ML20079G536
| ML20079G536 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/30/1982 |
| From: | Byrne S, Jacques R, Koziol J ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20079G528 | List: |
| References | |
| TAC-53512, TR-N-MCM-008, TR-N-MCM-8, NUDOCS 8401200168 | |
| Download: ML20079G536 (117) | |
Text
_
TR-N MCM 008 NORTHEAST UTILITIES SERVICE COMPANY Millstone Nuclear Unit no.2 evaluation irradiated capsule w-s7 REACTOR VESSEL MATERIALS IRRADI AT!ON SURVEILLANCE PRO'GR AM APRIL 1982 H POWER SYSTEMS COMBUSTION ENGINEERING. INC.
NORTHEAST UTILITIES SERVICE COMPANY MILLSTONE NUCLEAR UNIT NO. 2 POST-IRRADIATION EVALUATION OF REACTOR VESSEL SURVEILLANCE CAPSULE W-97 April 1982 0-Prepared by:
0*
Date: #-/- f2.
S. T. Byrne, Co nig/nt Engineer Approved by:
E Date:
' /d f L J. J.
ziol, pran[Janager
'/i" E Approved by:
/
4 7 1-Date:
^-
R. C. Jagues, Millstone Unit #2 Project flanager Combustion Engineering, Inc.
Nuclear Power Systems Windsor, Connecticut I
l
f TABLE OF CONTENTS Section Title Page No.
I Summary 1
II Introduction 2
III Surveillance Program Description 3
IV Capsule Withdrawal and Disassembly 16 V
Test Results 18 VI Data Analysis 70 VII References 80 Appendix A Tensile Tests - Description and Equipment A-1 Appendix B Charpy Impact Tests - Description and Equipment B-1 Appendix C Instrumented Charpy V-Notch Data Analysis C-1 s
ii
1 List of Tables Table No.
Title Page No.
III-l Reactor Vessel Beltline Plates 5
III-2 Reactor Vessel Beltline Welds 6
III-3 Reactor Vessel Beltline Plates Chemical Analysis 7
III-4 Surveillance Plate and Weld Metal Chemical 8
Analysis III-5 Millstone Unit 2 Reactor Vessel Surveillance 14 Capsule Removal Schedule Y'
III-6 Type and Quantity of Specimens in W-97 Capsule 15 IV-1 Mechanical Test Specimens Removed from W-97 Capsule 17 f(3 hk V-1.
Composition and Melting Points of Temperature 19 Monitor Materials l.y.4 m m.
V-2 Neutron Flux Monitors 23
.. hij V-3 Millstone Unit 2 Flux Spectrum Monitor Activities, 26 is t
Compartment 6214.
V-4 Millstone Unit 2 Flux Spectrum Monitor Activities, 27 Compartment 6241.
V-5 Millstone Unit 2 Flux Spectrum Monitor Activities, 28 Compartment 6273.
\\
V-6 Flux Monitor Activities 30 V-7 Millstone Unit 2 Fast Neutron Flux and Fluence 37 Values V-8 Irradiated Plate and Weld Chemical Analysis 40 V-9 Post-Irradiation Tension Test Properties 42 V-10 Pre-Irradiation Tension Test Properties 43 V-ll Charpy Impact Results, Base Metal (WR) 50 V-12 Charpy Impact Results, Base Metal (RW) 51 V-13 Charpy Impact Results, Weld Metal 52 V-14 Charpy Impact Results, HAZ 53 iii
List of Tables (cont'd)
Table No.
Title Page No.
VI-1 Summary of Toughness Property Changes 77 VI-2 Projected NDTT Shift and Adjusted RTNDT 78 for Controlling Material VI-3 Proposed New Capsule Removal Schedule 79 C-1 Instrumented Charpy Test, Base Metal (WR)
C-3 C-2 Instrumented Charpy Test, Base Metal (RW)
C-4 C-3 Instrumented Charpy Test, Weld Metal C-5 C-4 Instrumented Charry Test, HAZ C-6 C-5 Toughness Property Changes Based on Instrumented C-7 Charpy Impact Test iv
List of Figures-Figure No.
Title Page No.
! urveillance Capsule Assembly 10 III-1 S
III-2 Charpy Impact Compartment Assembly 11
-III-3
. Tensile-Monitor Compartment Assembly 12 LIII Location of Surveillance Capsule Assemblies 13
~V-1
- Temperature Monitors, Compartment 6214 20
-V-2.
Temperature Monitors, Compartment 6241 20
.V-3
. Temperature Monitors, Compartment 6273.
21 V-4' 558F and 590F Monitors, Compartment 6214 21
-. V-5 Geometry Used in 00T 34 V-6
~ Dimensions-and Power Distributions Used in D0T 35 V Surveillance Capsule Location 36
.V-8 Stress-Strain Record, Base Metal. 72F 44
~
V-9
. Stress-Strain Record, Base Metal, 250F 44 V-10' Stress-Strain Record,. Base Metal, 550F 45 V-11' Stress-Strain Record, Weld Metal, 72F 45 V-12 Stress-Strain Record, Weld Metal, 250F 46 V-13
~ Stress-Strain Record, Weld Metal, 550F 46 V-14 Stress-Strain Record, HAZ Metal, 72F 47 V-15.
Stress-Strain Record, HAZ Metal, 250F.
47 V-16 Stress-Strain Record, HAZ Metal, 550F 48
~
.V-17.
' Fracture Surface of Irradiated Tension 49 Specimens.
f
v
T List of Figures (Cont'd)
Figure No.
Title Page No.
V-20 Charpy Shear Fracture, Base Metal (WR) 56 V-21 Charpy Impact Energy, Base Metal (RW) 57 V-22 Charpy lateral Expansion, Base Metal (RW) 58 V-23 Charpy Shear Fracture, Base Metal (RW) 59 V-24 Charpy Impact Energy, Weld fietal 60 V-25 Charpy lateral Expansion, Weld Metal 61 V-26 Charpy Shear Fracture, Weld Metal 62 V-27 Charpy Impact Energy, HAZ 63 V-28 Charpy lateral Expansion, HAZ 64 V-29 Charpy Shear Fracture, HAZ 65 V-30 Fracture Surfaces, Impact Specimens, 66 Base Metal (WR)
V-31 Fracture Surface, Impact Specimens, Base 67 Metal (RW)
V-32 Fracture Surfaces, Impact Specimens, 68 Weld Metal V-33 Fracture Surfaces, Impact Specimens, HAZ 69 VI-1 predicted NDTT Shift for the Millstone Unit 2 74 Reactor Vessel Beltline Vi
List of Figures (Cont'd.)
Figure No.
Title Page No.
A-1 Tensile Test System A-2 A-2 Typical Tensile Specimen A-3 A-3 Location of Tensile Specimens in Base A-4 Metal A-4 Location of Tensile Specimens in Weld A-5 Metal A-5 Location of Tensile Specimens in HAZ A-6 B-1 Charpy Impact Test System B-4 B-2 Typical Charpy V-Notch Impact Specimen B-5 B-3 Location of Charpy Specimens in Base Metal B-6 B-4 Location of Charpy Specimens in Weld Metal B-7 B-5 Location of Charpy Specimens in HAZ B-8 C-1 ICV Load vs. Temperature Diagram, Base Metal (WR)
C-8 C-2 ICV Load vs. Temperature Diagram, Base Metal (RW)
C-9 C-3 ICV Load vs. Temperature Diagram, W'ld Metal C-10 e
C-4 ICV Load vs. Temperature Diagram, HAZ C-ll i
vii
I.
SUMMARY
The first surveillance wall capsule (W-97) was removed from the Millstone Unit 2 reactor vessel in August 1980 after 3.0 effective
. full power years of reactor operation.
The surveillance test specimens and monitors were evaluated at C-E's Windsor, Connect-icut laboratory facility.
Post-irradiation evaluation of the temperature monitors indicated that the irradiation temperature was between 536 F and 558 F.
Analysis of the neutron threshold detectors provided a capsule fluence of 3.78 x 10 n/cm2 (E>l MeV), which corresponded to a 8
maximum fluence at the inside surface of the reactor vessel of 18 2
2.78 x 10 n/cm,
Radiation induced changes in the tension and impact properties were determined for the base metal, weld metal and heat-affected zone surveillance materials. Transition temperature shifts ranged from 76 F for the weld metal to 96 F for the base metal pl a te. The upper shelf impact energy after irradiation was in excess of 75 ft-lb for each of the surveillance materials, ranging from 79 ft-lb for the base metal to 98 ft-lb for the weld metal.
The base metal exhibited the greatest toughness property change even though the residual copper content of the weld (0.30 w/o) was higher than for the plate (0.14 w/o). This difference in copper content was confirmed by chemical analysis c-f the irradiated base and weld metal Charpy specimens. The analysis also confirmed that the chemistry of the irradiated materials was consistent with the chemistry originally reported for the. surveillance materials.
The base metal, weld metal and HAZ exhibited similar changes in tensile properties after irradiation. The yield strength and ultimate strength increased approximately 11%, while total elongation and reduction in area decreased 3% to 5% after irradiation.
1_
The NDTT shift prediction method from the Millstone Unit 2 Technical Specifications was found to be conservative by a factor of 25% for the base metal and 58% for the weld metal based on the W-97 surveillance capsule measurements.
In contrast, shifts predicted using Regulatory Guide 1.99 were 140% higher than measured for the weld metal and 35% less than measured for the base metal. The greater radiation resistance of the weld is attributed to the low nickel content (.06%). Based upon experimental data, the weld metal shift will continue to be less than that of the plate, despite the difference in copper content. Therefore, a more accurate shift prediction method was developed based on the measured shift for the controlling base metal and the slope of the Regulatory Guide expression. The predicted end-of-life (32EFPY) adjusted RTNDT at the quarter thickness location in the reactor vessel is 199 F using the revised method for the controlling plate. The predicted end-of-life adjusted RTNDT for the weld is 98 F at 1/4t using a similar projection method.
The predicted decrease in upper shelf energy at end-of-life based on the method given in Regulatory Guide 1.99 is 38% at the one-quarter thickness location in the vessel.
Using this conservative prediction, the upper shelf energy of the plates will remain above 65 ft-lb during the design life of the vessel, and the weld shelf energy will remain above 80 ft-lb. These projected values are well in excess of the 50 ft-lb value currently considered to be a reasonable lower limit for continued safe operation.
Recomended changes to the surveillance capsule withdrawal schedule were provided to meet the requirements of 10CFR50, Appendix H.
If the proposed schedule is implemented, the next capsule will be withdrawn after 7 effective full power years of operation. Additional surveillance data would then be available before the current operating limits would require updating.
II.
Introduction The purpose of the Millstone Unit 2 surveillance program is to monitor the radiation induced changes in the mechanical properties _.
s '
g ti of ferritic materials in the reactor vessel beltline during the 7
ef operating lifetime of the reactor vessel. The surveillance program includes the determination of the preirradiation (baseline)
]
strength and toughnes properties and periodic determinations of g
the property changes following neutron irradiation. These property changes are used to verify and update the operating limits (heat-(
=
up and cool down pressure / temperature limit curves) for the
['
primary system.
The Millstone Unit 2 Surveillance program (I) is based upon ASTM E185-70, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels." The pre-irradiation (baseline) evaluation
~[
results from the Millstone Unit 2 reactor vessel surveillance
=
materials are described in C-E report 18767-TR-MCD-009.(2) The F
following report describes the results obtained from evaluation s
of irradiated materials from capsule W-97 which was removed from i
the reactor in August 1980.
Ik III Surveillance Program Description f
The Millstone Unit 2 reactor pressure vessel was designed and fabricated by Combustion Engineering, Inc. The reactor vessel
[
beltline, as defined by 10CFR50, Appendix H, consists of the six 7_[_
plates used to form the lower and intermediate shell courses in f
V the vessel, the included longitudinal seam welds and the lower to 3
intermediate sheil girth seam weld.
The plates were manufactured i
Gom SA533 Grade B Class 1 quenched and tempered plate.
