ML20100H458
| ML20100H458 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 02/28/1985 |
| From: | Congedo T, Meyer T, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20100H438 | List: |
| References | |
| TAC-59962, TAC-59963, TAC-61512, TAC-61513, WCAP-10786, NUDOCS 8504090187 | |
| Download: ML20100H458 (88) | |
Text
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WCAP-10786 ANALYSIS OF CAPSULE U FROM THE DUKE POWER COMPANY MCGUIRE UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko T. V. Congedo W. T. Kaiser February 1985 APPROVED:
7[/ Ilvet A T. A. Meyer, Manager Structural Materials and Reliability Technology Work performed under Shop Order No. DVFJ-106 Prepared by Westinghouse Electric Corporation for the Duke Power Company Although information contained in this report is nonproprietary, no distribution shall be,made outside Westinghouse or its licensees without the customer's approval.
WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 8504090187 850405 8485B:1b-022585 DR ADOCK 050 9
p
PREFACE This report has been technically reviewed and verified.
Reviewer Sections 1 through 5, 7 and 8 R. S. Boggs Y) b Section 6 S. L. Anderson gicf b c7L Appendix A F. J. Witt N//lM
/
4 i
v J84858:1b-031985
TABLE OF CONTENTS Section Title Page 1
SUMMARY
OF RESULTS 1-1 2
INTRODUCTION 2-1 3
BACKGROUND 3-1 4
DESCRIPTION OF PROGRAM 4-1 5
TESTING 0F SPECIMENS FROM CAPSULE U 5-1 5-1.
Overview 5-1 5-2.
Charpy V-Notch Impact Test Results 5-3 5-3.
Tension Test Results 5-18 5-4.
Compact Tension Test Results 5-27 6
RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 6-1.
Introduction 6-1
~
6-2.
Discrete Ordinates Analysis 6-1 6-3.
Neutron Dosimetry 6-7 6-4.
Transport Analysis Results 6-11 6-5.
Dosimetry Results 6-19 7
SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8
REFERENCES 8-1 APPENDIX HEATUP AND C00LDOWN LIMIT CURVES FOR A
NORMAL OPERATION A-1 A-1.
Introduction A A-2.
Fracture Toughness Properties A-5 A-3.
Criteria for Allowable Pressure -
Temperature Relationships A-5 A-4.
Heatup and Cooldown Limit Curves A-11 8485B:1b-032185 v
LIST OF ILLUSTRATIONS Figures Title Page 4-1 Arrangement of Surveillance Capsules in McGuire Unit 1 Reactor Vessel-(Updated Lead Factors for the Capsules Shown in Parentheses) 4-2 4-2 Capsule U Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-5 5-1 Charpy V-Notch Impact Data for McGuire U' nit 1 Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) 5-9 5-2 Charpy V-Notch Impact Data for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-10 5-3 Charpy V-Notch Impact Data for McGuire Unit 1 Reactor Vessel Weld Metal 5-11 5-4 Charpy V-Notch Impact Data for McGuire Unit 1 Reactor Vessel Weld HAZ Metal 5-12 5-5 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) 5-13 5-6 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-14 5-7 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Weld Metal 5-15 i
j 5-8 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Weld HAZ Metal 5 5-9 Comparison of Actual versus Predicted 30-ft-lb Transition Temperature Increases for the McGuire
-Unit 1 Reactor Vessel Material-based on the Prediction Methods of Regulatory Guide 1.99, Revision 1 5-17 5 Tensile Properties for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) 5-20
-8485B:1b-032185 vii
LIST OF ILLUSTRATIONS (Cont)
Figures Title Page 5-11 Tensile Properties for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-21 5-12 Tensile Properties for McGuire Unit 1 Reactor Vessel Weld Metal 5-22 5-13 Fractured Tensile Specinens from McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) 5-23 5-14 Fractured Tensile Specimens from McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-24 5-15 Fractured Tensile Specinens from McGuire Unit 1 Reactor Vessel Weld Metal 5-25 5-16 Typical Stress-Strain Curve for Tension Specimens 5-26 6-1 McGuire Unit 1 Reactor Geometry 6-2 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-4 6-3 Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel - Surveillance Capsule Geometry 6-12 6-4 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-13 6-5 Relative' Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-14 6-6 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Surveillance Capsules 6-15 A-1 Predicted Adjustment of Reference Temperature, as a Function of Fluence, Copper, and Phosphorus Contents A-2 A-2 Fast Neutron Fluence (E > 1.0 Mev) as a Function of Full Power Service Life (EFPY)
A-4 A-3 McGuire Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 10 EFPY A-9 8485B:1b-032185 viii
LIST OF ILLUSTRATIONS (Cont)
Figures Title Page 1
McGuire Unit 1 Reactor Coolant System Cooldown A-4 Limitations Applicable for the First 10 EFPY A-10
.i 1
1 I
J
' 8485B:1b-032185-ix
=
i l-
!L-LIST OF TA8LES f
Table Title P_aie a
4-1 Chemical Composition and Heat Treatment of The McGuire-Unit 1 Reactor Vessel Surveillance Materials 4-3 5-1 Charpy.V-Notch Impact Data for the McGuire Unit 1 Intermediate Shell Plate 85012-1, Irradiated at 550*F, Fluence 4.'14 x 10 n/cm2 (E > 1 Mev) 5-4 10 5-2 Charpy V-Notch Impact Data for the McGuire Unit 1 Reactor Vessel Weld Metal and HAZ Metal, Irradiated at 550*F, Fluence 4.14 x 1018 (E > 1 Mev) 5-5 5-3 Instrumented Charpy Impact Test Results for McGuire Unit 1 Intermediate Shell Plate B5012-1, Irradiated at 4.14 x 10 n/cm2 (E > 1 Mev) 5-6 18 5-4 Instrumented Charpy Impact Test Results for McGuire Unit 1 Weld Metal and HAZ Metal, Irradiated at 4.14 x 10 n/cm2 (E > 1 Mev) 5-7 18 18 2
5-5 Effect of 550*F Irradiation at 4.14 x 10 n/cm j
(E > 1 Mev) on the Notch Toughness Properties of McGuire Unit 1 Reactor Vessel Materials 5-8 5-6 Tensile Properties for McGuire Unit 1 Reactor Vessel 18 2
Material Irradiated at 550*F to 4.14 x 10 n/cm (E > 1 Mev) 5-19 6-1 47 Group Energy Structure 6-5 6-2 Nuclear Constants for Neutron Flux Monitors Contained in the McGuire Unit 1' Surveillance Capsules 6 6-3 Calculated Fast Neutron Flux (E > 1.0 mev) and Lead-Factors for-McGuire Unit 1 Surveillance Capsules 6-16 6-4 Calculated Neutron Energy Spectra at the Center of j
the McGuire Unit.1 Surveillance Capsule U 6-17
'84858:lb-032185-xi
LIST OF TABLES (Cont)
Table Title Page 6-5
. Spectrum-Averaged Reaction Cross Sections at the Center of McGuire Unit 1 Surveillance Capsule U (0 = 56')
6-18 6-6 Irradiation History of McGuire Unit 1 Surveillance Capsule U 6-20 6-7 Comparison of Measured and Calculated-Fast Neutron Flux Monitor Saturated Activities for Capsule U 6-21 6-8 Results of Fast Neutron Dosimetry for Capsule U 6-22 6-9 Results of Thermal Neutron Dosimetry for Capsule U 6-23 6-10 Summary of Fast Neutron Dosimetry Results for
~
Capsule U 6-24 6-11 Calculated Current and EOL Vessel Exposure for McGuire Unit 1 6-25 A-1 McGuire Unit 1 Reactor Vessel Toughness Table A-3 t
'8485B:1b-032185
'xii
SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule U, the first capsule to be removed from the McGuire Unit i reactor pressure vessel, led to the following conclusions.
~
o The capsule received an average fast neutron fluence (E > 1.0 Mev) 18 2
of 4.14 x 10 n/cm,
o Irradiation of the reactor vessel intermediate shell plate B5012-1 to 18 2
4.14 x 10 n/cm resulted in 30 and 50 ft-lb transition temperature increases of 50*F and 40*F, respectively, for specimens oriented normal to the principal rolling direction of the plate, and 30 and 50 ft-lb transition temperature increases of 45'F for specimens oriented parallel to the plate principal rolling direction.
18 2
o Weld metal irradiated to 4.14 x 10 n/cm resulted in 30 and 50 ft-lb transition temperature increases of 160*F and 170*F, respectively.
o Weld HAZ metal showed a 30*F and 50*F transition temperature increase 18 2
of 90*F and 95'F, respectively, after irradiation to 4.14 x 10 n/cm,
o Plate B5012-1, weld metal, and HAZ metal all showed upper shelf energy 18 2
levels well above 50 f t-lb af ter irradiation to 4.14 x 10 n/cm,
1 84858:1b-021885 1-1
SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule U, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the McGuire Unit i reactor pressure vessel materials under actual operating conditions.
The surveillance program for tne McGuire Unit i reactor pressure vessel materials was designed and recommended by the Westinghouse Electric '
Corporation. A description of the surveillance program and the preirradiation mechanical. properties of the reactor vessel materials are presented by Davidson and Yanichko[1] - The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based un ASTM E-185-73, " Recommended Practice for Surveillance Tests'for Nuclear Reactor Vessels"[2]
Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens were performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance Capsule U removed from the McGuire Unit I reactor vessel and discusses the analysis of these data.
1 8485B:1b-021385 2-1
SECTION 3 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 ensuring safety in the nuclear in'dustry. 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 533 Grade B Class 1 (base material of the McGuire Unit I reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil ductility temperature (RTNDT)*
RT is deformed as the greater of either the drop weight nil-ductility NDT trensition temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material.. The RT f a given material is used to index that NDT material to a reference stress intensity factor curve (K curve) which IR appears in Appendix G of the ASME Code. The K curve is a lower boSnd of IR dynamics, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K curve, allowable stress intensity factors can be obtained for this IR material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.
