ML18096B076
| ML18096B076 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/30/1992 |
| From: | James Anderson, Chicots J, Madeyski A WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML18096B075 | List: |
| References | |
| WCAP-13366, NUDOCS 9211040255 | |
| Download: ML18096B076 (134) | |
Text
9211040255 PDR ADOCK p
9
r P1 WESTINGHOUSE CLASS 3 WESTINGHOUSE CLASS 3 ANALYSIS OF CAPSULE X FROM THE PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM J. M. Chicots S. L. Anderson A. Madeyski June 1992 Work Performed Under Shop Order PQPP-106 fol Prepared by Westinghouse Electric Corporation for the Public Service Electric and Gas Company Approved by: 7-a' ~
J1.
T. A. Meyer>Maml er Structural Reliability and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
© 1992 Westinghouse Electric Corp.
All Rights Reserved J.-
- ~
PREFACE This report has been technically reviewed and verified.
Sections 1 through 5, 7, 8 and Appendix A Section 6 Appendix B Reviewer:
E. Terek E. P. Lippincott M. J. Malone i
TABLE OF CONTENTS Section Title Page 1.0
SUMMARY
OF RESULTS 1-1
2.0 INTRODUCTION
2-1
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE X 5-1 5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-4 5.3 Tension Test Results 5-6 5.4 Wedge Opening Loading 5-7 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-7
'1. 0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1
8.0 REFERENCES
8-1 APPENDIX A -
LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS APPENDIX B -
HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION ii
4-1 4-2 4-3 LIST OF TABLES Chemical Composition of the Salem Unit 2 Reactor Vessel Surveillance Material Heat Treatment History of the Salem Unit 2 Reactor Vessel Surveillance Material Chemical Composition of Four Salem Unit 2 Charpy Specimens Removed from Surveillance Capsule X 4-3 4-4 4-5 4-4 Chemistry Results from the NBS Certified Reference Standards 4-6 5-1 Charpy V-Notch Impact Data for the Salem Unit 2 5-8 Intermediate Shell Plate B4712-2 Irradiated at 550°F, Fluence 1.16 x 1019 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Salem Unit 2 5-9 Reactor Vessel Weld Metal and HAZ Metal Irradiated 5-3 5-4 5-5 at 550°F, Fluence 1.16 x 1019 n/cm2 (E > 1.0 MeV)
Instrumented Charpy Impact Test Results for the Salem Unit 2 Intermediate Shell Plate B4712-2 Irradiated at 550°F, Fluence 1.16 x 1019 n/cm2 (E > 1.0 MeV)
Instrumented Charpy Impact Test Results for the Salem Unit 2-Weld Metal and Heat-Affected-Zone (HAZ) Metal, Irradiated at 550°F, Fluence 1.16 x 1019 n/cm2 (E > I. 0 MeV)
Effect of 550°F Irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV) on the Notch Toughness Properties of the Salem Unit 2 Reactor Vessel Surveillance Materials i i i 5-10 5-11 5-12
5-6 5-7 6-1 6-2 6-3 6-4 6-5 6-6 6-7 6-8 LIST OF TABLES (Continued)
Comparison of the Salem Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions Tensile Properties for the Salem Unit 2 Reactor Vessel Surveillance Materials Irradiated at 550°F to 1.16 x 1019 n/cm2 (E > 1.0 MeV)
Calculated Fast Neutron Exposure Parameters at the Surveillance Capsule Center Calculated Fast Neutron Exposure Rates at the Pressure Vessel Clad/Base Metal Interface Relative Radial Distributions of Neutron Flux (E > 1.0 MeV) within the Pressure Vessel Wall Relative Radial Distributions of Neutron Flux (E > 0.1 MeV) within the Pressure Vessel Wall Relative Radial Distributions of Iron Displacement Rate (dpa) within the Pressure Vessel Wall Nuclear Parameters for Neutron Flux Monitors Monthly Thermal Generation During the First Five Fuel Cycles of the Salem Unit 2 Reactor Measured Sensor Activities and Reactions Rates iv 5-13 5-14 6-14 6-15 6-16 6-17 6-18 6-19 6-20 6-21
6-9 6-10 6-11 6-12 6-13 6-14 6-15 LIST OF TABLES (Continued)
Summary of Neutron Dosimetry Results Comparison of Measured and FERRET Calculated Reaction Rates at the Surveillance Capsule Center Adjusted Neutron Energy Spectrum at the Surveillance Capsule Center Comparison of Calculated and Measured Exposure Levels for Capsule X Neutron Exposure Projections at Key Locations on the Pressure Vessel Clad/Base Metal Interface Projected Neutron Exposure Values Updated Lead Factors for Salem Unit 2 Surveillance Capsules v
6-23 6-24 6-25 6-26 6-27 6-28 6-29
Figure 4-1 4-2 5-1 5-2 5-3 5-4 5-5 5-6 5-7 LIST OF ILLUSTRATIONS Title Arrangement of Surveillance Capsules in the Salem Unit 2 Reactor Vessel Capsule X Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Longi~udinal Orientation)
Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Transverse Orientation)
Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Surveillance Weld Metal Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Longitudinal Orientation)
Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Transverse Orientation)
Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Surveillance Weld Metal vi 4-7 4-8 5-15 5-16 5-17 *~
5-18 5-19 5-20 5-21
- Figure 5-8 5-9 5-10 5-11 5-12 5-13 5-14 6-1 LIST OF ILLUSTRATIONS {Continued)
Title Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal Tensile Properties for Salem Unit 2 Reator Vessel Intermediate Shell Plate 84712-2 {Transverse Orientation)
Tensile Properties for Salem Unit 2 Reactor Vessel Surveillance Weld Metal Fractured Tensile Specimens from Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2
{Transverse Orientation)
Fractured Tensile Specimens from Salem Unit 2 Reactor Vessel Surveillance Weld Metal Engineering Stress-Strain Curves for Plate 84712-2 Tensile Specimens JT7 and JT8 {Transverse Orientation)
Engineering Stress-Strain Curves for Weld Metal Tensile Specimens JW7 and JW8 Plan View of a Reactor Vessel Surveillance Capsule vii 5-22 5-23 5-24 5-25 5-26 5-27 5-28 6-13
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule X, the third capsule to be removed from the Public Service Electric and Gas Company Salem Unit 2 reactor pressure vessel, led to the following conclusions:
o The capsu~e received an average fast neutron fluence (E > J.O MeV) of 1.16 x 1019 n/cm2 after 6.20 EFPY of plant operation.
o Irradiation of the reactor vessel intermediate shell plate 84712-2 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction (longitudinal orientation), to 1.16 x 1019 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb.transition temperature increase of 80°F and a 50 ft-lb transition temperature increase of ll0°F. This results in a 30 ft-lb transition temperature of ll0°F and a 50 ft-lb transition temperature of
- 165°F for longitudinally oriented specimens.
o Irradiation of the reactor vessel intermediate shell plate 84712-2 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction (transverse orientation), to 1.16 x 1019 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 125°F and a 50 ft-lb transition temperature increase of 130°F.
This results in a 30 ft-lb transition temperature of 135°F and a 50 ft-lb transition temperature of 190°F for transversely oriented specimens.
o The weld metal Charpy specimens irradiated to 1.16 x 1019 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 195°F and a 50 ft-lb transition temperature increase of 205°F. This results in a 30 ft-lb transition temperature of 165°F and a 50 ft-lb transition temperature of 205°F for the weld metal.
1-1
o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 1.16 x 1019 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 150°F and a 50 ft-lb transition temperature increase of 175°F. This results in a 30 ft-lb transition temperature of 25°F and a 50 ft-lb transition temperature of 80°F for the weld HAZ metal.
o Irradiation of intermediate shell plate 84712-2 (longitudinal orientation) to 1.16 x 1019 n/cm2 (E > 1.0 MeV) resulted in a 2.0 ft-lb decrease in average upper shelf energy. This results in an upper shelf energy of 120 ft-lb for the intermediate shell plate 84712-2 (longitudinal orientation).
o Irradiation of intermediate shell plate 84712-2 (transverse orientation) to 1.16 x 1019 n/cm2 (E > 1.0 MeV) resulted in an 8 ft-lb decrease in average upper shelf energy. This results in an upper shelf energy of 89 ft-lb for the intermediate shell plate 84712-2 (transverse orientation).
o
- The average upper shelf energy of the weld metal decreased 25 ft-lb after irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV).
This results in an upper shelf energy of 86 ft-lb for the weld metal.
o The average upper shelf energy of the weld HAZ metal decreased 31 ft-lb after irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV).
This results in an upper shelf energy of 89 ft-lb for the weld HAZ metal.
o The surveillance capsule X test results indicate that the weld metal and plate 84712-2 longitudinal specimens, the 30 ft-lb transition temperature changes and upper shelf energy decreases are less than the Regulatory Guide 1.99 Revision 2 predictions.
1-2
o The surveillance capsule X test results indicate that the plate 84712-2 transverse specimens upper shelf energy decrease is less than the Regulatory Guide 1.99, Revision 2 prediction and the 30 ft-lb transition temperature increase exceeds the Regulatory Guide 1.99 Revision 2 prediction by 41°F.
o The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life (32 EFPY) of the vessel as required by 10CFR50, Appendix G.
o The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the Salem Unit 2 reactor vessel is as follows:
Vessel inner radius * - 1.76 ~ 1019 n/cm2 Vessel 1/4 thickness - 9.27 x 1018 n/cm2 Vessel 3/4 thickness - 1.85 x 1018 n/cm2
- Clad/base metal interface 1-3
SECTION
2.0 INTRODUCTION
This report presents the results of the examination of Capsule X, the third capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Public Service Electric and Gas Company Salem Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Salem Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-8824 entitled "Public Service Electric and Gas Company Salem Unit No. 2 Reactor Vessel Radiation Surveillance Program" by J. H. Phillips, et. al.[l] The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM El85-73, "Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Power Systems personnel were contacted to aid in the preparation of procedures for removing capsule "X" from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where, the postirradiation mechanical testing* of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained
- from surveillance capsule "X" removed from the Public Service Electric and Gas Company Salem Unit 2 reactor vessel and discusses the analysis of the data.
2-1
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry.
The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment.
The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as A533 Grade B Class 1 (base material of the Salem Unit 2 reactor pressure vessel intermediate shell plate 84712-2) are well documented in the 1iterature. 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[4].
The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNoT>*
RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NOTT per ASTM E-208)[5] or the temperatur~ 60°F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KrR curve) which appears in Appendix G to the ASME Code.
The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results -Obtained from several heats of pressure vessel steel.
When a given material is indexed to the KrR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
3-1
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties.
The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, such as the Salem Unit 2 Reactor Vessel Radiation Surveillance Program[!], in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested.
The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNoT> due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement.
This adjusted RTNDT (RTNDT initial +
ARTNoT> is used to index the material to the KrR 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.
3-2
SECTION
4.0 DESCRIPTION
OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the Salem Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up. The eight capsules were positioned in the reactor vessel between the thermal shield 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 X was removed after 6.20 Effective Full Power Years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and wedge opening loading (WOL) specimens (Figure 4-2) from the intermediate shell plate 84712-2, charpy V-notch and tensile specimens from submerged arc weld metal representative of the beltline region weld metal and charpy V-notch specimens from weld Heat-Affected-Zone (HAZ) of plate 84712-2. All hea~-affected zone specimens were obtained from within the HAZ of plate 84712-2 of the representative weld.
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 th~ plate.
Base metal Charpy V-notch impact specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation) and also with the longitudinal axis of the specimen normal to the major working direction (transverse orientation).
- Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction.
The WOL specimens were machined such that the simulated crack in the specimen would propagate parallel to the major working direction for the plate specimen and parallel to the weld direction.
4-1
The chemical composition and heat treatment of the surveillance material is presented iri Tables 4-1 and 4-2.
The chemical analysis and heat treatment reported in Tables 4-1 and 4-2 were obtained from unirradiated material used in the surveillance program[1l.
In addition, a chemical analysis performed on four irradiated Charpy specimens from the weld and base metal plate 84712-2 is reported in Table 4-3.
The chemistry results from the NBS certified reference standards are given in Table 4-4.
Capsule X contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt wire (cadmium-shielded and unshielded).
In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were placed in the capsule to measure the integrated flux at specific neutron energy levels.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. These thermal monitors were used to define the maximum temperature attained by the tes~ specimens during irradiation. The composition of the two alloys and their melting points are as follows:
2.5% Ag, 97.5% Pb 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point:
579°F (304°C)
Melting Point:
590°F (310°C) 4-2
TABLE 4-1 CHEMICAL COMPOSITION OF THE SALEM UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIAL[!]
Chemical Composition Cwt%)
Intermediate Shell Element Plate 84712-2 Weld Metal Westinghouse Lukens Steel Westinghouse Analysis Analysis Analysis c
0.23 0.22 0.10 Mn 1.34 1.37 1.27 p
0.015 0.011 0.017 s
0.010 0.015 0.011
. Si 0.30 0.24 0.29 Ni 0.61 0.60
- 0. 71 Mo 0.55 0.55 0.45 Cr 0.089 (a) 0.015 Cu 0.10 (a) 0.23 Al 0.030 (a) 0.007 Co 0.015 (a) 0.024 v
<0.010 (a) 0.001 Sn 0.008 (a) 0.005 N
0.004 (a) 0.007 Notes:
(a)
Not measured.
