ML20058C002
| ML20058C002 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/31/1990 |
| From: | Albertin L, Shaun Anderson, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20058C000 | List: |
| References | |
| WCAP-12685, NUDOCS 9010310180 | |
| Download: ML20058C002 (86) | |
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HESTINGHOUSE CLASS 3 WCAP-12685 ANALYSIS OF CAPSULE U FROM THE COMMNHEALTH EDISON COMPANY BRAIDWOOD UNIT 1 REACTOR VESSEL
)
RADIATION SURVEILLANCE PROGRAM E. Terek S. L. Anderson L. Albertin August 1990 i
Hork Performed Under Shop Order BMVP-106 Prepared by Westinghcuse Electric Corporation for the Commonwealth Edison Company N
Approved by:
T. A. Meyer, Mandger Structural Materials and Reliability Technology HESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division-P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 e 1990 Hestinghouse Electric Corp.'
PREFACE This report has been technically reviewed and verified.
1
}
Reviewer m
69' Sections I through 5, 7, and 8 Ni K. Ray
[o Ar a l,
Section 6 E. P. Lippincott y I
i i
4 i
l 0083D:10/081490 i
TABLE OF CONTENTS SElian Titie EASA.
l t
1.0
SUMMARY
OF RESULTS 1-1
2.0 INTRODUCTION
2-1 l
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE U 5-1 i
5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-3 4
5.3 Tension Test Results 5-6 5.4 Compact Tension Tests 5-6 j
6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 1 6.1 Introduction 1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-7 i
7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 i
8.0 REFERENCES
8-1 i
II
-l
-06830:10/100190 11
l LIST OF ILLUSTRATIONS Eiaure Title hgg 4-1 Arrangement of Surveillance Capsules in the Braidwood 4-6 I
Unit 1 Reactor Vessel 4-2 Capsule U Diagram showing location of specimens, Thermal 4-7 Monitors and Dosimeters l
5-1 Charpy V-Notch Impact Properties for Braidwood Unit 1 Reactor 5-14 Vessel Shell Forging 490867-1/49C813-1 (Tangential Orientation) 5-2 Charpy V-Notch Impact Properties for Braidwood Unit 1 Reactor 5-15 Vessel Shell Forging 490867-l!*9C813-1 -(Axial Orientation) 5-3 Charpy V-Notch Tr.; ct Properties for Braidwood Unit 1 Reactor 5-16 t
Vessel Held Metal 5-4 Charpy V-Notch Impact Properties for Braidwood Unit 1 Recctor 5 Held Heat Affected Zone Metal
)
- l 5-5 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit'l 5-18 Reactor Vessel Shell Forging 490867-1/49C813-1 (Tangential Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for Braidwocd Unit 1-5-19 Reactor Vessel Shell Forging 49D867-1/49C813-1 (Axial Orientation)
I 5-7 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 1 5-20 Reactor Vessel Held Metal 5-8 Charpy Impact Specimen Fracture. Surfaces for Braidwood Unit 1 5-21 Reactor Vessel Held Heat Affected Zone (HAZ) Metal 0683D:1D/092590 111 l
LIST OF ILLUSTRATIONS. (Cont)
Fiaure Title PJLER 5-9 Photographs of Capsule U Specimens (a) ET11 and (b) ET14 5-22 5-10 (a) Photograph of Capsule U Specimen EH2,-and (b) Photograph 5-23 of Capsule U Specimen EH14 i
5-11 Tensile Properties for Braidwood Unit 1 Reactor Vessel Shell 5-24' Forging 490867-1/49C813-1 (Tangential Orientation) 5-12 Tensile Properties for Braidwood Unit 1 Reactor Vessel.Shell-5,
Forging 49D867-1/49C813-1 (Axial Orientation) 5-13 Tensile Properties for Braidwood Unit 1 Reactor Vessel Held 5-26 Metal 5-14 Fractured Tensile Specimens from Braidwood Unit 1 Reactor 5-27
)
Vessel Shell forging 490867-1/49C813-1 l-(Tangential Orientation) l 5-15 Fractured Tensile-Specimens from Braidwood Unit.1 Reactor-5-28 Vessel Shell Forging 490867-1/49C813-1 (Axial Orientation)-
5-16 Fractured Tensile Specimens from BraidwNd Unit 1 Reactor 5-29 Vessel Held-Metal 5-17 Typical Stress-Strain Curve for Braidwood Unit 1 5-30 Shel1 Forging 490867-1/49C813-1~ Tension Specimens 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 6-2 Core Power 01stributions used in Transport Calculations 6-14 for Braidwood Unit 1 0683D:lD/100190 iv
i LIST OF TABLES Idle Ittle Pne 4-1 Chemical Composition and Heat Treatment of the Braidwood Unit 14-3 j
Reactor Vessel Surveillance Materials 1
4-2 Chemical Composition of Braidwood Unit 1 Capsule U Irradiated 4-4 1
Charpy Impact Specimens i
4-3 Chemistry Results from the NBS Certified Reference Standards 4-5 1
5-1 Charpy V-Notch Impact Data for the Braidwood Unit 1 5-7 Forging 49D867-1/49C813-1 Irradiated at 550*F, 18 Fluence 3.79 x 10 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Braidwood Unit 1 Reactor 5-8 Vessel Held Metal and HAZ Metal Irradiated at 550'F, Fluence I0 3.79 x 10 n/cm2 (E > 1.0 MeV) 5-3 Instrumented Charpy Impact Test Results for the Braidwood 5-9 Unit 1 Shell forging 490867-1/49C813-1 Irradiated at 550*F, 18 Fluence 3.79 x 10 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results for the Braidwood 5-10 Unit i Held Metal and HAZ Metal Irradiated at 550'F, Fluence 3.79 x 1018. /cm2 (E > 1.0 MeV) n 18 5-5 Effect of Irradiation to 3.79 x 10 n/cm2 (E > 1.0 MeV) 5-11 at 550*F on Notch Toughness Properties of Braidwood Unit 1 Reactor Vessel Surveillance Materials 06830:10/092590 v
l
LIST OF TABLES (Cont)
Table Title EAge 5-6 Comparison of Braidwood Unit 1 Surveillance Material-5-12 30 ft-lb Transi. tion Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-7 Tensile Properties for Braidwood Unit 1 Reactor Vessel 5-13 18 2:
Surveillance Material Irradiated at 550'F to 3.79 x 10 n/cm (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Parameters at the
~ 6-15 Surveillance Capsule Center 6-2 Calculated Fast Neutron Exposure Parameters at the 6-16 Pressure-Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux 6-17 (E > 1.0 MeV) within the Pressure Vessel Hall 6-4 Relative Radial Distributions of Neutron Flux 6-18 (E > 1.0 MeV) within the Pressure Vessel Hall 6-5 Relative Radial Distributions of Iron Displacement Rate 6-19 (dpa) within the Pressure Vessel Hall i
6-6 Nuclear Parameters for Neutron Flux Monitors 6-20 6-7 Irradiation History of Neutron Sensors' Contained in Capsule U 6-21 6-8 Measured Sensor Activities and Reaction Rates 6-22 6-9 Summary of Neutron Dosimetry Results 6-24 0683D:10/092590 vi
1 i
LIST OF TABLES (Cont) l Table Title EASA i
6-10 Comparison of Heasured and Ferret Calculated Reaction Rates 6-25 4
at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance Capsule 6-26 l
Center 6-12 Comparison of Calculated and Measured Exposure Levels for 6-27 Capsule U 6-13 Neutron Exposure Projections at Key Locations on the 6-28 Pressure Vessel Clad / Base Metal Interface for Braidwood Unit 1 6-14 Neutron Exposure Values for use in the Generation of 6-29 Heatup/Cooldown Curves 6-15 Updated Lead Factors for Braidwood Unit 1 Surveillance Capsules 6-30 I
i i
06830:10/081490 vii.
i I
i SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule U, the first capsule to be removed from the Commonwealth Edison Company Braidwood Unit i reactor pressure vessel, led to the following conclusions:
o The capsule received an average fast neutroc fluence (E > 1.0 MeV) 18 2
of 3.79 x 10 n/r.m af ter 1.10 EFPY of plant operation, o
Irradiation of t'ie reactor vessel icwer shell forging 490867-1/49C813 l Charpy speci mns to 3.79 x 10 n/cm2 (E >
18 1.0 MeV) resulted in a 30 and 50 ft-lb transition temperature increase of 5 and 10*F, respectively, for specimens oriented i
parallel to the major working direction (tangential orientation) and no transition temperature increase for specimens oriented normal to the major working direction (axial orientation).
18 2
o The weld metal Charpy specimens irradiated to 3.79 x 10 n/cm (E > 1.0 MeV) resulted in a 30 and 50 ft-lb transition temperature increase of 10*F.
This results in a 30 ft-lb transition temperature of -10*F and a 50 f t-lb transition temperature of 35'F for the weld metal.
o Irradiation of the reactor vessel weld HAZ metal Charpy specimens to 18 2
3.79 x 10 n/cm resulted in a 30 ft-lb transition temperature increase of 20*F and a 50 ft-lb transition temperature increase of 35'F.
o The average upper shelf energy of the lower shell forging 490867-1/49C813-1 showed no decrease in energy af ter irradiation to 18 2
3.79 x 10 n/cm for specimens oriented parallel to the major working direction (tangential orientation) and a decrease of 14 f t-lb for specimens oriented normal to the major working direction (axial or'entation).