8485B:1b-021885 3-1 m
n
l RT and, in turn, the operating limits of nuclear power plants can be NDT adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement of changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the McGuire Unit 1 Reactor Vessel Radiation Surveillance Prog' ram (1), in which a surveillance capsule is periodically removed from'the operating nuclear reactor and the encapsulated 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 RTNDT (RTNDT NDT initial + a RTNDT) is used to index the material to the KIR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
J E
l l
i i
8485B:1b-022285 3-2
SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the McGuire 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 shielding pads and the vessel wall as shown in figure 4-1.
The vertical center of the capsules is opposite the vertical center of the core.
Capsule U was removed from the reactor after 1.06 Effective Full Power Years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and Compact Tension (CT) specimens from submerged arc weld metal representative of the reactor vessel core region weld metal, and Charpy V-notch, tensile, CT, and bend bar specimens from the intermediate shell plate
~
B5012-1. The capsule also contained Charpy V-notch specimens from weld Heat Affected Zone (HAZ) metal. All heat affected zone specimens were obtained from the weld HAZ of plate B5012-1. The chemistry and heat treatment of the program surveillance materials is presented in table 4-1.
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. Some base metal Charpy V-notch and tensile specimens were oriented with the longitudinal axis of the specimens normal to (transverse orientation) and some parallel to (longitudinal orientation) the major working direction of the plate. The CT test specimens were machined so that the crack of the specimen _would propagate normal to (longitudinal specimens) and parallel to (transverse specimens) the major working direction of the plate. All specimens were fatigue precracked per ASTM E399-72. ~ The precracked bend bar was machined in the transverse orientation. Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens normal to (transverse orientation) the weld direction. Tensile specimens were oriented with the longitudinal axis of the specimens normal. to (transverse orientation) the weld direction.
l 8485B:1b-021885 4-1
12216-1a REACTOR VESSEL O.
CORE BARREL NEUTRON PAD Z (4.76) i U (4.76)
CAPSULE l
l (TYP)
V (4.06) 56' ~ l ' ~ 56
- d
- 38. *bf
~
\\
~
90' 270*
N s-
[
Y (4.06)
Q x (4.76)
I W (4.76) iso-Figure 41. Arrangement of Surveillance Capsules in McGuire Unit 1 Reactor Vessel (Updated Lead Factors for the Capsules Shown in Parentheses) 42
l t
TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE MCGUIRE UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Chemical Composition (wt%)
Element Plate B5012-1 Weld Material C-0.21 0.10 Mn 1.26 1.36
'1.19(a]
P 0.010 0.011 0.010(a]
j S
0.016 0.008 I
Si 0.23 0.24 0.23(a)
Cr 0.068 0.04 0.05(a]
4 Mo 0.57 0.55 0.54(a]
Ni 0.60 0.88 0.91(a) 1 Cu 0.087 0.21 0.20(a]
V 0.003 0.04 Sn 0.007 0.007 I
B
<0.003
<0.001 Cb
<0.001
<0.010 Ti 0.005
<0.010 W
<0.001
<0.010
-As 0.008 0.009 t
Zr 0.003
<0.001 Sb
<0.001 0.002 Pb 0.001
<0.001 N
0.003 0.008 2
Co 0.016 0.014 Heat Treatment History Material Temperature (*F)
Time (Hr)
Coolant Shell Plate B5012-1 1550-1650
- 4 Water quenched
-1200-1250' 4
Air cooled 1125-1175 40 Furnace cooled Weld 1125-1175 40 Furnace' cooled i
a.
Analysis performed on irradiated Charpy specimen DW-15 i
84858:1b-022285 4-3
+
^
. Capsule U contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15% cobalt (cadmium-shielded and unshielded).
In addition, cadmium 237 238 shielded dosimeters of Np and U were contained in the capsule.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. The composition of the two alloys and their melting points are as follows.
2.5% Ag, 97.5% Pb Melting point: 579*F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting point: 590'F.(312'C)
The arrangement of the various mechanical specimens, dosimeters and thermal j
. monitors contained in Capsule U are shown in figure 4-2.
1 t
4 8485B:1b-021385
.4-4
p.~
f e
)
i i
4 I
t o
5 h
I l
i 1
l h
i I
i CChPACT CChPACT CCbPACT COIPACT BEle BAR TENSILE TENSI m TENSIG4 CHMFY CHMFY OIMFY TEISIG4 TEISICH I
DW3 OWIS OW12 OHl2 OWS De 0E G415 OTI DW2 DW4 DW3 DW2 DWI OWie mile DWil mil l DES me (L4 CL3 DL2 OLI DE l
DWI OW13 CHIS DWlO OHIO OFF GC di i
a g3 e u a
I Cu Hd ll !! bW Al
.453C Ce 1
m, il i Fe 6 48 ll l LJLJLJ 7
s l]n if'iI fp
>-- Al
.151C Co (Cd)
,a m;;-.
l l tLf Ni a
1 f aP a f
k
-TO TOP OF VESM L i
t t
3x
12216-2 l
_l CORE i@
SPECIMEN nub 8ERIF., CODE:
oL - INTERMEDIATE SHELL PLATE B5012-1 (LONGITUDINAL)
DT - INTERMEDIATE SFELL PLATE B5012-1 (TRANSVERSE)
DW - WELD METAL DH - FEAT-AFFECTED-ZONE MATERIAL
&@a.
@ks
]
9,,
I Tl APERTUM CARD Also AvaHable On Aperture Card CAPSULE U I
Np237 I
u23e HARPY CHARPY TENSILE CHAFPY CHARPY CHARPY CHARPY CHARPY DH DN TENSILE
[
DH6 DW3 DH3 DL3 DTIS CLIS OTl2 DLl2 DT9 DL9 DTS DLS DT3 CI 3 DT3
[
DHS DW2 DH2 306 DL2 DTl4 DLl4 DTil DLil DTS DL8 DT5 DL5 DT2 DL2 DT4 DT3 DT2 DTI DT2 i
DH4 DWI DHI DLt DT13 DLl3 DTIO DLIO OT7 DL7 DT4 DL4 DTI r.A_ t DTI JL db
& ----e e- -- A l
. 3 5% C.
Cu W e---
Al
. 4 5% Ce 17e F*
Il I i.i l
i i LJLJLJ bdLJbJ s ll ]ll r
gr7r]
e----- A l
.15% Ca (Cd) g G-AI
.15% Co (Cd)
M --a II 8
ii i S?7Da --e l
! !! 'h '
!,, !h '
N1 Nt CENTER REDION OF VESSEL
?
TD eDTTDu Dr VESSEL 5
Figure 4-2. Capsule U Diagram Showing Loca-g' tion of Specimens, Thermal Moni-tors, and Dosimeters Mayo 90R'H3l
-5 4
i
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I L_
SECTION 5 TESTING OF SPECIMENS FROM CAPSULE U 5.1.
OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinahouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel._ Testing was performed in accordance with 10CFR50, Appendices G and H, ASTM Specification E185-82, and Westinghouse Procedure RMF-8402, Revision 0.
Upon receipt of the capsule at the laboratory, the specimens and spacer blocks
.were carefully removed, inspected for identification number, and checked aga' inst the master list in WCAP-9195W No discrepancies were found.
Examination of the two low-melting point 304*C (579"F) and 310*C' (590*F)
B' sed i
eutectic alloys indicated no melting of either type of thermal monitor.
a on this examination, the maximum temperature to which the test specimens were exposed was less than 304*C (579'F).
4 The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74,358J machine. The tup (striker) of i
the Charpy machine is instrumented with'an Effects Technology Model 500 j
instrumentation. system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E ).
From the load-time curve, the load of general yielding (Pg9),the D
time to general yielding (tgy), the maximum load (P ), and the time to.
l y
maximum load (t ) can be determined. Under some test conditions, a sharp y
i drop in load indicative of fast fracture was observed. The load at which fast-1 l
fracture was initiated is identified as the fast fracture load.(P ), and the p
load at which fast fracture terminated is-identified as_the' arrest load (P )*
A The energy at maximum load (E ) was determined by comparing the energy-time M
i record and the load-time record. The energy at maximum load is roughly equivalant to the energy required to initiate a crack in the specimen.
84858:1b-021885 1
=
I 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.
l Percent 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 the same specification.
Tension tests were performed on a 20,000 pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-81 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to Re 45. 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.
Deflection measurements were made with a linear variable displacement transducer-(LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen. failure. The i
extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 l
per ASTM E83-67.
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 each end of the gage section of a dummy specimen and in each grip.
In the i
test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and-lower grip and controller temperatures was developed over the range room temperature to 550*F (288'C)..The upper grip was used to control-the furnace temperature. During the actual testing the 8485B:1b-021885 5-2
7 grip temperatures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to +2*F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly.from the load extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fr'acture 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 i
The results'of Charpy V-notch impact tests performed on the various materials 18 2
contained in Capsule U irradiated at 4.14 x 10 n/cm are presented in tables'5-1 through 5-5 and figures 5-1 through 5-4.
The fractured surfaces of
}
the impact specimens are shown in figures 5-5 through 5-8.
Irradiation of Charpy V-notch impact specimens from the reactor vessel 18 2
intermediate shell plate, 85012-1, to 4.14 x 10 n/cm as shown in figure 5-1 resulted in 30 and 50 ft-lb transition temperature increases of 50*F and 40*F, res'pectively, for specimens oriented normal to the principal rolling direction (transverse orientation) of the plate. Specimens oriented parallel to the principal rolling direction'(longitudinal orientation) of the plate as shown in figure 5-2 exhibited a transition temperature increase of 45'F at both the 30 and 50 ft-lb index temperatures.