4-3
Material Intermediate Shell (Plate 84712-2)
Weldment TABLE 4-2 HEAT TREATMENT HISTORY OF THE SALEM UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIAL[!]
Temperature (° F) 1550 - 1650 4
1225 +/- 25 4
1150 +/- 25 40 Time (hrs)
Water Quenched Air Cool
- Furnace Cool 1150 +/- 25 40 Furnace Cool 4-4
TABLE 4-3 CHEMICAL COMPOSITION OF FOUR SALEM UNIT 2 CHARPY SPECIMENS REMOVED FROM SURVEILLANCE CAPSULE X Metals Fe Mn Cr Ni Mo Co Cu p
v c
s Si Analyses Metals Carbon Sulfur Silicon Concentration in Weight Percent JT-42 JW-38 JW-42 JW-47 Base Metal Weld Weld Weld MATRIX ELEMENT: Remainder by Difference 1.317 1.211 1.392 1.224 0.109 0.026 0.032 0.026 0.625 0.728 0.734 0.728 0.572 0.456 0.473 0.454 0.014 0.015 0.015 0.014
. 0.122 0.267 0.244 0.247 0.008 0.014 0.014 0.014 0.001 0.003 0.003 0.003 0.229 0.112 0.095 0.100 0.0103 0.0081 0.0100 0.0093 0.348 0.439 0.173 0.168 Method of Analysis ICPS, Inductively Coupled Plasma Spectrometry EC-12, LECO Carbon Analyzer Combustion/titration Dissolution/gravimetric 4-5
TABLE 4-4 CHEMISTRY RESULTS FROM THE NBS CERTIFIED REFERENCE STANDARDS MATERIAL ID Low Alloy Steel:
NBS Certified Reference Standards Concentration in Weight Percent NBS 361 NBS 362 Metals Certified Measured Certified Measured Fe 95.6 (Matrix) 95.3 (Matrix)
Mn 0.66 0.668 1.04 1.074 Cr 0.694 0.677 0.30 0.306 Ni 2.00 1.993 0.59 0.584 Mo 0.19 0.193 0.068 0.065 Co 0.032 0.031 0.30 0.302 Cu 0.042 0.042 0.50 0.494 p
0.014 0.013 0.041 0.033 v
0.011 0.012 0.040 0.039 c
0.383 0.379 0.160 0.157 s
0.014 0.036 0.0356 Si 0.222 0.231 0.39 0.415 s
NBS 12ld 0.013 0.0136
=============================================================
MATERIAL ID Low Alloy Steel:
NBS Certified Reference Standards Concentration in Weight Percent NBS 363 NBS 364 Metals Certified Measured Certified Measured Fe 94.4 (Matrix) 96.7 (Matrix)
Mn 1.50 1.524 0.255 0.255 Cr 1.31 1.322 0.063 0.062 Ni 0.30 0.301 0.144 0.134 Mo 0.028 0.026 0.49 0.492 Co 0.048 0.045 0.15 0.149 Cu 0.10 0.098 0.249 0.249 p
0.029 0.024 0.01 0.009 v
0.31 0.326
-0.105 0.109
=============================================================
- Matrix element calculated as difference for material balance. ( )
4-6
CAPSULE (TYP) w--
x 270° z
s Figure 4-1.
Arrangement of Surveillance Capsules in the Salem Unit 2 Reactor Vessel 4-7
CAPSULE SPECIMEN NUMBERING. CODE:
JT -
PLATE B 4712-2 (transverse orientation)
JL -
PLATE 8 4712-2 (longitudinal orientation)
JW -
CORE REGION WELD METAL JH -
HEAT-AFFECTED-ZONE METAL TEMS I LE WOL WOL WOL WOL TEMS I LE
~000001
~
~
~
~
~
~
C*
Cot Cd, I
I I I I
I I I I I HO*
I I
I I Ho I
I I UONITO" I I I
I I
I I I I I
.. u
~
w
- t...I o I I
CHARPY JW48 JH8 JW47 JH7 Cu I
I I
I I.___..., __,,
I I
I I I
I I
DOS IMElEP.
CHAR PY BLOCI JWij6 JH6 C:J JW115 JH5 CHAR PY CHARPY JWqq
- Jwq3 I I I I I I JJqq JTq1 I I I
u u
n I I Jwqz JWql Cu I...._-lo--**
I I I I JH2 JT4 I CHARPY CHAR PY CHAR PY JHq8 Jl32 JHq7 JL31 I I I I
- I I I I
I I I I I I u
u I I
,. --+--<..., :
I I CHAR PY JHqq Jl28 JH43 Jl27 Cu CHAR PY CHAR PY JH42 JL26 JH40 JH38 JHq I Jl25 JH39 JH37 Co CofCdJ I
I I I I I I u
I I h --+--<... I I I I
I
=- C[NT(RIA(CIO,..Orvrssu ----------------------*-------------
TO TQP O' VfSSlL To 1onow o* vrssn SI APERTURE CARD A~~ Avaiiable On Apef't~~e~~
Figure 4-2 Capsule X Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters
SECTION 5.0 TESTING OF SPECIMENS FROM CAPSULE X 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Power Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H[2l, ASTM Specification E185-82[6l, and Westinghouse Remote Metallographic Facility (RMF) Procedure MHL 8402, Revision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were.carefully removed, inspected for identification number, and checked against the master list in WCAP-8824[1].
No discrepancies were found.
Examination of the two low-melting point 579°F (304°C) and 590°F (310°C) eutectic alloys indicated no melting of either type of thermal monitor.
Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579°F (304°C).
The Charpy impact tests were performed per ASTM Specification E23-88[?] and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J machine.
The tup (striker) of the Charpy machine is instrumented with a GRC 8301 instrumentation system,, feeding information into an IBM XT computer.
With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E0).
From the load-time curve (Appendix A), the ~oad of general yielding (PGy), the time to general yielding (tGy), the maximum load (PM), and the time to maximum load (tM) can be determined.
Under some test conditions, a sharp drop in load indicative of fast fracture was observed.
The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA)*
5-1
The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record.
The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen.
Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (Eo) and the energy at maximum load.
The yield stress (ay) was calculated from the three-point bend formula having the following expression:
ay = P6y * {L/[B*(W-a)2*C]}
(1) where L = distance between the specimen supports in the impact testing machine; B = the width of the specimen measured parallel to the notch; W = height of the specimen, measured perpendicularly to the notch; a = notch depth.
The constant C is dependent on the notch flank angle
(~), notch root radius (p), and the type.of loading (i.e., pure bending or three-point bending).
In three-point bending a Charpy specimen in which ~ = 45° and p =
0.010", Equatipn 1 is valid with C = 1.21. Therefore (for L = 4W),
For the Charpy specimens, B = 0.394 in., W = 0.394 in., and a= 0.079 in.
Equation 2 then reduces to:
ay = 33.3 x P6y where ay is in units of psi and P6y is in units of lbs. The flow (2)
(3) stress was calculated from the average of the yield and maximum loads 1 also using the three-point bend formula.
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-a9[8].
The lateral expansion was measured using a dial gage rig similar to that shown in the same specif1cation.
5-2
Tension tests were performed on a 20,000-pound Instron Model 1115, split-consol~ test machine, per ASTM Specification E8-89b[9] and E21-79 (1988)[10], and RMF Procedure 8102, Revision 1.
All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45.
The upper pull rod was connected through a universal joint to improve axiality of loading.
The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.
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 extensometer gage length is 1.00 inch.
The extensometer is rated as Class B-2 per ASTM E83-85[ll].
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 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 of room temperature to 550°F (288°C).
The upper grip was used to control the furnace temperature.
During the actual testing the grip temperatures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to +/-2°F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve.
The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area.
The final diameter and final gage length were determined from post-fracture photographs.
The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5-3
5.2 Charpy-V-Notch Impact Test Results The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule X, which was irradiated to 1.16 x 1019 n/cm2 (E > 1.0 MeV}, are presented in Tables 5-1 through 5-4 and are compared with unirradiated results[l] as shown in Figures 5-1 through 5-4.
The transition temperature increases and upper shelf energy decreases for the Capsule X materials are summarized in Table 5-5.
Irradiation of the reactor vessel intermediate shell plate 84712-2 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 1.16 x 1019 n/cm2 (E > l.~ MeV) at 550°F (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 80°F and in a 50 ft-lb transition temperature increase of ll0°F. This resulted in a 30 ft-lb transition temperature of ll0°F and a 50 ft-lb transition temperature of 165°F (longitudinal orientation).
The average Upper Shelf Energy (USE) of the intermediate shell plate 84712-2 Charpy specimens (longitudinal orientation) resulted in a energy decrease of 2.0 ft-lb after irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at 550°F. This results in an average USE of 120 ft-lb (Figure 5-1).
Irradiation of the reactor vessel intermediate shell plate 84712-2 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation) to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at 550°F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 125°F and in a 50 ft-lb transition temperature increase of 130°F. This resulted in a 30 ft-lb transition temperature of 135°F and a 50 ft-lb transition temperature of 190°F (transverse orientation).
5-4
The average USE of the intermediate shell plate 84712-2 Charpy specimens (transverse -orientation) resulted in an energy decrease of 8 ft-lb after irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at SS0°F. This resulted in an average USE pf 89 ft-lb (Figure S-2).
Irradiation of the reactor vessel core region weld metal Charpy specimens to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at SS0°F (Figure S-3) resulted in an increase of 19S°F in 30 ft-lb transition temperature and a SO ft-lb transition temperature increase of 20S°F.
This resulted in a 30 ft-lb transition temperature of 16S°F and the SO ft-lb transition temperature of 20S°F.
The average USE of the reactor vessel core region weld metal resulted in an energy decrease of 2S ft-lb after irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at 550°F.
This resulted in an average USE of 86 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at SS0°F (Figure S-4) resulted in a 30 ft-lb transition temperature increase of 1S0°F and a SO ft-lb transition temperature increase of 17S°F. This resulted in a 30 ft-lb transition*
temperature of 2S°F and the SO ft-lb transition temperature of 80°F.
The average USE of the reactor vessel weld HAZ metal experienced an energy decrease of 31 ft-lb after irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at S50°F.
This resulted in an average USE of 89 ft-lb (Figure 5-4).
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures S-S through 5-8 and show an increasingly ductile or tougher appearance with increasing test temperature.
The surveillance capsule X test results indicate that the weld metal and plate 84712-2 longitudinal specimens, the 30 ft-lb transition temperature changes and upper shelf energy decreases are less than the Regulatory Guide 1.99 Revision 2 predictions.
5-5
The surveillance capsule X test results indicate that the plate B4712-2 transverse specimens upper shelf energy decrease is less than the Regulatory Guide 1.99, Revision 2 prediction and the 30 ft-lb transition temperature.
increase exceeds the Regulatory Guide 1.99 Revision 2 prediction by 41°F.
The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life (32 EFPY) of the vessel as required by 10CFR50, Appendix G.
The load-time records for the individual instrumented Charpy specimen tests are shown in Appendix A.
5.3 Tension Test Results The results of the tension tests performed on the various materials contained in Capsule X irradiated to 1.16 x 1019 n/cm2 (E > 1.0 MeV) are presented in Table 5-7 and are compared with unirradiated results[!] as shown in Figures 5-9 and 5-10.
The results of the tension tests performed on the intermediate shell *plate B4712-2 (transverse orientation) indicated that irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at 550°F caused less than a 12 ksi increase in the 0.2 percent offset yield strength and less than a 12 ksi increase in the ultimate tensile strength when ~ompared to unirradiated data[l] (Figure 5-9).
The results of the tension tests performed on the reactor vessel core region weld metal indicated that irradiation to 1.16 x 1019 n/cm2 (E > 1.0 MeV) at 550°F caused a 15 to 20 ksi increase in the 0.2 percent offset yield strength and a 13 to 18 ksi increase in the ultimate tensile strength when compared to unirradiated data[l] (Figure 5-10).
The fractured tension specimens for the intermediate shell plate B4712-2 material are shown in Figures 5-11; while the fractured specimens for the weld metal are shown in Figure 5-12.
5-6.
The engineering stress-strain curves for the tension tests are shown in Figures 5-13 and 5-14.
5.4 Wedge Opening Loading Specimens Per the surveillance capsule testing contract with the Public Service Electric and Gas Company the wedge opening loading fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center.
5-7
Sample No.
JL32 JL28 JL27 JL25 JL30 JL29 JL31 JL26 JT48 JT44 JT43 JT38 JT42 JT46 JT40 JT41 JT37 JT39 JT47 JT45
- No data.
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE SALEM UNIT 2 INTERMEDIATE SHELL PLATE 84712-2 IRRADIATED AT 550°F, FLUENCE 1.16 x 1019 n/cm2 (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion 121 i:ill.