The weld metal showed no decrease in upper 0683D:10/081490 1-1 1
s 1
18 2
shelf energy after irradiation to 3.79 x 10 n/cm Both i
materials exhibit a more than adequate upper shelf energy level for.
continued safe plant operation and are expected to maintain an upper shelf-energy of no less than 50 lb-ft throughout the life of the vessel as required by 10CFR50, Appendix G.
o The surveillance capsule test-results do not indicate any significant changes in the RT values'of the reactor vessel ~
NDT surveillance material, o
The calculated end-of-life (32 EFPY) maximum neutron' fluence (E >-
I 1.0 MeV) for the Braidwood Unit i reactor vessel clad / base metal interface is as follows:
I9 2
Vessel inner radius - 3.03 x 10 n/cm I9 2
Vessel 1/4 thickness - 1.66 x 10 n/cm 18 2
Vessel 3/4 thickness - 3.57 x 10 n/cm The end-of-life (32 EFPY) maximum neutron fluence values presented above are expected to drop with the implementation of a low leakage fuel management program.
i
.i l
06830:10/092590 1-2.
i SECTION 2.0 INTRODUCTION This report presents the results of the examination of Capsule V, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Braidwood Unit i reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Braidwood Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation I
mechanical properties of the reactor vessel materials are presented in WCAP-9807 " Commonwealth Edison Company Braidwood Station Unit No. 1, Reactor Vessel Radiation Surveillance Program" by Yanichko and Singer.U3 The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E-185-73, " 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 "U" from the reactor and its shipment to the Westinghouse Science and Technology Center where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed at the Westinghouse Science and Technology Center Hot Cell.
This report summarizes the testing of and the postirradiation data obtained from surveillance Capsule "U" removed from the Braidwood Unit I reactor vessel and discusses the analysis of these data.
06830:10/092590 2-1 l
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in a
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 SA 508 Class 3 (base material of the Commonwealth Edison Company Station Braidwood Unit i reactor pressure vessel lower shell forging) 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 have been presented in " Protection Against Nonductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code.
The method uses 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 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 (axial) to the major working direction of the material.
The RT of a given material is used to index that material to a NDT reference stress intensity factor curve (K curve) which appears in IR Appendix G of the ASME Code.
The K curve is a lower bound of dynamic, IR crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K IR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
0683D:10/082890 3-1
1 I
RT and, in turn, the operating limits of nuclear power plants can be NDT j
adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program, in which a surveillance capsule is l
periodically removed from the operating nuclear reactor and the encapsulated I
specimens are tested.
The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RT for radiation embrittlement.
This adjusted RT NDT NDT (RT initial + ARTNDT) is used to index the material to the KIR E
NDT curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials, i
06830:1D/082890 3-2
SECTION 4.0 DESCRIPTION OF PROGRAM 1
Six surveillance capsules for monitoring the effects of neutron exposure on-the Braidwood Unit i reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup.
The six capsules were positioned in the reactor vessel between the neutron shield pads and the vessel wall as shown in Figure 4-1.
The vertical center of the capsules is opposite the vertical center of the core.
Capsule U was removed after 1.10 effective full power years 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 forging i
490867-1/49C813-1 and weld metal representative of the intermediate to lower shell beltline weld seam 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 forging 490867-1/49C813-1 of the representative weld.
The chemical composition and heat treatment of the survelliance material is presented in Table 4-1.
The chemical analyses reported in Table 4-1 were obtained from unirradiated material used in the survelliance program.
In addition, a chemical analysis using Inductively Coupled Plasma Spectrometry (ICPS) was performed on irradiated specimens from forging 490867-1/49C813-1 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 forging.
Test specimens represent material taken at least one forging thickness from the quenched end of the forging.
Base metal Charpy V-notch impact and tension specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction of the forging (tangential orientation) and also normal to the major working direction (axial orientation).
Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding 0683D:10/092590 4-1
direction.
The Compact Tension test specimens in Capsule V are from the lower shell course forging and were machined in both tne axial and tangential orientations.
Thus, the simulated crack in the specimen will propogate normal and parallel to the major working direction of forging 490867-1/49C813-1.
Compact Tension Test specimens from the weld metal were machined normal to the teld direction with the notch oriented in the direction of the weld. Thus, the simulated crack in the specimen will propogate parallel to the weld direction. All CT specimens were fatigue precracked according to ASTM E399.
Capsule U contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15% colbalt (cadmium-shielded and unshielded).
In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were contained in the capsule.
Thermal monitors made from the two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule.
The composition of the two alloys and their melting points are as follows:
2.5% Ag, 97.5% Pb Melting Point:
579'F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Helting Point:
590*F (310*C) l The arrangement of the various mechanical specimens, dosimeters and thermal I
monitors contained in Capsule U are shown in Figure 4-2.
1 0683D:lD/081490 4-2
TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE BRAIDHOOD UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS (1) 1 Chemical Composition (wt%)
Lower Shell Forging Element 49DB67-1/49CB13-1 Weld Metal L
I C
.20
.066 Mn 1.33 1.44 P
Js07
.015 S
006 012 Si
.28
.48 NI
.73
.67 Mo
.52 44 Cr
.11
.10 Cu
.03
.04 Al
.018
.004 Co
.011
.011 Pb
.0003
.0006 W
.005
.010 Tl
.005
.007 Zr
.005
.003 V
.01
.005 Sn
.008
.005 As
.008
.004-Cb
.005
.004
.0096
.013 N2 8
.0001
.0007 HEAT TREATMENT HISTORY-Temperature Time Material
(*P)
(hr)
Cooling Austenitizing:
1600 1652 0%
Water quenched Lower Shell Forging Tempered:
490867 1/
1202 1229 7%
Air cooled 49C8131 Stress Relief:
1100 1150 12 %
Fumace-cooled Weidment 1100 1150 12 %
Fumace cooled 0683D:10/080890 4-3
TABLE 4-2 i
CHEMICAL COMPOSITION OF BRAIDWOOD UNIT 1 CAPSULE U IRRADIATED CHARPY IMPACT SPECIMENS t
Chemical Composition (wt.%)
Specimen No.
Metal [L-6 EW-4 EW-15 EW-1 EW-2 EW-3 EW-5 EW-6 EW-7 EW-8 EW-9 EW-10 EW-Il EW-la EW-13 EW-14 Fe (Matrix Element:
Remainder by Difference) l Mn 1.300 1.470 1.490 Cr 3.118 0.084 0.086 Ni
.0.746 0.666 0.666 0.723 0.709 0.728 0.699 0.751 0.683 0.673 0.668 0.686 0.616 0.651 0.598 0.656 Mo 0.574 0.444 0.444 Co
<0.01
<0.01
<0.01 Cu 0.052 0.035 0.033 0.034 0.035 0.034 0.035 0.035 0.031 0.032 0.029 0.029 0.034 0.033.0.033 0.031 P
<0.01
<0.01
<0.01 V
<0.01
<0.01
<0.01 C
0.207 0.065 0.063 S
0.005 0.013 0.013-Si 0.298 0.495 0.488 Analyses-Method of Analysis
- Metals ICPS, -Inductively Coupled Plasma Spectrometry.
Carbon EC-12, LECO Carbon Analyzer.
' Sulfur Combusion/ titration Silicon Dissolution /gravinetric l
4-4
. 06830:1D/082390 i
l i
i 1
l TABLE 4-3 CHEMISTRY RESULTS FROM THE NBS CERTIFIED REFERENCE STANDARDS Material ID Low Alloy Steel: NBS Certified Reference Standards f
NBS 361 NBS 362 i
Certified Measured (a)
Certified Measured (a)
Metals Concentration in Weight Percent Fe 95.60 (matrix) 95.30 (matrix)
Mn 0.660 0.643 1.040 1.010 Cr 0.694 0.668 0.300 0.289 Ni 2.000 2.030 0.590 0.591 Mo 0.190 0.205 0.068 0.054 Co 0.032 0.035 0.300 0.298 Cu 0.042 0.046 0.500 0.505 P
0.014 0.018 0.041 0.039 Y
0.011 0.007 0.040 0.044 1
C 0.383 0.160 0.160/0,159 S
0.014 0.036 0.037 Si 0.222 0.240 0.390 0.397 Material ID Low Alloy Steel: NBS Certified Reference Standards i
NBS 363 Certified Measured (a) i Metal s Concentration in Weight Percent Fe 94.4L NA Mn 1.500 NA Cr 1.310 NA Ni 0.300 0.307 Mo 0.028 NA Co 0.048 NA i
Cu 0.100 0.100 P
0.029 NA V
0.310 NA C
0.620 NA S
0.0068 NA
- 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, S and Si.
1 4-5
O' REACTOR VE888L CORE 8ARREL NElffRON PAD (301.5') Z CAPSud U (68.6')
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68.5' 61' i
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I REACTOR l
VESSEL 180' l
i PLAN VIEW -
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EWVATION VIEW Figure 4-1.
Arrangement of Surveillance Capsules in the Braidwood. Unit 1 Reactor Vessel l
06830:10/080890 4-6
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LEGENO. EL LOWER SHELL FORGING 490867 It49C8131 (TANGENTIAL)
ET LOWER SHELL FORGING 4908671149C813-1 (AXIAll EW WELD METAL EH HEAT.AFFECTE4 ZONE MATERIAL j
l
.' R/
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T8 ftMIS CHAAPTS DeMPTS CNAAPf3 M3 MTS tetNCTS CetNCTS TENett8,
113 (TIS ELil 1712 (L12 (11 RLt (f4 Ett gT3 EL3 gg3
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(W2 EH2 432 (L2 tq g(j g 1111 EL11 (f4 EL8 (f5 EL5 (T2 (L2 114 ff3 tif tif (T2 IWt (X) tot (113 EL13 !!!0 (L10 (T7 (L? (14 EL4 til (Lt gy g. i l i 1 sr<- 7 SI q i APERTURE Figure 4-2. Capsule U Diagram Showing Location i CARD } pf.3,
- S of Specimens, Thermal p-
'Also Avaliable On Monitors and Dosimeters i Aperture (.ard M103/0IO~0I 4 i.