The upper shelf energy.
of the shell plate showed a 1 ft-lb decrease in the transverse direction and a 7 ft-lb decrease in the longitudinal direction.
18 2
Weld metal specimens irradiated to 4.14 x 10 n/cm resultad in 30 ft-lb-and 50 ft-lb transition. temperature increases of 160*F and 170*F, respectively, as shown in figure'5-3.
Irradiation caused the upper shelf energy of the weld metal to decrease 37 ft-lb to a shelf energy level of 75 ft-lb. A comparison of the 30 ft-lb transition temperature increase shown in figure 5 9 with the predicted increase based on U.S. Nuclear Regulatory i
I33 Commission-Regulatory Guide 1.99, Revision I indicates that the weld metal is more sensitive than predicted at a fluence of 4.14 x 1018.n/cm,
2 8485B:1b-032085 5-3_
t
TABLE 5 CHARPY V-NOTCH IMPACT DATA FOR THE MCGUIRE UNIT 1 INTERMEDIATE SHELL PLATE B5012-1 IRRADIATED AT 550*F, FLUENCE 4.14 x 10 n/cm2 (E > 1 Mov) 18 Temperature Impact Energy Lateral Expansion Shear Sample No.
(*F)
('C)
(ft-lb)
(J)
(mils)
(mm)
(%)
{
Longitudinal Orientation DL11 0
-18 17.0 23.0 14.5 0.37 1
DL10 50 10 23.0 31.0 21.5 0.55 23 DL5 50 10-35.0 47.5 29.5 0.75 22
~
DL6 76-24 78.0 106.0 52.5 1.33 39 DL14 76
.24 55.0 74.5 42.5 1.08 30 DL2
.76 24 45.0 61.0 33.5 0.85 31 DL4-100 38 54.0 73.0 46.0 1.17 32 4
i DL1 100 38 78.0 106.0 58.5 1.49 48 DL7 125 52 82.0 111.0 61.0 1.55 54 DL12
~150 66 109.0 148.0 78.0 1.98 69 DL9 200 93 119.0 161.5 81.0 2.06 85 c
DL15 250 121 135.0 183.0 87.5 2.22 100 DL3 300 149 133.0 180.5 84.0 2.13 100 i
DL8 400 204 130.0 176.5 79.0 2.01 100 Transverse Orientation DT7
-50'
-46 5.0 7.0 5.5 0.14 1
i DT15 0
-18 8.0 11.0 7.5 0.19 5
DT13 25
-4 24.0 32.5
- 21.0 0.53 6.
DT8 50 10 37.0 50.0-29.5' O.75 9
DT4 50 10 31.0 42.0 29.0 0.74 16 i
DT1 76 24 46.0 62.5 36.5 0.93 35 DT3 76 24 35.0 47.5 30.5 0.77 33'
'DT6 100 38
'47.0 63.5 42.0 1.07 44 DT10 100 38 42.0 57.0 37.0 0.94 43 i
DT2 125 52 58.0 78.5 44.0 1.12 53 DT12 150 66 72.0 97.5-57.5 1.46 73 l
DT11 200 93 81.0 110.0 59.5 1.51 85 DT5-250 121 102.0 138.5 72.0 1.83 100 DT9 300 149 104.0 141.0 72.0 1.83 100 DT14 400 204 95.0 129.0 65.0 1.65 100 4
l' d
84858:lb-021885 5-4i
- ~.. -. - --
TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE MCGUIRE UNIT 1 REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 4.14 x 10 n/cm2 (E'> 1 Mev) 18 Temperature Impact Energy Lateral Expansion Shear Sample No.
(*F)
(*C)
(ft-lb)
(J)
(mils)
(mm)
(%)
Weld Metal DW8 0
-18 4.0 5.5 5.0 0.13 4
DW12 76 24 12.0 16.5 11.0 0.28 19 DW11 120 49 24.0 32.5 22.5 0.57 27 DW4 140 60 35.0 47.5 33.0 0.84 35 DW15 150 66 33.0 44.5 32.0 0.81 49 DW6 150 66 36.0 49.0 33.0 0.84 51 DW9 175 79 31.0 42.0 30.5 0.77 54 DW10 175 79 34.0 46.0 33.5 0.85 67 DW13 200 93 37.0 50.0 36.0 0.91 75 DW5 220 104 68.0 92.0 55.5 1.41 87 DW2 220 104 71.0 96.5 60.0 1.52 98 DW1 250 121 71.0 96.5 61.5 1.56 98 DW14 300 149 71.0 96.5 71.5 1.82 98 DW7 400 204 79.0 107.0 68.5 1.74 100 DW3 450 232 78.0 106.0 69.5 1.77 100 HAZ Metal DH1
-50
-46 7.0 9.5 9.0 0.23 2
DH9 0
-18 29.0 39.5 23.5 0.60 27 DH8 50 10 37.0 50.0 35.5 0.90 48 DH13 50 10 42.0 57.0 32.5 0.83 36 DH10 76 24 35.0 47.5 35.0 0.89 45 DH12 76 24 35.0 47.5 29.0 0.74 38 DH2 100 38 78.0 106.0 62.5 1.57 69 DH7 100 38 52.0 70.5 46.0 1.17 57 DH11 120 49 77.0 104.5 58.5 1.49 76 DH6 150 66 49.0 66.5 43.5 1.10 64 DH15 150 66 83.0 112.5 65.0 1.65 80 DH14 200 93 78.0 106.0 62.0 1.57 85 DH4 250 121 100.0 135.5 77.5 1.97 100 DH3 300 149 106.0 143.5 77.0 1.96 100 DHS 400 204 119.0 161.5 82.5 2.10 100 8485B:1b-021885 5-5
TA8LE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR IACGUIRE UNIT 1-INTERIAEDI ATE SHELL PLATE 85012-1 IRRADIATED AT 4.14 x 10 n/cm2 (E > 1 IIev) 18 Normallzed Energies Test Charpy Charpy Mantmum Prop Yield Time Maximum Time to Fracture Arrest Y teld -
Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yleid Load Blaximum.
Load Load Stress Stress Number
(*F)
(ft-Ib)
(ft-ID/in2)
(ktps)
(usec)
(ktps)
(Usec)
(ktps)'
(kips)
(kst)
(kst)
Longitudinal Orientation DL11 0
17.0 137 110 27 3.35 110 3.90 295 3.80 111 120 DL5 50 35.0 282 214 68 3.05 90 4.15 505 4.10 100 119 DL10 50-23.0 185 120 65 3.15 90 3.75 310 3.75 0.40 104-114 DL2 76 45.0 362 208 154 2.75 75 3.95 505 3.80 1.60 90
-111 DL6 76 78.0 628 365 263 2.85 80 4.25 815 4.05 1.30 95 117 DL14 76 55.0 443 289 154 S.00 90 4.25 655 4.20 1.35 100 120 DL1 100 78.0 628 332 296 2.85 85 4.10 770 3.75 1.20 94 115 DL4 100 54.0 435 327 108 2.85 95 4.05 770 4.05 1.10 95 114 DL7 125 82.0 660 269 392 2.70 80 3.95 650 3.50 1.85 90 110 DL12 150 109.0 878 319 559 2.75 85 4.05 755 3.15 1.80-91 112 DL9 200 119.0 958 268 690 2.60 80 3.95 655 2.55 2.25 86 108 DL15 250 135.0 1087 290 797 2.15 50 3.75 725 71 98 DL3 300 133.0 1071 255 816 2.30 80 3.65 665 76 98 m
DL8 400 130.0 1047 285 742 1.90 55 3.50 770 63 89 ch Transverse Orientation DT7
-50 5.0 40 25 15 2.80 70 3.40 95 3.40 0.15 92 102 O
8.0 64' 39 25 3.10 85 3.45 130 3.45 0.15 102 108 DT15
. 25 24.0 193 146 48 3.10 85 4.00 355 3.80 103 118 DT13 DT8 50 37.0 298 219 79 3.10 95 4.25 510 4.20 103 122 DT4 50 31.0 250 185 65 3.00 80 4.10 435 4.05 99 117 DT1 76 46.0 370 241 130 2.80 75 4.10 560 3.80 1.15 92 114 DT3 76-35.0 282 172 110 3.15 90 4.10 410 4.00 0.90 104 120 DT6 100 47.0 378 211 168 2.90 85 4.05 505 3.90 1.10 96 115 DT10 100 42.0 338 202 136 2.85 80 4.00 485 4.00 1.65 95 114 DT2 125 58.0 467 241 226 2.70 75 4.05 570 3.85 1.70 88 111 DT12 150
/2.0 580
'277 303 2.85 85 4.10 650 3.65 2.15 93 114-DT11 200 81.0 652 270 383 2.65 85 3.95 655 3.85 2.80 87 109 DT5 250 102.0 821 253 568 2.25 55 3.80 630 74 100 DT9 300-104.0 837 247 591 2.20 65 3.70 640 72 97 DT14' 400 95.0 765 211 554 2.15 60 3.50 570 71 93
=
84858:Ib-021885
TABLE 5-4 INSTRUMENTED CHARPY' IMPACT TEST RESULTS FOR MCGUIRE UNIT'1 WELD METAL AND HAZ METAL IRRADIATED 30 AT 4.14 x 10 n/cm (E > 1 Mev)
Normaltzed Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yteld Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2
. haber.
(*F)
(ft-Ib)
(ft-Ib/in )
(ktps)
(psec)
(kips)
(psec)
(ktps)
(kips)
(kst)
(kst)
Weld Metal DW8 0
4.0 -
32 25 7
3.55 95 3.45 LDW12 76 12.0
. 97 33 64 3.10 100 3.30 125 3,30 0.75
'102 105 DW11 120'
-24.0 193 139 54 3.00 85 3.75 355 3.65 1.05 99 111 DW3 140 35.0 282 192 90 3.00 85 3.95 460 4.00 1.40 99 115 DW15 150 33.0-266 179 87 2.9 90 3.80 450 3.80 1.10 95 111.