(£t-lb2 ill (mils2 i!!!!!!l Longitudinal Orientation 25
(- 4) 11
( 15) 10 (0.25) 75
( 24) 16
( 22) 13 (0.33) 110
( 43) 33
( 45) 28 (0. 71) 150
( 66) 46
( 62) 36 (0. 91) 190
( 88) 62
( 84) 52 (1. 32) 200
( 93)
( **)
(**)
225 (107) 118 (160) 83 (2.11) 275 (135) 123 (167) 84 (2.13)
Transverse Orientation
-25
(-32) 4
( 5) 4 (0.10) 50
( 10) 15
( 20) 10 (0.25) 100
( 38) 14
( 19) 18 (0.46) 115
( 46) 28
( 38) 26 (0. 66) 125
( 52) 27
( 37) 24 (0. 61) 150
( 66) 45
( 61) 44 (1.12) 175
( 79) 36
( 49) 32 (0. 81) 175
( 79) 39
( 53) 34 (0. 86) 200
( 93) 47
( 64) 38 (0. 97) 215 (102) 58
( 79) 50 (1. 27) 225 (107) 89 (121) 62 (1.57) 275 (135) 89 (121) 69 (1. 75)
Computer malfunction.
5-8 Shear (3) 10 15 30 50 70 100 100 5
15 20 25 25 45 50 65 65 95 100 100
Sample No.
JW45 JW39 JW40 JW44 JW47 JW46 JW38 JW37 JW42 JW43 JW41 JW48 JH38 JH44 JH48 JH42 JH45 JH41 JH46 JH40 JH39 JH47 JH37 JH43 TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE SALEM UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550°F, FLUENCE 1.16 x 1019 n/cm2 (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shea.r i:n !:.Ql.
(ft-lb) ill ' (mils)
~..la Weld Metal
-75
(-59) 3
( 4) 1 (0.03) 1 25
(- 4) 7
(
- 9) 4 (0.10) 5 75
( 24) 19
( 26) 22 (0.56) 15 100
( 38) 23
( 31) 18 (0.46) 20 125
( 52) 24
( 33) 17 (0.43) 20 150
( 66) 19
( 26) 18 (0.46) 20 175
( 79) 31
( 42) 25 (0.64) 35 200
( 93) 42
( 57) 31 (0. 79) 60 225 (107) 65
( 88) 52 (1.32) 95 250 (121) 73
( 99) 53 (1.35) 100 300 (149) 68
( 92) 58 (1.47) 100 350 (177) 118 (160) 86 (2.18) 100 HAZ Metal
-100
(-73) 9
( 12) 5 (0.13) 10
-50
(-46) 19
( 26) 15 (0.38) 20 0
(-18) 23
( 31) 15 (0.38) 20 15
(- 9) 33
( 45) 23 (0.58) 25 25
(- 4) 42
( 57) 29 (0. 74) 40 50
( 10) 44
( 60) 29 (0. 74) 45 75
( 24) 48
( 65) 36 (0. 91) 60 100
( 38) 71
( 96) 52 (1.32) 90 125
' ( 52) 34
( 46) 31 (0. 79) 40 150
( 66) 81 (110) 56 (1.42) 95 200
( 93) 92 (125) 67 (1. 70) 100 275 (135) 95 (129) 65 (1.65) 100 5-9
U'I I......
0 Teet Sample Temp Number
.!..71.
JL32 26 JL28 7q JL27 110 JL26 160 JL30 190 JL20 200 JL31 226 JL26 276 JT48
-26 JT44 60 JT43 100 JT38 116 JT42 126 JT46 160 JT40 176 JT41 176 JT37 200 JT30 216 JT47 226 JT46 276
- Fully ductile TABLE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE SALEM UNIT 2 INTERMEDIATE SHELL PLATE 84712-2 IRRADIATED AT 550°F, FLUENCE 1.16 X 1019 n/cm2 (E > 1.0 MeV)
Normalized Enersiee Charpy Charpy Maximum Prop. Yield Time Maximum Time to Fracture Arrest Yield Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress (ft-lb)
(ft-lb/in2)
(lbs)
(msec)
(lbs)
(msec)
(lbs)
(lbs)
(ksi)
Lonsitudinal Orientation 11 80 62 36 3636 0.14 3730 0.18 1119 178 117 16 120 67 62 3460 0.13 3879 0.21 1164 890 116 33 266 176 91 3626 0.14 4467 0.41 1340 822 120 46 370 246 126 3060 0.12 4378 0.67 1313 1627 101 62 400 231 268 3268 0.13 4363 0.66 1306 2310 108 118 060 326 624 2966 0.13 4661 0.72 1370 98 123 ooo 316 676 3627 0.27 4612 0.74 1364 120 Transverse Orientation 4
32 17 16 2664 o.oo 2823 0.10 847 88 16 121 33 88 3431 0.12 3602 0.14 1108 248 114 14 113 34 78 3608 0.12 3627 0.14 1088 062 117 28 226 114 111 2770 0.13 3666 0.36 1070 883 92 27 217 110 108 2697 0.12 3688 0.33 1076 1077 90 46 362 166 196 2972 0.13 4006 0.46 1202 1642 99 36 200 128 162 2660 0.11 3646 0.38 1094 674 89 39 314 181 133 2922 0.12 4038 0.46 1211 1840 07 47 378 93 286 2718 0.12 3371 0.31 1011 1276 90 68 467 218 249 2011 0.13 4048 0.66 1221 2621 97 80 717 303 414 3237 0.13 4389 0.67 1317 108 89 717 294 422 3093 0.14 4071 0.60 1221 103 fracture.
No arrest load.
- No data.
Computer. malfunction.
Flow Stress (ksi) 121 122 134 123 126 126 136 91 118 119 106 104 116 106 116 101 116 127 110
Teet Sample Temp Number 121 JW46
-76 JW30 26 JW40 76 JW44 100 JW47 126 JW46 160 JW38 176 JW37 200 JW42 226 JW43 260 (J'1 JW41 300 I
JW48 360 JH38
-100 JH44
-60 JH48 0
JH42 16 JH46 26 JH41 60 JH46 76 JH40 100 JH3o 126 JH47 160 JH37 200 JH43.
276
- Fully ductile TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE SALEM UNIT 2 WELD METAL AND HEAT-AFFECTED-ZONE (HAZ} METAL, IRRADIATED AT 550°F, FLUENCE 1.16 x 1019 n/cm2 (E > 1.0 MeV}
Normalized Energies Oharp;y Oharp;y Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress (ft-lb}
(ft-lblin2}
ill!!l (meec}
(lbs}
(meec}
(lbs}
(lbs}
(ksi}
Weld Metal
- 'a 24 12 12 1763 0.07 2016 0.00 606 68 7
66 27 30 2021 0.11 3320 0.13 006 61 07 10 163 06 68 3417 0.16 3878 0.28 1164 667 114 23 186 147 30 3811 0.13 4603 0.33 1381 361 127 24 103 137 66 3223 0.12 3030 0.37 1182 660 107 19 163 90 63 3631 0.13 4144 0.26 1243 733 121 31 260 104 146 3002 0.13 3628 0.31 1068 891 100 42 338 236 102 3443 0.16 4319 0.62 1300 929 114 66 623 194 329 2668 0.14 3664 0.63 1100 979 86 73 688 181 406 2863 0.12 4166 0.47 1261 96 68 648 181 366 3142 0.31 3860 0.61 1161 104 118 960 333 617 3166 0.41 3928 1.06 1170 106 RAZ Metal 9
72 46 28 3921 0.12 4337 0.16 1301 130 19 163 72 81 4071 0.14 4463 0.20 1336 1371 136 23 186 163 32 3382 0.11 4406 0.36 1349 112 33 266 162 104 3187 0.17 4122 0.46 1237 706 106 42 338 260 80 3633 0.13 4827 0.64 1448 67 121 44 364
- n7 137 2902 0.12 4017 0.64 1206 181 96 48 387 212 174 3782 0.13 4627 0.46 1388 2223 126 71 672 227 346 3606 0.13 4476 0.61 1346 1393 116 34 274 224 60 3682 0.32 4171 0.70 1261 2081 119 81 662 306 346 3066 0.13 4016 0.74 1206 461 102 02 741 236 606 3228 0.12 4397 0.66 i320 107 96 766 308 467 3180 0.13 4466 0.70 1340 106
- fracture.
No arrest. load.
Flow Stress (ksi) 63 104 121 140 110 129 108 129 103 117 116 118 137 142 131 121 141 116 140 133 129 118 127 127
TABLE 5-5 EFFECT OF 550°F IRRADIATION TO 1.16 x 1019 n/cm2 (E > 1.0 MeV)
ON THE NOTCH TOUGHNESS PROPERTIES OF THE SALEM UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS Average 30 ft-lb(l)
Average 35 mil(l)
Average 50 ft-lb(l)
Average Energy(l)
Transition Lateral Expansion Transition Absorption at Temperature ( °F)
Temperature ( ° F)
Temperature ( °F)
Full Shear (ft-lb)
Material Un irradiated Irradiated A.T Un irradiated Irradiated A.T Unirradiated Irradiated In Un irradiated Irradiated A(ft-lbl Plate 64712-2 30 110 80 45 150 105 55 165 110 122 120 2.0 (Longitudina 1)
U'I Plate 64712-2 10 135 125 40 170 130 60 190 130 97 89 8
I.....
N (Transverse)
Weld Metal
- 30 165 195
- 10 205 215 o_
205 205 111 86
- 25 HAZ Metal
-125 25 150
- 65 65 130
- 95 80 175 120 89
- 31 (1) "AVERAGE" is defined as the value read from the curve fitted through the data points of the Charpy tests (Figures 5-1 through 5-4).
U1 I......
w TABLE 5-6 COMPARISON OF THE SALEM UNIT 2 SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS Material Capsule Pl ate B4712-2 T
(Longitudinal) u x
Pl ate B4712-2 T
(Transverse) u x
Weld Metal T
u x
Fluente 1019 n/cm2 0.276 0.570 1.16 0.276 0.570 1.16 0.272 0.570 1.16 30 ft-lb Transition Temp. Shift R.G. 1.99 Rev. 2 (Predicted)
Measured (of)
(oF) 52 50 68 70 84 80 52 70 68 95 84 125 129 155 167 190 206 195 Upper Shelf Energy Decrease R.G. 1.99 Rev. 2 (Predicted)
(%)
14 17 20 14 17 20 30 35 41 Measured
(%)
6 8
1 8
13 8
29 33 22
U1 I
.p.
Material Plate (Transverse)
Weld Sa.mp le Number JT7 "JT8
.JV/7 JW8 Teet Temp.
rn 136 660 180 660 TABLE 5-7 TENSILE PROPERTIES FOR THE SALEM UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS IRRADIATED AT 550°F TO 1.16 X 1018 n/cm2 (E > 1.0 MeV) 0.2" Yield Ultimate Fracture Fracture Fracture Uniform Total Strength Strength Load Streee Strength Elongation Elongation (kei)
(kei)
(kiel (kei)
(kei)
(%)
(%)
76.9 97.8 3.60 183.6 71.3 9.9 18.9 72.8 96.8 3,90 127.0 79.6 9.6 16.6 86.1 99.8 3.68 167.2 72.9 9.0 20.0 77.9 96.7 3.89 171.6 79.3 9.0 16.7 Reduction in Area.
(%}
61 37 66 64
100
,-.. 80
~
0:::
60
<l:
w
- r:
(/) 40 20 0
100 tfl 80
'E.
Q 60 x w 40
_J t-20
<l:
_J 160 140 120
_Q I
100
~
<.+...
80 L:J 0:::
60 w
- z w
40 20 0
Figure 5-1.
(O[)
-150
-100
-50 0
50 100 150 200
~*
I 3,1 I
r 2
-~/ ~.,,
L
~/ /
I.--------
2.5
~
t
__J 2.0 t
j i
I I
---1 1.5 I
I E
r-
__j E
o o r
1.0 I
I i
I
--i L-0.5 I l
! 0 0 UNIRRADIATED e IRRADIATED <550°Fl. FLUENC[
!.]6 x 1019 n/CM?
200 i
I i
i I
_j 160 I
0 I I 120 -,
80 40 3
0
-200
-100 0
100 200 300 400 TEMPERATURE (°F)
Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Longitudinal Orientation) 5-15
Cl::'.
<I:
w I (/)
LI)
"E.
Q x w
-1 t-
<I:
-1
_Q I
~
'-+-
L'.J Cl::'.
w z
w Figure 5-2.
(O[)
-J50
-JOO
-50 0
50 JOO J50 200 100 80 60
~
40 I
20 I-2 I
0 100 2.5 80 2.0 60 1.5 E
40 1.0 E
20
- J-0.5
\\
0 0
0 UNJRRADIATED e IRRADIATED <550a Fl, FLUENCE l.16 x 1019 n/cn2 J20 L 160 100 0
120 80
)
60 80 40 40 20 0
0
-200
-100 0
100 200 300 400 500 TEMPERATURE (OF)
Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Transverse Orientation) 5-16
Figure 5-3.
Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Surveillance Weld Metal 5-17
Figure 5-4.
Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-18
JL32 JL30 Figure 5-5.
J128 JL27 JL25 J129 J131 J126 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Longitudinal Orientation) 5-19
JT48 JT42 JT37 Figure 5-6.