The heat f_{
treatment consisted of austenization at 1600 1 50F for four p
hours, water quenching and tempering at 1225 1 25F for four f
hours. The ASME Code qualification test plates were stress 3
relieved at 1150 1 25F for forty hours, and furnace cooled to 600F. The longitudinal and girth seam welds were fabricated
.g-using E8018-C3 manual arc electrodes and Mil B-4 submerged arc weld wire with Linde 124 and Linde 0091 flux.
The post weld heat treatment consisted of a forty hour 1150 1 25F stress relief heat l _-
treatmentxfollowed by furnace cooling to 600F.
The beltline
,p
~~
k k b b
+
materials are identified in Tables III-l and III-2.
The chemical analysis of the six beltline plates is given in Table III-3. The materials included in the surveillance program were obtained from the actual t-eactor vessel beltline materials. The base metal surveillance material, a section from plate C-506-1, was selected from the six beltline plates on the basis of the highest initial drop weight NOTT. The heat treatment of the surveillance plate duplicated that of the reactor vessel ASME Code qualification test plates. The surveillance weld material was fabricated by welding together sections of plates C-506-2 and C-506-3 using the same weld procedure and wire-flux combination used for the inter-mediate to lower shell girth seam weld. Mil B-4 submerged arc filler wire and Linde 0091 flux was used. The post-weld heat treatment consisted of a forty hour stress relief at 1125 + 25F followed by furnace cooling to 600F. The surveillance heat-affected zone material was fabricated by welding together sections of plate.c C-506-1 and C-506-3 in the same manner as the surveillance weld material with the same postweld heat treatment. The chemical analysis of the surveillance plate and weld (2) is given in Table III-4.
TABLE'III-l REACTOR VESSEL BELTLINE PLATES
' Location Piece Number Code Number Heat Number Supplier Intermediate 215-02A C-505-1 C-5843-1 Lukens
- Shell Intermediate-215-02B C-505-2 C-5843-2 Lukens
.Shell Intermediate 215-02C C-505-3 C-5843-3 Lukens Shel1
. Lower Shell 215-03A C-506-1 C-5667-1 Lukens Lower-Shel1 215-03C C-506-2 C-5667-2 Lukens Lower Shell' 215-03B C-506-3 A-5518-1 Lukens o I' n
TABLE III-2 REACTOR VESSEL BELTLINE WELDS Location Weld Seam No.
Wire Heat No.
Flux Type Flux Batch Intermediate 2-203A A-8746 Linde 124 3878 Shell Longi-tudinal Seam Intermediate 2-203B
.A-87a6 Linde 124 3878 Shell Longi-tudinal Seam Intermediate 2-203C A-8746 Linde 124 3878 Shell Longi-M/A L0BI*
tudinal Seam Lower Shell 3-203A A-8746 Linde 124 3878 Longitudinal M/A BOIA
- Seam M/A BOLA
- Lower Shell 3-203B A-8746 Linde 124 3878 Longitudinal M/A CAFJ
- Seam M/A DBIJ
- M/A EODJ
- Lower Shell 3-203C A-874G Linde 124 3878 Longitudinal M/A BOIA
- Seam M/A BOLA
- Intermediate 9-203 10137 Linde 0091 3999 to Lower Girth M/A BOLA
- Seam 90136 Linde 0091 3998 M/A LAGJ
- M/A JACJ *
~
Manual shielded metal arc electrode (all others automatic submerged arc wire).
(
l TABLE III-3 REACTOR VESSEL BELTLINE PLATES CHEMICAL ANALYSIS Element C-505-1 C-505-2 C-505-3 C-506-1 C-506-2 C-506-3 Si
.18
.18
.18
.12
.20
.21 S
.015
.015
.018
.014
.018
.01 2 P
.006
.008
.007
.006
.007
.005 I
Mn 1.27 1.25 1.26 1.30 1.30 1.32 C
.26
.21
.21
.21
.21
.27 Cr
.10
.10
.10
.10
.10
.19 Ni
.64
.64
.65
.61
.61
.70 Mo
.60
.60
.62
.63
.63
.62 V
.003
.003
.004
.005
.005
.004 Cb
<.01
<.01
<.01
<.01
<.01
<.01 B
.0004
.0005
.006
.004
.0006
.0003 Co
.009
.009
.010
.011
.011
.012 Cu
.13
.13
.13
.14
.14
.13 Al
.024
.030
.029
.020
.028
.020 W
<.01
<.01
<.01
<.01
<.01
<.01 Ti
<.01
<.01
<.01
<.01
<.01
<.01 As
.012
.014
.014
.011
.01 2
.012 Sn
.006
.006
.006
.009
.010
.007 Zr
.002
.002
.002
.002
.002
.002 N
.009
.008
.008
.009
.008
.009 2 -.
l TABLE III-4 SURVEILLANCE PLATE AND WELD METAL CHEMICAL ANALYSIS Weight Percent Plate 1/4 T - ID Weld 1/4 T - OD Weld Element C-506-1(a)
C-506-2/C-506-3(b)
C-506-2/C-506-3(c)
Si
.12
.17
.15 S
.014
.013
.013 P
.006
.015
.01 6 Mn 1.26 1.13 1.13 C
.21
.12
.12 Cr
.10
.04
.05 Ni
.61
.06
.06 Mo
.62
.54
.53 V
.004
.006
.007 Cb
<.01
<.01
<.01 B
.0006
.0003
.0003 Co
.011'
.009
.009 Cu
.14
.30
.21 Al
.020
<.001
<.001 W
<.01
.01
<.01 Ti
<.01
<.01
<.01 As
.011
.011
.012 Sn
.009
.004
.003 Zr
.002
.002
.002 N
.009
.008
.009 2
a Heat C-5667-1 b
Mil B-4 wire heat 90136, Linde 0091 flux lot 3998 c
Mil B-4 wire heat 10137, Linde 0091 flux lot 3999..
Drop weight, Charpy impact and tension test specimens were machined from the surveillance materials as described in ref-erence 1.
In addition to the surveillance material specimens, Charpy impact specimens were machined from a section of plate 01 from the Heavy Section Steel Technology (HSST) program to serve as standard reference material (SRM).
The surveillance and SRM test specimens were enclosed in six capsules for irradiation in the Millstone Unit 2 reactor vessel.
The surveillance capsule assembly is shown in Figure III-1.
Each assembly consists of four compartments containing Charpy impact specimens (Figure III-2) and three compartments (Figure III-3) containing tension specimens and mcnitors (flux and temperature).
Each capsule is positioned in a holder tube attached to the reactor vessel cladding to irradiate the specimens in an environ-ment which duplicates as closely as possible that experienced by the reactor vessel. Capsule locations are shown in Figure III-4.
The axial portion of each capsule is bisected by the midplane of the core. The circumferential locations were selected to coincide with the peak flux regions of the reactor vessel.
The existing Technical Specification withdrawal schedule for the surveillance capsules is given in Table III-5.
The type and quantity of test specimens contained in the W-97 capsule are given Table III-6. __
Lock Assembly Nif b4 Wedge Coupling Assembly
/ (:
Tensile -Monitor-Compartment 2
- y:.
Charpy impact Compartments Tensile -Monitor Compartment %
N N, s
/sy N(
N T
s i Charpy lmpact Compartments 9
'Nt s si Tensile-Monitor Compartment g
I NORTHEAST UTILITIES pj urG g
Nucleo er $tation Typical Surveillance Capsule Assembly Unit No. 2 III_1
-Wedge Coupling - End Cap l
Charpy impact Specimens
\\
Spacer d\\s a
-s
%/
/\\
- sse,
" Rectangular Tubing i
O
-Wedge Coupling - End Cap 1
N NORTHEAST UTILITIES p
Nucleo e[9ation 0
bb Y
Unit No. 2 111-2
-1 1 -
f Wedge Coupling - End Cap
/
Flux S0ectrum Monitor Ca mium Shielded Flux Monitor Housing h
iN Stainless Steel Tubing Stainless Steel Tubing
/
Cadmium Shield Threshold Detector L
Threshold Detector
\\
Flux Spectrum Monitor
'q-Quartz Tubing
(.
e h]LN[
Temperature Monitor-(jN Weight Temperature Monitor-Low Melting Alloy
\\'
Housing Tensile Specimen r
L Split Spacer Bk Tensile Specimen Housing
" Rectangular Tubing
\\9
-Wedge Coupling - End Cap NORTHEAST UTILITIES F.'au re Miiistone Typical Tensile-Monitor Compartment Assembly Nucleggog,r patin 111-3.
Z Z E. 9 150 c*
r ~ -- - - 'l Outlet Nozzle g, g
=
Tj E4
\\
f
,gc i
i t
wwc I
- a. c, Vessel
/-
<=
0 8
m 104 A
-s
's
\\ Inlet
/
I d I
\\ Nozzle E
(
Core Shroud
)
\\
[ Core Support Barrel
/
nm s
m s
i 4vThNd Vessel
' Vessel Sh
[
97 %
Reactor Vessel
' 263
! q 0
! bES!*'ie l
E' ""
a vessei vess/ei u
! !s Asseinbly E
83
' Thermal Shield
' 277U f
0
~
g_
2 o
- s (s
/
1 p
\\
Vessel
.i n
{
lj j
c[
L
_J
'y 284 Hi
\\
l l
/
l N
L-s _
,.,/
l
[
I
-Core E.
//
Reactor Su port
)
g Ba el 3
(
j Vessel
\\
/
'i
(
s~_-----
~ 2-go c
Enlarged Plan View Elevation
~
2, View a
m
TABLE III-5 EXISTING TECHNICAL SPECIFICATION SCHEDULE FOR MILLSTONE UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL Removal Azimuthal Approximate Target 2
Sequence Location Removal Time (Years)
Fluence (n/cm )
18 1
97 8
3.2 x 10 18 2
104 16 5.7 x 10 18 3
282.'
23 8.3 x 10 I9 4
263 30 1.2 x 10 I9 5
277 35 1.4 x 10 I9 6
83 40 1.7 x 10 l
1 _
-TABLE III-6 TYPE AND QUANTITY OF SPECIMENS IN W-97 CAPSULE
. Material Charpy Impact Tension
-Base Metal 12-(Transverse)
Base Metal 12 3
(longitudinal)
Weld Metal 12 3
Heat-Affected Zone 12 3
Total 48' 9
i.
IV.
CAPSULE WITHDRAWAL AND DISASSEMBLY The Millstone Unit 2 reactor vessel was shut down for refueling at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> on August 17, 1980. The W-97 surveillance capsule was subsequently inspected using an under-water video system.
The capsule was found to be securely locked ia position.
The lock assembly adapter (ACME threaded nose cone which serves as the point of attachment for the retrieval tool) was intact.
Following the video inspection, a retrieval tool was attached to the W-97 capsule adapter to disengage the latches, and the capsule was withdrawn from its holder and out of the reactor vessel. The W-97 capsule was transferred to the spent fuel pool where it was sectioned into lengths for insertion into a shipping cask. Sectioning was accomplished by drilling to separate the wedge assembly halves, leaving the specimen compa-tments intact.
The surveillance capsule was shipped to Neutron Products, Inc.
in Dickerson, Maryland, for inspection, disassembly and specimen removal in the hot cell facility.
No unusual features or damage were revealed by visual inspection. A remote control circular saw was used to open the capsule compartments.
Each compartment was identified and inspected prior to cutting, and the contents were removed and verified against the original loading records.
An inventory of the mechanical test specimens removed from the W-97 capsule is given in Table IV-1.
h TABLE IV-1 MECHANICAL TEST SPECIMENS REMOVED FROM W-97 CAPSULE Compartment Material and Number Specimen Type Specimen Identification 6214 HAZ Tension 4K7, 4J2, 4KA 6224 HAZ Charpy 435, 43U, 455 44D, 464, 432 447, 451, 420 45A, 44M, 45M 6232 Base Metal Charpy 23U, 25C, 252 (Transverse) 23M, 23K, 21 K 251, 25J, 21M 21E, 23T, 23A 6241-Base Metal Tension 1KC, lJU lJA 6251 Base Metal Charpy 147, 122, 14D (Longitudinal) 135, llB, llY 14P, 137, 16B 14M, 14U, 114 6263 Weld Metal Charpy 31L, 352, 31D 33J, 34D, 34A 31C, 360, 317 33B, 36M, 33C 6273 Weld Metal Tension 3JY, 3J2, 3JJ l
l L - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
V.
