x10' 10 l
17 6.39x10'I 1.47x10 44 8.32x10-6 8.11x10' 10 18 4.98x10-I 1.04x10 45 5.04x10
7.67x10' 10 19 3.88x10-I 1.27x10 46 3.06x10
7.29x10' 10 20 3.02x10'I 1.66x10 47 1.86x10
6.85x10' 10 21 1.83x10"I 1.51x10 48 1.13x10-6 5.91x10' 10 22 1.11x10'I 1.23x10 49 6.83x10~7 5.73x10' 10 23 6.74x10-2 9.58x10' 50 4.14x10-7 8.60x10' 24 4.09x10-2 6.48x10' 51 2.51x10-7 9.47x10' 10 25 2.55x10-2 6.21x10' 52 1.52x10-7 1.05x10 10
-26 1.99x10-2 4.22x10' 53 9.24x10
3.20x10 27 1.50x10-2 6.32x10' NOTE: Tabulated energy levels represent the upper energy of each group.
9225e:1d/102089 6-29 l
l
~ _.
4 l
TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED j
ExpotuRE trvEtt FOR tuRVEILLANCE CAPSULE Sl CAPSULE Y Calculated Maksured flji l
l II 2
1.58 x 10 '
1.48 x 10 1.07 6(E) 1.0 MeV) (n/cm )
I I
t 4.71 x 10 '
4.76 x 10 '
0.99 4(E) 0.1 MeV) (n/ca }
2.55 x 10-2 2.46 x 10-2 1.04 i
dpa CAPSULE T I
Calculated Measured flji
- l t
- f 2
I8 I8 4(E) 1.0 MeV) (n/cm )
6.70 x 10 7.81 x 10 0.86 4
I 2
II 2.51 x 10 '
O.80 6(E) 0.1 MeV) (n/cm )
2.00 x 10 1.08 x 10-2 1.29 x 10-2 0.84 dpa l
l CAPSULE U
+
Calculated Measured flji 2
I8 18 4(E) 1.0 MeV) (n/cm )
2.34 x 10 2.73 x 10 0.86 2
18 I8 4(E) 0.1 MeV) (n/cm )
6.98 x 10 8.56 x 10 0.82 3.78 x 10-2 4.43 x 10~3 0.85 dpa 9225e:1d/102089 6-30
j TA8LE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRrtttlRE VESSEL CLAD / RASE METAL INTERFACE AIIKITHAL ANCLE r
O*
N A
t.18 EFPY 18 18 18 16 4(E) 1.0 MeV) 1.73 x 10 2.85 x 10 3.48 x 10 5.34 x 10 l
I (n/cm2) i 6(E) 0.1 MeV) 4.38 x 10 7.13 x 10 8.93 x 10 1.47, x 10 (n/cm2) 5.63 x 10 '
8.70 x 10
~
~
~
~
dpa 2.83 x 10 4.58 x 10 20.0 EFPY
$(E) 1.0 MeV) 3.47 x 10 5.85 x 10 7.20 x 10 1.11 x 10 l
O (n/cm2) 4(E) 0.1 MeV) 8.77 x 10 1.47 x 10 1.85 x 10 2.97 x 10 (n/cm2)
~
~
~
dpa 5.67 x 10 9.41 x 10 1.16 x 10 1,81 x 10
~
32.0 EFPY 6(E) 1.0 MeV) 5.41 x 10 9.18 x 10 1.13 x 10 1,74 x 10 (n/cm2) 4(E) 0.1 MeV) 1.36 x 10 2.30 x 10 2.91 x 10 4,65 x 10 (n/cm2)
~
~
~
dpa 8.83 x 10 1.47 x 10 1.83 x 10 2.84 x 10
~
9225e:1d/102089 6-31
TAetE 6-14 NEUTRON EXPoaser VALUES FOR USE IN THE C_F_mFRATION OF NEATUP/(DOLD(BOI CURVES 20 EFPY l
NEUTHON FLUENCE (E > 1.0 MeV) SLOPE dea SLDPE 2
2 (n/cm )
(equivalent n/cm )
l Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T l
N N
U II 0*
3.47 x IO 1.83 x IO 3.78 x 10 3.47 x 10 "
2.22 x 10 "
8.05 x 10 I8 N
U N
N N
15*
5.85 x 10 3.11 x IO 6.44 x 10 5.85 x IO 3.74 x IO 1.34 x IO 30*
7.20 x 10 "
3.57 x 10" 8.13 x IO 7.20 x 01 "
4.70 x 10 "
1.74 x 10" N
45' 1.11 x 10 5.84 x 10 1.16 x 10 1.11 x 10 7.10 x 10 "
2.45 x 10 "
D I8 32 EFPY b
NEUTRON FinnrulT (E > 1.0 MeV) SLOPE lba_5 LEE 2
2 (n/cm )
(equivalent n/cm )
I Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 0*
5.41 x 10 "
2.86 x 10 "
5.89 x 10 "
5.41 x 10 3.46 x 10 1.26 x 10 "
I8 I8 15' 9.18 x 10 4.88 x 10 "
1.01 x 10 9.18 x 10 5.87 x 10 2.11 x 10 "
0 I8 0
I8 N
1.13 x 01" 7.37 x IO 2.73 x 10 "
N 30*
1.13 x 10 6.07 x 10 1.27 x IO I8 D
I8 45*
1.74 x 10 9.15 x 10 1.82 x 10 1.74 x 10 1.11 x 10 3.84 x 10 i
i 9225e:1d/102089 I
)
TABLE 6-15 r
UPDATED LEAD FACTORS FOR ZION UNIT 2 SURVEILLANCE CAMULES Cansula Lead Facter U
Mithdrawn I
T Nithdrawn I
Y Mithdrawn X
2.97 l
N 1.01 V
1.01 Z
1.01 S
1.01 i
9 9225e:1d/102089 6-33
~
v-a w
.,a
,.n- - -.