JT44 JT43 JT38 JT46 JT40 JT41 JT39 JT47 JT45 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Transverse Orientation) 5-20
JW45 JW47 JW42 Figure 5-7.
JW39 JW40 JW44 JW46 JW38 JW37 JW43 JW41 JW48 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Surveillance Weld Metal 5-21
JH38 JH45 JH39 Figure 5-8.
JH44 JH48 JH42 JH41
- JH46 JH40 JH47 JH37 JH43 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-22
120 110 100 r-.. 90
-Vi
.:::c.
'.../
80 V?
V?
w Cl<'.
r-70 V?
60 50 40 80 70 60 r-.. 50 r-40
_J r-u 30
- J i:=i 20 10 0
Figure 5-9.
(OC) 0 50 100 150 200 250 300 800 700 ULTIMATE TENSILE STRENGTH 600 0
~2 CL
~
500 0.2/. YIELD STRENGTH 400 300
!::,, 0 UNIRRADJATED A e IRRADIATED AT 550cF, FLUENCE 1,61 x1019 n/cM2 8
~
TOT AL ELDNGA TJDN
~
UNIFORM tCDNGA TIDN 0
100 200 300 400 500 TEMPERATURE (°F)
Tensile Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Transverse Orientation) 5-23
120 110 100 90 U1 80
(/)
(/)
w 0,::
t--
70
(/)
60 50 40 80 70 60 50 t--
40
_J t--
LJ 30
- J i::::l 20 10 0
Figure 5-10.
(OC) 0 50 100 150 200 250 300 800 700 ULTIMATE TENSILE STRENGTH __.
~
600 d
0...
500 0.2/. YIELD STRENGTH
~
400 300
!:::.. 0 UNIRRADIATED A e IRRADIATED AT SS0°F, FLUENCE l.16 x1019 n!cM2 0
0 A-
/2 TOTAL ELONGATION 1~;
£!,
2~
~
UNIFORM ELDNGA TIDN 0
100 200 300 400 500 TEMPERATURE (°F)
Tensile Properties for Salem Unit 2 Reactor Vessel Surveillance Weld Metal 5-24
Figure 5-11.
Specimen JT7 135°F Specimen JT8 550°F Fractured Tensile Specimens from Salem Unit 2 Reactor Vessel Intermediate Shell Plate 84712-2 (Transverse Orientation) 5-25
Figure 5-12.
Specimen JW7 Specimen JW8 550°F Fractured Tensile Specimens from Salem Unit 2 Reactor Vessel Surveillance Weld Metal 5-26
I 100.00 90.00 80.00 iii 70.00
~
60.00 cti 50.00
(/)
w 0:::
I-40.00
(/)
30.00 20.00 10.00 0.00 0.0 100.00 90.00 80.00 70.00 en
~ 60.00 en*
en 50.00 w
a:
f-40.00 en 30.00 20.00 10.00 0.00 0.0 0.04 STRESS-STRAIN CURVE SALEM UNIT 2 "X" CAPSULE JT7 135 F 0.08 0.12
- STRAIN, IN/IN STRESS-STRAIN CURVE SALEM UNIT 2 "X° CAPSULE JT8 550 F 0.04 0.08.
0.12 STRAIN, IN/IN
' \\
0.16 0.20
\\
0.16 Figure 5-13.
Engineering Stress-Strain Curves for Plate 84712-2 Tensile Specimens JT7 and JT8 (Transverse Orientation) 5-27
STRESS-STRAIN CURVE SALEM UNIT 2 "X" CAPSULE 120.00 100.00 17i 80.00
~
ui 60.00 en w
~
I-en 40.00 JW7 20.00 180 F 0.00 0.0 0.10 0.20
- STRAIN, IN/IN STRESS-STRAIN CURVE SALEM UNIT 2 "X" CAPSULE 100.00 90.00 80.00 70.00 (j)
~ 60.00 CJ)
(j) 50.00 LU a:
I-40.00 (j) 30.00 JW8 20.00 550 F 10.00 0.00 0.0 0.04 0.08 0.12 0.16 STRAIN, IN/IN Figure 5-14.
Engineering Stress-Strain Curve for Weld Metal Tensile Specimens JW7 and JW8 5-28
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of Light Water Reactor (LWR) reactor pressure vessel surveillance programs for two reasons.
First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known.
Second, in order. to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens.
The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.
The latter information is derived solely from analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for LWR applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition.
In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pr~ssure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, "Analysis and Interpretation of Light Water Reactor 6-1
Surveillance Results," recommends reporting iron atom displacements (dpa) alon~
with fluence (E > 1.0 MeV) to provide a data base for future reference.
The energy dependent dpa function to be.used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom 11
[ 22l. The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide *1.99, "Radiation Damage to Reactor Vessel Materials 11
[ 3]
This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance Capsule X.
Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history.
The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided.
6.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1.
Eight irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program.
The capsules are located at azimuthal angles of 4.0°, 40.0°, 140.0°, 176.0°,
184.0°, 220.0°, 320.0°, and 356.0° relative to the core cardinal axes as shown in Figure 4-1.
A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 6-1.
The stainless steel specimen containers. are 1.0 inch square and approximately 38 inches in height.
The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core.
6-2
From a neutron transport standpoint, the surveillance capsule structures are significant: They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel.
In order to properly determine the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations for the surveillance capsules and react~r vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {~(E > 1.0 Mev), ~(E > 0.1 Mev), and dpa} through the vessel wall.
The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e.,
dpa/~(E > 1.0 MeV), within the pressure vessel geometry.
The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e.,
the 1/4T, 1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core.
The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement.
These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects of varying neutron yield 6-3
per fission and fission spectrum introduced by the build-in of plutonium as the burnup of individual fuel assemblies increased.
The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the forward calculation provided the means to:
- 1.
Evaluate neutron dosimetry obtained from surveillance capsule locations.
- 2.
Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.
- 3.
Enable a-direct comparison of analytical prediction with measurement.
- 4.
Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R, 8 geometry using the DOT two-dimensional discrete ordinates code[12] and the SAILOR cross~section library[13J.
The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for LWR applications.
In these analyses anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an s8 order of angular quadrature.
The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop plants.
Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery.
Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used.
Since it is unlikely that a single reactor would have a power distribution at the nominal +2a 6-4
level for a large number of fuel cycles, the use of this reference distribution is expected -to yield somewhat conservative results.
All adjoint analyses were also carried out using an s8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library.
Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as-the geometric center of each surveillance capsule.
Again, these calculations were run in R, 9 geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case,¢ (E > 1.0 MeV).
Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:
R (r, 8) =Ir Ia IE I(r, 9, E) S (r, 8, E) r dr d9 dE where:
R (r, 9)
I (r, 9, E)
S (r, 9, E)
¢ (E > 1.0 MeV) at radius r and azimuthal angle 8
= Adjoint importance function at radius, r, azimuthal angle 9, and neutron source energy E.
= Neutron source strength at core location r, 9 and energy E.
Although the adjoint importance functions used in the analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/¢ (E > 1.0 MeV) is insensitive to changing core source distributions.
In the application of these adjoint importance functions to the Salem Unit 2 reactor, therefore, the iron displacement rates (dpa/sec) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/¢ (E > 1.0 MeV) and
¢ (E > 0.1 MeV)/¢ (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific¢ (E > 1.0 MeV) solutions from the individual adjoint evaluations.
6-5
The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design reports for the first six operating cycles of Salem Unit 2[14 through 19].
Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5.
The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall.
In Table 6-1, the calculated exposure parameters [¢ (E > 1.0 MeV),
¢(E > 0.1 MeV), and dpa] are given at the geometric center of the two symmetric surveillance capsule positions for both the design basis and the plant specific core power distributions.
The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared.
Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the Cycles 1 through 6 plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad/base metal interface; and, thus, represent the maxi~um exposure levels of the vessel wall itself.
Radial gradient information for neutron flux (E > 1.0 MeV),
neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively.
The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.
6-6
F9r example, the neutron flux (E > 1.0 MeV) at the l/4T position on the 45° azimuth is given by:
where:
¢ (220.27,45°)
F (225.75, 45°)
= ¢(220.27, 45°) F (225.75, 45°)
Projected neutron flux at the l/4T position on the 45° azimuth
= Projected or calculated neutron flux at the vessel inner radius on the 45° azimuth.
= Relative radial distribution function from Table 6-3.
Similar expressions apply for exposure parameters in terms of¢ (E > 0.1 MeV) and dpa/sec.
6.3 Neutron Dosimetry The passive neutron sensors included in the Salem Unit 2 surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest [¢ (E > 1.0 Mev), ¢ (E >
0.1 MeV), dpa].
The relative locat~ons of the neutron sensors within the capsules are shown in Figure 4-2.
The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules.
The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.
6-7
The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest.
Rather, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period.
An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.
In particular, the following variables are of interest:
0 The specific activity of each monitor.
0 The operating history of the reactor.
0 The energy response of the monitor.
0 The neutron energy spectrum at the monitor location.
0 The physical characteristics of the monitor.
The specific activity of each of the neutron monitors was determined using established ASTM procedures [20 through 331.
Following sample preparation and weighing, the activity of each monitor was determined by means of a
_lithium-drifted germanium, Ge(Li), gamma spectrometer.
The irradiation history of the Salem Unit 2 reactor during Cycles 1 through 6 was obtained from NUREG-0020, 11 Licensed Operating Reactors Status Summary Report 11 [37]
for the applicable period.
The irradiation history applicable to Capsule Xis given in Table 6-7.
Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8.
Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code [341 The 6-8 I
FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data.
The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra.
In the FERRET evaluations, a log-normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations.
In general, the measured values fare linearly related to the flux ¢ by some response matrix A:
(s,ex:)
f
= L:
g (s)
A ig (ex:)
¢ g
where i indexes the measured values belonging to a single data set s, g designates the energy group and ex: delineates spectra that may be simultaneously adjusted.
For example, R = L:
i g
(]
ig
¢ g
relates a set of measured reaction rates Ri to a single spectrum ¢g by the multigroup cross section <Jig*
(In this case, FERRET also adjusts the cross-sections.) The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.
In the FERRET analysis of the dosimetry data, the continuous quantities (i.e.,
fluxes and cross-sections) were approximated in 53 groups.
The calculated
\\
fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [35J.
This procedure was carried out by first expanding the a pri?ri spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed 6-9
to the group scheme used in FERRET.
The cross-sections were also collapsed into the 53 energy-group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions.
The cross sections were taken from the ENDF/B-V dosimetry file.
Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.
For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight.
In some cases, as for the cross sections, a multigroup covariance matrix is used.
More often, a simple parameterized form is used:
where RN specifies an overall fractional normalization uncertainty (i.e.,
complete correlation) for the corresponding set of values.
The fractional uncertainties Rg specify additional random uncertainties for group g that are correlated with a correlation matrix:
-(g-9')2 (1 - 9) ogg' + 9 exp [
2
]
21 The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 1 (9 specifies the strength of the latter term).
For the a priori calculated fluxes, a short-range correlation of 1 = 6 groups was used.
This choice implies that neighboring groups are strongly correlated when 9 is close to 1.
Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E.
Maerker[36J. Maerker's results are closely duplicated when 1 = 6.
For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.
6-10
Using this approach results in the dpa equivalent fluence values listed in Table 6-14.
In Table 6-15 updated lead factors are listed for each of the Salem Unit 2 surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules.
6-12
Results of the FERRET evaluation of the Capsule X dosimetry are given in Table 6-9.
The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1.16 x 1019 n/cm2 (E > 1.0 MeV) with an associated uncertainty of+/- 8%.
Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa).
Summaries of the fit of the adjusted spectrum are provided in Table 6-10.
In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental r~action rates.
The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.
A summary of the measured and calculated neutron exposure of Capsule X is presented in Table. 6-12.
The agreement between calculation and measurement falls within+/- 11% for all fast neutron exposure parameters listed.
Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13.
Along with the current (6.20 EFPY) exposure derived from the Capsule X measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EFPY).
In the evaluation of the future exposure of the reactor pressure vessel th~ exposure I
rates averaged over the first six cycles of operation were employed.
In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Salem Unit 2 reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were also evaluated.
Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14.
In order to access RTNDT vs. fluence trend curves, dpa equivalent fast neutron fluence levels for the l/4T and 3/4T positions were defined by the relations:
r/> I (l/4T) dpa (l/4Tl
= rf> (Surface) {dpa (Surface)}
r/>' (3/4T) dpa (3/4Tl
= r/> (Surface) {dpa (Surface)}
6-11 i
FLUX W!~ES FLUX ~l~ES CA?SULE CENiER -
FLUX WIRES-3 49 2.63
~
r 2.48 1.60 L~
1'7 _' I I
0.27 ~ ~
THERMAL SHIELD r
- I Racius (cm) 213.85 213.01
~
212.68 212.18 211.68 211.41 211.18
-- 210.68 210.14 209.81
-- 208.97 Figure 6-1.
Plan View of a Reactor Vessel Surveillance Capsule
~
6-13
DESIGN BASIS CYCLE 1 O"I I.....