DW6.,
ISO 36.0 290 204 85 2.95 85 3.90' 495 3.80 1.40 98 114 DW9"
- 75 31.0 250 156 93 2.95 85 3.75 395 3.75 1.60 97 110 DW10 175 34.0 274 178 96 2.90 85' 3.80 445 3.80 2.00 96 111 DW13 200 37.0 298 159 139 2.9 80 3.75 400 3.75 2'.05 95 110 DW5 220-68.0 548 203 345' 2.60 70 3.85 495 86 107 DW2 220 71.0 572 228' 344 2.65 85 3.85 560 88 108 DW1 250 71.0 572 201 370 2.75-85 3.80 505 91 108 DW14 300 71.0 572 194 378 2.60 80 3.60 505 85
'102 DW7 400 79.0 636 191 445 2.35 70 3.60 505 78 99 O
DW3 450 78.0 628 217 411 2.15 60 3.55 580 70 94
~
HAZ Metal DH1
-50 7.0 56
-11 46 2.50 60 2.50 0.70 DH9-0 29.0 234 147 86 3.45 90 4.10 350 4.10 1.15 113 124 DH8 50 37.0 298 103 195 3.35 85 3.85 260 3.75 2.25 111 119 DH13 50 42.0 338 184 155 3.35 85 4.15 420 4.20 2.90 til 124 DH10 76 35.0 282 99 183 3.10 80 3.65 260 3.45 2.10 102 111 DH12 76-35.0 282 138
~ 14 4 3.35 90 4.00 335 3.80 2.05 111 121
-DH2 100 78.0 628 295 333 3.15 90 4.25 665 4.00 2.75-105 122 DH7 100 52.0 419 143 276 3.15 90 3.90 355 105 117
-DH11
.120
.77.0 620 279 341 3.10 85 4.00 650 102 117 DH6-150 49.0 395 179-216 3.10 95 3.85 445 3.70 2.65 102 115
.DH15 150 83.0 668 287 381 2.90 90 4.20 665 96 117.
-DH14 200 78.0 628 203 425 2.90 85 3.80 500 97 111 DH4 250' 100.0 805 323 482 2.65 85 3.95 770 88 109 DH3 300 106.0 854 334 519 2.50 75 3.85 815 82 105 DHS 400 119.0 958 286 672 2.30 55 3.70 725 76 99 84858:ib-021885
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Average 50 f t-Ib Temperatt re (*F )
Li f
l Material Unteradiated Irradisted AT U
Plate B5012-1 75 115 40 (Transverse)
\\
Rate 85012-1 35 80 45 i
(Loagttudinal)
Weld metal 20 190 170 HAZ metal
-5 90 95 i
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TABLE'5-5
- T OF 550*F IRRADIATIOri AT 4.14 x 10 n/cm2 (E > 1 Mev) 10 ON NOTCH TOUGHt4ESS PROPERTIES OF MCGUIRE UNIT 1 RE ACT OR VESSEL MATERIALS Average 35 ml]
Average Average Energy teral Expansion Temperature (* F )
30 ft-lb Temperature (*F)
Absorbtion at Full Shear (ft-ID) birradiated Irradiated M Unteradiated Irradiated AT Unirradiated Irradiated Art Ib 50 85 35 0
50 50 101 100
-1 25 70 35 5
50 45 140 133
-7 0
170 170
-5 155 160 112 75
-37 J
-15 70 85
-50 40 90 118 108
-10 Mao Available On Apertilre Card l
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Figure 5-1. Charpy V-Notch impact Data for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) l 5-9
12216-4 (OC) 50 0
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$ 80 e
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.:* 45 F' 0
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0 0
100 0
100 200 300 400 500 TEMPERATURE (OF) l Figure 5-2. Charpy V Notch Impact Data for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-10
i 12216-5 (OC)
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Figure 5-3. Charpy V-Notch Impact Data for McGuire Unit 1 Reactor Vessel Weld Metal 5-11
i
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Rgure 5-4. Charpy V-Notch impact Data for McGuire Unit 1 Reactor Vessel Weld HAZ Metal 5-12
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-g
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1 3
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Charpy impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Transverse Orientation) 5-13 l
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Charpy impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 5-14
l 12216C-6 1
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1
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Charpy impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Weld Metal 5-15
12216C-7 ww.
p 3
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1sk __ '
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l f
a f
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Charpy impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Weld HAZ Metal l
l l
5-16
12216-7 l
s 400 1
300 gd*
c g.G s.*
S
{200 A
E WELD METAL E
d
S M100
- ,0 g
U g 80 s
9 O
PLATE B50121 g 40 E
H CODE m
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O PLATE B5012-1 (TRANS.)
9 PLATE B5012-1 (LONG.)
u.
g 20 6 WELD METAL V HAZ METAL I
I I
I I i 11 I
I I
I IIll 30 1018 2
4 6
8 1019 2
4 6
8 1920 2
FLUENCE (n/cm )
Figure 5-9. Comparison of Actual versus Predicted 30 ft lb Transition Temperature Increases for the McGuire Unit 1 Reactor Vessel Material Based on the Prediction Methods of Regulatory Guide 1.99, Revision 1 5-17
This type of behavior is not unusual at neutron fluence levels less than 19 2
1 x 10 n/cm for relatively high copper reactor vessel weld and base materials (greater than 0.15 percent copper).
18 2
Weld HAZ specimens irradiated to 4.14 x 10 n/cm resulted in 30 ft-lb and 50 ft-lb transition temperature increases of 90*F and 95'F, respectively, as shown in figure 5-4.
The upper shelf energy of the HAZ metal decreased by 10 ft-lb due to the irradiation.
The fracture appearance of each irradiated Charpy specimen from the various inadiated materials is shown in figures 5-5 through 5-8.
Each of the vessel materials shows an increasing ductile or tougher appearance with increasing test temperature.
5-3.
TENSION TEST PESULTS The results of tension tests performed on material from the reactor vessel 18 3
intermediate shell plate B5012-1 and weld metal irradiated to 4.14 x 10 2
n/cm are shown in table 5-6 and' figures 5-10 through 5-12. Plate B5012-1 test results are shown in figures 5-10 and 5-11 and indicate that irradiation 18 2
to 4.14 x 10 n/cm caused ~6 ksi increase in 0.2 percent yield strength and a 5 ksi increase in ultimate tensile strength. Weld metal tension test results presented in figure 5-12 show that the 0.2 percent yield strength and ultimate tensile strength increased - 18 ksi with irradiation.
The fractured tension specimens for the plate material are shown in figures 5-13 and 5-14, while the fractured tension specimens for the weld metal are shown in figure 5-15. A typical stress strain curve for the tension tests are shown in figure 5-16.
t 8485B:~1b-021885 5-18
~
A TABLE 5-6 TENSILE PROPERTIES FOR MCGUIRE UNIT 1
. REACTOR VESSEL MATERIAL IRRADIATED AT 550*F TO 4.14 x 10 n/cm2 (E > 1.0 Mev) 18 Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction-Temperature Strength Strength Load Stress Strength Elongatton Elongation in Area Number' Materia).
(*F)
(kst)
(ksi)
(kip)
(ksi)
(ksi)
(%)
(%)'
(%)
DT3 Plate 85012-1 100 71.3
.91.3 3.06 173.2 62.3 11.7 23.7 64 CT1
-(transverse 200 68.8 87.6 3.18 152.2
.64.7 10.5 20.6 57 orientatlon)
CT2 550 64.7 87.6 3.22 126.5 65.6 10.5 19.0 48 DL1 Plate B5012-1 100 71.3 91.7 2.92 174.4 59.5 11.4 25.7 66 DL3' (longitudinal 200 68.8 86.6 2.75 155.6 56.0 10.5 23.4-64 orientatlon)
DL2 550 63.7 87.6 3.00 181.7 61.1 9.7 20.3 66 DW3 Weld metal
. 175 78.9 93.1 3.30 214.4 67.2
.11.6 22.7 69 DW1 Weld metal 225 77.4 92.1 3.20 169.9 65.2 11.3 21.7 62 Dw2 Weld metal 550 71.3 90.7 3.45 195.2 70.3 9.6 18.0 64 us Le 4
8485B:1D-021385 I
12216-8 (OC) 0 50 100 150 200 250 300 l
l l
1.