- i i l SECTION-5.0 i TESTING OF SPECIMENS FROM CAPSULE U i 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Hestinghouse Science and Technology Center with consultation by Westinghouse Power Systems personnel. - Testing was performed in accordance with 10CFR50, Appendices G and H,[2] ASTM Specification. 1 I63 E185-82 , and Westinghouse Procedure MHL 8402, Revision 1 as modified t'y Hestinghouse RMF Procedures 8102, Revision 1 and 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 U3 against the master list in HCAP-9807 No discrepancies were found. Examination of the two low-melting point 304'C (579'F) and 310*C (590'F) eutectic alloys indicated no melting of either type of thermal monitor._ 8ased on this examination, the maximum temperature to which the test specimens were exposed was less than 304*C (579'F). The Charpy impact tests were performed per ASTM Specification E23-88I73 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology Model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy _l energy (E ). From the load-time curve, the load of general yielding' D (PGY), the time to general 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 fracture load (P ), and the load at which fast fracture terminated is F identified as the arrest load (P )* A i i 1 0683D:lD/082890 5-1 l
The energy at maximum load (E ) was determined by comparing the energy-time i M record and the load-time record. The energy at maximum load is roughly j equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E ) is the difference p between the total energy to fracture (E ) and the energy at maximum load. D The yield stress (o ) is calculated from the three-point bend formula. y The flow stress is calculated from the average of the yield and-maximum loads, also using the three-point bend formula. Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-89[8) j The lateral expansion was measured using a dial. gage-rig similar to that shown { in the same specification. Tension tests were performed on a 20,000-pound Instron, split-console test I93 and E21-79 (1988)D0] machine (Model 1115) per ASTM Specification E8-89 and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test. i Deflection measurements were made with a linear variable displacement transducer (LVDT) extansometer. The extensometer knife edges were: spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as'. Class B-2 per ASTM E83-85UU. f Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All-tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature: Chromel-alumel thermocouples were inserted in shallow holes in the center and 0683D:1D/082890 5-2
I each end of the gage section of a dummy specimen and in each grip. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was i developed over the range of room temperature to 550'F (288'C). The upper grip t$as used to control the furnace temperature. During the actual testing the I grip temperatures were used to obtained desired specimen temperatures. Experiments indicated that this method is accurate to 2*F. The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true C *ess at fracture) and percent reduction in area was computed using the final diameter measurement. 5.2 Charov V-Notch Imoact Test Results The results of Charpy V-notch impact tests performed on the various materials 18 contained in Capsule U irradiated to 3.79 x 10 n/cm2 (E > 1.0 MeV) are presented in Tables 5-1 through 5-5 and are compared with unitradiated results in Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-6. Irradiation of Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major working direction (tangential orientation) of I8 the reactor vessel lower shell forging 50D102-1/50D97-1 to 3.79 x 10 2 (E > 1.0 MeV) at 500*F (Figure 5-1) resulted in the following; a 30 n/cm ft-lb transition temperature increase of 5*F which results in a 30 ft-lb transition temperature of -60*F and a 50 ft-lb tr3nsition temperature increase of 10'F which results in a 50 ft-lb transition temperature of -30*F. 1 06830:1D/100990 5-3 l l
~ l The average upper shell energy (USE) of Charpy specimens oriented with the longitudinal axis of the specimen parallel;to the major working direction (tangential orientation) of the lower shell forging 500102-1/50097-1 resulted i 18 in no energy decrease after irradiation to 3.79 x-10 n/cm2 (E > 1.0 MeV) at 550'F. This results in an average USE of 168 ft-lb (Figure 5-1). Irradiation of Charpy specimens oriented with the longitudinal axis of the specimen normal to the major working direction (axial orientation) of the 18 2 reactor vessel lower shell forging 500102-1/50097-1 to 3.79 x 10 n/cm (E > 1.0 MeV) at 550'F (Figure 5-2) resulted in no 30 and 50 ft-lb transition temperature increases. This results in a 30 ft-lb transition temperature of -45*F and a 50 ft-lb transition temperature of -15'F. The average upper shelf energy (USE) of Charpy specimens oriented with the longitudinal axis of the specimen normal to the major working direction (axial i orientation) of the lower shell forging 50D102-1/50097-1 resulted in an 18 average USE decrease of 14 ft-lb after irradiation to 3.79 x 10 n/cm2 (E > 1.0 MeV) at 550*F. This results in an USE of 138 ft-lb (Figure 5-2). Irradiation of the reactor vessel core region weld mett' Charpy specimens to 18 3.79 x 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-3) resulted in 30 and 50 f t-lb transition temperature increases of 10'F. This'resulted in a 30 i f t-lb transition temperature of -10'F and a 50 f t-lb transition temperature of 35'F. The average upper shelf energy (USE) of the reactor vessel core region weld metal Charpy specimens resulted in no energy decrease af ter irradiation to 18 3.79 x 10 n/cm2 (E > 1.0 MeV) at 550*F. This resulted in an average USE of 70 ft-lb (Figure 5-3). Irradiation of the reactor vessel weld metal Heat-Affected Zone (HAZ) Charpy 18 specimens to 3.79 x 10 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-4) resulted in 30 ft-lb transition temperature increase of 20*F which results in a 30 ft-lb transition temperature of -150'F and a 50 f t-lb transition temperature increases of 35'F which results in a 50 ft-lb transition temperature of 105'F. 06830:10/100990 5-4
i The average upper shelf energy (USE) of the reactor vessel HAZ metal Charpy specimens resulted in an increase of 3 ft-lb after irradiation to 3.79 x 18 10 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average USE of 112 j ft-lb. ? The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test temperature. A comparison of the 30 ft-lb transition temperature increases and the upper shelf energy decreases for the various Braidwood Unit i surveillance materials eith predictions using the methods of NRC Regulatory Guide 1.99, Revision 2[3] is presented in Table 5-6. This comparison indicates that the transition temperature increases and the upper shelf energy decreases 18 2 resulting from irradiation to 3.79 x 10 n/cm are less than the Guide predictions. Ouring the testing of the Braidwood Unit 1 surveillance capsule U Charpy specimens the machine loader malfunctioned. This resulted in two of the Charpy specimens ET14 and ET11, to be improperly tested. Figure 5-9 provides photographs of the specimens that were affected. During the loading of these two specimens, the loading device flipped the specimens such that the hammer struck 90' to the notched surface. The cause of the malfunction is believed to be associated with excursions in pneumatic pressure of the loading device. The loader was repaired and tested. The specimens tested prior to and following the malfunction were carefully examined to-insure that they had been properly loaded. i Figure 5-10 (a) is a photograph of HAZ metal specimen EH-2, which failed at a high energy level. The large branching cracks at the root of the notch verify that the specimen was very tough. The darker side of the specimen is base metal and the lighter side of the specimen is weld metal. It appears that the notch was placed in the base metal and that the fracture proceeded through the base except for the final 1/3 of the specimen which would be a small part of the energy absorbed. It is difficult to reconstruct the exact shape of the weld / base metal interface because of the deformation incurred during the 06830:10/100190 5-5 4 a
test. However, it appears-that'there may have been a protrusion where the specimen was cut. The expected weld / base metal interface is shown in Figure-5-10 (b) for comparison purposes. l 5.3 Tension Test Results The results of tension tests performed on shell forging 490867-1/490813-1 (tangential and axial orientation) and the weld metal irradiated to 3.79 x 18 2 10 n/cm are shown in Table 5-7 and are compared with unirradiated l results as shown in Figures 5-9, 5-10 and 5-11. Forging 490867-1/490813-1 test results are shown in Figures 5-9 and 5-10 and indicated 18 2 that irradiation to 3.79 x 10 n/cm caused a less than 8 ksi increase in the 0.2 percent offset yield strength and ultimate tensile strength. Held metal tension tests results shown in Figure 5-11, show that the ultimate tensile strength and the 0.2 percent offset yield strength increased by less- { then 10 ksi with irradiation. The small increases in 0.2'/. yield strength and tensile strength exhibited by the forging material and weld metal indicate 18 that these materials are not highly sensitive to radiation at 3.79 x 10 2 n/cm, as is also indicated by the Charpy impact' test'results.- The fractured tension specimens for the forging material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal. are shown in Figure 5-14. A typical stress-strain curve for the tension tests is shown in Figure 5-15. 5.4 Comoact Tension Tests Per the surveillance capsulo testing program with the Commonwealth Edison Company, 1/2 T-compact tension fracture mechanics specimens will'not be tested and will be stored at the Westinghouse Science and Techiology Center Hot Cell. 0683D:10/100190 5-6
i TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE BRAIDWOOD UNIT 1 FORGING 49D867-1/490813-1 IRRADIATED AT 550*F, i-l FLUENCE 3.79 x 10 n/cm2 (E > 1.0 MeV) [ 18 I Tem $erature In act Energy Lateral Expansion Shear Sample No. (*F ('C) (f -Ib) H), (milsY ,(gQ, (0 l Tanzential Orientation l EL13 -150 (-101 12.0 16.5 10.0 0.25 10 EL15 -100 23.0 31.0 16.0 0.41 15 ( EL4 -85 18.0 24.5 20.0 0.51 15 l EL9 -50 11.0 15.0 9.0 0.23 10 79.0 107.0 55.0 1.40 50 EL5 0 EL1 0 116.0 157.5 68.0 1.73 65 EL12 25 130.0 176.5 73.0 1.85 80 'l EL7 50 113.0 153.0 67.0 1.70 75 EL14 50 124.0 168.0 72.0 1.83 90 EL6 100 122.0 165.5 71.0 1.80 85 EL3 150 162.0 219.5 87.0 2.21 100 EL2 150 163.0 221.0 89.0-2.26 100 ELIO 200 171.0-232.0 87.0 2.21 100 EL11 200 172.0 233.0 88.0-2.21 100 EL8 250 1 174.0 236.5 86.0 _ 2.18 100 Axial Orientation ET14 -125 BAD TEST (MACHINE WALFUNCTION) ET10 - 80 19.0 26.0 13.0 0.33 15 ET1 - 80 42.0 57.0 27.0 0.69 25 ET3 - 50 24.0 32.5 15.0 0.38 20 ET2 - 50 34.0 46.0 31.0 0.79 25 ET9 - 20 50.0 68.5 32.0 0.81 25 ET12 0 75.0 101.5 51.0-1.30 45 ET11 25 BAD TEST (MACHIFE MALFUNCTION) ET6 50 98.0 133.0 63.0 1.60 75 ET5 50 102.0 138.5 64.0-1.63 80 ET4 100 104.0 141.0 69.0 1.75 90 ET8 150 135.0 183.0 84.0. 2.13 100 ET7 150 142.0 192.5 84.0 2.13 100-ET13 200 .136.0 184.5 85.0 2.16 '100-ET15 225 1 136.0 184.5 75.0 1.91 100 f 'l 0683D:1D/081490 5-7 l
7_ L l L TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE BRAIDH000 UNIT 1 REACTOR L VESSEL HELD METAL'AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 3.79 x 10 n/cm2 (E > 1.0 MeV) I0 i 'l I Temperature Impact Energy Lateral Expansioni Shear Sample No. ('F) ('0) (ft-lb) H), (mils) ,[EE}. (%) l Weld Wetal EW4 -55 -48) 16.0 21.5) 14.0 (0.36) 10 EW14 -25 -32) 14.0 19.0) 13.0 -(0.33) 15 EW13- -10 -23) -28.0 38.0) 22.0 0.56) 20 EW11 -10 -23) 39.0 53.0) 31.0-0.79) 25 EW15 25 - 4) 43.0 58.5)
- 35.0 0.89 35 EW6 25
- 4) 46.0 62.5) -40.0 (1.02 35 t EW10 75 24) 64.0 87.0)- 52.0 (1.32 95 EW1 75 24 67.0 91.0) 55.0 (1.40) 100 i EW2 100 ( 38 63.0 85.5)- 51.0-(1.30) 100-EW9 100 ( 38 73.0 99.0) 59.0 (1.50) 100 EW12 150 (66 64.0 87.0) +$1.0 (1.30) 100-EWS 150 66) 72.0 ( 97 5) 59.0 (1.50) 100 3 EW7 175 79) 69.0 (93.'5). 61.0-(1.55) 100-EW3 215 102) 76.0 (103.0) 61.0 (1 55). 100 EW8 215 (102) 78.0 (106.0) 61.0 (1.55) 100 HAZ Metal EH14 -150 (-101) 23.0 ( 31.0) 10.0 (0.25) 10 EH3 -150 (-101) 56.0 76.0) 30.0 (0.76) 25 EH13 -100 (-73) 26.0 35.5) 20.0 (0.51) 15 -EH7 - 50 (-46) 32.0-43,5) 26.0 (0.66) 20 t EH1 - 50 (-46) 85.0 115.0) - 46.0. .(1.17) 50 EH8 0 (-18) 78.0 106.0) 4",0 (1.19) ~60 EH2 0 (-18) -- 144.0 195.0). 76.0 -(1.93) 75 EH4 50 (-10) 72.0 ( 97.5). 44.0 (1.12) 60 EH10 50 ( 10) 72.0 ( 97.5) 44.0 (1.12) 60 i EH11 100 ( 38) 102.0 (138.5) .59.0 (1.50)~ 95 EH5 100- ( 38) 116, (157.5) 65.0 (1.65) 95 i EH6 175 (79) 94.0 '(127.5) 62.0. (1.57) 100 EH9 200 ( 93) 113.0-(153.0) 77.0-(1.96) 100 EH15 250 (121) 118.0 (160.0)- 65.0- .(1.65) 100 EH12 250 (121) 126.0 (171.0) 71.0 (1.80) 100 0683D:10/081490 5-8 1 3
TABLE S-3 i INSTRUMENTED CHARPY IMPACT TEST RESULTS FO THE BRAIDH000 U4II 1 SHELL FORGING 490867-1/490813-1 IRRADIATED AT 550 F, FLUENCE 3.79 x 1018,7c,2 (E > 1.0 MeV) brealised Enerstics Test Charpy Charpy Mamieve Prop Tield Time Maxieue Time to Fracture Arrest Tield Flow Sample Temp Energy Ed/A Ee/A Ep/A Load to Tield Load Maxiome lead Load Stress Stress haber (*F) (ft-Ib) (f t-lbfin ) (hips) (psec) (hips) (psec) (hips) (hips) (ksi) (&si) 8 Tamstential Grientation EL13 -150 12.0 97 79 18 4.10 120 4.65 210 4.65 136 145 ELIS -100 23.0 185 154 31 3.60 75 4.70 330 4.70 120 138 EL4 - 85 18.0 145 122 23 2.85 90 4.40 300 4.30 93 119 EL9 - 50 11.0 89 64 24 3.35 110 3.80 200 3.80 111 119 ELS 0 79.0 636 311 325 3.40 110 4.60 670 4.25 113 133 ELI O 116.0 934 317 617 3.55 110 5.05 630 4.15 1.95 118 143 EL12 25 130.0 1047 396 650 3.90 190 4.85 800 3.75 1.85 128 145 EL7 50 113.0 910 325 585 3.05 165 4.55 745 3.95 1.40 100 126 EL14 50 124.0 998 310 689 3.2 150 4.55 725 3.70 1.75 105 128 EL6 100 122.0 982 377 606 3.15 125 4.45 845 3.70 2.05 104 125 EL3 150 162.0 1304 358 946 2.95 140 4.30 850 97 119 EL2 150 163.0 1313 316 997 3.15 70 4.70 700 0.35 0.35 103 124 EL10 200 171.0 1377 354 1023 2.75 150 4.20 875 91 115 EL11 200 172.0 1385 273 1112 2.8E 30 4.25 620 0.30 0.30 95 118 EL8 250 174.0 1401 195 1206 1.45 150 2.70 805 49 69 Avial orientation ET14 -125 BAD TEST (MACEINE MALFUNCTION) ET10' - 80 19.0 153 128 25 3.80 120 4.40 310 4.3 126 135 ET1 - 80 42.0 338 279 59 3.45 65 4.80 550 4.70 0.15 113 136 ET3 - 50 24.0 193 160 33 2.85 150 4.45 420 4.40 95 121 i ET2 - 50 34.0 274 242 32 3.60 120 4.70 520 4.65 119 137 ET9 - 20 50.0 403 283 120 3.70 140 4.70 625 4.55 0.25 123 i39 ET12 0 75.0 604 333 271 3.55 190 4.70 730 4.25 0.30 117 136 ET11 25 BAD TEST (MACRINE MALFliNCTION) ETS 50 98.0 786 324 465 2.85 145 4.55 730 3.75 1.15 94 123 ETS 50 102.0 821 304 517 3.25 150 4.50 695 3.90 0.90 104 129 ET4 100 104.0 837 294 543 3.10 120 4.30 690 3.80 2.25 102 122 ET8 150 135.0 1087 321 766 3.00 130 4.20 765 99 119 ET7 150 142.0 1143 324 820 2.7 135 4.30 805 88 116 ET13 200 136.0 1995 322 773 2.85 175 4.25 805 94 117 ET15 225 136.0 1995 316 779 2.80 125 4.00 785 92 112 0683D:1D/081490 5-9 l
TABLE 5-4 INSTRUMEDTED CHARPY IMPACT TEST RESULTS F O THE BRAIDWOOO OIT 1 HELD METAL AND HAZ METAL IRRADIATED AT 550'F FLUENCE 3.79 x 13 n/cm2 (E > 1.0 MeV) 18 brealised Energies Test Charpy Charpy Manisme Prop Yield Time Namieue Time to Fracture Arrest Yield Flow Emeple Temp Energy Ed/A Em/A Ep/A Load to Yield lead Manisme Load Load Stress. Stress haber (*F) (ft-lb) (f t-lb/in') (hips) (msec) (hips) (psec) (nips) (nips) (ksi)* (nsi) Weld Metal EW4 -55 16.0 129 108 21 3.30 100 4.00 280 4.00 109 12I EW14 -25 14.0 113 41 72 2.45 55 3.45 135 3.45 0.50 80 97 EW13 -10 28.0 225 162 63 2.75 115 4.20 410 4.20 0.80 90 115 EW11 -10 39.0 314. 229 85 3.30 100 4.30 520 4.20 0.95 109 125 EWIS 25 43.0 346 166 280 2.95 85 4.15 375 4.10 1.35 97 118 EW6 25 46.0 370 216 154 3.25 120 4.10 525 3.85 1.25 108 122 EWlo 75 64.0 515 216 300 3.2 130 4.15 525 3.95 3.55 105 121 EW1 75 67.0 540 214 326 3.15 120 4.05 520 105 119 EW2 100 63.0 507 215 293 3.30 250 4.15 560 100 123 EW9 100 73.0 588 212 376 2.85 100 4.10 520 94 115 EW12 150 64.0 515 207 306 2.30 75 3.90 520 75 102 EWS 150 72.0 580 251 329 2.15 100 4.10 625 72 103 EW7 175 69.0 556 203 352 2.75 100 3.85 525 91 109 EW3 215 76.0 612 188 424 1.90 85 3.90 505 63 96 EWS 215 78.0 628 204 424 2.90 130 3.90 535 96 113 BAZ Metal EE14 -150 23.0 185 117 68 3.90 90 4.70 255 4.80 0.15 129 143 EN3 -150 56.0 451 341 110 3.85 130
- 95 995 4.85 0.15 128 146 EB13
-100 26.0 209 185 24 3.80 110 -35 400 4.75 Ils 138 EE7. - 50 32.0 258 205 52 3.25 130 . 80 475 4.75 0.20 108 233 E51 - 50 85.0 684 337 347 3.85 120 4.90 680 4.35 127 145 EB8 0 78.0 628 362 266 3.35 85 4.95 705 4.50 0.15 111 137 EE2 0 144.0 1160 323 837 3.70 110 4.75 670 2.40 1.10 122 139 E84 50 .72.0 580 313 267 3.25 170 4.65 705 4.05 1.85 los 130 BEIO 50 72.0 580 317 263 3.60 85 4.80 625 4.25 1.80 119 139 Egli 100 102.0 821 291 531 3.20 110 4.50 640 3.15 2.00 106 128 E55 100 116.0 934 355 579 3.15 110 4.55 775 3.35 2.10 105 128 EM6 175 94.0 757 371 386 2.95 100 4.30 845 97 120 EN9 200 113.0 910 333 577 3.20 160 4.35 800 106 125 EEIS 250 118.0 950 348 602 2.75 65 4.35 795 91 117 EN12 250 126.0 1015 334 680 2.55 70 4.20 775 85 112 0683D:1D/081490 5-10
TA8t* 5-5 EFFECT OF IRRADIATION TO 3.79 m 10 n/cm (E > 1.0 MeV) AT 550*F ON 810TCH TOUG W ESS PROPERTIES OF BRAIDWOOO tmIT 1 REACTOR VESSEL SURYEILLANCE METERIALS Average 35 mil Average Upper Average 30 ft-lb Lateral Espansion Average 50 ft-1b Shelf Energy at Temperatore (*F) Temperature (*F) Temperature (*F) Fv11 Shear (f t-1b) Material Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT unirradiated Irradiated A(ft-1b) Forging -65 .-60 5 -45 -45 0 -40 -30 10 168 168 0 490667-1/490813-1 (Tangential)- Forging -45 ' 45 0 -15 -15 0 -15 -15 0 152 138 -14 490867-1/490813-1 (Aslal) -Weld Metal -20 ' 10 - 10 0 15 15 . 25 35 10 70 70 0 .HAZ_ Metal -170 -150 20 -180 -170 10 -140 -105 35 109 112 0 0683D:1D/092590 5-11
TABLE 5-6 COMPARISON OF BRAIDH000 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-Ib Transition Temp. Shift _tlpper Shelf Eneroy Decrease Fluence R.G. 1.99 Rev. 2 Capsule U R.G. 1.99 Rev. 2 Capsule U (Predicted) (Predicted) 8 2 Haterial 10 n/cm (.7) (.7) (g) (g) Forging A90867-1/49D813-1 (Tang.) 3.79 15.0 5 15 0 Forging 490867-1/490813-1 3.79 15.0 0 15 9 (Axial) Held Metal 3.79 40.0 10 15 0 a) Cu and Ni values from Table 4-1 were used to determine R.G. 1.99 predictions. 0683D:10/082890 5-12