.e a
-a v
a.,
,- - - +,,,-
r.----
__ -~ - - =
'O.
b "c.
SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE
[
The following removal schedule meets ASTM E185-82 and is recommended for.
io' future capsules to be removed from the Zion Unit 2 reactor vessel:
i l
Capsule Estimated i-Location Lead Fluence 2
(;
Capsule' (deg.)
Factor Removal Time (a)
(n/cm )
I 18 l'
U 140 2.97 1.27 (Removed) 2.73 x 10 T
40 2.97 3.56(Removed) 7.81 x 1018(b)
Y-320 2.97 9.18 (Removed 1.48 x 1019(c)
[
19 X
220 2.97 15 2.42 x 10
_ W 184 1.01 Standby
-[
V 176 1.01 Standby I
Z 356 1.01 Standby S
4 1.01 Standby t
t (a) Effective full power years from plant startup.
I (b) Approximate fluence at 1/4 thickness reactor vessel wall at end of life.
(c) Approximate fluence at reactor vessel inner wall at end of life.
1
,s e
p w.n m so 73
o
- d
8.0 REFERENCES
1..
S. E. Yanichko and D. L. Lege, " Commonwealth Edison Co. Zion Unit No. 2 Reactor Vessel Radiation Surveillance Program", WCAP-8132, May 1973.
o 2.
Perrin, farmelo, Jung and Fromm, " Zion Nuclear Plant Reactor Pressure Vessel Surveillance Program: Unit No. 1 Capsule T, and Unit No. 2 Capsule U", BCL-585-4 March 1978.
3.
E. B. Norris, "Resetor Vessel Material Surveillance Program for Zion Unit No. 2 Analysis of Capsule T", SwRI Project No. 06-6901-001, July 1983.
4.
Code of Federal Regulations, 10CFR50, Appendix G. " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.
5.
Regulatory Guide 1.99, Proposed Revision 2, " Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, February 1986.
6.
R. G. Soltesz, 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.
7.
"0RNL RSCI Data Library Collection DLC-76, SAILOR Coupled Self-Shielded,
~47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".
8.
S. Altomore, et al., " Core Physics Characteristics of the Zion Station Nuclear Plants (Units 1 and 2, Cycle 1), WCAP-7675, November 1971.
(Proprietary) 4 I
l l
s m.ne m eio 8-1 l
9.
T. L. Wheeler, et al., " Core Physics Characteristics of the Zion Station Nuclear Plant (Unit 2, Cycle 2)," WCAP-8881, November 1976.
(Proprietary) 4 10.
T. L. Wheeler, et al., " Core Physics Characteristics of the Zion Station Nuclear Plant Unit 2 (COM), Cycle 3," WCAP-9246, December 1977.
[
(Proprietary) 11.
W. F. Staley, et al., " Core Physics Characteristics of the Zion Station Nuclear Plant Unit 2 (COM), Cycle 4 " WCAP-9458, January 1979.
(Proprietary) 12.
C. A. Grier, et al., " Core Physics Characteristics of the Zion Station Nuclear Plant Unit 2 (COM), Cycle 5 " WCAP-9687, April 1980.
(Proprietary) 13.
D. L. Maret, et al., " Core Physics Characteristics of the Zion Station Nuclear Plant (Unit 2, Cycle'6)," WCAP-9959, August 1981.
(Proprietary) 14.
D. J. Wanner, et al., " Core Physics Characteristics of the Zion Station Nuclear Plant (Unit 2, Cycle 7)," WCAP-10282, February 1983.
(Proprietary) 15.
D. K. Lee, et. al., " Nuclear Design Report for Zion Unit 2, Cycle 8,"
NFSR-0024, May 1984.
(CECO. Proprietary) l 16.
D. K. Lee, et al., " Nuclear Design Report for Zion Unit 2, Cycle 9,"
NFSR-0037, December 1985.
(CECO. Proprietary) l 17.
D. K. Lee, et al., " Nuclear Design Report for Zion Unit 2, Cycle 10,"
NFSR-0052, December 1985.
(CECO. Proprietary) 18.
ASTM Designation E482-82, " Standard Guide for Application of Neutron L
Transport Methods for Reactor Vss.31 Surveillance", in ASTM Standards, l'
Section 12, American Society for Testing and Materials, Philadelphia, l
PA, 1984.
I l
l
>$47 ne:7esio 8-2
,o e
19.
ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM t
Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984, 20.
ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)",
in ASTM Standards, Section 12, American Society for Testing end Materials, Philadelphia, PA, 1984.
21.
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.
22.
ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
23.
ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphie, PA, 1984.
24.
ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
i 1
1 25.
ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron l
Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
26.
ASTM Designation E264-82, " Standard Method for Determining fast-Neutron l*
Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
w mnu io 8-3 1
I o
c 27.
ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, e
PA, 1984.
28.
ASTM Designation E523-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
29.
ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
30.
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.
31.
ASTM Designation E1005-84, " Standard Method for Application and Analysis l
of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
32.
F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford I
Engineering Development Laboratory, Richland, WA, September 1979.
33.
W.- N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative l
Method of Neutron Flux Spectra Determined by Foil Activation, l
AFWL-TR-67-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
34.
EPRI-NP-21BB, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al., 1981.
i no, nome io 8-4
,