TEST RESULTS A.
Irradiation Environment 1.
Temperature Monitors Each Tensile-Monitor Compartment (Figures III-l and III-3) in the capsule assembly contained a set of four temperature monitors to provide an indication of the maximum temperature in the capsule during irradiation.
The composition and melting point of each eutectic alloy monitor is given in Table V-1.
Each monitor consisted of a helix of the eutectic alloy and a stainless steel weight encapsulated in a quartz tube. Each set of four temperature monitors was inserted into a stainless steel housing, and the temperature monitors were irradiated in the top, middle and bottom surveillance capsule compartments.
Post-irradiation examination of the temperature monitors was performed at C-E's Windsor, Connecticut facility.
Each temperature monitor was identified by length, photographed, and inspected to determine whether the eutectic alloy helix had melted and been crushed by the weight.
Photographs of the three sets of monitors at 5X are shown in Figures V-1 through V-3.
The 536 F monitors (80% Au - 20% Sn) were completely melted. The 558 F monitors (90% Pb - 5% Sn - 5%
Ag) helices were distorted but exhibited only localized melting. The 580 F and 590 F monitors were intact and exhibited no distortion. A 558 F monitor is compared with a 590 F monitor in Figure V-4 to illustrate the differences in behavior. Each set of monitors exhibited similar features, indicating that the maximum irradiation temperature was in the 550 F - 558 F range and uniform along the length of the surveillance capsule.
TABLE V-1
-COMPOSITION AND MELTING POINTS OF TEMPERATURE MONITOR MATERIALS Composition Melting Temperature (Weight %)
F
- 80 Au, 20 Sn 536
~90 Pb, 5 Sn 5 Agl 558 97.5 Pb, 2.5 Ag 580 97.5 Pb, 0.75 Sn, 1.75 Ag 590 l
FIGURE V-1 TEMPERATURE MONITORS COMPARTMENT 6214 (5x)
Y 7
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590F 580F 558F 536F
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FIGURE V-2 TEMPERATURE MONITORS P_
COMPARTMENT 6241 (5x) p
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-.3 1
- p F
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2.
Neutron Dosimetry Each Tensile-Monitor compartment (Figures III-l and III-3) in the capsule assembly contained two sets of neutron flux monitors as described in Table V-2.
Each flux monitor was encapsulated in a stainless steel sheath (except for the sulfur which had a quartz sheath); in addition, cadmium covers were placed around the uranium, cobalt, nickel and copper monitors which have competing thermal activities.
Each set of nine flux monitors was inserted into two stain-less steel housings, one set for each of the top, middle and bottom surveillance capsule compartments.
The flux monitors were removed from the capsule compartments in the hot cell.
Each monitor was inspected and its position in the housing verified by the number of grooves in the stainless steel sheath. The monitors were then repackaged and shipped to C-E's Windsor, Connecticut facility for radiochemical analysis.
a.
Radiochemical Analysis Radiochemical analysis of the flux monitors was performed in accordance with C-E Procedure 00000-FMD-401, Rev. O, November 1,1978 (" Standard Method for the Analysis of Radioisotopes in Reactor Irradiation Surveillance Detectors and Flux Distribution Monitors").
Each monitor was removed from its sheath and inserted in a glass vial.
Recovery of the uranium, titanium and cadmium shielded monitors was complicated by oxidation and contamination of the monitors.
In order to assure that the subsequent activity measurements from these monitors would be accurate, specific steps were taken to either remove or identify the source of the conta-mination.
_22-
=M
7 r
TABLE V-2 NEUTRON FLUX MONITORS Material Reaction Threshold Energy (Mev)_
Half-Life Cobalt CoS9(n.t)Co60 Thermal 5.3 Years
.(Cadmium Shielded)
. Uranium
- U238(n f) Cs137 0.7 30.i. Years Titanium T146(n.p)Sc46 8.0 84 days
' Iron Fe54(n.p)Mn54 4.0 314 days Cobal t CoS9(n,y)Co60 Thermal 5.3 Years Uranium
- U238(n,f) Cs137 0.7 30.2 years (Cauni rn Shielded)
Nickel NiS8(n.p) CoS8 5.0 71 days l-(Cadnium Shielded) l Copper-Cu63(n.a) Co60 7.0 5.3 years (Cadmium Shielded)
Sulfur S32(n p) P32 2.9 14.3' days
- U-238 foil depleted in U-235 to.05 -w/o 1
.e.
The uranium foil had converted to a black powder, assumed to be U 0.
Therefore, instead of using a simple grayimetric 38 measurement, the amount of uranium recovered was determined by atomic absorption spectroscopy.
The titanium wires were embrittled, but otherwise they presented no handling or counting problems. The cadmium shield on the cobait, copper and nickel wires had apparently melted and fused to the wire during irradiation. The cadmium shields were mechanically removed by stripping, scraping and filing.
Final monitor weights were based on elemental analysis using atomic absorption spectroscopy. The sulfur monitor from compartment 6241 contained hydrogen sulfide gas at a pressure high enough to cause a violent separation of the glass capsule during the opening process. As a result, the data were not used for analysis since it was not certain that the sulfur which remained in the capsule was a representative sample; the phosphorus-32 activity was about a decade lower than that found for the other two sulfur monitors.
The remainder of the samples for radio-chemical analysis were prepared using standard methods.
Gamma counting was performed with a 4096 channel gamma spectrometer system coupled with a hyperpure germanium detector.
The system was calibrated at 0.5 Kev per channel to span the gamma energy range from 0.05 tG 2 Mev.
Efficiency calibration was performed using eight (8) gamma energies h
emitted from an NBS traceable mixed isotope standard.
Phosphorus-32 beta activity was measured on a gas flow proportional counter which was calibrated with NBS traceable beta standards.
Physical constants used in the calculation of radioisotope activity levels are as follows:
I.
Isotope Half-Life Gama Energy (Mev)
Intensity Cobal t-58 70.8 days 0.8108 0.99 Cobalt-60 5.27 years 1.3325 1.00 Cesium-137 30.2 years 0.6616 0.85 Manganese-54 312.5 days 0.8348 1.00 Scandium-46 83.8 days 0.8890 1.00 Phosphorus-32 14.3 days 1.00 Flux spectrum monitor activity levels are presented in Tables V-3 through V-5.
All values are decay corrected to the time of reactor shutdown, 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />, August 17, 1980.
The uncertainty listed with each result is the 2-sigma counting error only. An additional error of 120% for uranium monitors and 15% for all other metal monitors is estimated from volumetric and gravimetric operations and from the certified uncertainties of calibration isotopes. The additional error associated with the sulfur monitor results is 18%.
The shutdown activities determined from gamma ray emission rates were calculated as follows:
A=
EWBC (exp-At) where: A shutdown activity in disintegrations per
=
minute per milligram of material (dpm/mg) radioisotope net counts per minute Np
=
full energy peak effiency (counts per gamma E
=
ray emitted) weight of monitor sample (milligrams)
W
=
radioisotope gamma ray branching ratio (gama B
=
raysperdisintegration) correction for coincident or random st.mning C
=
radioisotope decay constant A
=
elapsed time between plant shutdown and t
=
counting __-___-_ - _____
l TABLE V-3 MILLSTONE UNIT 2 FLUX SPECTRUM MONITOR ACTIVITIES, COMPARTMENT 6214 Monitor Number of Weight Measured Activity at End Material Grooves (mg)
Isotope Irradiation (Dpm/mg) 5 Cobalt 0
7.7 Co-60 1.76 + 0.01 x 10 (shielded)
~
4 Uranium 1
17.5 Cs-137 2.7010.02 x 10 Titanium 2
9.9 Sc-46 3.69 1 0.08 x 10 4
Iron 3
24.3 Mn-54 1.121 1 0.004 x 10 5
Cobalt 4
8.1 Co-60 1.331 1 0.006 x 10 6
Uranium 5
15.6 Cs-137 1.06 + 0.01 x 104 (shielded)
~
Nickel 6
21.4 Co-58 1.803 + 0.002 x 106 (shicided)
Copper 7
20.0 Co-60 6.7 !- 0.1 x 10 3
(shielded)
Sulfur 18.5 P-32 2.167 1 0.009 x 10 6
l _ ______
TABLE V :
MILLSTONE UNIT 2 FLUX SPECTRUM MONITOR
,1 ACTIVITIES, COMPARTMENT 6241
. Monitor Number of
' Weight Measured Activity at End of Material Grooves (mg)
Isotope Irradiation (D/M/mg) 5 Cobalt:
0
- 8. 3' Co-60 1.81 + 0.01 x 10 (shielded).
4
' Uranium.
.1 31.9 Cs-137 2.37 1 0.03 x 10 4
Titanium 2
13.4 Sc-46
'3.29 1 0.06 x 10 5
l Iron 3
26.5 Mn-54 1.055 1 0.004 x 10 6
- 4 8.4 Co-60 1.51410.007 x 10 4
-Uranium.
5 11.1 Cs-137
.1.15 + 0.02 x 10 (shielded) 6 Nickel.
6 22.5 Co-58 1.668 +.0.002 x 10 (shielded) 3 Copper 7
~25.0 Co-60 6'.610.1 x 10 (shielded) 5 10.7 P-32 2.84 1 0.04 x 10 Sulfur 4
+ -
- . x.
, =,... -.
.... - -.. -., -,, =.. _....
TABLE V-5 MILLSTONE UNIT 2 FLUX SPECTRUM MONITOR ACTIVITIES, COMPARTMENT 6273 Monitor Number of Weight Measured Activity at End of Material Grooves (mg)
Isotope Irradiation (D/M/mg) 5 Cobalt 0
7.9 Co-60 1.84 + 0.01 x 10 (shielded)
~
4 Uranium 1
4.9 Cs-137 2.23 1 0.04 x 10 4
Titanium 2
12.5 Sc-46 3.43 1 0.07 x 10 5
Iron 3
24.8 Mn-54 1.11610.004 x 10 6
Cobalt 4
8.3 Co-60 1.022 + 0.006 x 10 4
Uranium 5
5.6 Cs-137 1.12 + 0.03 x 10 (shielded) 6 Nickel 6
20.2 Co-58 1.828 + 0.002 x 10 (shicided)
~
3 Copper 7
23.2 Co-60 7.6 + 0.1 x 10 (shielded) 6 Sulfur 17.3 P-32 1.927 1 0.008 x 10
7_.
b.
Threshold Detector Analysis The SAND-II(3) and DOT-III I4) computer codes were used to calculate the fast flux and fluence at the surveillance capsule assembly location and at the reactor vessel.
The SAND-II computer code is used to calculate a neutron flux spectrum from the measured activities of the flux monitors. SAND-II requires an initial flux spectrum estimate; this is calculated using DOT-III.
The measured activities must be adjusted before they can be put into SAND. The various steps of the procedure are descrioed below.
The measured activities were decay corrected to reactor shutdown. The foils irradiated and the shutdcwn activities are shown in Table V-6.
Before being used by SAND, the foil activities must be converted to saturated activity with units of disintegrations per second per target atom (dps/a).
The following equation was used for the conversion:
M A 16.67 A
=
sat NIS where A
Saturated activity (dps/a)
=
sat M
Measured activity at shutdown (dpm/mg)
=
A Atomic weight
=
N Avoga ho's number
=-
I Isotopic abundance of target isotope
=
S Saturation factor, explained below
=
238 For U fission product activities, the required SAND input 238 has dimensions of fissions per second per U atom (fps /a).