CYCLE 2
~
CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER
(,b(E > 1.0MeV)
(,b(E > O.lMev)
Iron Displacement Rate Cn/cm2-secl
[n/cm2-secl Cdpa/secl 4.0° 40.0° 4.0° 40.0° 4.0° 40.0° 2.82 x 1011 9.05 x 1010 8.15 x 1010 3.04 x 1011 4.58 x lo-11 1.55 x 10-10 2.08 x 1010 6.66 X lolO
- 6.01 X lolO 2.24 x 1011 3.37 x lo-11 1.14 x 10-10 2.35 x 1010 8.12 x 101° 6.79 x 1010 2.73 x 1010 3.81 x 10-11 1.39 x 10-10 2.19 X lolO 5.23 x 1010 6.33 x 101°
- 1. 76 X lolO 3.55 x lo-11
- 8. 94 x lo-11 1.89 X lOlO 4.98 X lolO 5.46 x 101° 1.67 X lOlO 3.06 x 10-11 a.52 x 10-11 2.13 x 1010 4.49 X iolO 6.16 x 1010
- 1. 51 x 1010 3.45 x 10-11 7.68 x 10-11 2.01 x 1010 4.56 X lolO 5.81 x 1010
- 1. 53 x 1011 3.26 x 10-11 7.80 x 10-11
DESIGN BASIS CYCLE 1
CYCLE 2
CYCLE 3
CYCLE 4
CYCLE 5
CYCLE 6
DESIGN BASIS CYCLE 1
CYCLE 2
CYCLE 3
CYCLE 4
CYCLE. 5 CYCLE 6
DESIGN BASIS CYCLE 1
CYCLE 2
CYCLE 3
CYCLE 4
CYCLE 5
CYCLE 6
TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD/BASE METAL INTERFACE
¢CE > l.OMeV) fn/cm2-sec]
0.0° 15.0° 30.0° 45.0° 8.43 x 10°9 1.36 x 1010 1.72 x 1010 2.68 x 1010 6.21 x 1009 9.98 x 1009 1.25 x 1010 1.93 x 101° 7.03 x 10°9 1.15 x lolO 1.48 x 1010 2.33 x 1010 6.48 x 1009 1.02 x 1010 1.01 x 101° 1.55 x 1010 5.62 x 10°9 8.82 x 1009 9.83 x 1009 1.47 x 1010 6.28 x -10°9 9.69 x 10°9 9.30 x 1009 1.34 x 1010 5.94 x 1009 9.08 x 10°9 9.44 x 10°9 1.36 x 1010 0CE > O.lMeV)
[n/cm2-sec]
0.0° 15.0° 30.0° 45.0° 2.11 x 1010 3.41 x 1010 4.34 x lolO 6.96 x 1010 1.55 x 1010 2.50 x 1010 3.15 x 1010 5.02 x 1010 1.16 x 101° 2.89 x 101° 3.73 x 1010 6.06 x 1010 1.62 x 101° 2.56 x 101° 2.10 x 1010 4.03 x 1010 1.41 x 1010 2.21 x lolO 2.48 x 1010 3.82 x 1010 1.57 x 1010 2.43 x 1010 2.34 x 101° 3.48 x 1010 1.49 x 101° 2.28 x 1010 2.38 x 1010 3.54 x 1010 Iron Atom Displacement Rate [dpa/sec]
0.0° 15.0° 30.0° 45.0° 1.37 x lo-11 2.19 x 10-11 2.73 x lo-11 4.26 x lo-11 1.01 x lo-11 1.61 x 10-11 1.99 x 10-11 3.07 x 10-11 1.15 x lo-11 1.85 x 10-11 2.35 x 10-11 3.70 x 10-11 1.06 x lo-11 1.64 x 10-11 1.70 x 10-11 2.46 x 10-11 9.16 x lo-12 1.42 x 10-11 1.56 x lo-11 2.34 x 10-11 1.02 x lo-11 1.56 x lo-11 1.48 x lo-11 2.13 x lo-11 9.68 x lo-1 2 1.46 x 10-11 l.5o x 1--11 2.16 x lo-11 6-15
TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius (cm) oo 220.21<1>
1.00 220.64 0.977 221. 66 0.884 222.99 0.758 224.31 0.641 225.63 0.537 226.95 0.448 228.28 0.372 229.60 0.309 230.92 0.255 232.25 0.211 233.57 0.174 234.89 0.143 236.22 0.117 237.54 0.0961 238.86 0.0783 240.19 0.0635 241. 51
- 0. 0511 242.11<2>
0.0483 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 15° 30° 1.00 1.00 0.978 0.979 0.887 0.889 0.762 0.765 0.644 0.648 0.540 0.545 0.451 0.455 0.373 0.379 0.310 0.315 0.257 0.261 0.. 212 0.216 0.175 0.178 0.144 0.147 0.118 0.121 0.0963 0.0989 0.0783 0.0807 0.0632 0.0656 0.0501 0.0519 0.0469 0.0487 6-16 45° 1.00
- 0. 977 0.885 0.756 0.637 0.534 0.443 0.367 0.303 0.250 0.206 0.169 0.138 0.113 0.0912 0.0736 0.0584 0.0454 0.0422
TABLE 6-4
- RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius
.(gnL oo 15° 30° 45° 220.21< 1>
1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 221.66 1.00 0.996 1.00 0.994 222.99 0.965 0.958 0.968 0.953 224.31 0.916 0.906 0.919 0.898 225.63 0.861 0.849 0.865 0.838 226.95 0.803 0.790 0.809
- 0. 777 228.28 0.746 0.732 0.752
- 0. 717 229.60 0.689 0.675 0.695 0.657 230.92 0.633 0.619 0.640 0.600 232.25 0.578 0.565 0.586 0.544 233.57 0.525 0.513 0.534 0.490 234.89 0.474 0.463 0.483 0.437 236.22 0.424 0.414 0.433 0.387 237.54 0.375 0.367 0.385 0.338 238.86 0.328 0.322 0.338 0.291 240.19 0.283 0.277 0.292 0.244 241. 51 0.239 0.232 0.245 0.196 242.17<2>
0.229 0.220 0.232 0.183 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-17
TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa)
WITHIN THE PRESSURE VESSEL WALL Radius limL oo 220.270) 1.00 220.64 0.983 221.66 0.913 222.99 0.818 224.31 0.728 225.63 0.647 226.95 0.574 228.28 0.510 229.60 0.453 230.92 0.402 232.25 0.356 233.57 0.315 234.89 0.277 236.22 0.243 237.54 0.212 238.86 0.182 240.19 0.155 241. 51 0.131 242.11<2>
0.125 NOTES:
- 1) Bas~ Metal Inner Radius
- 2) Base Metal Outer Radius 15° 30° 1.00 1.00 0.983 0.984 0.914 0.918 0.819 0.827
- o. 728.
0.739 0.646 0.659 0.573 0.587 0.507 0.523 0.450 0.466 0.399 0.414 0.353 0.368 0.312 0.327 0.275 0.289 0.241 0.254 0.210 0.222 0.181 0.192 0.154 0.164 0.128 0.137 0.122 0.130 6-18 45° 1.00 0.983 0.915 0.820 0.730 0.647 0.573 0.507 0.449 0.397 0.349 0.307 0.269 0.233 0.201 0.170 0.141 0.113 0.106
TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Weight
Response
Product Yield Material Interest Fraction Range Half-Life
(%)
Copper cu63(n,a)Co60 0.6917 E > 4.7 MeV 5.272 yrs Iron Fe54(n,p)Mn54 0.0582 E > 1.0 MeV 312.2 days Nickel Ni58(n,p)Co58 0.6830 E > 1.0 MeV 70.90 days O'I Uranium-238*
u238(n,f)Cs137 1.0 E > 0.4 MeV 30.12 yrs 5.94 I..... '° Neptunium-237*
Np237(n,f)Csl37 1.0 E > 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum*
co59(n,a)co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobalt-Aluminum co59(n,a)co60 0.0015 E > 0.015 MeV 5.272 yrs
- Denotes that monitor is cadmium shielded.
TABLE 6-7 MONTHLY THERMAL GENERATION DURING THE FIRST FIVE FUEL CYCLES OF THE SALEM UNIT 2 REACTOR THERMAL THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION GENERATION MONTH (MW-hr}
MONTH (MW-hr)
MONTH (MW-hr)
MONTH (MW-hr) 6/81 78098 2/84 4879 10/86 110662 6/89 2070988 7/81 737059 3/84 2059562 11/86 0
7/89 2475830 8/81 1906942 4/84 406510 12/86 326002 8/89 2384116 9/81 1252728 5/84 1927080 1/87 2346533 9/89 2429035 10/81 1260696 6/84 1345678 2/87 2288412 10/89 983484 11/81 1805837 7/84 399636 3/87 1863264 11/89 1997556 12/81 2041486 8/84 1823350 4/87 1431199 12/89 2524596 1/82 2238970 9/84 2013103 5/87 2139360 1/90 1874868 2/82 2182846 10/84 276166 6/87 1766332.. 2/90 2262746 3/82 2481187 11/84 0
7/87 1571339 3/90 2418187 4/82 2174431 12/84 0
8/87 788479 4/90 0
5/82 2508634 1/85 0
9/87 2434121 5/90 0
6/82 2448542 2/85 0 10/87 1882250 6/90 125880 7/82 1887012 3/85 0 11/87 0
7/90 1798 8/82 2066801 4/85 407818 12/87 1135051 8/90 908050 9/82 1640998 5/85 1867901 1/88 2490826 9/90 1990867 10/82 1989686 6/85 2210338 2/88 2298118 10/90 2470128 11/82 1916650 7/85 1485823 3/88 2532218 11/90 2403902 12/82 1872773 8/85 2128176 4/88 2249256 12/90 2485303 1/83 1109383 9/85 2346017 5/88 2323214 1/91 2526556 2/83 0 10/85 2306441 6/88 2170375 2/91 2260361 3/83 0 11/85 2455831 7/88 2388288 3/91 2428037 4/83 0 12/85 810317 8/88 2012440 4/91 2456870 5/83 0
1/86 1108313 9/88 0
5/91 1585037 6/83 0
2/86 2066002 10/88 0
6/91 2410973 7/83 53664 3/86 2477746 11/88 10756 7/91 2540352 8/83 816137 4/86 2257325 12/88 539083 8/91 2529274 9/83 447922 5/86 2324078 1/89 1512676 9/91 2372525 10/83 527417 6/86 2429813 2/89 1708567 10/91 2318374 11/83 0
7/86 2147440 3/89 2150445 11/91 689894 12/83 0
8/86 1397742 4/89 1978951 1/84 0
9/86 728568 5/89 2170394 6-20
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Capsule Center Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis/sec-gm)
(dis/sec-gm)
(RPS/NUCLEUS)
Cu-63 (n,a) Co-60 Top-Middle 1.10 x 105
- 2. 57 x 105 Middle 1.13 x 105 2.62 x 105 Bottom-Middle 1.15 x 105 2.67 x 105 Average 1.13 x 105 2.62 x 105 3.85 x 10-17 Fe-54(n,p) Mn-54 Top 1.24 x 106
- 2. 37 x 106 Top-Middle 1.15 x 106 2.20 x 106 Middle 1.14 x 106 2.18 x 106 Bottom-Middle 1.18 x 106
- 2. 26 x 106 Bottom 1.13 x 106 2.16 x 106 Average 1.17 x 106
- 2. 23 x 106 3.76 x 10-15 Ni-58 (n,p) Co-58 Top 7.90 x 106 3.10 x 107 Middle
- 7. 97 x 106 3.13 x 107 Bottom
- 7. 75 x 106 3.04 x 107 Average 7.87 x 106
Middle
- 3. 90 x 105 3.09 x 106 2.06 x 10-14 6-21
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd Monitor and Axial Location Np-237{n,f) Cs-137 (Cd)
Middle Co-59 (n,a) Co-60 Top Bottom Average Co-59 (n,a) Co-60 (Cd)
Top Bottom Average Measured Activity
{dis/sec-gm) 6 3.38 x 10 no data no data no data no data
- 6. 71 x 106
- 6. 71 x 106 6-22 Saturated Activity
{dis/sec-gm)
. 2.68 x 107
- 1. 56 x 107
- 1. 56 x 107 Capsule Center Reaction Rate CRPS/NUCLEUSl 1.62 x 10-13 1.18 x 10-12
TABLE 6-9
SUMMARY
OF NEUTRON DOSIMETRY RESULTS TIME AVERAGED EXPOSURE RATES
1.0 MeV} {n/cm2-sec} 5.93 x 1010 +/- 8%
0.1 MeV} {n/cm2-sec} 2.02 x 1011 +/- 15% dpa/sec 9.92 x 10-11 +/- 10% INTEGRATED CAPSULE EXPOSURE ~ (E > 1.-0 MeV) {n/cm2} 1.16 x 1019 +/- 8% ~ (E > 0.1 MeV} {n/cm2} 3.95 x 1019 +/- 15% dpa
- 1. 94 x 10-2
+/- 10% NOTE: Total Irradiation Time = 6.20 EFPY 6-23 TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER Adjusted Reaction Measured Calculation Cu-63 (n,a) Co-60 3.85x10-17 3.85x10-l7 Fe-54 (n,p) Mn-54
- 3. 76x10-15 3.80x10-15 Ni-58 (n,p) Co-58 5.19xI0-15
I.62x10-13 l.64xio-13 Co-59 (n,8) Co-60 (Cd) l.18xI0-12 l.18xI0-12 6-24 C/M 1.00 1.01 1.00 0.96 1.01 1.00 TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy Adjus2ed Flux Energy Adjus2ed Flux Group (Mev) (n/cm -sec) Group (Mev) (n/cm -sec) 1 I. 73xl01 3.23xl06 28 9.12x10-3 8.19xl09 2 l.49xl01 7.88xl06 29 5.53xI0-3 l.03xlo10 3 l.35xl01