I I
l-800 110 i
~
100
-ULTIMATE TENSILE i
STRENGTH A-600 g l 80 4
E m
$ 70
' N g __
g0.2% YlELD STRENGTH 500 60 Q-400 2
i m
300 80 l
2 70
'C n
e S
60 REDUCTION IN AREA
,w
' E 50 CODE e
SO UNIRRADIATED A G IRRADIATED AT 4.14 X 1018 n/cm2 3 40 Po TOTAL ELONGATION 3 30 O
u A
3
~
20 A
a O
n n
10 O
5 CUNIFORM ELONGATION l
I I
I I
o O
100 200 300 400 500 600 TEMPERA,TURE (OF)
Figure 5-10. Tensile Properties for McGuire Unit 1 Reactor Vessel Shell Plate 85012-1 (Transverse Orientation) 5-20
12216 9 (OC) 0
,50 100 150 200 250 300 120 l
l l
l l
l l-800 110 700 100 ULTIMATE TENSILE STRENGTH
- :- 90 Q-3- =0,
e s 80 m
E y
2 SM
$ 70 8
%._N O
O-4@
0.2% YlELD STRENGTH i
50 300 40 l
80 t
70 REDUCTION IN AREA 8-* ~.
c
-e O
60 CODE O
6 50 60 UNIRRADIATED M
A9 IRRADIATED AT 4.14 X 1018 n/cm2 3 40 P
4 o
@ 30 4A_
TOTAL ELONGATION
_A-A a
20 g
n n
- ~
-8 g
l 10 i
UNIFORM ELONGATION l
l l
O 0
100 200 300 400 500 600 TEMPERATURE ( F)
Figure 5-11. Tensile Properties for McGuire Unit 1 Reactor Vessel Shell Plate B50121 (Longitudinal Orientation) i 5-21 l
l
\\
f 12216-10
( c) 0 50 100 150 200 250 300 120 l
l l
l l
l l-800 110
~
100
= 90 A
600 2
2 ULTIMATE TENSlLE STRENGTH g
Me, 4
i i
$ 70 g-500
=
0.2% YEILD STRENGTH Q-400 50 300 40 90 80 70 O
g\\*-
_-J n
g REDUCTION IN AREA
~
50 D
AO UNIRRADIATED
=f
&9 IRRADIATED AT 4.14 X 1018 n/cm2 y 40 2
30 A
TOTAL ELONGATION A
A ---
n e-e-0 2
UNIFORM ELONGATION I
I I
I I
O 100 200 300 400 500 600 TEMPERATURE (OF)
I Figure 5-12. Tensile Properties for McGuire Unit 1 Reactor Vessel Weld Metal l
5-22 i
12216C-1 Mty "7 g wr m - v
,..... p..,~.4~..
r.
- y
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SPECIMEN DT-2 550 F 1
l Figure 5-13. Fractured Tensile Specimens From McGuire Unit 1 Reactor Vessel Shell Plate 85012-1 (Transverse Orientation) 5-23
12216C-2 m, n r e n n m m e n w ~ v x,~ r -
s.
4.-
e.;
J f
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SPECIMEN DL-2 550 F Figure 5-14. Fractured Tensile Specimens From McGuire Unit 1 Reactor Vessel Shell Plate B5012-1 (Longitudinal Orientation) 1 5 24
12216C4 i
.,-r~. wn w,.,ner,-e,-~~ r u..
.4,
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- -. mmm wn,v..r;
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- . 7 : ;
o-t
.Ic.,-s 9.3 cc s-.g
- r<me -
1 :-
t -
_uar amwau mlL
.4...
.a&
m -
0 SPECIMEN DW-2 550 F l
[
Figure 5-15. Fractured Tensile Specimens From McGuire Unit 1 Reactor l
Vessel Weld Metal 5-25
12216-11 120 108 -
96 -
84 72 d
$ 60 I
~
E
$ 48 36 24 SPECIMEN DW 1 2250 F 12 l
i I
I I
I I
I I
a O
0.03 0.06 0.09 0.12 0.15 0.18 0.21 0.24 0.27 0.30 STR AIN (in./in.)
Figure 5-16. Typical Stress - Strain Curve for Tension Specimens 5-26
o 5-4.
COMPACT TENSION TEST RESULTS The 1/2 T compact tension fracture mechanics specimens that were contained in Capsule U have been stored at the Westinghouse Research and Development Laboratory and will be tested and reported on at a later time.
t 8485B:1b-021885 5-27
'SECTION 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1.
INTRODUCTION I
Knowledge of th'e neutron environment within the pressure vessel / surveillance capsule geometry is required as an integral part of LWR pressure vessel l
surveillance programs for two reasons. First, in the interpretation of
' radiation-induced properties, changes observed in materials test specimens and the neutron environment (fluence, flux) to which the. test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship must be established between the environment at various -
positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.
Tho latter information is derived solely from analysis.
l This section describes a discrete ordinates S transport analysis performed n
for the McGuire Unit I reactor to. determine the fast neutron (E > 1.0 Mev) flux and fluence, as well as the neutron energy spectra within the reactor vessel and surveillance capsules. The analytical data were then used to I'
develop lead factors-for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross sections derived from this calculation, the analysis.of the neutron dosimetry contained in Capsule U is presented.
6-2.
DISCRETE ORDINATES ANALYSIS A plan view of the McGuire Unit i reactor geometry at~the. core midplane is
-l shown in figure 6-1.
Since the reactor exhibits 1/8th core symmetry, only a i
0- to 45-degree sector is depicted. Six irradiation capsules attached to 1
8485B:1b-021385' 6-1
d 4
0=O' 122i8-12 g
e CORE BARREL
/
0=25*
i 0=31.5*
0=34' I
NEUTRON PAD 0=45'
[
/
/
//
/'/
)
/ {
/j
[/
j i
l
//
I l
//
\\
1.0 Mev) to j
measured material property changes. To properly account for' burnout of the product isotope generated by fast neutron' reactions,. it is <necessary to also determine the magnitude of the thermal neutron flux at the monitor locat' ion.
l Therefore, bare and cadmium-covered cobalt-aluminum monitors were also I
included.
}
TABLE 6-$
n i
l-NUCLEAR CONSTANTS'FOR NEUTRON FLUX' MONITORS. CONTAINED IN
,/
THE McGUIRE UNIT 1 SURVEILLANCE CAPSULES
(
c I
Target Fission l
Weignt Product Yield Moni tor 'Ma t'erial Reaction of Interest-Fraction Half-Life
(%)
. Cub 3(n,a)Co60 0.6917 5.27 years' Iron Fe54 (n p) Hn54 0.0585 314 days Nickel
.NiS8 (n,p) CoS8 0.6777 71.4 days I~
Uranium-238(a)
.U238 (n,f) Cs137 '
1.0 30.2 years 6.3 Neptunium-237(a]
.Np237 (n,f) Cs137 1.0 30.2 years 6.5 Cobalt-aluminum (a]
CoS9 (n,r) Co60 0.0015 5.27 years Cobalt-aluminum
.CoS9(n,r)Co60 0.0015 5.27 years
.a.
Denotes that monitor is= cadmium-shielded 84858:1b-0213851 6-7
~
- s-
The relative locations of the various monitors within the surveillance capsule are shown in figure 4-2.
The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in table 6-2 does not yield a direct measure of the energy-dopendent flux level at the point of interest.
Rather, the activation or fission process is a measure of the integrated s
effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of
.the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.
In particular, the followin'g variables are of interest.
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 analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two operations. First, the disintegration rate of product isotope per unit mass of monitor must be determined. Second, in order to define a suitable spectrum-averaged reaction cross section, the neutron energy spectru;a at the monitor location must be calculated.
The specific activity of each of the monitors is determined using established ASTM procedures (7,8,9,10,11]
Following sample preparation, the activity s
' of each monitor is determined by means of a lithium-drif ted germanium, Ge(Li),
aamma spectrometer.
The overall standard deviation of the measured dcta is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. for the samples removed from 84858:1b-021885 6-8 o
McGuire Unit 1, the overall 2e deviation in the measured data is determined to be +10 percent. The neutron energy spectra are determined analytically using the method described in section 6-1.
Having the measured activity of the monitors and the neutron energy spectra at the locations of interest, the calculation of the neutron flux proceeds as follows. The reaction product activity in the monitor is expressed as follows.
R=[N n
P.
-At ),-Atd 4
(6-2)
(E)e(E)dE E / max (1 e 3
fy E
J=1 where R
induced product activity
=
Avogadro's number N
=
g A
atomic weight of the target isotope
=
f weight fraction of the target isotope in t % target material
=
3 Y
number of product atoms produced per reaction
=
o(E)-
energy dependent reaction cross section
=
o(E) energy dependent neutron flux at the monitor location with
=
the reactor at full power P
3 average core power level during irradiation period j
=
P,,x maximum or reference core power level
=
A
=
- decay constant of the product isotope length of irradiation period j t
=
3 decay time following irradiation period j t
=
d Because neutron flux distributions are calculated using multigroup transport
. methods and, further, because the prime intarest is in the fast neutron flux above 1.0 Mev, spectrum-averaged reaction eross sections are defined so that the integral term in equation 6-2 is replaced by the following relationship.
,j
~
8485B:1b-022285 6-9
7_
E..o(E)#(E)dE = o # (E > 1.0 Mev) where N
.["a(E)#(E)dE I e#99
, Jo
, a=1 N
- (E)dE I
'9 1.0 Nov 9'9.0 Mev 1
Therefore, equation 6-2 is rewritten N
P R=[N h(1-e-it ),-Atd f Y
- (E > 1.0 Mev) I 3
3 i
j=1 max i
or, solving for the neutron flux, l
R
- (E > l'.0 Mev) =
I6~3)
-it ),-Xtd f Y3I (1-e 3
g j=1 max The total fluence above 1.0 Mev is then given by n
P.
6(E> 1.0 Mov) = #(E> 1.0 Mev) I htj
.(6-4) a=1 max
- where i
n P
total effective full power seconds of reactor j[1 3 = operation up to the time of capsule removal t
,, x i
An assessment'of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered CoS9 (n,r) Co60 data by
-means of cadmium ratios and the use of a 37-barn, 2,200 m/sec cross section.
i Rbare
'Th
- n f
y,
(.,-it ),-A td (6-5) 5 3
J=1 max where D=Rbare/ rcd covered -
l=
8485B:1b-022285 6-10 t
6-4.
TRANSPORT ANALYSIS RESULTS Results of the S. transport calculations for the McGuire Unit I reactor are n
summarized in figures 6-3 through 6-6 and in tables 6-3 through 6-5.
In figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius,1/4-thickness location, and 3/4-thickness location are presented as a function of azimuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident.
In figure 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 Mev) through the thickness of the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel is given in figure 6-5.
Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6-3 or 6-4 by the appropriate values from figure 6-5.
In figure 6-6, the radial variations of fast neutron. flux within the surveillance capsules are presented. These data, in conjunction with the ma'ximum vessel flux, are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E > 1.0 Mev) at the dosimeter block location (capsule center) to the maximum fast' neutron flux at the pressure vessel inner radius. The updated lead factors for the McGuire Unit 1 surveillance capsules are listed-in table 6-3.