TABLE V-6 FLUX MONITOR ACTIVITIES Monitor Material Measured Isotope 1
Cobult(shielded)
Co-60 2*
Titanium Sc-46 4
Iron Mr.-54 5
Uranium (shielded)
Cs-137 7
Nickel (shielded)
Co-58 8
Copper (shielded)
Co-60 9*
Sulfur P-32
- These monitors were not included in the SAND analysis.
I _
TABLE V-6 (cont'd)
FLUX MONITOR ACTIVITIES Shutdown Saturated Compartment (c)
Monitor Activity (dpm/mg)
Activity (dps/a)(b) 6214 1
1.76+5 I")
2(d) 9.475-16 2.70+4 4.344-14 3
3.69+4 9.154-16 4
1.121+5 4.638-15 5
1.331+6 7.165-15 6
1.06+4 1.706-14 7
1.803+6 6.348-15 8
6.7+3 5.629-17 9
2.167+6 2.166-15 6241 1
1.81+5 9.745-16 2(d) 2.37+4 3.813-14 3
3.29+4 8.162-16 4
1.055+5 4.365-15 5
- 1. 514+6 8.151-15 6
1.15+4 1.850-14 7
1.668+6 5.873-15 8
6.60+3 5.545-17 9
2.84+5 2.838-16 6273 1
1.84+5 9.906-16 2(d) 2.23+4 3.587-14 3
3.43+4 8.509-16 4
1.116+5 4.618-15 5
1.022+6 5.502-15 6
1.12+4 1.801-14 7
1.828+6 6.436-15 8
7.6+3 6.386-17 9
1.927+6 1.926-15 (a) Denotes power of 10 (b) Uranium Foils are (fps /a)
(c) Used compartment number 6273 to obtain the fast flux in the surveillance capsule given in Table V-7.
The relative axial variation in the unfolded fast flux was as follows:
6214/6273
= 0.94; 6241/6273 = 0.985 (d) U-238 fission yield of Cs-137 =.0626
This is obtained by dividing A by the fractional fission sat yield of the fission product whose activity was measured.
The saturation factor, S, converts the measured activity to a saturated activity. The actual reactor operating history was used to calculate the saturation factor. The reactor was assumed to operate for several periods of constant power. Then, for each isotope, S was calculated.
S=Ihexp(-AT)[1-exp(-At)]
j j
i where Pi = Power of ith interval Po = Full Power A = isotope decay constant Tj = Time between end of ith operating period to reactor shutdown tg = length of ith operating period The saturated a:tivities are given in Table V-6.
The combination of a cadmium shielded and unshielded U-238 foil is included in the flux monitor set. The activity of the unshielded foil is used to correct for U-235 fissions in the shielded foil. As a result of this calculation, the U-238 fission rate was determined to be 94% of the shielded uranium foil activity in Table V-6.
No correction for photo fission of U-238 was included.
SAND requires an initial estimate of the neutron flux spectrum.
This initial estimate was calculated using D0T-III, a two-dimensional discrete ordinate code.
The DLC-23 CASK, 22
\\ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
group neutron cross section library was used. The reactor geometry is shown in Figure V-5.
Dimensions (5) and power distributions used are as shown in Figure V-6.
The distribution
.of pin power was obtained from an end of cycle 3 PDQ calcu-lated with the assembly power distribution adjusted to correspond to an average over the time from startup to the end of cycle 3.
Figure V-7 shows the surveillance capsule detail used in the D0T model.
SAND uses an iterative technique to calculate the neutron flux spectrum. The activities of the set of flux monitors dnd an initial flux spectrum are the input required -by SAND.
[
Activities are calculated for each foil for the flux spectrum l
using the following equation
[
A = g o(E )0(E )aEj 1
where o(E ) is reaction cross section at energy E, barns.
4 2
f O(E ) is the flux at E, n/cm -s mey 4
AE is width of energy band at E, mev.
j 4
The flux spectrum is adjusted by an iterative technique until the calculated and measured activities agree within a standard deviation of five percent.
The result of this is a-620 group neutron flux.
In addition to DOT-III being used for the initial flux estimate for SAND, it was also used to determine the azimuthal flux distribution on and in the vessel relative to that at the surveillance capsule. With the D0T-III results, Lead Factors were calculated as the ratio of the peak fast flux (En > 1.0 MeV) at the surveillance capsule assembly to the peak fast flux at _the vessel-clad interface and 1/4 and 3/4,
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Figure V-7 MILLSTONE UNIT 2 SURVEILLANCE CAPSULE LOCATION CAPSULE W-97 ALL DIMENSIONS IN l
CENTIMETERS CAPSULE AND CLAD ARE STAINLESS STEEL VESSEL CLAD l
CAPSULE s
WATER 3.81 70 1
2 3
4 5
6 RADI AL DISTANCE (CENTIMETERS) 1 214.65 2
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).
vessel thickness locations. These Lead Factors (1.36, 2.62 and 13.9, respectively) were then multiplied by the fast flux generated by SAND to give us the fast flux and fluence at the vessel-clad interface and 1/4 and 3/4 thickness into the vessel respectively. The fluence was calculated for the end of cycle 3 (3.0 Effective Full Power Years at 2700 MWt) and end of life (32 EFPY at 2700 MWt).
The maximum fast flux at the Millstone-II W-97 Surveillance 2
capsule assembly location is 4.0 x 10+10 n/cm -s.
The highest flux was measured in the bottom compartment (6273).
The neutron flux in the top compartment (6214) and middle 10 compartment (6241) was approximately 3.76 x 10 and 3.94 x 10 2
10 n/cm -s, respectively. The maximum fast flux at the vessel-clad interface is deduced from this analysis to be 2
2.94 x 10+10 n/cm -s thus resulting in an updated estimate of the end of life fluence of 2.96 x 10+I9 n/cm. This 2
assumes a 40 year life with an 80% capacity factor.
TABLE V-7 MILLSTONE UNIT 2 FAST NEUTRON FLUX AND FLUENCE VALUES (a)
Fast Flux (E>1.0 Mev) 2 Location Maximum Flux (n/cm -s) b Surveillance Capsule 4.00(+10 Vessel-clad interface 2.94(+10 1/4 thickness of vessel 1.53(+10 f
3/4 thickness of vessel 2.88(+9) t l
Fast Fluences (E>1.0 MeV) c d
Location End of Cycle 3 End of Life Surveillance Capsule 3.78(+18) 4.03(+19)
Vessel-clad interface 2.78(+18) 2.96(+19) 1/4 thickness of vessel 1.44(+18) 1.54(+18) 3/4 thickness of vessel 2.72(+17) 2.90(+18)
(a) Uncertainty in fluence values +30%
(b) Denotes power of 10 (c) 3.0 Effective Full Power Years (EFPY) (2700 mwt)
(d) 32.0 EFPY (2700 mwt),.
This updated estimate of the end of life fluence at the vessel-clad interface is a factor of 1.55 times higher than the original I
2 FSAR predicted value of 1.91 x 10 n/cm. The increase in predicted vessel fluence is the result of a number of factors such as stretch power level (2700 Mwt versus 2560 Mwt), differences in assumed coolant temperature conditions, vessel dimensions, and core wide power distributions (i.e., fuel management strategies).
It should be noted that the end of life vessel fluence predicted as a result of using those updated parameters in a predictive I9 2
mode calculation yields a vessel fluence of 4.0 x 10 n/cm (assuming a 1.15 time-averaged axial peak), which differs from that inferred from the surveillance capsule by 35%. See Table V-7 for additional fluence data.
Reference 3 states that the SAND code will give fluxes that are accurate to within + 10% to + 30% if the errors in the measured activities are within similar limits.
In Tables V-3 through V-5, the quoted uncertainties are at a 2-Sigma level and are determined from counting statistics. An additional error of i 20% for uranium monitors and i 5% for all other metal monitors is estimated from volumetric and gravimetric operations and from the certified uncertainties of calibration isotopes.
The additional error associated with the sulfur monitor results in i 8%. Therefore, it is estimated that the uncertainty in the measured flux at the surveillance capsule location is about + 20% to + 30%. The extrapolated flux in the vessel will be slightly higher, so a reasonable value to use for the uncertainty of the fluence at the vessel ID is 130%. This uncertainty does not include variations that may occur in the Lead Factor due to changes in the azimuthal distributions which could result from different fuel management strategies..
B.
Chemical Analysis Six Charpy specimens each of the base metal (transverse orientation) and weld metal were chemically analyzed by X-ray fluorescence for molybdenum, copper, nickel, manganese, silicon, sulfur and phosphorous content.
Each Charpy specimen was placed in a carrier with a graphite mask for analysis.
Calibration curves were initially established for the seven elements using nine plate and weld specimens with a known chemical content. One of these specimens (11E) was used to check for reproducibility with copper and phosphorous as the selected elements. Twelve separate measurements yielded a copper reproducibility of 11% and a phosphorous reproduc-ibility of 17% at one standard deviation. Specimen 11E was also used as a control for each set of irradiated specimen measurements.
!Results of the analysis of the irradiated specimens and the cont.rol specimen are given in Table V-8.
The base mecal specimens represent four different sections of the surveillance plate; specimens 23A, 23K and 23M represent the same plate section. The weld metal specimens represent six separate layers through the thickness of the weld.
-C.
Strength and Toughness Properties 1.
Tension Tests Tension tests were conducted in accordance with applicable ASTM standards and C-E laboratory procedures. The test method and equipment are described in Appendix A.
The.three irradiated specimens from each material (base metal, weld metal and heat-affected zone) were tested at
TABLE V-8 1RRADIATED PLATE AND WELD MATERIAL CHEMICAL ANALYSIS Chemical Content (Weight Percent)
Specimen Lab Material Cu Ni Mn Si S
P ID Number Mo 21E 6025 Plate (WR) 0.60 0.15 0.61 1.28 0.20 0.016 0.008 23A 6026 Plate (WR) 0.59 0.14 0.60 1.26 0.38 0.015 0.007 23K 6020 Plate (WR) 0.59 0.14 0.60 1.28 0.28 0.016 0.008 23M 6019 Plate (WR) 0.59 0.14 0.60 1.27 0.28 0.15 0.007 251 6015 Plate (WR) 0.60 0.14 0.60 1.27 0.61 0.017 0.009 25C 6018 Plate (WR) 0.59 0.14 0.59 1.29 0.51 0.017 0.008 317 6023 Weld 0.54 0.23 0.065 1.12 0.17 0.015 0.014 31L 6024 Weld 0.52 0.24 0.071 1.14 0.34 0.016 0.014 338 6022 Weld 0.54 0.28 0.059 1.09 0.24 0.015 0.012 34D 6021 Weld 0.54 0.29 0.055 1.08 0.40 0.015 0.014 352 6016 Weld 0.54 0.30 0.057 1.07 0.21 0.015 0.013 36M 6017 Weld 0.53 0.31 0.044 1.10 0.20 0.015 0.01 2 llE
- 1 Control 0.57 0.16 0.57 1.27 0.21 0.014 0.008 llE
- 2 Control 0.57 0.16 0.54 1.24 0.20 0.014 0.008 llE
- 3 Control 0.57 0.16 0.55 1.24 0.20 0.014 0.007 llE
- 4 Control 0.56 0.16 0.54 1.25 0.20 0.014 0.008 40-
room temperature, 250'F and 550F. The tensile properties are listed in Table V-9, and the stress-strain curves are shown in Figure V-8 through V-16.
The pre-irradiation tensile properties (2) are summarized in Table V-10 (each value average of three tests). Photographs of the fracture surface of the broken irradiated specimens are shown in Figure V-17.
2.
Charpy V-Notch Impact Tests Charpy V-notch impact tests were conducted in accordance with applicable ASTM standards and CE laboratory procedures.
The test method and equipment are described in Appendix B.
Twelve irradiated specimens from each material (transverse and longitudinal base metal, weld metal and heat-affected zone) were tested at a series of temperatures to establish the transition temperature behavior. The impact data (impact energy, lateral expansion and fracture appearance as a function of test temperature) are shown in Tables V-11 through V-14 and Figures V-18 through V-29.