- 3. 63xl07 30 3.36xI0-3 3.18xl09 4
l.16xl01 9.2lxl07 31 2.84xlo-3 3.03xl09 5 l.00xl01 2.23xl08 32 2.40xI0-3
- 2. 93xl09 6
- 8. 6lx10° 4.03xl08 33
- 2. 04xI0-3 8.45xl09 7
7.4lxI0°
- 9. 77xl08 34 I. 23xI0-3 8.19xl09 8
6.07xI0° l.42xl09 35 7.49xI0-4 7.87xl09 ,9
- 4. 97xI0°
- 2. 95xl09 36
- 4. 54xI0-4 7.63xl09 10
~.68xl0°
- 3. 76xl09 37
- 2. 75xI0-4 8.04xl09 11 2.87xI0°
- 7. 56xl09 38 l.67xl0-4 8.80xl09 12
- 2. 23xI0°
- 9. 55xl09 39 l.OlxI0-4 8.67xl09 13 I. 74xI0° I. 25x1010 40 6.14xl0-5 8.6lxl09 14 l.35xI0°
- 1. 25x1010 41
- 3. 73xI0-5 8.44xl09 15
- l. llxlOO 2.13xlOlO 42
- 2. 26xI0-5 8.23xl09 16 8.2lxI0-1
- 2. 26x10lO 43 l.37xI0-5 8.0lxl09 17 6.39x10-l
- 2. 2lx10lO 44 8.32x10-6 7.68xl09 18
- 4. 98xl0- 1
- 1. 57xlOlO 45 5.04xI0-6 7.17x109 19 3.88x10-1 2.06xlOlO 46 3.06x10-6
- 6. 77x109 20
- 3. 02xl0-l
- 2. 26x10lO 47 l.86xI0-6 6.29x109 21 l.83x10-l 2.14x10lO 48 l.13xI0-6 4.9lx109 22
- l. llxI0-1 l.68xI010 49 6.83xI0-7 5.07x109 23 6.74xI0-2
- 1. 2lxl010 50 4.14xI0-7 7.36x109 24 4.-09x10-2*
7.29xl09 51
- 2. 5lxI0-7 6.69x109 25 2.55x10-2
- 8. 72xl09 52 l.52xI0-7 6.10xl09 26
- 1. 99x10-2
- 4. 73xl09 53
- 9. 24xI0-8 l.4lx1010 27
- 1. 5ox10-2 6.39xl09 NOTE:
Tabulated energy levels represent the upper energy of each group. 6-25 ~ ( E > I. 0 Me V) ~(E > 0.1 MeV) dpa {n/cm2} {n/cm2} TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE X Calculated Measured 1.05 x 1019 1.16 x 1019
- 3. 53 x 1019 3.95 x 1019 1.80 x 10-2 I. 94 x 10-2 6-26 C/M 0.91 0.89 0.93
TABLE 6-14 PROJECTED NEUTRON EXPOSURE VALUES oo 15° 30° 45° oo 15° 30° 45° 16 EFPY NEUTRON FLUENCE {E > 1.0 MeV) SLOPE (n/cm2) Surf ace lLU ill_! 3.41 x 1018 1.81 x 1018 3.72 x 1017 5.38 x 1018 2.86 x 1018 5.91 x 1017 5.96 x 1018 3.20 x 1018 6.74 x 1017 8.82 x 1018 4.64 x 1018 9.26 x 1017 32 EFPY NEUTRON FLUENCE {E > 1.0 MeV) SLOPE (n/cm2) Surf ace ill._I ill_! 6.83 x 1018 3.61 x 1018 7.44 x 1017 1.08 x 1019 5.72 x 1018 1.18 x 1018 1.19 x 1019 6.41 x 1018
- 1. 35 x 1018
- 1. 76 x 1019 9.27 x 1018 1.85 x 1018 (a) Maximum point on the pressure vessel 6-28 Surf ace 3.41 x 1018 5.38 x 1018 5.96 x 1018 8.82 x 1018 Surface 6.83 x 1018 1.08 x 1019 1.19 x 1019
- 1. 76 x 1019 dpa SLOPE (equivalent n/cm2)
.uu llU 2.19 x 1018 7.92 x 1017 3.44 x 1018 1.24 x 1018 3.89 x 1018 1.44 x 1018 5.64 x 1018
- 1. 95 x 1018 dpa SLOPE (equivalent n/cm2) ill._I
.ill....I 4.37 x 1018
- i. 58 x 1018 6.87 x 1018 2.47 x 1018
- 7. 78 x 1018 2.89 x 1018 1.13 x 1019 3.90 x 1018
TABLE 6-15 UPDATED LEAD FACTORS FOR SALEM UNIT 2 SURVEILLANCE CAPSULES Capsule Lead Factor T Withdrawn u Withdrawn x 3,39(b) y 3,39(b) s 1.31 (b) v
- 1. 31 (b) w i.31(b) z 1.31 (b)
(a) Plant specific evaluation based on end of Cycle 6 calculated fluence. 6-29 SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Salem Unit 2 reactor vessel: Capsule T u x y s v w z Location (deg.) 40 140 220 320 4 176 184 356 Capsule Lead Factor withdrawn withdrawn 3.39 3.39 1.31 1.31 1.31 1.31 Removal Time (a) 1.16 (Removed) 2.70 (Removed) 6.20 (Removed) 11 EOL Standby Standby Standby (a) Effective Full Power Years (EFPY) from plant startup. Estimated Fluence (n/cm2) 2.76 x 1018 (Actual) 5.07 x 1018 (Actual) 1.16 x 1019 (Actual) 2.05 x 1019 2.30 x 1019 7-1
- 1.
- 2.
SECTION 8. 0 REFERENCES J. H. Phillips, et al., "Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveil 1 ance Program," WCAP-8824, January 1977. Code of Federal Regulations, 10CFR50, Appendix G, "Fracture Toughness Requirements", and Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington,. D.C.
- 3.
Regulatory Guide 1.99, Proposed Revision 2, "Radiation Damage to Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May 1988.
- 4.
Section III of the ASME Boiler and Pressure Vessel Code, Appendix G, "Protection Against Nonductile Failure."
- 5.
ASTM E208, "Standard Test Method for Conducting Drop-Weight Test to Determine Ni 1-Duct il ity Trans it ion Temperature of Ferri tic Steels."
- 6.
ASTM El85-82, "Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)."
- 7.
ASTM E23-88, "Standard Test Methods for Notched Bar Impact Testing of Metallic Materials."
- 8.
ASTM A370-89, "Standard Test Methods and Definitions for Mechanical Testing of Steel Produ_cts."
- 9.
ASJM E8-89b, "Standard Test Methods of Tension Testing of Metallic Materials."
- 10.
ASTM E21-79(1988), "Standard Practice for Elevated Temperature Tension Tests of Metallic Materials." 8-1
- 11.
ASTM E83-85, "Standard Practice for Verification and Classification of Extensometers."
- 12.
R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, "Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970.
- 13.
"ORNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".
- 14.
L. T. Mayhue, et. al., "The Nuclear Design of the Salem Unit 2 Nuclear Power Plant - Cycle l", WCAP-9374, August 1978. (Proprietary)
- 15. J. E. Pritchett, et. al., "The Nuclear Design of the Salem Unit 2 Nuclear Power Plant - Cycle 2, Rev l", WCAP-10248, Rev l, May 1983. (Proprietary)"
- 16.
M. M. Weber, et. al., "The Nuclear Design of the Salem Unit 2 Nuclear Power Plant - Cycle 3", WCAP-10790, February 1985. (Proprietary)
- 17.
M. M. Weber, et. al., "The Nuclear Design of the Salem Unit 2 Nuclear Power Pl ant - Cycle 4 11 , WCAP-11218, October 1986. (Proprietary)
- 18.
D. J. Augustine, et. al., "The Nuclear Design of the Salem Unit 2 Nuclear Power Plant - Cycle 5", WCAP-11920, October 1988. (Proprietary)
- 19.
D. J. Augustine, et. al., "The Nuclear Design of the Salem Unit 2 Nuclear Power Plant - Cycle 6", WCAP-12534, April 1990. (Proprietary)
- 20.
ASTM Designation E482-89, "Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991. 8-2
- 21.
- 22.
ASTM Designation E560-84, "Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991. ASTM Designation E693-79, "Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 23.
ASTM Designation E706-87, "Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 24.
ASTM Designation E853-87, "Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 25.
ASTM Designation E261-90, "Standard Method for Determining Neutron Flux, Fluence, ~nd Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 26.
ASTM Designation E262-86, "Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 27.
ASTM Designation E263-88, "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 28.
ASTM Designation E264-87, "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991. 8-3
- 29.
ASTM Designation E481-86, "Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 30.
ASTM Designation E523-87, "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 31.
ASTM Designation E704-90, "Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 32.
ASTM Designation E705-90, "Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 33.
ASTM Designation ElOOS-84, "Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Se~tion 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
- 34.
F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
- 35.
W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 36.
EPRl-NP-2188, "Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al., 1981.
- 37.
NUREG-0020, "Licensed Operating Reactors Status Summary Report" - 6/81 through 11/91. 8-4 APPENDIX A Load-Time Records for Charpy Specimen Tests A-0 I... a <{ 0 ..J PGY *GENERAL YIELD LOAD PM== MAXIMUM LOAD I i I I I I I TIM[ Vigurc A-1. Idealized load-time record I P1
- rnACTuAE LOAD I
I /PA= ARREST LOAD 0 0., ~- ~ ~ A ~- ~ ~- 0 .b 0 0., 0 0.. ~ a.. 0 0 ~.. 0 ~ 0 0 0 .b Figure A-2. D'MIT\\.P I ~/\\Al\\ 1.0 D'MIT\\.P 1.0 SOL..L3a I I 2.0 Tl!£ SOL.Ll!ll 2.0 Tit£ 3.0 ( llSEC ) 3.0 ( llSEC ) I 4.0 4.0
- s.o
- s.o Load-time records for Specimens JL32 and JL28 A-2
0.. 0 0.. 0 0 0.n 0.n 1.0 1.0
- 2. 0 TUE 2.0 3.0
( itm:) 3.0 TllE < itm: > 4.0 s.o 4.0 s.o Figure A-3. Load-time records for Specimens JL27 a.nd JL25 A-3
- .,..-IMll,;..._1\\.P~~-.~~~-sa._:,.~..l.3~0'--~~-..-~~~..l.3~0::__....,...~~~~~
n 0 0. 0 0 0 *' 2.0 3.0 4.0 s.o me c llS[C > ~...,...""";;.;;.;.~1\\.P:;.-~-...,...,r-r-~.,..;;;.;:,.::::....~~~..--~-==---.~~~~~ n 0 0. 0 0 a *' 2.0 3.0 4.0 s.o me < llS[C > Figure A-4. Loa.d-time records for Specimens JL30 and JL29 .~ 0.
- 0 I. a 2.0 3.0 4.0 TIP£
< l'ISEC > 2.0 3.a s.o Tll£ ( "5EC ) Figure A-5. Load-time records for Specimens.1131 and.1126 A-5 L D IMIATl.P Siii..JT48 JT.JT48 T 0.. ~- A ~- ~.. ~ ~- N ~- I .b 1.0 z.o 3.0 4.0 s.o TIIE ( llSEC ) 0 IMIATIP W..JT44 .JT44 0 : I.. 0.; ~.. 0 ~
- b 1.0 2.0 3.0 4.0 s.o TIIE C llSEC >
Figure A-6. Load-time records for Specimens JT48 and JT44 A-8 i. 5 ~ 0 amoTl.P SAL Jf'3 JT43 0 0.. 0
- i N
0-1-~--=..?:.?~""""""~~~-r-~~-.-~~--l- .u 1.0 2.0 3.0
- o s.o TU£
< P1SEC >
- -.-D'mA~-ll.P~~~.--~~~-siw....,...JT311~~~~-,-~~~~~~~~~-,
0 0 .u 1.0 2.0 TllE 3.0 ( llSEC ) 4.0 s.o Figure A-7. Load-time records for Specimens JT43 and JT38 A-'1 0 0 n 0 ~ ~ 'ii 0 0 5... 0 ~ 0 N 0 0 0 -~ 0 0 n 0 0.. I 'ii 0 0 ~... 0 ~ 0.. 0 2 °""'T\\P 1.0 °""'T\\P 1.0 Siii.. JT42 2.0 3.0 TUE < G:c > Siii.. JT.. 2.0 TllE 3.0 ( llS[l:) JT<<! 4.0
- s. 0 JT46 4.0 Figure A-8.