The neutron flux monitors contained within the surveillance capsule are all located at the same radial location, the capsule center. Had they been located at different radial locations within the capsules it would have been necessary to adjust the disintegration rates for the gradients that exist within the capsules.
In the present analysis, the point of comparison for all.
reaction rates is, of course, the capsule center.
In order to derive neutron flux and fluence levels from the measured disin_tegration rates, suitable spectum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center of the McGuire Unit 1 surveillance Capsule U is listed-in table 6-4.
The associated spectrum-averaged cross sections for each of the fast neutron reactions are.given.in table 6-5.
'8485B:1b-022285 6-11
-..... -... ~
b
1221614 l
100.0 70.0 50.0 o
"o 20.0 SURVEILLANCE CAPSULE R= 207.31 cm I 10.0 5
5.0 PRESSURE VESSEL s
1 INNER RADIUS (IR) 2.0
/ T LOCAT 4
1.0 a
d 0.7
.5 3/ T LOCATION 4
0.2 l
l I
I I
I I
I 0.i 45 50 55 60 65 70 75 80 85 90 AZIMUTHAL ANGLE (0) l Figure 6-3. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel Surveillance Capsule Geometry 6-12
1221615 100.0 70.0 219.71 50.0 l
30.0 225.19 m
l S 20.0 l
M
~
$ 10.0 s
=
c 7,0 4
5 5.0 236.14 B2 3.0 A 2.0 l
241.62 l
E 4
X t
31.0 l
Z
' O.7 0.5 0.;
g 0.2 INNER RADIUS OUTER RADIUS (IR) 1/4T 3/4T (OR)
I I
I I
0.1 215 220 225 230 235 240 245 RADIUS (cm)
Figure 6-4. Calculated Radial Distribution of Maximum Fast Neutron Flux (E >1.0 Mev) Within the Pressure Vessel 6-13
12216 16 1.000 0.700 -
0.500 -
0.200 0.100 X
3 0.070 u.
$ 0.050 E
D E
s P 0.020 3
a 0.010 0.007 0.005 CORE MIDPLANE 0.002 TO VESSEL lI CLOSURE HEAD l
l l
0.001
-300 200 100 0
100 200 300 DISTANCE FROM CORE MIDPLANE (cm)
Figure 6-5. Relative Axial Variation of Fast Neutron Flux (E >1.0 Mev) Within the Pressure Vessel 6-14 O
1 12216 17 100 70 50 CAPSULES U, W, X, Z "o
205.73 5
207.31 20 208.90 E
%2 10 CAPSULES V, Y
- s 7
-n E
5 5
am
~
IR Q.
OR l
I I
I I
i 200.0 202.5 205.0 207.5 210.0 212.5 215.0 RADIUS (cm)
Figure 6-6. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Surveillance Capsules 6-15
TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 Mev) AND LEAD FACTOR FOR McGUIRE UNIT 1 SURVEILLANCE CAPSULES Azimuthal 9(E > 1.0 Mev)
Lead 2
Capsule Location (*)
(n/cm - see)
Factor 11 V
56' 1.4231x10 4.76 11 W
124' 1.4231x10 4.76 11 X
236' 1.4231x10 4.76 ll 2
304' 1.4231x10 4.76 11 V
58.5' 1.21536x10 4.06 11 Y
238.5*
1.21536x10 4.06 8485B:1b-031985 6-16
TABLE 6-4 CALCULATED NEUTRCN ENERGY SPECTRA AT THE CENTER OF THE McGUIRE UNIT 1 SURVEILLANCE CAPSULE U Group Group 2
2 No.
4 (n/cm -sec)
No.
4 (n/cm sec) 7 10 1
2.33290 x 10 25 8.24716 x 10 7
10 2
8.56092 x 10 26 8.59363 x 10 8
10 3
2.98276 x 10 27 6.91292 x 10 8
10 4
5.45563 x 10 28 4.84202 x 10 8
10 5
9.11017 x 10 29 1.44982 x 10 9
9 6
2.02395 x 10 30 7.80675 x 10 9
10 7
2.81044 x 10 31 2.13876 x 10 9
10 8
5.80793 x 10 32 1.38042 x 10 9
10 9
5.41035 x 10 33 2.43642 x 10 9
10 10 4.54219 x 10 34 3.56461 x 10 9
10 11 5.47463 x 10 35 6.05357 x 10 9
10 12 2.74190 x 10 36 5.44679 x 10 9
10 13 8.41871 x 10 37 7.61645 x 10 9
10 14 4.24241 x 10 38 4.13856 x 10 10 10 15 1.14137 x 10 39 4.59692 x 10 10 10 16 1.55890 x 10 40 6.22153 x 10 10 10 17 2.43599 x 10 41 7.34671 x 10 10 10 18 5.51878 x 10 42 4.06677 x 10 10 10 19 4.25406 x 10 43 4.63747 x 10 10 10 20 2.01268 x 10 44 2.87056 x 10 10 10 21 7.41069 x 10 45 2.22168 x 10 10 10 22 5.81217 x 10 46 3.11690 x 10 10 10 23 7.26368 x 10 47 3.97555 x 10 10 24 7.08945 x 10 8485B:1b-021885 6-17
TABLE 6-5 SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF McGUIRE UNIT 1 SURVEILLANCE CAPSULE U (e=56')
[a]
3 (barns)
Reaction Fe54 (n,p) Mn54 0.0559 CuS3 (n,a) Co60 0.000479 NiS8 (n.p) CoS8 0.0779 Np237 (n,f) Cs137 3.338 U238(n,f)Cs137 0.313 CoS9 (n,r) Co60 23 o=f".o(E)$(E)dE o
a.
f"Mev
- (E)dE 1
8485B:1b-021885 6-18
6-5.
DOSIMETRY RESULTS The irradiation history of the McGuire Unit I reactor up to the time of removal of Capsule V is listed in table 6-6.
Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsule U based on the irradiation history shown in table 6-6 are listed in table 6-7.
The data are presented as measured at the capsule center.
The fast neutron (E > 1.0 Mev) flux and fluence levels derived for Capsule V are presented in table 6-8.
The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in table 6-9.
Due to the relatively low thermal neutron flux at the capsule location, no burnup correction was made to any of the measured activities.
The maximum error introduced by this assumption is estimated to be < 1 percent for the NiS8 (n,p)CoS8 reaction and even less significant for all of the other fast neutron reactions.
An examination of table 6-8 shows that the fast neutron flux (E > 1.0 Mev) 11 derived from the five threshold reactions ranges from 1.12 x 10 to 11 2
1.35 x 10 n/cm sec, a total span of less than 21 percent.
It may also 11 2
be noted that the calculated flux value of 1.42x10 n/cm sec exceeds all of the measured values, with calculation to experimental ratios ranging from 1.05 to 1.27.
Comparisons of measured and calculated current fast neutron exposures for Capsule U and the Inner Radius (IR) of the pressure vessel are presented in table 6-10.
Calculated current and E0L vessel exposures are presented in table 6-11, for vessel inner radius,1/4 thickness and 3/4 thickness. The measured value is given based on the average of all five threshold reactions.
Based on the data given in table 6-10, the best estimate exposure of Capsule U is as below.
18 4T = 4.13 x 10 n/cm2 (E > 1.0 Mev) 8485B:1b-031985 6-19
TABLE 6-6 IRRADIATION HISTORY OF McGUIRE UNIT 1 SURVEILLANCE CAPSULE U Irradiation Time [a]
Decay Time [b]
Pj Pmax Month Year (mw)
(mw)
P /P (day)
(day) 3 gx 10 1981 296 3565 0.083 31 1077 11 1981 1089 3565 0.306 30 1047 12 1981 112 3565 0.031 31 1016 1
1982 1771 3565 0.497 31 985 2
1982 1443 3565 0.405 28 957 3
1982 694 3565 0.195 31 926 4
1982 1660 3565 0.466 30 896 5
1982 2000 3565 0.561 31 865 6
1982 1804 3565 0.506 30 835 7
1982 755 3565 0.212 31 804 8
1982 1981 3565 0.556 31 773 9
1982 2040 3565 0.572 30 743 10 1982 1814 3565 0.509 31 712 11 1982 735 3565 0.206 30 682 12 1982 1702 3565 0.477 31 651 1
1983 1154 3565 0.324 31 620 2
1983 0
3565 0.000 28 592 3
1983 0
3565 0.000 31 561 4
1983 0
3565 0.000 30 531 5
1983 233 3565 0.065 31 500 6
1983 2783 3565 0.781 30 470 7
1983 2718 3565 0.762 31 439 8
1983 1547 3565 0.434 31 408 9
1983 3225 3565 0.905 30 378 10 1983 2682 3565 0.752 31 347 11 1983 2619 3565 0.735 30 317 12 1983 2804 3565 0.787 31 286 1
1984 3146 3565 0.882 31 255 2
1984 2692 3565 0.755 29 226 3
1984 0
3565 0.000 1
225 7
a.
Total irradiation time = 3.36 x 10 Effective Full Power Seconds (EFPS) b.