i TABLE 5-7 IECSILE PROPERTIES FOR BRAIDH000 ("3IT 1 REACIG3 VESSEL SURVEILLANCE MATERIAL IRRADIATED AT 550*F 10 3.79 x 1018,7c,2 (E > 1.0 MeV) Test 0.2% Yield Ultimate Fractere Fracture Fracture Uniform Total Beduction Sample Temp. Strength Strength Load Stress Strength Elongation Elongation in Area l Material No ber M (ksi) (ksi) (kip) (ksi) (nsi) (%) (%) (%) Forging ELI 70 67.7 88.6 2.50 188.3 50.0 13.5 28.2 73 l 490867-1/ EL2 300 61.1 79.5 2.30 149.4 46.9 10.5 24.6 69 490813-1 EL3 550 58.6 83.5 2.50 147.9 50.0 11.2 23.8 66 (Tangent. Drient.) Forging ETI 70 69.8 89.6 2.80 181.9 57.0 11.3 24.2 61 -49D867-1/ ET2 300 62.8 81.0 2.60 181.6 53.0 9.0 21.9 61 490813-1 ET3 550 58.6 84.5 2.80 153.3 57.0 9.8 20.4 63 6 (Arial Orient.) Weld EW1 70 73.8 90.7 3.25 117.7 66.2 10.5 21.2 61 Weld EW2 300 70.8 85.6 2.95 151.4 60.1 8.3 18.9 80 -Weld EW3 550 68.8 85.6 3.20 181.1 65.2 8.3 17.4 56 '0683D:ID/081490 5-13
(
- C)
-150 -100 - 50 0 50 100 150 200 250 I I I I I I I i i 100 A 3 80 y2 2 u e 2 # e'4 a m e o o ~ 0 i I g [ ,[ 55 0 ye ts, i 8O 1.03 (M oo 45 8 0 i e 0 200 I I I I i i l I i {. 180 20 l@ o 200 _ 10 8 --- Unirradiated o =! 120 o. 160 ~ 100 Irradiated at 550*F l / 18 120 C l g 80
- 3. 79 x 10 n/cm
"g ,/ ( E > I MeV)
- 80 l
10*F4 0 5' o 0 20 o 0 6-O' ' 1 0 - 200 -100 0 100 200 300 00 500 Temperature (* F) Figure 5-1. Charpy V-Notch Impact Properties for Braidwood Unit 1 Reactor Vessel Shell Forging 490867-1/49C813-1 (Tangential Orientation) i 06830:10/080890 5-14
1 (
- C) l
-150 -100 - 50 0 50 100 150 200 250 i i i i i i i i i i 100 2 3 80 / k 60 ,y i m m g / 20 0 F 100 2.5 j 2 /,^r~ ~, " = 80 20 e 2 - 60
- 1. 5 3
'2 $0
- 1. 0 *
-(20 0.5 0 I i i i i 0 200 180 240 \\ 2 160 t ,a -- *-- 200 _ 140 /* .Q / g120 160 o ~ 100 g --- Unirradiated o 1N ic u g 80 2-8 Irradiated at 550 F 60 - 18 2 80
- 3. 79 x 10 n/cm 40 (E ) I MeV)
- 40 20 0
i i i O - 200 -100 0 100 200 300 400 500 i l Temperature (* F) Figure 5-?.. Charpy V-Notch Impact Properties for Braidwood Unit 1 Reactor Vessel Shell Forging 490867-1/49C813-1-(Axial Orientation) 06830:10/080890 5-15
l ('C) -150 -100 -50 0 50 100 150 200 250 l 1 i i ): y ig i i i 100
- e
$ 80 gg / 5c [2 / e- / E 0 V l# 15 i i i i i i i i i 5 80 LO J' y' 60 1, 5 1.0g 8O 2 15'F ~2 gE a5 0 I i 0 -100 -50 0 50 100 150 200 250 80 i i i i i i i 3 9 70 ~ 60 [. g e 80 -y50 / 7, 10* F 5 40 --- Unirradiated o 60,.; E 30 ~ -y 10 F Irradiated at 550*F 40 8/ 18 2 20
- 3. 79 x 10 n/cm
( E > I MeV) e 20 10 f 0 M i i 0 - 200 -100 0 100 200 300 400 500 Temperature ( F) Figure 5-3. Charpy V-Notch Impact Properties for Braidwood Unit 1 Reactor Vessel Held Metal 06830:10/080890 5-16
l l (
- C)
-150 -100 - 50 0 50 100 150 200 250 i i i i o[i 3 i i i i 100 $ 80 2 $ 60 .4 2 6! O f 8 l m 0 'M' 100 2.5 m 80 10 g ' e.a. .. -+. - = -0 ( L51 o )'f0*F IO 1.0 3 "2 o 8[' ( 20 0.5 '7 l I i 0 0 i i i i i i i i i l 180 20 160 200 le 2 g120 o. 160 h100 8 m/ p 's ---.- 120 - --- Unirradiated o l b 80 o o c .o o g '2 Irradiated at 550 F 80 " 60 .,/ 18 2 1 0 / 20 3.79 x 10 n/cm ( E > 1 MeV)
- 20
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Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 1 Reactor Vessel Shell, forging 490867-1/49C813-1 (Axial Orientation) 5-19 RW-24on
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i l 90 80-g" 70- @60-g50-m y40- "30-SPEC ET-3 10-550 F 0 0.b5 0.'1 0.i 5 0.'2 0.25 STRAN, N/W l Figure 5-17. Typical Stress-Strain Curve for Braidwood Unit 1 Shell Forging 490867-1/49C813-1 Tension Specimens. 0683D:10/100190 5-30
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIHETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and i 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 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 cvntained in each of the surveillance capsules. The latter information is derived solely from analysis. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for di.'ferences 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 tell as to a more accurate evaluation of damage gradients through the pressure vessel wall. Because of this potential shif t away from a threshold fluence.toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence 4 0683D:10/080890 6-1
(E > 1.0 HeV) to provide a data base for future reference. The energy detondent dpa function to be used for this evaluation is specified in ASTM Stancard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials." This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule U. 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 e:.posure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided. 6.2 Discrete Ordinates Analvsis 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', i 121.5', 238.5', 241.0', and 301.5' relative to the core cardinal axes as shown in Figure 4-1. l 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 l l-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. 1 0683D:10/092590 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 geometr.y as well as to establish relative radial distributions of exposure parameters {4(E > 1.0 Mev), $(E > 0.1 Mev), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/$(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 locattions. The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for the cycle 1 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 l l 06830:10/080890 6-3
of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased. The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the forward calculation provided the means to: 1. Evaluate neutron dosimetry obtained from surveillance capsule locations. 2. Exttc7otate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall. 3. Enable a direct comparison of analytical prediction with measurement. 4. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves. The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R. O geometry using the DOT two-dimensional discrete ordinates code [12) and the SAILOR cross-section library (13). 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. g The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2e i 06830:10/080890 6-4
level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results. 1 order of angular All adjoint analyses were also carried out using an S8 quadrature and the P cross-section approximation from the SAILOR library. 3 Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R O geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, 4 (E > 1.0 MeV). Having the impor-tance functions and appropriate core source distributions, the response of interest could be calculated as: R (r 0) = f I I I(r, 0, E) S (r. 9. E) r dr de dE r O E 4 (E > 1.0 MeV) at radius r and azimuthal angle 0 there: R(r, 0) = Adjoint importance function at radius, r, azimuthal I (r, 0, E) = angle 0, and neutron source energy E. Neutron source strength at core location r, O and S (r O, E) energy E. Although the adjoint importance functions used in the Braidwood Unit I analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/$ (E > 1.0 MeV) is insensitive to changing core source distributions. In the application of these adjoint important functions to the Braidwood 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/$ (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. 0683D:10/080890 6-5
The reactor core power distribution used in the plant specific adjoint ealculations was taken from the fuel cycle design report for the first operating cycle of Braidwood Unit 1 (143. The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of the pressure vessel and surveillance capsules are summarized in Figure 6-2. For comparison purposes, the core power distribution (design basis) used in the reference forward calculation is also illustrated in Figure 6-2. Selected results from the neutron transport analyses performed for the Braidwood Unit I reactor are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurerint 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 ($ (E > 1.0 MeV), $ (E > 0.1 MeV), and dpal 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 I provided as a point of reference against which plant specific fluence j evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycle 1 plant specific power distributions. It is important to note that the data for the vessel inner I radius were taken at the clad / base metal interface: i, 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-a, 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 I 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. i 1 0683D:10/080890 6-6
Values of key fast neutron exposure parameters were_ derived from the measured reaction rates using the FERRET least square aOustment code (29), The FERRET approach used the measured reaction r te data and the calculated neutron energy spectrum at the the center of th_e 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 reart; ion rate data. The exposure t prr e stors along with associated uncertainties where then obtained from the adjust 4dspectra. In the FERRET evaluations, a log normal least-squares algorithm weights both the a priori values and thc measured data in accordance with the assigned j uncertainties and correlations. In general, the measured values f are linearly related to the flux 4 by some response matrix A: i (s,a) (s) (a) f -I A 4 i ig 9 9 there i indexes the tieasured 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 e 4 i ig g g l relates a set of measured reaction rates R to a single spectrum 4 by g _g the multigroup cross section og. (In this case, FERRET also adjusts the g cross-sections.) The lognormal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties, y l i