Load-time records for Specimens JT42 and JT46 A-8
- -.-°"""~Tl.P~~~~~~~-SSIL-r-ft-40~~~.....,...~~~ft-40~......--~...:_~~~
n 0.. 0 0 0 .n a n 0 0 a
- n
- 1. 0 1.0
- 2. 0 3.0 Til'E
< PISEC > S<IL ft41 ft41 2.0 3.0 ~.o Tl!£ ( llSCC ) Figure A-9. Load-time records for Specimens JT40 and JT41 A-G D'l'ttATlP SAL JT37 "'""T'""~~~~-r~~~~~~~~~T""""~~~~T""""~~~~""T'"" n ~.. 0..
- b I..
~ ~ 1.0 . IJ'llllTIP 0 " 0.. 0 0.. 0 - 1.0 2.0 TII£ SOI.. JT39 2.0 3 ** ( !<SEC ) 3.0 TII£ < !<SEC > s.o s.o Figure A-10. Load-time records for Specimens JT37 and JT39 A-10 0.. 0 0 0 .b 1.0 1.0 2.0 3.0 Tll£ ( llSEC ) 2.0 3.0 Tl!£ ( llSEC ) JT47 4.0 s.o 4.0 s.o Figure A-11. Load-time records for Specimens JT47 and JT45 A-11 J 3 ~ ~,--D'mA~TIJ'~~~-,-,~~~-SOL---,,r.M<l5~~~~-.~~~..IM--'S~,-~~~~~,.... n ~- 0 .b 1 ** 2.0 3.o 4.0 s.o TUE C PISEC > 0 D'mATI.P SOL.Kn ..Kn 0 I I I I ~- ~- ~-.. ~- 0 I I I I .b 1.0 2.0 3.0 4.0 s.o 111£ (~) Figure A-12. Load-time records for Specimens JW45 a.nd JW39 A-12 0 """'TIJ' 0 ~ i. 0 0 ~... 0 ~ 0.. 0 1.0 1.0 ........ o 2.~ TIP£ 2.0 3.0 < IGEC > 3.0 Tt.W: ( l'tSEC ) .... 0 4.D s.o
- o s.o Figure A-13.
Load-time records for Specimens JW40 and JW44 A-13 a a.. a a ~ 1.0 1.0 2.0 3.0 TII£ < llSEC l 2.0 TII£ 3.0 ( llSEC ) 4.0 s.o 4.0 s.o Figure A-14. Load-time records for Specimens JW47 and JW46 A-14 0 0 0 ,b 0 n 0 0 0 .b
- 0 1.0 2.0 TlllC 2.0 3.0
( llSEC. ) 3.0 TillE ( llSEC. )
- 0
- 0
'* 0 Figure A-15. Load-time records for Specimens JW38 and JW37 A-15 ll'IM'lll' 0.. .b
- l. 0 1.0 2.0 3.0 Tll'£ C tl$EC >
2.0 TllE 3.0 < llS[C ) 4.0 s.o 4.0 s.o Figure A-17. Load-time records for Specimens JW41 and JW48 A-17
- -.-O'l'llA'---1'1.1'~~-,-~~~-SAL--.--""2~~~~..-~~~---,~~~~--,
n 0 0.. 0 1.0 2.0 3.0 4.0 s.o Tl1£ ( llSEC )
- ""T""O'l'llA~-1\\P~~-,-~~~-SAL---y-..uo:J~~~~T"""~~~--.~~~~.,
n 1.0 2.0 3.a Tll£ ( llSEC ) 4.0 s.o Figure*A-16. Load-time records for Specimens JW42 and JW43 A-18 'II ~- - ~- ~ ~ ~ ~- 0 0 0.. 0 0.. 0 0... 0 0.. 0 *' IMllTl.P 1.0 2.0 3.0 4.0 TUE <~> 1.0 2.0 3.0 4.0 5.0 Tll£ ( llSEI: ) Figure A-18. Load-time records for Specimens JB38 and JH44 A-18 'It II 'i ~ !I ~ I 0 im-n.F $QL.- 0 I I I I n ~- J ~- ~- N J\\ 0 I I .b 1.0 2.0 3.0 4.0
5.0 TlllE
( llSEC )
~..... ll'IM~-n.p~~---.....-~~~--""T'".IN~~~~-.-~~~~.....-~~~~..,...
n
~ :;
- b 1.0
2.0 TlllE
3.0
( llSEC )
4.0 5.0 Figure A-19.
Load-time records for Specimens JH48 and JH42 A-19
I
- b
~.
~ ::
1.0 1.1 e.o 3.0
- o s.o Tll'I:
< !!SEC >
... u 2.1 3.0
- o s.o Tiii:
< llSIEC >
Figure A-20.
Load-time records for Specimens JH45 and JH41 A-20
I i
~.
~..
1.0 Tll'I
< lllEC >
3.Q
( lllEC )
4.0
,.o Figure A-21.
Load-time records for Specimens JH46 and JH40 A-31
~
~
CD I
i.li c.... >**-
UIAD
- ,,o 10.0 eo.o 30.0 50.0 I
I l:"4 0 g_
I ct' 0
0 El II
'1 CD
~
0
.. I"
.. I" 0
~ 0
~
M 0
~
'1 I
r:i.
f4 R
R 0 '1 w I" w I" 0
0 r.o
~
II 0..... s f'
f' Cl I:'
0 (I
-~
co I"
~
0 0
r:i.
~
~
0 n 1.0 1.0 2.0 3.0 TIP£
< llSEC >
2.0 TllE 3.0
( llSEC >
4.0 5.0 4.0 s.o Figure A-23.
Load-time records for Specimens JH36 a.nd JH43 A-23
APPENDIX B Heatup and Cooldown Limit Curves for Normal Operation B-0
TABLE OF CONTENTS Section Title Page 1
INTRODUCTION B-3 2
FRACTURE TOUGHNESS PROPERTIES B-3 3
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS B-4 4
HEATUP AND COOLDOWN LIMIT CURVES B-7 5
ADJUSTED REFERENCE TEMPERATURE B-8 6
REFERENCES B-16 B-1
LIST OF ILLUSTRATIONS Figure Title Page 1
2 3
4 1
Salem Unit 2 Reactor Coolant System Heatup Limitations B-14 (Heat up rate up to 60°F/hr and l00°F/hr) Applicable 2
for the First 15 EFPY (Without Margins For Instrumentation Errors)
Salem Unit 2 Reactor Coolant System Cooldown (Cooldown Rates up to l00°F/hr) Limitations Applicable for the First 15 EFPY (Without Margins For Instrumentation Errors)
LIST OF TABLES Title Salem Unit 2 Reactor Vessel Toughness Table (Un irradiated)
Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location Calculation of Adjusted Reference Temperatures for Limiting Salem Unit 2 Reactor Vessel Material - Lower Shell Longitudinal Weld Calculation of Adjusted Reference Temperatures for Limiting Salem Unit 2 Reactor Vessel Material -
Intermediate Shell Longitudinal Weld B-2 B-15 B-10 B-11 B-12 B-13
- 1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel.
The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced ARTNDT*
RTNDT is designated as the higher of either 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-mil lateral expansion (normal to the major working direction) minus 60°F..
RTNDT increases as the material is exposed to fast-neutron radiation.
Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT*
The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)[!]. Regulatory Guide 1.99, Revision 2 is used for the calculation of RTNDT values at l/4T and 3/4T locations (T is the thickness of the vessel at the beltline region).
- 2.
FRACTURE TOUGHNESS PROPERTIES The unirradiated RTNDT values for the beltline region materials in the Salem Unit 2 reactor vessel were established using the guidance provided in NUREG-0800, Branch Technical Position, MTEB 5-2[2], and subarticale NB-2331 of the ASME Boiler and Pressure Vessel Code, Section IIr[3]. The copper and nickel content for each material was determined by Combustion Engineering during the fabrication of the Salem Unit 2 reactor vessel. These pre-irradiation fracture-toughness properties of the Salem Unit 2 reactor vessel are presented in Table 1.
B-3
- 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, KI, for the combined thermal and pressure s_tresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time.
KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code[3]. The KIR curve is given by the following equation:
KIR = 26.78 + 1.223 exp [0.0145 {T~RTNDT + 160)]
{l) where KIR = referen~e 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[3] as follows:
where KIM = stress intensity factor caused by membrane {pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function. of temperature relative to the RTNDT of the material C
= 2;0 for Level A and Level B service limits C
1.5 for hydrostatic and leak test conditi6ns during which the reactor core is not critical B-4
{2)
At any time during the heatup or cooldown transient, KIR is determined by the meta 1 tempel"ature
- at the tip of the postulated fl aw, the appropriate va 1 ue for RTNDT' and the reference fracture toughness curve.
The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT' for the reference flaw are computed.
From equation 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 reference 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 the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperatur~ at the tip of the assumed flaw.
During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID.
This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of 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 KIR exceeds KIT' the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various B-5
intervals along a cooldown ramp.
The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three se~arate calculations are required to determine the limit curves for finite heatup rates.
As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4 T crack during heatup is lower than the KIR for the 1/4 T crack during steady-state conditions at the same coolant temperature.
During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1R*s do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the 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 increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-pojnt comparison of the steady-state and finite heatup rate data.
At any given temperature, the B-6
- I I
allowable pressure is taken to be the lesser of the three values taken from the curves undeY consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the 1983 Amendment to 10CFR5o[4J has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Salem Unit 2).
Table 1 indicates that the initial RTNDT of 28°F occurs in the vessel flange of Salem Unit 2, so the minimum allowable temperature of this region is 148°F.
These limits are shown in Figures 1 and 2 whenever applicable.
- 4.
HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary reactor pressure vessel have been calculated using the methods discussed in Section 3.
Figure 1 contains the heatup curve for 60°F/hr.
Figure 2 contains the cooldown curves up to l00°F/hr.
Both Figures 1 and 2 are applicable for the first 15 EFPY of operation and include no margins for possible instrumentation errors.
Allowable combinations of temperature and pressure for specific temperature change rates are b~low and to the right of the limit lines shown in Figures 1 and 2.
This is in addition to other criteria which must be met before the reactor is made critical.
The leak limit curve shown in Figure 1 represents minimum temperature requirements at the leak test pressure specified by applicable codes[2,3J.
The leak test limit curve was determined by methods of References 2 and 4.
B-7
The criticality limit curve shown in Figure 1, specifies pressure-temperature limits for eore operation to provide additional margin during actual power production as specified in Reference 4.
The pressure-temperature limits for core operation (ex~ept for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum pressure-temperature curve for heatup and cooldown calculated as described in Section 3.
The maximum temperature for the inservice hydrostatic test for the Salem Unit 2 reactor vessel is 284°F.
A vertical line at 284°F on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 1 and 2 define limits for ensuring prevention of nonductile failure for the Salem Unit 2 reactor vessel.
- 5.
ADJUSTED REFERENCE TEMPERATURE From Regulatory Gujde 1.99 Rev. 2 [1] the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
ART = Initial RTNDT + ARTNDT + Margin (3)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB2331 of Section III of the ASME Boiler and Pressure Vessel Code.
If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
ARTNDT = [CF]f(0.28-0.10 log f)
(4)
B-8.
~I
To calculate ARTNDT at any depth (e.g., at l/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
f
- f (e-.24x)
(depth X) -
surface
{5) where x (in inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then put into equation (4) to calculate ARTNDT at the specific depth.
CF (°F) is the chemistry factor, obtained from Reference 1. All materials in the beltline region of Salem Unit 2 were considered for the limiting material.
RTNDT at l/4T and 3/4T are summarized in Table 2.
From Table 2, it can be seen that *the limiting material is lower shell for heatup and cooldown curves applicable up to 15 EFPY.
Sample calculations for the RTNDT for 15 EFPY are shown in Tables 3 and 4.
B-9
TABLE 1 SALEM UNIT 2 REACTOR VESSEL TOUGHNESS TABLE (Unirradiated) cu NI I-RTNDT (a)
Material Description
(%)
(%)
( o F)
Closure Head Flange, 84702-1 28 (b)
Vessel Flange, 85001 12 (b)
Intermediate Shell, 84712-1 0.13 0.56 0
Intermediate Shell, 84712-2 0.14 0.60 12 Intermediate Shell, 84712-3 0.11 0.57 10 Lower Shell, 84713-1 0.12 0.60 8
Lower Shell, 84713-2 0.12 0.57 8
Lower Shell, 84713-3 0.12 0.58 10 Intermed. Shell Longitudinal 0.23 0.73
-40 Welds, 2-442A, B and C Lower Shell Longitudinal 0.20 0.86
-56 (c)
Welds, 3-442A, B and C Circumferential Weld, 9-442 0.18 0.20*
-56 (c)
- a.
The initial RTNDT (I) values for the plates and welds are measured values based on transverse data.
- b.
To be used for considering flange requirements for heatup/cooldown curves[4J.
- c. Generic mean value per 10CFR50.61[6J.
Estimated upper limit value for wire heat type B-4.