(Also shown in each of the figures is the unirradiated transition temperature curve from the baseline evaluation.(2)) Fracture surface photographs of the broken irradiated specimens are shown in Figures V-30 thro 9gh V-33.
Each impact test was instrumented. Additional data related to instrumented impact testing are presented in Appendix C.
TABLE V-9 POST-IRRADIATI0fi TEriSION TEST PROPERTIES Yield Ultimate Elongation Reduction (1-inchgage)
Test Strength Tensile Fracture Fracture (a)
Fractu{g)
Specimen Temp.
0.2% Offset Strength Load Strength Stress of Area TE/UE Material Code
( F)
(ksi)
(ksi)
(lb)
(ksi)
(ksi)
(%)
(%)
Base Metal iXC 72 73.3 96.0 2940 59.9 181 67 26/9.3 lJA 250 70.9 92.5 2880 58.7 174 66 26/10.0 IJU 550 61.7 86.7 2820 57.5 158 64 27/9.3 Weld Metal 3JY 72 86.8 102.1 3180 64.8 213 70 27.5/9.0 3JJ 250 79.5 92.8 2880 58.7 174 66 25/8.8 3J2 550 72.8 90.4 2820 57.0 168 66 22.5/8.5
'?
HAZ 4K7 72 75.8 100.1 2940 59.9 185 68 27/7.0 4KA 250 69.3
- 97. 2 2940 59.9 181 67 22.5/8.5 4J2 550 68.4 87.8 2820 57.9 168 66 22/6.6 a - Fracture strength is the fracture load divided by initial cross sectional area b - Fracture stress is the fracture load divided by final cross sectional area
TABLE V-10 PRE-IRRADIATION TENSION TEST PROPERTIES Yield U1timate Elongation Strength (b) Fracturic) Reduction (1-inch gage)
Fracture 0.2% Offset (a) Tensile Strength Fracture Test Stress of Area TE/UE Strength Load Temp.
Material
( F)
(ksi)
(ksi)
(lb)
(ksi)
(ksi)
(%)
(%)
l Base Metal 71 67.1/63.5 85.7 2620 53.5 183 71 29/11.7 250 61.4/59.4 79.3 2500 51.0 171 70 26/10.2 550 56.3 82.9 2680 54.3 179 69 26/10.1 Weld Metal 71 76.1/73.0 85.8 2540 51.8 201 74 28/11.2 250 73.7/69.1 80.5 2460 50.2 176 71 25/9.3 550 66.8 84.9 2880 58.8 169 65 25/9.9 HAZ 71 67.8/63.7 87.6 2760 56.3 188 70 24/9.2 250 60.4/59.0 80.4 2540 51.8 186 72 25/7.8 550 61.7 83.1 2700 55.1 173 68 21/7.3 l
a - Upper / Lower yield strength where observed.
b - Fracture strength is the Fracture Load divided by initial cross sectional area.
c - Fracture stress is the Fracture Load divided by final cross sectional area.
1 1
i
100,000 l
l l
l l
l l
l l
l l
l l
l 80,000 -
g 60,000 -
8 E
$ 40,000 --
20,000 --
I 0 0 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STRAIN, IN/IN Figure V-8 STRESS STRAIN RECORD OF TENSION TEST, BASE METAL PLATE C 506-1 SPECIMEN No.1KC, TEST TEMPERATURE R.T.
F 100,000 i
i i
i i
i i
i i
i i
i i
80,000 -
G 60,000 --
8 E
b 40,000 20,000 0 O 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STR AIN, IN/IN Figure V-9 STRESS STRAIN RECORD OF TENSION TEST, BASE METAL PLATE C-5061 SPECIMEN No.1JA, TEST TEMPERATURE 250 F - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
1 M,000 i
i i
i g
i i
g l
80,000
- 60,000 E
d w
K b
40,000 20,000 I
I I
I I
I I
I I
I I
I I
I 0
O 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STR AIN, IN/IN Figure V 10 STRESS STRAIN RECORD OF TENSION TEST, BASE METAL PLATE C-506-1 0
SPECIMEN No.1JU, TEST TEMPERATURE 550 F 100,000 i
i i
i i
i i
80,000 -
- 60,000 -
E i
E
$ 40,000 -
20,000 - -
I I
I I
I I
0 O
0.04 0.08 0.12 0.16 0.20 0.24 0.28 STRAIN, IN/IN Figure V 11 STRESS STRAIN RECORD OF TENSION TEST, WELD METAL PLATE C-506-2/3 SPECIMEN No. 3JY, TEST TEMPERATURE R.T. 0F,
100,000 j
l l
l t
l l
l 1
1 I
I I
80,000
_ 60,000 l E
S j
ua CC
$ 40,000 20,000 l
l l
l l
l l
l l
l l
l l
1 0 0 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STR AIN, IN/IN Figure V 12 STRESS STRAIN RECORD OF TENSION TEST, WELD METAL PLATE C-506 2/3 0
SPECIMEN No.3JJ, TEST TEMPERATURE 250 F 100,000 g
g i
i g
g g
i g
g 80,000 60,000 - -
g n.
N
.E
$ 40,000 '
~
20,000 l
l 0O 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STRAIN, IN/lN Figure V-13 STRESS STRAIN RECORD OF TENSION TEST, WELD METAL PLATE C 506-2/3 0
SPECIMEN No. 3J2, TEST TEMPERATURE 550 F __
i i
100,000 g
i i
l l
l l
l l
g i
l 80,000 g 60,000 -
Ch.
N E
$ 40,000 - -
20,000 - -
0 0
0.04 0.08 0.12 0.16 0.20 0.24 0.28 STRAIN, IN/IN Figure V 14 STRESS STRAIN RECORD OF TENSION TEST, HAZ METAL PLATE C-506-1 SPECIMEN No. 4K7, TEST TEMPERATURE, R.T. 0F 100,M 80,000 -
~
g 60,000 -.
N E
40,000 - -
20,000 I
I I
I I
0 O 0.04 0.08 0.12 0.10-0.20 0.24 0.28 STRAIN, IN/lN Figure V-15 STRESS STRAIN RECORD OF TENSION TEST, HAZ METAL PLATE C-506-1 0
SPECIMEN No.4KA,TESTTEMPERATURE 250 F.-
.-. - -_=
140,000 g
g g
g g
g g
g y
y 120,000 100,000
+
g 80,000 E
G 60,000 40,000 20,000 f
I I
I I
I I
I I
I I
I I
I I
0 l
0 0.04 0.08 0.12 0.16 0.20 0.24 0.28 STRAIN, IN/IN Figure V 16 STRESS STRAIN RECORD OF TENSION TEST, HAZ METAL PLATE C-506-1 0
SPECIMEN No.4J2, TEST TEMPERATURE 550 F l
l l
t l -.
l FIGURE V-17 FRACTURE SURFACE OF IRRADIATED TENSION SPECIMENS Base Metal Specimen No:
1KC IJA IJU Test Temp.:
72 F 250 F 550*F Weld Metal g.
.M.
.2,:, l,
'l
~
)
P n
{?j x.'
1
..J, _ 'Q.._
')
_ 3 ' " N-adP3
'Nw
_ q' -
Specimen No:
3JY 3JJ 3J2 Test Temp.:
72 F 250 F 550 F Heat-Affected Zone Lpecimen No:
4K7 4KA 4J2 Test Temp.:
72 F 250 F 550*F ___
TABLE V-11 CHARPY V-h0TCH IMPACT RESULTS IRRADIATED BASE METAL (TRANSVERSE)
(PLATE C-506-1)
Specimen Test Impact Lateral Fracture Identification Temperature Energy Expansion Appearance Number
(*F)
(Ft-lbs.)
(mils)
(% Shear) 25't 0
10 9
0 21M ti0 13 16 10 21E 100 26 26 20 230 100 30 30 20 21K 160 44 44 40 23M 10J 54 49 40 23T 200 64 62 70 25J 240 71 64 100 23A 240 83 74 100 251 280 76 72 100 23K 280 80 72 100 25C 320 87 75 100 1
,e TABLE V-12 CHARPY V-NOTCH IMPACT RESULTS l
L IRRADIATEDBASEMETAL(LONGITUDINAL)
~
,s
.(PLATEC-506-1)
Specimen' Test Impact-Lateral Fracture
-Identification-Temperature.
Energy Expansion Appearance Number
( F)
(Ft-lbs.)
(mils)
(% Shear) j
.114.
40-11 11 0
7168 80
'29:
28 20 11Y~
120 33 36 20 b.
"14P 120 36' 36
' 2 10 l135 160 48
.45-
'. 30 147-160'
.52 45 30' 122 r
160 53 51 30 14M 200 72 66 70
/
137 240-98 -
83 1 00
~- 11 B 280 92 78 100 14U 280 97 79 100 2
14D 350
-90 80 100 e
/
d h
h.
W t......_..
TABLE V-13 CHARPY V-NOTCH IMPACT RESULTS IRRADIATED WELD METAL Specimen Test Impact lateral Fracture Identification Temperature Energy Expansion Appearance Number
( F) __
(Ft-lbs.)
(mils)
(% Shear) 33B 0
17 15 0
33J 40 23 24 20 31D*
40 47 41 30 33C 60 38 37 40 31L*
80 74 57 60 36M 80 80 66 70 352 120 71 59 80 31 7*
120 99 78 90 34A 160 95 78 100 31C*
160 105 83 100 34D 200 93 79 100 36U 240 106 85 100
- Specimens from outside diameter region of weldment; all other from inside diameter region. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.,-n TABLE V CHARPY V-NOTCH IMPACT RESULTS IRRADIATED HAZ METAL ~
(BASEMETALPLATEC-506-1)
Specimen Test Impact.
Lateral Fracture Identification Temperature Energy Expansion Appearance Number-( F)
(Ft-lbs.)
(mils)
(% Shear) 455-0 14 14-10 45A 40 19 18 20 430::
40 20 19 10
-45M 60 26 25 20 42D 80-47 43 50 44D 100 48 41 40 4 51 100 51 47 60
-435; 120 76 51-60 464 160 92 75 90 447 -
200 89 70 100 44M
.200 92 73 100 432-240
.92 72 100 o
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FIGURE V-30 FRACTURE SURFACES OF CHARPY V-NOTCH IMPACT TEST SPECIMENS BASE PETAL (TRANSVERSE)
- n-
% T -
^'
w.
~
.a-
... -.,na Specimen No.:
252 21M 21E Test Temperature:
0F 60 F 100 F
.--.,m
,y,, _ -
_a 4
16' h )* * '
%=-.
3
.g
.
- 4
//
. ha
[tlN".
,e
.,{
.,+ _~
g (,.:.. }.w _
ar- --
h Specimen No.:
23U 21K 23M Test Temperature:
100 F 160 F 160 F Specimen No.:
23T 25J 23A Test Temperature:
200 F 240 F 240 F EB EEN EB Specimen No.:
251 23K 25C Test Temperature:
280 F 280 F 320 F
FIGURE V-31 FRACTURE SURFACES OF CHARPY V-NOTCH IMPACT TEST SPECIMENS BASE METAL (LONGITUDINT.L) n--(
,x-Lh'khhh Yh
- ~ fr ll h h'$ ' ~
$$ ~ W
~
y
.fh:
$ fff k
~
Specimen No.:
114 16B llY Test Temperature:
40 F 80 F 120cF-
, ;,,' i.M, 3,
..w~
.,s\\
- w. 4-/
[. 7
}
f*
r b. n+.m. t Q'
.a <g (~.. --.;
- ,e, x
~
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' S.