- i 0683D:10/080890 6-9
In the FERRET analysis of the dosimetry data,- the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53' groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code (30). This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed to the group scheme used in FERRET. The cross-sections were also collapsed into the 53 energy-group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/B-V dosimetry file. Uncertainty estimates and 53 x 53 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 H 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, = RN+R R,P N g g gg, there R specifies an overall fractional normalization uncertainty N (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties R specify additional random uncertainties for g group g that are correlated with a correlation matrix: Pgg, - (1 - 0) 6gg, + 0 exp [- ) 0683D:10/080890 6-10
C For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45 azimuth is given by: I $(220.27, 45') F (225.75, 45') $1/41(45') Projected neutron flux at the 1/4T position on where $1/4T(45') the 45' azimuth Projected or calculated neutron flux at the. $ (220.27, 45') vessel inner radius on the 45' azimuth. Relative radial distribution function from - F (225.75, 45') Table 6-3. Similar expressions apply for exposure parameters in terms of 4(E > 0.'l MeV) and dpa/sec. The 00T calculations were carried out for a typical octant of the reactor.. However, for the neutron pad arrangement in Braidsood Unit 1, the pad extent for all octants is not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron pad extending from 32.5' to 45' (12.5') which produces the maximum vessel flux. Other octants have neutron pads extending 22.5* or 20' which provide more shielding. For the. octant with' the 12.5' pad, the maximum flux to the vessel occurs near 25* and the values in the tables for the 25' angle are vessel maximum values. Exposure values for O*, 15', and 45' can be used for all octants; values in the tables for 25* and 35' are maximum values and only apply to octants with a 12.5' neutron pad extent. 6,3 Neutron Dosimetrv The passive neutron sensors incleded in the Braidwood Unit I surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest (4 (E > i.0 Mev), 4 (E > 0.1 HeV), dpa]. p 0683D:lD/080890' 6-7
The relative locations of the neutron sen Nrs within the capsules are shown in Figure 4-2. The iron, nickel, copper, an;i 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 cere 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 i 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 (15 through 28). Following sample preparation and 1 eeighing, the activity of each monitor was determined by means of a lithium-drif ted germanium, Ge(Li), gamma spectrometer. The irradiation history of the Braidwood Unit 1 reactor during cycle I was obtained from NUREG-0020 " Licensed Operating Reactors Status Summary Report" for the applicable period. i The irradiation history applice.ble to capsule U 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. 1 06830:10/080890 6-8
The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 8 (e 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. Macrker [31]. Maerker's results are closely duplicated when 8 - 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 U dosimetry are given in Table 6-9. The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 3.79 x 10 n/cm2 (E > 1.0 MeV) with an I0 associated uncertainty of z 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 U is presented in Table 6-12. The agreement between calculation and measurement falls within 13% for all fast neutron exposure parameters listed. The l thermal neutron exposure calculated for cycle 1 undepredicted the measured value by 62 percent. Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13.. Along with the current (1.10 EFPY) exposure derived from the capsule U measurements, projections are also provided for an-exposure poriod of 16 EFPY and to end of vessel design life (32 EFPY), The calcula.ted design basis exposure rates given in Table'6-2 were used to perform
- rujections beyond the end of cycle 1.
0683D:1D/080890 6-11
In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Braidwood Unit I reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel tall 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 positions were defined by the relations 4' (1/4T) - 4 (Surface) { d a ( rfa e) } a(u a e) I 4' (3/4T) = 4 (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 Braidwood Unit i surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules. 0683D:10/080890 6-12
i (TYPICAL) 58.58 -61.08 8"1h v 7 usc l _.2.. m. z -f h l 3} m y I NEUTRON PAD N-Figure 6-1. Plan View of 'a Dual. Reactor Vessel Surveillance Capsule l 1 1 0683D:10/071390 6-13 l
F I. f i 8.74-0.70 0.78 0.59 Cycle-I s 1.01 1.84 0.96 9.77 Design Basis -r 0.99 1.82 0.97 9.95 0.04 0.57 1.02 1.10 1.80 1.95 1.10 0.71 1.13 1.09 1.97 1.05 0.98 1.01 1.05 0.87 0.07 1.07 f.ge 1.33 1.14 1.13 1.13-1.14' 1.g3 1.09 1.06 0.00 1.18 f.e4 1.10 1.14-1.14-1.20 0.90 1.94 1.12 9.92 i l l l I I I l l l Figure 6-2. Core Power Distributions Used in Transport f Calculations for Braidwood Unit 1 t l 06830:10/071390 6-14 i r w
1 TABLE 6-1 CALCULATED FAST NEUTRON CXPOSURE PARAMETERS-AT THE SURVEILLANCE CAPSULE CENTER DESIGN BASIS CYCLE 1 g. g. g. .g. ll II 10 10 4 (E > 1.0 MeV) 1.13 x 10 1.21 x 10 '8.84 x 10 9.51 x 10 2 (n/cm -sec) II Il ll II $ (E > 0.1 HeV) 5.07 x 10 5.44 x 10 3.97 x 10 4.28 x 10 i 2 (n/cm -sec) dpa/sec 2.21 x 10-10 2.37 x 10-10 1.73 x 10-10 1.86 x 10-10 i I 'l Y b r 06830:10/080890 6-15
[ TA8LE 6-2 i CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD /8ASE METAL-INTERFACE i DESIGN B SIS i 0* 15' 25' 35' 45'- 4(E) 1.0Mev) 1.78 x 10 2.66 x 1010 3.01.x 1010 10 10-10 2.45 x 10 2.81 x 10 2 (n/cm -sec) i i 4(E) 0.1Mev) 3.70 x 10 5.60 x 1010 8.22'x 1010 10 ~7.04 x 10 .l 10 10 6.96 x'10 2 (n/cm -sec) dpa/sec 2.77 x 10'II 4.12 x 10-II 5.04 x 10-II 4.15 x 10~II 4.48 x.10-II~ Q CLE 1 SPECIFIC. O' 15' 25' 35' 45* 4(E) 1.0Mev) 1.32 x 10 2.06 x 10 2.38 x 10 1.98 x 1010.2.31 x 1010 10 10 10 2 (n/cm -sec) 4(E) 0.1Mev) 2.74 x 10 4.34 x 1010' 6.50 x 1010 5.62x10)0 5.79'x 1010 J 10 2 (n/cm -sec) dpa/sec 2.05 x 10-II 3.19 x 10-II 3.99 x 10~II 3.35 x 10-II 3.68 x 10-II 06830:10/082390 6-16
s TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0.MeV) HITHIN THE PRESSURE VESSEL HALL Radius _(cm)_ O' 15* 25' 35' 45' 220.27(I) 1.00 1.00 1.00 1.00 1.00 220.64 0.976 0.979 0.980 0.977 0.979 221.66 0.888 0.891 0.893 0.891 0.889 222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655-0.648 225.63 0.551 0.550 0.554 0.552 0.543 226.95 0.462 0.460 0.465 0.463 0.452 228.28 0.386 0.384 0.388 0.386 0.375 229.60 0.321 0.319 0.324 0.321-0.311 230.92 0.267 0.265 0.271 0.267 0.257 232.25 0.221 0.219 0.223 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 1 237.54 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' O.0471-242.17(2) 0.0506 0.0488 0.0518 0.0521 0.0438 l l NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius l
l i 0683D:10/080890-6-17 l
TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 HeV) HITHIN THE PRESSURE VESSEL HALL-Radius (cm) O' 15' 25' 35' 45' 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 O 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 'O.792 0.782 228.28 0.761 0.750 0.716 0.734 0;721-229.60 0.705 0.693 0.704 0.677 0.662-230.92 0.649 0.637 0.649 0.621 0.605 232.25 0.594 0.582 0.594 0.567 0.549 I 233.57 0.540 0.529 0.542 0.515 0.495 234.89 0.487 0.478 0.490 0.465-0.443 236.22 0.436 0.428 0.440 0.416 0.392 237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201 242.17(2) 0.233' O.226 0.237 0.223 0.188 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius i
0683D:10/080890 6-18 ~
i TA8LE 6 ! RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL HALL Radius 3 (cm) O' 15' 25' __ 35' 46* - 220.27(I} 1.00 1.00 1.00 1.00 1.00 220.64 0.984 0.981 0.984 0.983 0.984 221.66 0.912 0.909 0.917 0.921 0.915 222.99 0.815 0.812 0.826 0.833 0.821-224.31 0.722 0.719 0.737 0.747 0.730-225.63 0.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572 228.28 0.497 0.493 0.519 0.533 0.506 i 229.60 0.439 0.435 0.462 0.475 0.447 230.92 0.387 0.383 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 t
06830:10/080890 6-19
TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS i Reaction Target Fission Monitor of Height
Response
Product Yield tiittarial Interest Fraction Ranae Hal f-Li fe. (1.) Copper Cu63(n a)Co60 .0.6917 E)4.7-MeV 5.272 yrs i Iron Fe54(n.p)Mn54 0.0582 E > 1.0 MeV 312.2 days Nickel NiS8(n,p)CoS8 0.6830 E >.l.0~HeV 70.90 days Uranium-238* U238(n,f)Cs137 1.0 E.) 0.4 MeV 30.12 yrs' 5.99 Neptunium-237' Np237(n,f)CsI37 1.0 E > 0.08 HeV 30.12 yrs 6.50 Cobalt-Aluminum
- CoS9(n,8)Co60 0.0015 0.4ev>E) 0.015 MeV.5.272 yrs Cobalt-Aluminum CoS9(n,8)Co60 0.0015 E') 0.015 MeV 5.272 yrs
- Denotes that monitor is cadmium shielded.