Decay time is referenced to 10/12/84 8485B:1b-022285 6-20
TABLE 6-7
~
COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX HONITOR SATURATED ACTIVITIES'FOR CAPSULE U f
~
Saturated Activity Reaction Radial (dps/gm) and Location Actual for Calculated for
=
Axial Position (cm)
Capsule U Capsule U
'Fe54 (n p) Mn54 r
3.97x10l Top 207.31 Middle 207.31 4.10 x 100 Bottom 207.31 4.16 x 10 6
6 AVERAGE 4.08 x 10 5.19 x 10 Cu63(n,a)Co 60' 3.96x10f Top 207.31 Middle 207.31 4.16 x 10 5 Bottom 207.31 3.98 x 10 5
5 AVERAGE 4.03 x 10 4.50 x 10 NiS8 (n,p) CoS8 7
Top 207.31 6.04 x 10 7 Middle 207.31 6.30 x 10 7 Bottom 207.31 6.52 x 10 7
7 AVERAGE 6.29 x 10 7.81 x 10 Np237 (n,f) Cs'137 7
7 Middle 207.31 7.01 x 10 7.80 x 10 U238(n,f)Cs137 6
6 Middle 207.31 7.65 x 10 7.10 x 10 8485B:1b-021885 6-21
l TABLE 6-8 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE U f (E > 2).0 Mev)
- 1 Adjusted Saturated Activity
@ ( E >2 1.0 Mev)
(n/cm (dos /gm)
(n/cm -sec)
Reactton Measured Calculated Measured Calculated Me5sured Calculated Fe (n,p)Mn 4.08 x 10 5.19 x 10 1.12 x 10 '
1.42 x 10 '
3.75 x 10 4.78 x 10 Cu (n.a)Co 4.03 x 10 4.50 x to 1.27 x 10 1.42 x 10 4.27 x 10 4.78 x 10 10 '
3.85 x 10 4.78 x 10 N1 (n.p)Co 6.29 x 10 7.81 x 10 1.15 x 10' 1.42 x Np (n,f)Cs 7.01 x 10 7.80 x 10 1.27 x 10 '
1.42 x 10 4.27 x 10 4.78 x 10 6.73 x 10 7.10 x 10 1.35 x 10 1.42 x 10 4.52 x 10' 4.78 x 10 (n.f)Cs' U
235
- a. U238 adjusted saturated activity has been multiplied by O.88 to correct for 350 ppm U 1mpurity
?
l3 84850:1b-022285
TABLE 6-9 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE U Saturated Activity (dps/gm)
'Th 2
Axial Location Bare Cd-covered (n/cm -sec) 8 7
11 Top' 1.13 x 10 5.65 x 10 1.59 x 10 7
7 11 Middle 9.45 x 10 5.59 x 10 1.09 x 10 7
7 11 Bottom 9.37 x 10 5.68 x 10 1.05 x 10 8
7 11 AVERAGE 1.00 x 10 5.64 x 10 1.24 x 10 J
6 i
i l
8485B:1b-031985 6
c TABLE 6-10
SUMMARY
OF FAST NEUTRON DOSIMETRY RESULTS FOR CAPSULE U Calculated Irradiation vessel vessel Itme
((E>1.p Mev) 4(E>1.0,Mev) tead F luenp)e F luey)e (n/cm (n/cm Basis (EFPS)
(n/cm -sec)
(n/cm )
Factor 5
54 11 18 17 18 F e' 4( n.p )Mn 3.36 x 10' 1.12 x 10 3.76 x 10 4.76 7.90 x 10 1.01 x 10 18 f.verage of all 3.3G x 10 1.23 x 10 4.13 x 10 4.76 8.68 x to 1.01 x 10 dostmeters m
l 84HSB:1b-031985 1
5
-r
I l
TABLE 6-11
. CALCULATED CURRENT AND E0L(a). VESSEL EXPOSURE FOR McGUIRE UNIT 1 Calculated Calculated Current # (E > 1.0 Mev)
EOL 4 (E > 1.0 Mev) 2 2
location (n/cm )
(n/cm )
18 19 Vessel IR~
1.01 x 10 3.02 x 10 17 19 Vessel 1/4T 5.59 x 10 1.68 x 10 17 18 Vessel 3/4T 1.13 x 10 3.40 x 10 a.
EOL fluences_are based on operation ~at 3565 MWt for 32 EFPY 8485B:1b-031985 6-25 1
SECTION 7 SURVEILLANCE CAPSULE R'EMOVAL SCHEDULE The following removal schedule is recommended for future capsules to be removed from the McGuire Unit I reactor vessel.
Lead Removal Estimated Fluence Caosule Factor Time [a]
(n/cm x 10 1 2
19 U
4.76 Removed (1.06) 0.414 (Actual)
X 4.76 4
1.80[b]
V 4.06 8
3.07[c]
Y 4.06 15 5.75 W
4.76 Standby Z
4.76 Standby a.
EFPY from plant startup b.
Approximates vessel end of life 1/4 thickness wall location fluence c.
Approximates vessel end of life inner wall location fluence 8485B:1b-031985 7-1
SECTION 8 REFERENCES 1.
Davidson, J. A., and Yanichko, S.E., " Duke Power Company William B. McGuire Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-9195, November, 1977.
2.
ASTM Standard E185-73, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" in ASTM Standards, Part 10 (1973), American Society for Testing and Materials, Philadelphia, 1973.
3.
Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, April 1977.
4.
Soltesz, R. G., Disney, R. K., Jedruch, J., and Zeigler, S.L.,
" 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.
5.
SAILOR RSIC Data Library Collection DLC-76, " Coupled, Self-shielded, 47 Neutron, 20 Gamma-ray, P3, Cross Section Library for Light Water Reactors."
6.
Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology
-- to be published.
7.
ASTM Designation E261-77, Standard Practice for Measuring Neutron Flux, Fluence, and Spectra by Radioactivation' Techniques," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 915-926, American Society for Testing and Materials, Philadelphia, li
8485B:1b-021885 8-1
8.
ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 927-935, American Society for Testing and Materials, Philadelphia, 1981.
9 '. ASTM Designation E263-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation of Ircn," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 936-941, American Society for Testing and Materials, Philadelphia, 1981.
- 10. ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 1063-1070, American Society for Testing and Materials, Philadelphia, 1981.
- 11. ASTM Designation E264-77, " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 942-945, American Society for Testing and Materials, Philadelphia, 1981.
8485B:1b-021885 8-2 A
APPENDIX A HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL GPERATION A-1.
INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced /.RT RT is designated as the higher of either NDT.
NDT the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35 mi!
lateral expansion (normal to the major working direction) minus 60*F.
RT increases as the material.is exposed to fast-neutron radiation.
NDT Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ART due to the radiation exposure associated with that NDT time period must be added to the original unirradiated RT The extent of NDT.
the shift in RT is enhanced by certain chemical elements (such as copper NDT and phosphorus) present in reactor vessel steels. Design curves which show the effect of fluence and copper and phosphorus contents on ART IT NDT reactor vessel steels are shown in figure A-1.
Given the copper and phosphorus contents of the most limiting material, the radiation-induced ARTNDT can be estimated from figure A-1.
The most limiting material occurs in the lower shell longitudinal weld seam no. 3-442A listed in table A-1, and it is at an azimuthal angle of 60'.
Fast neutron fluence (E > 1 Mev) for an azimuthal angle of 60' at the vessel inner surface, the 1/4 T (wall. thickness), and 3/4 T (wall thickness) vessel locations are given as a function of full power service life in figure A-2.
The data for all other ferritic materials in the reactor coolant pressure boundary are examined to ensure that no other component will be limiting with respect to RTNDT*
8485B:1b-021885 A-1
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TABLE A-1 MCGUIRE UNIT 1 REACTOR VESSEL TOUGHNESS TABLF Materlai Specification Code Cu P
NI T
RT USE NDT NOT Component Number Number
(%)
(%)
(%)
(*F)
(*F)
(ft-lb)
Closure head dome A533BCL.1 05086-1 0.11 0.010 0.48 20 37[c]
65[c]
Closure head segments A533BCL.1 B5087 0.11 0.008 0.62 10 10[c]
89(c]
ICI 40(c]
101 ECI Closure head flange A508CL.2 85002 0.010 0.75 40 Vessel flange A508CL.2 64701 O.010 0.73 29[c]
29(c]
101 ICI lc]
60[c]
89(c]
Inlet nozzle A506CL.2 05003-1 0.12 0.010 0.68 60 lc]
60(c]
88[c]
Inlet nozzle A508CL.2 B5003-2 0.10 0.012 0.71 60 Inlet nozzle A508CL.2 65003-3 0.10 0.009 0.69 60[c]
60[c]
79[c]
Iniet nozzle A506CL 2 F5003-4 0.10 0.010 0.09 60[c]
60[c]
77(c)
Outlet nozzle A508CL.2 B5004-1 0.005 0.74 60(c]
60[c]
82 ICI Outlet nozzle ASOGCL.2 D5004-2 O.007 0.74 60[c]
60!c]
75[c]
Outtet nozzle A508CL.2 B5004-3 O.005 O.71 60[c]
60ICI 90[c]
ICI 60[c]
81[c]
Outlet nozzle A508CL.?
05004-4 O.006 0.79 60
.p Upper shell A533BCL.1 B'i453-2 O.14 0.011 0.58 to IS[cl 73(c]
O Upper shell A533BCL.I B5011-2 0.10 0.011 0.54 10 27[c]
68[c]
Upper shell A533BCL.1 B5011-3 0.13 0.010 0.56 0
O(c) 95[c]
Intermediate shell A533BCL.1 85012-1 0.13 0.010 0.60
-30 34 101 Intermentate shell A533BCL.1 05012-2 0.13 0.011 0.62 O
O 104.5 Intermediate shell A533DCL.1 B5012-3 O.10 0.013 0.66
-20
-13 109 Lower shell A533BCL.1 05013-1 0.14 0.009 0.56
-10 0
94 Lower shell A533BCL.1 P~~)13-2 0.10 0.010 0.52
-10 30 115 Lower shell A533BCL.1 s:s013-3 0.10 0.010 0.55 O
15 104 ICl 90[c]
Bottom head segment A533BCL.1 05458-1 0.14 0.011 0.60
-70
-26 Dottom head segment A533BCL.1 05458-2 0.15 0.014 0.54
-30
-15[c]
96(c)
Bottom head segment A533BCL.1 05458-3 0.13 0.012 0.56
-20 2[c]
82[c]
Bottom head dome A5330CL.1 05085-1 0.13 0.010 0.53 0
10(c]
79[c]
I8I O.21 0.011 0.88
-60
-50
>110 Intermediate shell longitudinal weld seams M1.22 Intermediate shell to lower shell weld G1.39 0.05 0.006
-70
-70
>126 Lower shell tongitudinal weld seams Mt.32 0.20 0.015 O[c]
O[c]
90 tower shell longitudinal weld seams M1.33(b]
O.21 0.016 0.68 O
O lc]
ICI Lower shell longitudinal weld seams M1.34 O.30 0.013 0.64 O(c]
O[c]
IDI a.