M i 4, I
, '1 -Ti? :-< j-
,9 F
- 3.
g.z r >. ;
~
_u m-.a. a a,= w. w n:o :n -
y Specimen No.:
14P 135 147 cn Test Temperature:
120 F 160 F 160 F
' W Z:
&bh
' x<~
Specimen No.:
122 14M 137 Test Temperature:
160 F 200 F 240 F R,$~
3.W.'
w l,4
~.::
Y k~,
2, (Liiy
,, 4
.: > 7.
g.6 g1, -f; g
.s
- .g
' :y sa?. Sin $ a 5
. :,wwa.e
,u Specimen No.:
llB 14U 140' Test Temperature:
280 F 280 F 350 F
FIGURE V-32 FRACTURE SURFACES OF CHARPY V-NOTCH IMPACT TEST SPECIMENS WELD METAL
"""?7r mm,
- -w7-m yy,._,y sg' ^ ' }
),,..
v y.
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.14
~
g-
,..j g
g
')
s
' _ ~ ' }.
ki.-
k
..~
- j A,^
a u
Specimen No.:
33B 33J 31D Test Temperature:
0F 40 F 40 F v & -un w ?'- err e mynmyy, -~
IN h
Specimen No.:
33C 31L 36M Test Temperature:
60 F 80 F 80 F Specimen No.:
352 31 7 34A Test Temperature:
120 F 120 F 160 F
. )
~,
ms'.
g:4
.c.
.es Specimen No.:
31C 34D 36U Test Temperature:
160 F 200 F 240 F
FIGURE V-33 FRACTURE SURFACES OF CHARPY V-NOTCH IMPACT TEST SPECIMENS HEAT-AFFECTED-ZONE l
E '*".97 m..
r Ah % s:-
i
-1 w
,,z s}:.
jf:: S$N,
'l ' l1' Q' ' h y-Q
'._b
- ,? #
Specimen No.:
455 45A 430 Test Temperature:
0F 40 F 40F
,7,_7._ g_ - a r~r a---r-- --
4 M,
l
~^ z
- 1. ;,
- g; qq,
-c p.?
,S P,: iQ.
.' c y:
s:; ;i 3
' K. ~ '.;
2 :.., g, t_
+'
~ ?!r gy __
- ~ -
c f
Specimen No.:
45M 420 44D Test Temperature:
60 F 80 F 100 F Specimen No.:
451 435 464 Test Temperature:
100 F 120 F 160 F my, r mem7 n,, ~,w
" ~
Y
,~P~
a Specimen No.:
447 44M 432 Test Temperature:
200 F 200 F 240 F
.A
l VI.
DATA ANALYSIS A.
Irradiation Exposure The W-97 surveillance capsule was removed from the Millstone Unit 2 reactor vessel following plant shutdown on August 17, 1980, 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />. The surveillance specimens were irradiated at approximately 550 F; the temperature variation along the length of the capsule was negligible based on the similarity in appearance of the three 558 F monitors.
The surveillance capsule was exposed to a maximum neutron 10 2
flux of 4.0 x 10 n/cm -s (E>l.0 MeV). The highest flux was measured in the bottom compartment (6273). The neutron flux in the top compartment (6214) and middle compartment 10 2
10 (6241) was approximately 3.76 x 10 n/cm -s and 3.94 x 10 2
n/cm -s, respectively, indicating a total axial sariation of 6%. The maximum surveillance capsule fluence after 3.0 effective full power years (EFPY) operation at 2700 MWT was, 18 2
therefore, 3.78 x 10 n/cm (E>l.0 MeV).
Lead factors between the surveillance capsule and the maximum azimuthal position on the reactor vessel were calculated to be 1.36 at the vessel-clad interface, 2.62 at the quarter thickness location in the vessel, and 13.9 at the three-quarter thickness location.
The maximum neutron flux at the 10 2
vessel-clad interface was, therefore, 2.94 x 10 n/cm -s.
The predicted end-of-life fluence at the vessel-clad interface I9 is 2.96 x 10 n/cm2 (+30%), usir.g the same power level (2700 Mwt),
coolant inlet temperature *, and fuel management strategy as for the first 3.0 EFPY of operation. This projected estimate I9 is only 55% higher than the FSAR estimate of 1.91 x 10 2
n/cm which was based on 2560 Mwt power level and assumed
- Coolant inlet temperature increased from 542F in Cycle 2 to 549F in Cycle 3.
Although the projected fluence-is based on the lower temperature, the higher inlet temperature will not cause a significant increase in vessel wall fluence.
design parameters. As noted previously, when the DOT-III computer code and the DLC-23 CASK cross section library were used in a predictive mode (ie, without using the surveillance dosimeter measurements for nonnalization) and actual operating parameters were factored in, the end-of-life fluence was I9 2
conservatively predicted to be 4.0 x 10 n/cm.
B.
Chemical Analysis Chemical analysis of the irradiated base metal and weld metal Charpy specimens demonstrated that the encapsulated surveillance specimens had the same chemical composition as that originally' reported in the baseline evaluation. (2)
Furthermore, analysis of the weld specimens enabled delineation between the high copper (inner) and low copper (outer) sections of the surveillance weldment. As can be noted in Table V-8, the copper content of the first two specimens (317 and 31L) is significantly lower than for the remaining The weld metal specimen layout drawing (6)was then four.
used to identify which of the six remaining irradiated Charpy specimens were from the lower copper portion of the weldment - specimen number 317, 31C, 31D and 31L were all from the first two layers on the outside of the weld.
C.
Uniaxial Tension Properties Radiation induced changes in uniaxial tension properties of the Millstone Unit 2 surveillance materials were determined from a comparison of Tables V-9 and V-10.
The yield strength and ultimate strengtn increased approximately 11% following 18 2
irradiation to 3.78 x 10 n/cm. Uniform elongation decreased approximately 11% following irradiation, while reductior, in area and total elongation decreased only 3 to 5%. Fracture stress (fracture load divided by final cross l - _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ _
sectional area) was not changed significantly by irradiation.
In general, property changes were similar for each of the materials despite the large difference in copper content between base metal and weld metal.
Post-irradiation room temperature yield strength values ranged from 73,300 psi for the base metal to 86,800 psi for the weld metal. Total elongation of the irradiated materials ranged from 22% to 27.5%.
D.
Charpy Impact Toughness Properties The radiation induced changes in toughness of the Millstone Unit 2 surveillance materials are summarized in Table VI-1.
Index temperature shifts were measured using the average curves at the 30 ft-lb level (Cv30), 50 ft-lb level (Cv50),
and 35 mils lateral expansion level (Cv35). Upper shelf energy changes were based on the average impact energy for each set of test specimens exhibiting 100% shear fracture measured before and after irradiation. The unirradiated Charpy impact data were obtained from the baseline evaluation report. (2)
The base metal (transverse orientation) exhibited the greatest shift in the 30 ft-lb index temperature (96 F). The shift for the heat-affected-zone material was similar (94 F),
whereas the weld metal shift was only 76 F at the 30 ft-lb ievel. The decrease in upper shelf energy was similar for each of the irradiated materials, ranging from 26 to 29%.
The Millstone Unit 2 design curve for prediction of transition temperature shift as a function of neutron fluence is iven in Figure B3/4.4-1 of the Technical Specificatiors. (7 The design curve prediction for the surveillance capsule exposure 18 2
of 3.78 x 10 n/cm is 120 F, or 25% higher than the base
~
i - _ _ - _ - _ _ - _ _ _ _ _ _ _.
metal shift and 58% higher than the weld metal shift.
In contrast, shifts predicted using NRC Regulatory Guide 1.9950) based on measured copper and phosphorus content are 61 F for the base metal (35% less than measured) and 181 F for the weld metal (140% more than measured).
In order to provide an improved means of predicting transition temperature shift of the controlling beltline material for Millstone Unit 2, Figure VI-1 was developed. This figure should be used to adjust the reactor coolant system pressure-temperature operating limit curves in place of the current Figure B3/4.4-1 of the Millstone Unit 2 Technical Specifications.
Figure VI-l was developed based on the methodology of Regulatory Guide 1.99(8)using the measured value of shift at the 30 ft-I8 2
lb level for the base metal (96 F) at 3.78 x 10 n/cm.
The shift versus neutron fluence relationship is as follows:
ANDTT=156(f0I9) '
where 0 = neutron fluence, E > 1.0 MeV, and ANDTT = transition temperature shift.
TMs relationship should be evaluated again once results are l
available from surveillance materials irradiated to a higher fluence.
l The measured shift for the weld material was 20% lower than that for the base metal material even though the copper
[
content of the weld was significantly higher (0.30% Cu for the weld versus 0.14% Cu for the plate). The greater irradiation j
j resistance of the weld is a result of the low nickel content (0.06%).
Based on on empirical evaluation of weld metal j
irradiation sensitivity (9), low nickel weldments (eg, less than 0.2% Ni) cxhib!ted significantly smaller NDTT shifts,.
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L than high nickel weldments. For the higher copper (inner) section of the Millstone Unit 2 surveillance weld, the predicted shift based on the empirical weld model is 74 F versus the 76 F measured shift. Similarily, for the lower copper (outer) section of the weld, the predicted shift is 48 F versus the 37 F measured shift.
(Note that the measured shift for the outer section is based on only four Charpy testspecimens.) Therefore, the low weld metal shift determined from thew-97 surveillance capsule evaluation is consistent with experimental results. Furthermore, it is reasonable to expect that the base metal, rather than the weld metal, will continue to be the most limiting (highest shift) material in the reactor vessel beltline.
The current pressure / temperature operating limits (Amendment 24, June 24,1977) given in Figure 3.4-2b of the Technical Specifications (7)are based on an initial RT f 5 F and a NDT shift of 135 F after 10 EFPY (equivalent to the original I
2 design fluence of 5.3 x 10 n/cm ).
Given the revised shift prediction method and vessel fluence projection, the estimated base metal shift after 10EFPY is 150 F based on 18 2
surface fluence (9.27 x 10 n/cm ) and 108 F based on tFe 18 2
one-quarter thickness (1/4 t) fluence (4.83 x 10 n/cm ),
Conversely, the design base 135 F shift is predicted to 18 2
occur at a fluence of 7.5 x 10 n/cm, which corresponds to l
8.1 EFPY at the vessel inside surface and 15.6 EFPY at the vessel 1/4 t.
Projected values of NDTT shift and adjusted RT are given NDT in Table VI-2.
Based on the revised shift prediction method and vessel fluence, the predicted shift (vessel inside surface) after 32 EFPY is 268 F versus the original estimate (7) of 225 F.
Similarily, the 1/4 t shift is projected to be 194 F. L
l l
. The predicted decrease in upper shelf energy at end-of-life based on the method given in Regulatory Guide 1.99IO) is 38%
at the one-quarter thickness location in the vessel. Using this conservative prediction, the upper shelf energy of the plates will remain above 65 ft-lb during the design life of the vessel. The upper shelf energy of the weld will rcmain
.above 80 ft-lb throughout vessel. life. These projected values. of upper shelf energy are well in excess of the 50 ft-lb value currently considered to provide a reasonable margin for continued safe operation.
Recommended changes to the surveillance capsule removal schedule are contained in Table VI-3.
The schedule is designed based on 10CFR50, Appendix H and a lead factor
.(capsule to vessel ID)'of 36%.
(For example, the fourth capsule's exposure after 18 EFPY will be equivalent to that at the vessel.inside surface after 24 EFPY). Data from the second capsule will. be available in time for updating the pressure / temperature operating limits. The 104 and 284' capsules are designated for standby purposes.. _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
TABLE VI-1
SUMMARY
OF TOUGHNESS PROPERTY CHANGES FOR MILLSTONE UNIT 2' IO 2
SURVEILLANCE MATERIALS (550F IRRADIATION, 3.78 x 10 n/cm )
-Material Index Temperatures ( F)
Upper Shelf Energy Cv 30.