-i l i 0683D:10/080890 6-20 i
TABLE 6 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U P Irradiation Decay Irradiation - P) 3 Period (MH ) P Time (days) Time (days) t Ref. 7/87 134 .039 20 1035 8/87 286 .084 31 1004 9/87 568 .167 30 974 10/87 1255 .368 31 943 11/87 2084 .611 30 913 12/87 2548 .747 31 882 1/88 64 .019 31 851 2/88 0 .000 29 822 3/88 66 .019 31 791 4/88 215 .063 30 761 5/88 2203 .646 31 730 6/88 2351 .689 30 700 7/88 246 .072 31 669 8/88 2920 .856 31 638 9/88 2401 .704 30 608 10/88 2704 .793 31 577 11/88 2703 .792 30 547 12/88 3129 .917 31 516 1/89 1836 .538 31 485 2/89 2764 .810 28 457 3/89 2795 .819 31 426 4/89 1773 .520 30-396 5/89 2872 .842 31 365-6/89 2967 .870 30 335 7/89 -2252 .660 31 304 8/89 1851 .543 31 273 9/89 757 .222 2 271-NOTE: Reference Power - 3411 MHt s 0683D:1D/082390 6-21 t
.? -r TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES i Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec-am) (dis /sec-am) (RPS/ NUCLEUS) Cu-63 (n a) Co-60 4 5 Top 5.16 x 10 4.38 x 10 4 5 Middle 4.53 x 10 3.84 x 10 4 5 Bottom 4.50 x.10 3.82 x 10 Average 4.73 x 10 4.01 x 10 6.12 x 10-I7 4 5 Fe-54(n,p) Mn-54 6 6 Top 1.08 x 10 3.98 x 10 5 6' Middle 9.89 x 10 3.65 x 10 5 6 Bottom 9.74 x 10 3.59 x 10 Average 1.01 x 10 3.74 x.10 5.96 x 10-15 6 6 Ni-58 (n.p) Co-58 0 7 Top .2.65 x 10 5.56 x 10 6 7 Hiddle-2.43 x 10 5.10 x 10 6 7 Bottom 2.42 x 10 5.08 x 10 Average 2.50 x=10 5.25 x 10 7.49 x 10-15' 0 7 U-238 (n,f) Cs-137 (Cd) i 5 6 I4 Middle 1.33 x'10 5.46 x 10 3.60 x 10 l 0683D:10/080890 6-22 ---^--------------m
I TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd-Measured Saturated . Reaction Monitor and Activity Activity Rate Axial location (dis /sec-am) (dis /see-am)- (RPS/ NUCLEUS) Np-237(n,f) Cs-137 (Cd) Middle 1.32 x 10 5.411x 10 3.28 x 10-I3 6 7 Co-59 (n,8) Co-60 ( { 7 7 Top 1.06 x 10 8.95 x 10 7 7 Middle 1.06 x 10 8.99'x'10 7 7 Bottom 1.06 x 10 8.99 x 10 Average 1.06 x 10 8.98 x 10 5.86 x 10-12 7 7 Co-59 (n,8) Co-60 (Cc) 6 7 Bottom 5.43 x 10 4.61 x 10 Average 5.43 x 10 4.61 x 10 3.01 x 10-12 6 7 'I i l l 0683D:lD/080890 6-23 j
TABLE 6-9 i
SUMMARY
OF NEUTRON 00SIMETRY RESULTS TIME AVERAGED EXPOSURE RATFj 2 ll 4 (E) 1.0'MeV) {n/cm -sec). 1.09 x.10 8% 2 ll 4 (E) 0.1 MeV) {n/cm -sec) 4.80 x 10 1.15% dpa/sec 2;09 x 10-10 11%- 2 ll 4(E< 0.414 eV) (n/cm -sec) 1.17 x 10 i 21% INTEGRATED CAPSULE ~ EXPOSURE 2 18 o (E) 1.0 MeV) {n/cm ) 3.79 x 10 8% 2 I9 6 (E) 0.1 MeV) (n/cm ) 1.67 x 10 1 15% dpa 7.26 x 10-3 11% 2 18 o (E< 0.414 eV) (n/cm ) 4.06 x.10 21% i NOTE: Total Irradiation Time - 1.10 EFPY I i t j l 0683D:1D/082390 6-24
t i TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED i REACTION RATES AT THE SURVEILLANCE CAPSI".E CENTER ( Adjusted Reaction Measured Calculation Cid Cu-63 (n,a) Co-60 6.12x10-I7' -6.19x10-I7 1.01 Fe-54 (n.p) Mn-54 5.96x10-15 5.85x10-15 0.98 i Ni-58 (n.p) Co-58 7.49x10-15 7.66x10-15 1.02 U-238 (n,f) Cs-137 (Cd) 3.60x10-I4 L3.35x10-I4 0.93 Np-237 (n,f) Cs-137 (Cd) 3.28x10-13 3.37x10-13 1.03 Co-59 (n,8) Co-60 (Cd) 3.01x10-12 3.02x10 0.99 Co-59 (n,8) Co-60 5.86x10-12 5.81x10-12' 1.00 1 l f E 0683D:1D/082390 6-25 i
F TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy Adjusted Flux Energy Adjusted Flux j Gr_qu (Mev) (n/cm2-sec) Group (Mev) -(n/cm2-sec) I 6 10 1 1.73x10 8.93x10 28 9.12x10-3 2.24x10 2 1.49x10 2.01x10 29 5.53x10-3 2.90x1'010 I 7 I 7 9 3 1.35x10 7.72x10 '30 3.36x10-3 9.06x10 I 8 9 4 1.16x10 1.72x10 31 2.84x10~3 8.67x10 ' I 8 9 5 1.00x10 3.75x10 32 2.40x10 8.36x10 0 8 10 6 8.61x10 6.35x10 33 2.04x10-3 2.36x10 0 9 10 7 7.41x10 1.44x10 34 1.23x10-3 2.18x10 0 9 10 8 6.07x10 2.04x10 35 7.49x10~4 2.02x10 0 9 10 9 4.97x10 4.28x10 36 4.54x10~4 1.93x10 0 9 10 10 3.68x10 5.67x10 37 -2.75x10~4 2.08x10 0 10 10 11 2.87x10 1.19x10 38 1.67x10-4 2.25x10 0 10 10 12 2.23x10 1.65x10 39 1.01x10-4 2.25x10 0 10 10 13 1.74x10 2.33x10 40 6.14x10-5 2.23x10 0 10 I 14 1.35x10 2.60x10 41 3.73x10-5 2.18x10.0 0 10 1 15 1.11x10 4.76x10 42 2.26x10-5 2.11x10_0 10 10 16 8.21x10~I 5.45x10 43 1.37x10 2.05x10 10 10 17 6.39x10-I 5.68x10 44 8.32x10-6 1.95x10 10 10 18 4.98x10-I 4.13x10 45 5.04x10-6 1.80x10 10 10 l 19 3.88x10~I 5.81x10 46 3.06x10-6. 1.68x10 l 10 10 20 3.02x10-I 5.99x10 47 1.86x10-6 1.55x10 10 10 21 1.83x10-I 5.94x10 48 1.13x10-6 1.15x10 10 10 22 1.11x10-I 4.76x10 49 6.83x10~7 1.48x10 10 10 23 6.14x10-2 3.31x10 50 4.14x10~7 1.98x10 10 9 24 4.09x10-2 1.88x10 51 2.51x10-7 1~.99x10 - 10 9 25 2.55x10-2 2.47x10 52 1.52x10~7 1.91x10 10 10 26 1.99x10-2 1.22x10 53 9.24x10-8 5.85x10 10 27 1.50x10-2 1.55x10 j NOTE: Tabulated energy levels represent'the upper energy of each group. 0683D:1D/080890 6-26
f TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED 3 EXPOSURE LEVELS FOR. CAPSULE U Caleulated Measured Cfy 2 I8 18 4(E) 1.0 MeV) (n/cm ) 3.30 x 10 3.79 x 10 0.87' 2 I9 I9 4(E) 0.1 MeV) (n/cm ) 1,49 x 10 1.67 x 10 0.89 t 6.46 x 10~3 7.26 x 10-3 0.89 dp3 2 I0 18 4(E< 0.414 eV) (n/cm ) 3:.57 x 10 4.06 x 10 0.39 3 1 0683D:10/080690 6-27 j
.i i TABLE 6 NEUTRON EXPOSURE PROJECTIONS'AT KEY LOCATIONS 1 ON THE PRESSURE VESSEL CLAD / BASE HETAL INTERFACE FOR BRAIDH000 UNIT 1 -t A2IMUTHAL ANGLE _Z5'(a) 35' '4 5 ' O' 15' 1.10 Ef.Pl 4(E< 1.0 MeV) 5.27 x 10 8.22 x 10 9.50 x 10 7.90 x 10 9.22 x 10 (n/cm2) 4(E) 0.1 MeV) 1.07 x 10 1.69 x 10 2.54 x 10 2.19 x 10 2.26 x 10 (n/cm2) i ~ ~ ~ ~ dpa 8.00 x 10 1.24 x 10 1.56 x 10 1.31 x 10 1.44 x 10 16.0 EFPY 18 19 19 19 19 4(E) 1.0 MeV) 8.90 x 10 1.33 x 10 1.51 x 10 1.23.x.10 1,41'x 10 (n/cm2) 4(E) 0.1 HeV) 1.84 x 10 2.80 x 10 4.11-x 10 3.48 x 10 3.'53 x 10 (n/cm2) dpa 1.38x16 2.06x16 2.52x16 2.08 x 16 2.25x-16 32.0 EFPY 4(E) 1.0 MeV) 1.79 x 10 2.68 x 10 3.03 x 10 2.47 x 10 2.83 x 10 (n/cm2) 4(E) 0.1 MeV) 3.71 x 10 5.62 x 10 8.26 x 10 ~ 7.00 x 10 7.08 x 10 (n/cm2) ~ ~ ~ ~ ~ dpa 2.78 x 10 4.14 x 10 5.07.x 10 4.17 x 10 4'.51 x 10-(a)- Maximum point on the pressure vessel 0683D:10/082390 6-28 i