B-10
TABLE 2
SUMMARY
-OF ADJUSTED REFERENCE TEMPERATURE (ART) AT 1/4T and 3/4T LOCATION Component Intermediate Shell Pl ate, 84712-1 Intermediate Shell Pl ate, 84712-2 Intermediate Shell Plate, 84712-3 Lower Shell Pl ate, 84713-1 Lower Shell Pl ate, 84713-2 Lower Shell Plate, 84713-3 Intermed. Shell Longitudinal Welds Lower Shell Longitudinal Welds Circumferential Weld 15 EFPY RTNDT at 1/4T (on 3/4T
( 0 Fl 106 82 126 (107) 100 (81) 103 84 109 87 108 86 117 93 149 102 151 101 85 60 RTNDT numbers within ( ) are based on chemistry factor calculated using capsule data.
B-11
/
/
j
TABLE 3 C;ALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING SALEM UNIT 2 REACTOR VESSEL MATERIAL - LOWER SHELL LONGITUDINAL WELD Parameter Chemistry Factor, CF (°F)
Fluence, f (1019 n/cm2)(a)
Fluence Factor, ff Regulatory Guide 1.99 - Revision 2 1/4 T 203
.3338
.698 15 EFPY 3/4 T 203
.1186
.452 ARTNDT = CF x ff (°F)
Initial RTNDT* I (°F)
Margin, M ( ° F) ( b) 141
-56 65.5 92
-56 65.5 Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 151
Fluence, f, is based upon f surf (1019 n/cm2, E>l Mev) = 0.56 at 15 EFPY.
The Salem Unit 2 reactor vessel wall thickness is 8.625 inches at the beltline region.
(b)
Margin is calculated as, M = 2 [ 0 12 + oA2]0.5.
The standard deviation for the initial RTNDT margin term, 01, is assumed to be 0°F since the initial RTNDT is a measured value.
The standard deviation for ARTNDT term, oA, is 28°F for the weld, except that aA need not exceed 0.5 times the mean value of ARTNDT*
oA, is 14°F for the weld (cut. in half) when surveillance data is used.
Limiting value used in development of heatup and cooldown limit curves:
B-12
L__
TABLE 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING SALEM UNIT 2 REACTOR VESSEL MATERIAL - INTERMEDIATE SHELL LONGITUDINAL WELD Parameter Chemistry Factor, CF (°F)
Fluence, f (1019 n/cm2)(a)
Fluence Factor, ff Regulatory Guide 1.99 -
Revisi~n 2 1/4 T 191. l
.3338
.698 15 EFPY 3/4 T 191.1
.1186
.452' ARTNDT = CF x ff (°F)
Initial RTNDT, I (°F)
Margin, M (°F) (b) 133
-40 56 86
-40 56 Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 149 102
Fluence, f, is based upon fsurf (1019 n/cm2, E>l Mev) = 0.56 at 15 EFPY.
The Salem Unit 2 reactor vessel wall thickness is 8.625 inches at the beltline region.
(b)
Margin is calculated as, M = 2 [ aI2 + a82]0.5. The standard deviation for the initial RTNDT margin term, aI, is assumed to be 0°F since the initial RTNDT is a measured value.
The standard deviation for ARTNDT term, a8, is 28°F for the weld, except that a8 need not exceed 0.5 times the mean value of* ARTNDT*
a8, is 14°F for the weld (cut in half) when surveillance data is used.
Limiting value used in development of heatup and cooldown limit curves.
B-13
,'I MATERIAL PROPERTY BASIS LIMITING ART AFTER 15 EFPY:
l/4T, 151°F 3/4T, 102°F 2500 l':A I I~ ~ 'L I I I I I I I I I
I I I I 11 I I I I I I I
- LEAK TEST LIMIT 2250 2000 1750 1500
. UNACCEPTABLE
(/)
a..
- OPERATION I/
Lo.I a::
1250 I/
I
(/)
(/)
Lo.I I/
a::
a..
I/
1000 0
Lo.I I-ct u -
1-- 1--
I/
1-- 1--: HEATUP RATE UP
- to60 °F/HR I/
750 0
I/
z -
~
500 250 50 100 150 200 250 I
I I
I I
I I
I I
I I
I I
I I
I
. ACCEPTABLE :
I/
I
- OPERATION I/
I CRITICALITY LIMIT -
BASED ON INSERVICE - -
HYDROSTATIC TEST
-+-
TEMPERATURE (284 OF)
-+-
-+-
FOR THE SERVICE PERIOD-1----
UPTO 15EFPY I I I I I I I I I I I I I
I I I I I I I I I I I I I
300 350 400 450 500 INDICATED TEMPERATURE (DEG.F)
Figure 1.
Salem Unit 2 Reactor Coolant System Heatup Limitations (Heat up rate up to 60°F/hr) Applicable for the First 15 EFPY (Without Margins For Instrumentation Errors)
B-14
MATERIAL PROPERTY BASIS LIMITING ART AFTER 15 EFPY:
l/4T, 151°F 3/4T, 102°F
/,
2250H---;-t-++-t-++-t-++-";-++-+-+-+-+-+-t-7+-t-++-.-++-r-+~1-+-+-1-+-+-1--+-+-+-I-+-+--,--+-+-+-!-
1 7 5 0 l-+-----1-+-+-+......-+-
i UNACCEPTABLE OPERATION I
I
! I I
I i
I I
1500r+-:-t-t--+-l-++-il--++-+
1-+-+-+-+-+-+1-+-~-+-1-+-+--1il--t-/+-i-+-~-+-+-+-'-+-+-+-+-+-+-1-+-'-l-+-l--'!--'
I
~ 12501-+-_._!-+-+-11-++-if-+-+-+-+-+-+-+-~',+-+-+-+-+-+-~/+-+-+-+-+-+-+-+--l I
I ACCEPTABLE OPERATION Vl Vl w
~
10001-+-~
1 -+-+-+--+-+-+--+-+-+-+-1-+--+--i-+-+-+-+-+.~,,.,_.1-++-+1-+-+-+-+-+-+-+-~~~~~~,..-;...~!--1-~~
0 w
<(
(.)
0 z i
I
'./,
I I
7 5 o COOLDOWN RATES +-+-+-A'"""'/'.~~+--
1
,_;: ++-+-+-++
1
-+-+-+-!--1-+-+-+-+-+-+-1-<-+-+-+-+-,,,_,+.-1-~
°F/HR VVY//
I i
5 0 0 -+-+-+-
2 5 0 l-+-4 Q
/1 /
I I
,,-++-1--~~,j..4~~/~/~~A+-+-+-1i-++-l-+-+l-l-+-+-l-+-+-l-:--+-+-i-+-~--l-+-+--l-+-+--l-+-+~
20..,,,..~v=:,j...~v~~~~/:+-il--l-+-+-+-+-+-+-+-+-l-+-+-l-+-+-l-!-l--+-!-l-~--l-+-+--l--'-!--l-+.-+~
40_..~~qv~~v4y~/+-+-l-+-++-i-++-l-+-+-1-+-~-+-+-1-+-+-+-+-+-+-l-+-+-+-l--i--+-:--++...;
60,....,-+-+-:>"'"f-v+-+-+-+-+-+-+-+-+-1-+-+-1-+..,__,-++-i-++-i-+-+-+-+-+-+-+-+-+-+-+.-;--+-+-++-
~
100 -+-+-+-+-t-++-i-++l-+-+-+-+-1-++-il--++-i-++-i\\-++-+-+-+-+-+-+-+-+-+-+-+-+-+l-+-1--+--il I
I I
I I
50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG.F)
Figure 2.
Salem Unit 2 Reactor Coolant System Cooldown (Cooldown rates up to I00°F/hr) Limitations Applicable for the First 15 EFPY (Without Margins For Instrumentation Errors)
B-15
' 6.
REFERENCES 1
Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988.
2 "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.
3 ASME Boiler and Pressure Vessel Code,Section III, Division 1 -
Appendixes, "Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure," pp.. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986.
4 5
6 Code of Federal Regulations, 10CFR50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,
Federal Register, Vol. 48 No. 104, May 27, 1983.
WCAP-11554, "Analysis of Capsule U from the Public Service Electricand Gas Compnay Salem Unit 2 Reactor Vessel Radiation Surveillance Program", S. E.
Yanichko, et. al., September 1991.
"Fractufe Toughness Requirements for Protection Against Pressurized Thermal Shock Events", 10 CFR Part 50, Vol. 58, No. 94, May 15, 1991.
B-16
ATTACHMENT 1 DATA POINTS FOR HEATUP AND COOLDOWN CURVES (Without Margins for Instrumentation Errors)
The data points used in the development of the heatup and cooldown curves shown in Figures 1 and 2 are contained on the attached computer printout sheets, pages 8-17 through 8-23.
8-17
OJ I
CX>
INDICATED TEMPERATURE
. (DEG.F)
INDICATED TEMPERATURE (DEG.F)
.\\**:**1 se; :.-000,: .'
/:=,::.:\\190.00(;)":=
195.000 200.000
THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 3 Cf'.::\\?:/\\:/c}/(?:~\\?:=:/}:::
!"\\)
C>
- f.
CJ I
N -
to I
N N.*
HR COOLDOWN
OJ I
N w THE FOLLOWING DATA WERE CALCULATEDFOR THE INSERVICE HYDROSTATIC LEAK TEST.
MINIMUM INSERVICE LEAK TEST TEMPERATL~E~~'*;~.066 ~FPY~
- *.*.' ***: * * ' * : '.<<' ** * :: ; ' ::,. * * * * * * *.*. ',., ;::: ' > I. ** w:,, > *,.: :. : : PRf,s.'.*,*.*.,*... s
.. ':**.:,u.,:.,.,'.*.,R**.*.*.*.. ~.*.'.:,.,*.*., *. :.. *,(**.. P,*.* *
.*.. s.. i
..... :.). :,'.:. :.:.... '.*,..,'.*.*.. :::- ; E. MP g R ATU~ ~.:.* :*(.Q~d}~).?.:'.:.**.**.*,*.: *
.: ::.*,.. :.r.*.: ~: < ;:-:: =:-.:::: ::::*::*:*:=:>:..-:\\::::-: =:*>* *. -====:*.. =*:.:. *:.:=:* ::*.. -::.**.*,.::::::=:* :"*.
- =.
2000 262
OJ I
N
~
PNJ. 60 DEG~F/HR HEATUP. REG. GUIDE 1. ~9; REV' 2 WI niouT MAR<HN
- q13/06/!!l~.. < *... *.....
CDMPOSITi CURVE PLOTTED FOR HEATUP PROFILE 2 IRRADIATION PERIOD = 15.000 EFP YEARS F~AW DEPTH~ (1~AOWIN)T HEATUP RATE(S) (DEG.F/HR) 60.0
- .. *.:.. :./~:*:>.... :.* ::
INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F)
(PSI)
(DEG.F)
(PSI)
(orn.o (PSI) 1
- ~~,060. '
~3a.~s 11
.165.ooo
- J\\ 6aa:1f.l*
> 33
< ;45.oao.*... *
- . *.11
- .23)2
- . 32
- .*.*.*.'.*.~.*.*.**.~.* ***.**:.*... ***.\\ * * *.*. ****.* **
- 2
- ao;ooo**
s43.a3.. **
1e*
110:000.*.\\*.112.22**
- 34.**<250;000 c ;:.:*****
3 95.000 539.37 19 175.000 737.76 35 255.000 1375.21 4
100.000 537.63 20 180.000 765.27 36 260.000 1431.71 6
105.000 538.51 21 185.000 794.92 37 265.000 1492.26 61*...
... *11 !05.: oooooo *:,.,...... :,,;::: 55.. 4461*,,44.as**.... * ***** >> *~~ :*. 1199. as* ~oooo. oo.. :.'.::: :::,: \\J3a2s. 77. -~s7,,,, /. \\33. as) ? /~11g;oooooo.****.*..,**,*.:......... *
... *......11:s!56266.*** ;~7~ / )/.}3i.:.\\.t,/ >> ***'**.. *:.. *
........ /<***, '* ':::':i-.,..
... :.:.:.* *..* :.:*.*,,*.**1*2*'0******.*.*.***a**.o****a***.** *. *.=-.*.**,*... *... *,*,*.*.*.*.*
~*
..,. \\?.**.:.,... *,.; *.£,:;i:. *** ***.'.. *
.. *a*.o*.:*,*.. ***.*.**o*.o.*o:* *
- fa '/,t.*:.:.*.:::.*... *,*.. **.*.*.*.*.**.. *4.*o***.. :,*.*...* * ::::/.:::..~:..... :.=..:........
- =:*
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- ,a
<:?553;2a>
- ***.24.. ***
- = :aa'hd4,..........
...... ::2ao;ooo ::: /'A100.;10X**......... {\\j/§?
9 125.000 561.91 25 205.000 919.47 41 285.000 1780.33 10 130.000 572.12 26 210.000 954.03 42 290.000 1865.52 11 135.000 584.00 27 215.000 991.11 43 295.000 1956.61 15 155.000 647.06 31 235.000 1169.42 47 315.000 2387.69 16 160.000 667.02 32 240.000 1222.34
..