ACv 30 Cv 50 ACv 50 Cv 35 a Cv 35 Energy (ft-lb)
Change (%)
a b
c d
d d
d Base Metal (WR) 17
'55 25 108 113 96 168 113 123 98 79 27 d
d d
d Base Metal (RW)
.38 59 40 131 108 70 102 93 116 76 94 28 d
d d
d Wald Metal
-30
-10
-28 132 46 76 78 88 62 90 98 26 i
d d
Heat-Affected d
d O
129 Zone
-30 5
64 94 96 91 85 85 91 29 a - 30 ft-Ib Index Temperature b - 50 ft-lb Index Temperature c - 35 mils lateral expansion Index Temperature d - Unirradiated values
TABLE.VI PROJECTED NDTT SHIFT!AND ADJUSTED RTNDT FOR CONTROLLING MATERIAL Vassel Location
b a
Fluence NDTT Shift _
Adj. ' RTNDT Fluence NDTT Shift
'Adj. RTNDT a
18 2
I9 2
Inside Surface 9.27 x 10 n/cm 150 F 155 F 2.96 x 10 n/cm 268 F.
273 F 18 2
I9 2
1/4 t 4.83 x 10 n/cm 108 F 113 F 1.54 x 10 n/cm jg4 7 jggog 17 2
18 2
3/4 t 9.08 x 10 n/cm 47 F 52 F 2.90 x 10 n/cm 84*F 89 F h
a - Projected fluence assuming same power level. coolant inlet temperature and fuel management strategy as for first 3 EFPY of plant operation.
b - Adjusted RTNDT = Initial RTNDT (5 F) plus NDTT shift.
s
TABLE VI-3 PROPOSED NEW CAPSULE REMOVAL SCHEDULE /
FOR MILLSTONE UNIT 2 Removal Azimuthal Approximate Sequence Location Removal Time
- 1 97 3 EFPY 2
83 or 263*
7-8 EFPY 3
263* or 83' 12-14 EFPY 4
277*
17-20 EFPY 5
104*
Standby 6
284*
Standby 0 Time in effective full power years (EFPY) at 2700 MWt.
I l
l l
l - _ _ _ _ _ _ _ _ _ _ _ _ _ _
VII.
REFERENCES 1.
" Program for Irradiation Surveillance of Millstone Point Unit 2 Reactor Vessel Materials," Combustion Engineering, Inc., N-NLM-Oll, October 15, 1970.
2.
" Northeast Utilities Service Company, Millstone Nuclear Unit No. 2, Evaluation of Baseline Specimens, Reactor Vessel Materials Irradiation Surveillance Program," 18767-TR-MCD-009, October 1976.
3.
SAND Users Manual. AFWL-TR67-41, September 1967.
4.
DOT-III Users Manual, ORNL-TM-4280, September 1973.
5.
C-E Drawing J-18767-164-001, " Millstone Reactor Assembly,"
January 3, 1973.
6.
C-E Drawing E-18767-165-lll-02, " Weld Metal Test Specimens,"
June 20, 1972.
7.
Northeast Nuclear Energy Company, Millstone Nuclear Power Station, Unit No. 2, Waterford, Conn., Safety Technical Specifications, Docket 50-336.
8.
Regulatory Guide 1.99, Revision 1, " Effects of Residual
(
Elements on Predicted Radiation Damage to Reactor Vessel Materials," April 1977.
i 9.
J. D. Varsik and S. T. Byrne, "An Empirical Evaluation of the Irradiation Sensitivity of Reactor Pressure Vessel Materials," Effects of Radiation on Structural Materials _,
l' r
l >
MPENDIXA TENSION TESTS - DESCRIPTION AND EQUIPMENT The tension tests were performed using a Riehle universal screw testing machine with a maximum capacity of 30,000 lb and separate scale ranges between 50 lb and 30,000 lb. The machine, thown in Figure A-1, is capable of constant cross head rate or constant strain rate operation. The tension testing was covered by the certificate of calibration which is included at the end of the Appendix A.
Elevated temperature tests were performed in a 2-1/2" ID x 18" long high temperature tension testing furnace with a temperature limit of 1800F. A Richle high temperature, dual range extensometer was used for monitoring specimen elongation.
The tension specimen is depicted in Figure A-2.
Figures A-3 through A-5 are isometric drawings showing the orientation and location of the tension specimens in the base metal, weld metal and heat-affected-zone, respectively.
Tension testing was conducted in accordance with ASTM Method E 8-77, " Tension Tests of Metallic Materials" and/or Recommended Practice E 21-70, 'Short-Time Elevated Temperature Tension Tests of Materials," except as modified by Section 6.1 of Recommended Practice E 184-62, " Effects of High-Energy Radiation on the Mechanical Properties of Metallic Materials." Implementation of the ASTM Test Methods to the testing of irradiated tension. specimens is described in C-E Laboratory Procedure 00000-MCM-041, Revision 0, " Procedure for Tention Testing of Irradiated Metallic Materials," August 16, 1978.
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FIGURE A-1 TENSILE TEST SYSTEM WITH CONTROL CONSOLE AND ELEVATED TEMPERATURE TESTING EQUIPMENT A-2
t A
1.000'
/l
/
f6 - 14 NC 2 Thread O.250"Dia v
0.438 D h
FIGURE A-2 TYPICAL TENSILE SPECIMEN A-3
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bb!5 FIGURE A-4 LOCATION 0F TENSILE SPECIMENS WITHIN WELD METAL TEST MATERIAL A-5 t
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AFFECTED-ZONE TEST MATERIAL
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Acco industries Inc.
cortificate of calibration i
ggg Measurement Systems Division Richle* Testing Machinos 97J conner.ticut Avenue, Don 'J021, Dredgeport, Connecticut 06602 Telephone 203 33'a 2511
~
Ca!ibration Date January 12,1981 Machine Det.crip!!on Richle DS-30 Customer combustion Engineering Serial No.
RA-44372 Building #5
~
Prospect Ilill Road Windsor, CT Attn: D. Vielleux Measurement Systems Division of Acco Industries Inc. certifies that the machine described above has been calibra.ed to ASTM designa-tion E4 using calibrated weights and/or proving rings calibrated to National Bureau of Standards specification.
~
0NSYON Machino Range 30,000 achine Ranoe 3,000 Machine reaJing
Error Machine reading
% Error
~
6000
+.184 E00
+.114_
1200
+. 2 8 7_ _
12000
+ 092__
3 1800
+.251 18000
+.021 2400
.A.1.95_ _
240.0_0
+ 031 g
.1000_
+.156__
___ _.10.000
+.025_
Bachine reading Machine ri ading
% Error Machino Range achine Range 6,_QOO
%. Error 1200 4.292_
l 2400
+m244 _
_ _. 218_ _
3.60.0
+
4 8 0.0
_ -tt196_ _
6000
+.157 M
Machine Range achine Range 15,000 Machine reading
% Error
'. Error
_ achine reading M
3000
+,J1_7._
E 6000
+.J_7_7_
W 8.00.0
+. 0.6 6 _
M 12000
+.123 E
15666
+. 0'74
~
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- Calibrating apparatus used CaWat,n Engineer Capacity serial no Cat date Lab.no.
2,000 579 3-8-79 SJT.232.09/209275 [
10,000 4354 11-29-79 SJT.01/101631 M'I 4=
M-standard uanager 60,000 239 6-16-80 SJT.01/101738 A-7
1
^" ' " " " ' ' ' ' * * ' " "
A= =
929 Connecticut Avenue. Box 9021. Bridgeport, Connecticut 00602 RichIe* Testing Machines Measurement Systems Division Telephone 203 335-2511 Calibration Date
. January 13,1981 Instrument Description Riehle Extensometer, Richle Recorder Customer Combustion Engineering Serial No. DN-120007 R-67338 Prospect Hill Road Windsor, CT Measurement Systems Division of Acco Industries Inc. verifies that the attached graph is certification of calibration of the instrument described above. This instrument was calibrated to ASTM designa-tion E83.
Recorder Extensometer Calibrator EM 528864 Equipment used in calibration
.e >fL Calibrat6on Engineer M
Standards Manager A-8
APPENDIX B p
CHARPY IMPACT TESTS - DESCRIPTION AND EQUIPMENT The standard impact tests and instrumented tests were performed on a
. [,
[F
["k@/. y
[..I.j calibrated instrumented impact testing system, shown in Figure B-1.
C-E's instrumented impact test equipment provides for signal retention and the subsequent data analysis. The output signal from the instrumented tup is
,[M..;.M
<. q. a recorded by an oscilloscope. A permanent visual record was made of the
..O~
.t load signal, as it was displayed on the oscilloscope screen, with a polaroid p, y i,.:p.? A 25'I4d' camera.
y[g K. [(
n ;;n The system consists of the following elements:
k.8td _ *}<'
A Model SI-l BLH Sonntag Universal Impact Machine with a specific-a.
ally machined pendulum tup, instrumented with four resistance i:
N-4 strain gages in full bridge circuit.
This tup " load cell" is calibrated statically and dynamically to provide a given pounds /
volt sensitivity for known settic.gs of the balance and gain on the dynamic response system. The instrumented machine meets all impact test machine requirements of ASTM and is certified by AMMRC, the U.S. Army Materials and Mechanics Research Center
[M5f' M.t.l; A,.' N N-(WatertownArsenal). A copy of the certification papers is g%: w.7 included in this Appendix.
f.Q.g., 7. E k.f.htl '>Q
$h b.
A Model SCO Dynatup dynamic response system which supplies regulated
..2-,,
and constant dc excitation to strain gages on the pendulum tup,
' 3.k provides balancing, variable load sensitivity and calibration
- 4.. gff
@ql.y.D dOP functions, and amplifies load-time signal to a +10 volt, +100 milliampere level while preserving kHz frequency response and
--;.' i V-0.05 percent accuracy while simultaneously recording the area f[g';*{
NI beneath the load-time trace.
l 9
)
l
.J B-1
(
e F
b A photoelectric triggering device and velocometer composed of a 1 if E
c.
g high intensity light directed through a grid mounted on the pendulum of the impact tester, and passed to a photosensor through f
fiber optics.
A special circuit ensures accurate, reliable and m
f fail safe triggering of the oscilloscope recorder plus an accurate display of the average veloci ty of the pendulum during impact.
o f~
6 d.
A SiO3N Dual Beam Tektronix Storage Oscilloscope with a No. SA18N i
dual-trace amplifier plug-in unit and a No. SB12N dual tine base p ug-in unit.
Also included is a C-58 camera with mounting l
ddapter.
This device gives a display of each test trace for f
v;sual analysis of the load-time impluse recorded by the instrument.
L
[
The standard Charpy specimen is described in Fiqure B-2.
riqures B-3 through B-5 are isometric drawings showing the orientation and location E
of the Charpy impact specimens in the base retal, weld metal and heat-lj -ik,..$ ).
dffected-zone, respectively.
u
.- ' y:
.f
~..
.9.
3 All Charpy inmact tests were conducted in accordance with AS3 Method D,:;.,,:..c[
,^
, j.
a E 23-72, " Notched Bar Impact Testing of Metallic Materials."
Impl ementation s
g.-.
of ASTM E23 for the testing of irradiated Charpy specimens is described
'mj -.
4 %
.c in C-E Laboratory Procedure 00000-MCM-040, Revision 0, " Procedure for x,g Q.%/;
Instrumented Charpy Impact Tost;ng of Irradiated Metallic Materials,"
g jh.M.4 g4 h
Ju1y 31, 1978.
.yy.,7 ;.;,9
?A y'C$,e.
L d
E The constant tennerature necessary for conducting the Charpy impact jpy _6g tests was obtained from a series of circulating licuid baths capable of maintaining stable temperature throughout the rance of -150'F to A;5. - ' ; l c
+250'F.
Any selected tenperature in this range was maintained to an N'
accuracy of 2 For tests above 250^F, specimens were heated in a
-E.dh.4
.n&t e.
controlled circulation furnace where temperature was maintained to an f
accuracy of CF.
The temperature baths were composed of the following A
equipment:
I r
=
E B-2
?
Two Neslab Constant Temperature Circulating Baths - Model TEZ 10, with Model CT 150 Thermoregulators and Labline 11 inch diameter thermocups.,
Designated Baths 1 and 4.
l l
fledium:
Ethylene Glycol - room temperature to 250 F.