l-TABLE 6-14 l. - NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/C00LD0HN CURVES 16 EFPY NEUTRON FLUENCE (E) 1.0 MeV) SLOPE doa SLOPE 2 2 (n/cm ) (equivalent n/cm ) Surface-1/4 T 3/4 T Surface 1/4 T 3/4 T 18 IO I0 18 18 8 0* 8.90 x 10 4.83 x 10 1.03 x 10 8.90 x 10 5.61 x 10 1.95 x 10 I9 18 18 I9 18 18 15* 1.33 x 10 7.20 x 10 1.51 x 10 1.33 x 10 8.34 x 10 2.88 x 10 I9 18 18 I9 18 18 25*(a) 1.51 x 10 8.24 x 10 1.78 x 10 1.51 x 10 9.80 x 10 3.59 x 10 18 18 18 18 35* 1.23 x 10 6.69 x 10 1.43 x 10 1.23 x 10 8.15 x 10 3.05 x 10 0 18 18 18 45* 1.41 x 10 7.54 x 10 1.52 x 10 3,4j x 10 9.02 x 10 3.09 x 10 32 EFPY NEUTRON FLUENCE (E> 1.0 MeV) SLOPE dpa SLQP_E 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 18 18 D 18 0* 1.79 x~10 9.72 x 10 2.07 x 10 1.79 x 10 1.13 x 10 3.92 x 10 I9 18 D 18 15* .2.68 x 10 1.45 x-10 3.05 x 10 2.68 x 10 1.69 x 10 5.81 x 10 18 U D 18 25*(a)' 3.03 x 10 1.66 x 10 3.57 x 10 3.03 x 10 1.97 x 10 7.21 x 10 18 D D 18 35* 2.47 x 10 1.35 x 10 2.86 x 10 2.47 x 10 1.64 x 10 6.12 x 10 l9 I9 18 D D 18 45* 2.83 x 10 1.52 x 10 3.06 x 10 2.83 x 10 1.81 x 10 6.20 x 10 (a) Maximum point on'the pressure vessel 0683D:1D/080890 6-29
TABLE 6-15 ( UPDATED LEAD FACTORS FOR BRAIDH000 UNIT 1 l SURVEILLANCE CAPSULES j l Caosule Lead Factor t V 4.00(a) X 4.02 H 4.02 1 Z 4.02 l V 3.75 Y 3.75 'l 1 (a) Plant specific evaluation l I 'l l ) t l 1 1 I 06830:10/080890 6-30 i
1 t SECTION 7.0. SURVEILLANCE CAPSULE REMOVAL SCHEDULE i The removal schedule listed below was projected based on the calculated design l basis neutron flux levels given in Table 6-1. If'the plant operates with a low f leakage fuel management strategy, the neutron flux at the vessel wall and the-capsule locations would be reduced. Thus, the fluence levels reached at the target EFPY's would be correspondingly decreased. For plants similar in design to Braidwood Unit 1, operation with low leakage fuel management typically results in neutron fluence rates at the capsule 18 2 18 2 locations of approximately 2.7X10 n/cm -EFPY and 3.0X10 n/cm -EFPY at the 29' and 31.5' capsule locations, respectively. These values:can be used to readjust withdrawal schedules to accommodate low leakage operation. The following removal schedule meets ASTM E185-82, is based on the design base fluence calculations and is recommended for future capsules to be removed from the Braidwood Unit I reactor vessel: Capsule Estimated Location Lead Fluenge Capsule (deg.) Factor Removal Time (b) (n/cm) l U 58.5 4.00 1.10 (Removed)(a) l X 238.5 4.02 4.5 . 3.79 x 1 18 l 1.7 x 10 (c) l V 61.0 3.75 9.0 3.2 x 10 (d) Y 241.0 3.75 15 '5.3 x 10 W 121.5 4.02 Standby' Z 301.5 4.02 Standby i (a) Plant Soecific Evaluation (b) Effective full power years from plant startup. (c) Approximete fluence at 1/4 thickness reactor vessel wall at end of life (32 EFPY). (d) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY). I l 7-1. 1
1 SECTION
8.0 REFERENCES
1. Yanichko and Singer, " Commonwealth Edison Company Braidwood Station Unit No.1. Reactor Vessel Radiation Surveillance Program " HCAP-9807, February 1981. 2. 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. 3. Regulatory Guide 1.99, Proposed Revision 2, " Radiation Damage to Reactor l Vessel Materials". U.S. Nuclear Regulatory Commission, February,1986. 4. Section III of the ASME Boiler and Pressure Vessel Code,- Appendix G, i " Protection Against Nonductile Failure." 5. ASTM E208, " Standard Test Method for Conducting -Drop-Height : Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels." 6. ASTM E 185-82, " Standard Practice for Light-Hater Cooled Nuclear Power Reactor Vessels, E 706 (IF)." 7. ASTM E 23-88, " Standard Test Methods for Notched.Bar Impact Testing' of Metallic Materials." l 8. ASTM A 370-89, " Standard Test Methods and Definitions' for Mechanical Testing of Steel Products." l 4 E 8-89, " Standard Test Methods of Tension Testing of Metallic e als." l l ( l E 21-79, " Standard Practice for Elevated Temperature Tension Tests Metallic Materials." v i 1 06830:1D/082890 8-1 \\
- 11. ASTM E 83-85 " Standard Practice for Verification and Classification of.
Extensometers." 12. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Voi. 5--Two-Dimensional Discrete Ordinates Transport Technique", HANL-PR(LL)-034, Vol. 5. August 1970. 13. "0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, l 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Hater j Reactors". 14. J. V. Alexander, et. al., " Core Physics. Parameters and Plant Operations Data for the Braidwood Generating Station Unit 1 Cycle 1"..HCAP-10935, June 1986. (Proprietary) }
- 15. ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American h;.iety for Testing and Materials, Philadelphia, PA, 1984.
- 16. ASTM Designation E560-77, " Standard Recommended Practice for
[ Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM i Standards, Section 12, American Society for Testing and. Materials, i Philadelphia, PA, 1984.
- 17. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)",
o in ASTM Standards, Section 12, American Society for Testing and-Materials, Philadelphia, PA, 1984. 18. ASTM Designi'lon E706-81a, " Standard. Master Matrix for Light-Hater Reactor Pressure Vessel Surveillance Standard", in ASIM Standards', Section 12, American Society for Testing and Materials, 'Philadel'p'hia,.PA', 1984. 0683D:1D/092590 8-2
l
- 19. ASTM Designation E853-84, " Standard Practice for Analysis and
.i Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 20. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, l
Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, { Section 12, American. Society for Testing and Haterials, Philadelphia, PA, l 1984.
- 21. 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.
- 22. ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, 1
American Society for Testing and Materials, Philadelphia, PA,1984..
- 23. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
l
- 24. 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, PA, 1984.
- 25. 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.
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I 0683D:10/081490 8-3
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