Used'in reactor vessel survet11ance weldment b.
Used in weld root region only Estimated per U.S. NRC Standard Review Plan!II C.
84850:1h-031985
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- A-2.
FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the McGuire Unit I reactor vessel materials are presented in table A-1.
The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan (1)
The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the McGuire Unit 1 Vessel Material Surveillanca Program.
A-3.
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for t,he combined thermal and pressure stresses at any time during heatup g
or cooldown cannot be greater than the reference stress intensity factor, KIR, f r the metal temperature at that time. K is btained from the IR reference' fracture toughness curve, defined in Appendix G to the ASME Code.[2] The Kgp curve is given by the foll? wing equation.
Kyp = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)]
(A-1) where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code (2) as follows.
CKIM ' Elt
- KIR (A-2) 8185B:1b-022285 A-5 l.
where KIM = stress intensity factor caused by membrane (pressure) stress kit = stress intensity factor caused by the thermal gradients f the material KIR = function of temperature relative to the RTNDT C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test
'nditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K is determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, kit, f r the reference flaw are computed. From equation A-2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the re.ference flaw of Appendix G to the ASME Code is assumed to exist'at the inside of the vessel wall. During cooldown, the. controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are
. generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.
8485B:1b-021885 A-6
_-_--------------------.------.-------.---I
x s
=$ l 7
During cooldown, the 1/4 T vessel location is at a higher temperature'than the 4;_
l fluid adjacent to the vessel ID. This condition, of course, is not trus for j
the steady-state situation.
It follows that, at'any given reactor coolant h
temperature, the AT> developed during cooldown results in a higher value of j
KIR at the 1/4 T location for finite cooldown rates 'than for steady-state operation. Furthermore, if conditions exist so that the increase in K IR An exceeds kit, the calculated allowable pressure during cooldown will be y
greater than the steacy-state value.
g A
The above procedures are needed because there is no direct control on
(
temperafure at the 1/4 T location and, therefore, allowable pressures may z
unknowingly be violated if the rate of cooling is decreased at var".ous 5
intervals along a cocidown ramp. The use of the composite curve eliminates I
this problem and ensures conservative operation of the system for the entire cooldown period Three separate calculations are 7equired to determine the limit curves for ze finite heatup rates. As is done in the cooldown analysis, allowable g
pressure-temperature relationships are developed for steady-state conditions g
as well as finite heatup rate conditions assuming the presence of a 1/4 T E
defect at the inside of the vessel wall.
The thermal gradients during heatup
- 7 produce compressive stresses at the inside of the wall that alleviate the C
tensile stresses produced by internal pressure. The metal temperature at the
-lE 8
crack tip lags the coolant temperature; therefore, the K f r the 1/4 T g
IR crack during heatup is lower than the K f r the l'/4 T crack during d
IR steady-state conditions at the same coolant temperature. During heatup,1 especially at the end of the transient, conditions may exist so that the 4
effects of compressive thermal skresses and lower K I R
.t do not offset each w
other, and;the pressure-temperature curve based on steady-state conditions no g
longer represents a lower bound of all similar curves for finite ndetup rates when the 1/4 T flaw is considered. Therefore, both caset have to be analyzed
]
in order to ensure that at any coolant temperature, the lower value of the
]
alicwable pressure calculated for steady-ctate and finite heatup rates is J
ebtained.
i a
n g
B 8485B:1b-021885 A-7 J
s
=
~
s 55
- 2 n-The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present.
These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Since the thermal stresses at the outside are tensile and increase with g
increasing heatup rates, each heatup rate trust be analyzed on an individual 7
basis.
following the generation of pressure-temperature curves for both the steady-state andlinite heatup rate situations, the final limit curves are
.::e
[
produced by const'ructing a composite curve based an a point-by point
}{~ig comparison of the steedy-state and finite heatup rate data. At any given T.lH "1 i;=.
temperature, thE allo vable pressure is taken to be the lesser of the three 4( ^. c values taken from the curves under consiaeration.
The use of the composite b ?Q
[
curve it necessaff,to set conservative naatup limitations because it is
((['
E possible for conditions to exist wherein, over the course of the heatup ramp, lg~.
9,.>
'the controlling condition switches from the inside to the outside and the yK pressure limit must at all times be based on analysis of the most critical
[.$[ y b
criterion. Then, composite curves for thd 'teitup rate data and the cooldown
~
d.
-,v rate data are'adjuued fos possible< errors in the pressure an' temperature eeQ sensing instruments by the values indicated en figures A-3 and A-4.
I${
gf(
Finally, the new 10CFR50(3)' rule which addresses the metal temperature of 4"M".y' the closure head flange and vessel flange regions is considered. This 10CFR50
't i :
s %% ;
y rule states that the metal temperature of the closure flange regions must
(.4 y f
exceed the material RTND,. by at least 120 F for normal operation when the
'+.. _
3..
pressure exceeds 20 percent of the preservice hydrostatic test cressure (621 4.g
~
j psig for McGuire Unit 1). Table A-1 indicates that the limitir.g RT f
nab NDT i
40'F occurs in the closure head flange of McGuire Unit 1, and 'che minimum
[.; C M
allowable temperature of this re.gion is 160"F at pressures g eater' than 621 NM l
Psig.
$y..,
6 3
I 7.
84858:lb-021885
'A-8
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12216-20 3000.0 t
l l
1 1
I
}
}
LEAK TEST---*/
f f
LIMIT J
f f
}
(
l 2000.0 j
j
^m a
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9-
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1000.0
[
H EATUP--(
CURVE K
f i
7 r
p 2
T l
CRITICALITY LIMIT BASED ON INSERVICE HYDROSTATIC 0
TEST TEMPERATURE (323 F)
FOR THE SERVICE PERIOD UP TO 10 EFPY 0.0 O.0 100.0 200.0 300.0 400.0 500.0 INDICATED TEMPERATURE (OF)
MATERf AL PROPERTY BASIS CONTROLLING MATERIAL WELD METAL COPPER CONTENT 0.30 wt%
PHOSPHORUS CONTENT 0.013 wt%
RTNDT INITIAL 0"F 0
RTNDTAFTER 10 EFPY 1/4T,178 F
. 3/4, 83*F CURVE APPLICABLE FOR HEATUP RATES UP TO 60 F/HR FOR THE SERVICE PERIOD UP TO 0
10 EFPY AND CONTAINS MARGINS OF 10*F AND60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.
Figure A-3. McGuire Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 10 EFPY A-9 i
l
12216 21 3000.0 b
I J
l 2000.0
.?
a ua 5
f B
1 e
/
o 6
]
O
)
z
~
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1000.0 r
i i
)
\\
f COOLDOWN RATES
- F/H r
,d33//
WJ se::My 40 80 7i -,,
r
____ j oo 0.0 0.0 100.0 200.0 300.0 400.0 500.0 INDICATED TEMPERATURE ( OF )
MATERI AL PROPERTY BASIS CONTROLLING MATERI AL : WELD METAL COPPER CONTENT 0.30 wt%
PHOSPHORUS CONTENT 0.013 wt%
RTNDTINITIAL O'F 0
RTNDTAFTER 10 EFPY 1/4,178 F 3/4,830F 0
CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100 F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.
Figure A 4. McGuire Unit 1 Reactor Coolant System Cooldown Limitations Applicable for the First 10 EFPY A-10
i A-4.
HEATUP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in section A-3.
The derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan.E43 Transition temperature shifts occurring in the pressure vessel materials due
.to radiation exposure have been obtained directly from the reacter pressure vessel surveillance program. Charpy test specimens from Capsule U indicate that the surveillance weld metal and limiting core region intermediate shell plate heat no. 85012-1 exhibited shifts in RT f 160*F and 45'F, NDT 18 2
respectively. These shifts at a fluence of 4.14x10 n/cm are well within the appropriate design curve (figure A-1) prediction. As a result, the heatup and cooldown curves are based'on the ARTNDT given in figure A-1 for the most limiting beltline material which is in the lower shell longitudinal j
weld seam no. 3-442A. The resultant heatup and cooldown limit curves for normal operation of the reactor vessel are presented in figures A-3 and A-4 and represent an operational time period of 10 EFPY. These limit curves are not impacted by the new 10CRF50 rule.
i Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of tho criticality limit line shown in figure A-3.
This is in addition to other criteria which must be met befcre the reactor is made critical.
The leak test limit curve shown in figure A-3 represents miniwm temperature requirements at the leak test pressure specified by applicable codes. The-laak test limit curve was determined by methods of references 2 and 4.
Figures A-3 and A-4 define limits for ensuring prevention of nonductile failure.
8485B:1b-022285 A-11
APPENDIX A REFERENCES 1.
" Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
2.
ASME Boiler and Pressure Vessel Code,Section III, Division 1 -
Appendices, " Rules for Constru: tion of Nuclear Vessels," Appendix G,
" Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York, 1983.
l 3.
Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,
Amended May 17, 1983 (48 Federal Register 24010).
4.
" Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR l
Editicn, NUREG-0800, 1981.
~
- W I
8485B:1b-021385-A-12
_ _ _..