One Neslab Constant Temperature Circulating Bath - Model TEZ 10 with a Model CT 59 Tiermoregulator and a Labline 11 inch diameter thermo cup.
Designated Bath 2.
Medium:
Isopropanol - room temperature to -10 F.
Neslab Portable Bath Cooler, Model PCB-2 connected.
One Low Temperature Stirred Bath, one 11 inch diameter thermo cup, one Honeywell Controller and Solenoid control valves to Flexi-Cool cooling system. Designated Bath 5.
Medium:
Isopropanol - room temperature to -150 F.
Coolant: Freon One Grieve Industrial oven, controlled air circulation.
Designated Bath 3.
Medium: Air,100 F to 800 F.
2d All baths - Copper Constantan Thermocouple l.
Honeywell Six Point Temperature Chart Recorder l~,
Digitec Thermocouple Thermometer - Model 590 TF l
Standard Mercury Column Thermometer Bimetallic - spring Thermometer The temperature instruments were calibrated in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Paragraph 2360.
Copies of the applicable calibration certificates are provided at the end of this aopendix.
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TEMPEPTi^URF RTTV, AND INSTRUMENTED CHAPPY IMPACT DAT A P00CE SSING EOL'IPENT 4,_yt', '.,c%,
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FIGURE B-2 TYPICAL CHARPY V-NOTCH IMPACT SPECIMEN l
B-5
c 5"
=
,th tt x-x, e
p
+
===8*
em
=.
Transverse t
_m
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l g..
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- g,'x.;fy.Qa
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BE E
I FIGURE B-3 LOCATION OF CHARPY IMPACT SPECIMENS WITHIN BASE METAL TEST MATERIAL EM
mm Weld Metal 5n eID Side ANf' y
fi;
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4 11/4"
/
i
/
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s
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=
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t = Plate Thickness i
jf I/fl f/
f 4f.
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FIGURE B-4 LOCATION OF CHARPY IMPACT SPECIMENS WII,'"
i
~
WELD METAL TEST MATERIAL
%,a A.v: 1.i ~ '
,$4C g
- - cx y.y c.
,,..;+ _ -
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r
=
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!_ _ _ _ _'v' t = Plate Thickness ig I
l FIGURE B-5 LOCATION OF CHARPY IMPACT SPECIMENS WITHIN HEAT-AFFECTED-ZONE TEST MATERIAL
I ll.
4
. ?---
DEPARTMENT OF THE ARMY g
~
~g ARMY MATERIALS AND MECHANICS RESEARCH CENTER WATERTOWN. MASSACHUSETTS o2172
. r
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DRXMR-HQ S January 1981 Combustion Engineering, Inc.
ATTN:
Mr. Ray lluriburt Dept. 9454-501 1000 Prospect liill Road Windsor, CT 06095
^
Dear Mr. Hurlburt:
A set of Charpy impact test specimens broken on the 240 ft-lb capacity Satec machine, Serial No.
1366 has been received for evaluation along with the completed questionnaire.
i The results of the tests indicate the machine to be producing acceptable energy values at both energy levels (see inclosed table).
This machine satisfies the proof-test requirements of AS*IN Standard E-23.
If this machine is moved or undergoes any major repairs or adjustments, this certification becomes invalid and the machine must be rechecked.
Removal of F
the pendulum, replacement of anvils or adjusting the height of drop are 1 _:
examples of such major repairs or adjustments.
It should be noted that if a specimen requires over 80% of the machine capacity to fracture, the machine
~
should be checked to assure that the pendulum is straight, the anvils or striker have not been damaged and that all bolts are still tight. This certification is valid for one year from the date of the test.
~
Sincerely yours,
]Q 1 Inc1 PAUL W. ROLSTON Table Chief Quality Engineering Branch XMR Form IFL-3 7
B-9
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Nucle.it I.iteor.itoet~s INSTitut.1LtJT CAlllittAllON ItLOUliti.f.1LtJT SilLET
[
DATE: 1/22/81 e
[ EQUIPMENT Digital Thermocouple Thermometer AREA 235-5 r
ti-vb r
INSTRUMENT READADILITY CALIDRATION CilECKED MIN FUNCTION TYPE RANGE READAUILITY ACCURACY FREQUENCY DY c
r Thermometer Digital
-313
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+1 F 3 months
=
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INSTRUMENTED U1ARPY V-NOTCH DA'A ANALYSIS j
.%ff-lh ;
All baseline and irradiated Charpy impact tests in this program were performed
$:'4. O4.,9 m q1pa;? 9'.t on an instrumented test system.
Instrumented impact testing provides more
.* V/.,
g,.
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n.p -. c4 =
quantitative data from a Charpy specimen which enable a more detailed analysis of the surveillance material toughness behavior.
P # d QI.
'v+, y : y ;-s Q&rf:....,
19M I P Photograpns of the oscilloscope traces of load and energy varsus time were
..m v, 7 +
taken f or each test of the 'Jase plate (transverse and longitudinal ori-g+g>; p7 entation), weld, and heat-affected-zone.
From each t ace, the general
- p;).f ] 3.
<. dr f[; pNg yield load (PGY), maximum load (PM), and fracture load (PF) were detemined, as shown in Tables C-1 through C 4 For each material, the lois were
..p 2
plotted against the corresponding test temperature to generate the irradiated pr.a c:1. -
ph.}[.
load / temperature diagrams.
The post -irr 3diation load /tenperature resul ts are shown in Figures C-1 to C-4
[d((.p {-
-J.: Y.f-
- b.y,.7 W.9 Three index temperatures are of interest.
T, the brittle transition B
- 4.
ys if;.Ap: $
temperature, corresponds to the onset of ductile fracture; below T the B
fracture is completely brittle.
T, the ductility transition temperaturc, s'7hh<
N y;;!,3.gg.,
corresponds to the mid-transition region where the fracture has become
- .- {. g.J j$
4 predominantly ductile.
T, the ductility temperature, corresponds to the
. +.
D h
onset of the upper sbulf energy where fracture is completely ductile.
.c
, 4. v.c
.i,,
' ;:.g;.. i The radiation-induced toughness property changes of the surveillance materials are summar ized in Table C-5.
Standard Charpy impact data are
' C-C.
- ,, %., p.
',,[ 5,..
included with the instrumented data since each method represents a unique material property.
The standard Cnarpy test provides a bulk measurement of 7.f jr T4..Of[
m the energy to initiate and propagate a crack through to failure of the
},, p. M. M w.m..
s material.
In contrast, analysis of the instrumented data enables charac-
?* e D.
'(; h.gQs.r$
terization of the components of the dynamic load behavior prior to material failure. The shif t in the brittle transition temperature, T,
nd the i
B
. {' ;.
ductility transition temperature, Tg, are comparable to the shif t in the 30 f g' f ;h. -
1,qs ;.
j,' is u.c.% y.
>:y. Y;i,
?,\\Ol) '
y p !.
, )%Q+i Q
, 9.d ;f "
.? Q $ ?t! ?
C-1
1 The radiation induced changes in the ft-lb Charpy index temperature, Cv30 instrumented data therefore tend to corroborate the changes determined from the standard Charpy impact data; The third parameter obtainable from the instrumented data is T, the ductility D
temperature, which is given in Table C-5.
T corresponds closely with the D
onset of the upper shelf energy (minimum temperature for the material to exhibit 100% shear fracture). The agreement is seen to hold for both the unirradiated and irradiated data.
The instrumented Charpy analysis substantiates the results from the standard impact tests.
In particular, this approach provides a more quantitative means of measuring radiation induced property changes by analysis of the entire load record rather than using the single (bulk) measurement of impact energy. As more experience is gained with this technique, it offers the potential of providing a more quantitative measurement of toughness property changes than is possible with current impact testing.
\\
C-2
TABLE C-1 INSTRUMENTED CHARPY IMPACT TEST, MILLSTONE UNIT 2 IRRADIATED BASE METAL (TRANSVERSE)
Test Fast Fracture Specimen Temperature Yield Load, Maximum Load,
- Load, Identification
( F)
PGY (lb)
PM (lb)'
PF (1b) 252 0
3400 3300 21M 60 3200 3600 l
21E 100 3000 3890 l
23U 100 No Record 21K 160 2800 3800 l
23M 160 2900 3900 3800 23T 200 2900 3800 3600 25J 240 3000 3900 23A 240 2700 3900 251 280 2700 3800 23K 280 2600 3700 25C 320 2700 3800 I
C-3
TABLE C-2 INSTRUMENTED CHARPY IMPACT TEST, MILLSTONE UNIT 2 IRRADIATED BASEMETAL(LONGITUDINAL)
Test Fast Fracture Specimen Temperature Yield Load, Maximum Load, Load.
Identification
( F)
PGY (lb)
PM (lb)
PF (lb) 114 40 3000 3200 16B 80 2800 3600 11Y 120 2700 3500 14P 120 2900 3900 135 160 2800 3800 No Record 147 160 122 160 2800 4000 3900 No Record 14M 200 137 240 No Record 118 280 2800 4000 14U 280 2600 3900 No Record 14D 350 C-4
TABLE C-3 INSTRUMENTED CHARPY IMPACT TEST, HILLSTONE UNIT 2 IRRADIATED WELD METAL Test Fast Fracture e
Specimen Temperature Yield load, Maximum Load,
- Load, Identification
( F)
PGY (lb)
PM (lb)
PF (lb) 33B 0
3500 4000 33J 40 3400 3900 310 40 4400 33C 60 3300 4000 36M 80 3300 4300 3400 31L 80 3300 4300 4000 352 120 3100 4100 3900 317 120 3100 4100 3400 31C 160 3000 4100 34A 160 3000 4000 34D 200 2900 3900 36U 240 2800 3900 l
0 9
C-5
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- n c
,.. p..
INSTRUMENTED CHARPY IMPACT
.?s:.r: p.9 w.,.3, J
+;
TEST, MILLSTONE UNIT 2 IRRADIATED
.M...'k ' 0:
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- b t.
.. y."' L. O
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Test Fast Fracture
' Yr
~
Specimen Temperature Yield Load, Maximum Load,
- Load, I.
Identificatien
( F)
PGY (lb)
PM (lb)
PF (lb)
+ ;.?
t 455 0
3500 3900 7.-
e*. f.
.. e; c,
M 43U 40 3400 3900 J, .s y\\. te i
- f E
45A 40 3400 3800 i f ' 'l.g.
y t
.s.
f 45M 60 3400 4000 7,.,5s =d %. ;
e W.1 -.
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-t
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420 80 3300 3900 t..
..... y e e
451 100 3300 4300 4200 g? < ~..A4 x,
f 440 100 3300 4300 4100
.J, 7,.;.:. ).E; 435 120
-- No Record cr? rs
+
kied
- x. :;a. fiji,...
464 160 3000 4300 3100 w
?
44M 200 3000 4100
',,. y p... "
447 200 2900 4200
[
f$
Y 432 240 3000 4200
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?:J TABLE C-5 Y
. ;j TOUGHNESS PROPERTY CHANGES BASED ON
-t
- .7. 5 INSTRUMENTED CHARPY IMPACT TEST
~.l..
-?
m 3:;
Min. Teraperature For 3*
'E TB ( F)
ATB(F)
TN(F)
ATN(F)
ACv30( F)
ACv TD(.f) 100% Shear Fracture (*F)
%?i 50 Material
. +.
a, Base Metal (WR) i..
160 160
.6 unirrad
-76 60
.N irrad
-40 36 130 70 96 113 200 240 a-n
- 2. :
w Base Metal (RW)
/,
132 160 4-66
[R -
- g..
unirrad.
-60
.?
irrad 30 90 160 94 70 93 225 240
- t.i
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+ A.
m n
f a
Weld Metal 56 80 s g,
-16 unirrad.
-84 irrad.
-40 44 47 63 76 88 150 160 g
4 i
3 Heat-Affected Zone 70 80
.s;
-10
- 4,-
unirrad.
-150 irrad.
-45 105 85 95 94 91 135 200 g
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