Text

WCAP-12920, Analysis of Southern California Edison Co San Onofre Unit 3 Reactor Vessel Surveillance Capsule Removed from 97 F Location
	ML20073H818
Person / Time
Site:	San Onofre
Issue date:	03/30/1991
From:	Meyer T; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20073H781	List: ML20073H777; ML20073H818;
References
	WCAP-12920, NUDOCS 9105070274
	Download: ML20073H818 (153)
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WESTINGHOUSE CLASS 3 WCAP-12920 l

l ANALYSIS OF THE SOUTHERN CAllFORNIA EDISON COMPANY SAN ONOFRE UNIT 3 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVED FROM THE 97* LOCATION E. Terek E. P. Lippincott A. Madeyski March 1991 Work Performed Under Shop Order SJ0P-106 Prepared by Westinghouse Electric Corporation for the Southern California Edison Company Approved by: -

W9 Eb T.A.Meyer,Ma6ager Structural Reliability and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 I

C 1990 Westinghouse Electric Corp.

PREFACE This report has been technically reviewed and verified.

Reviewer Sections 1 through 5, 7 and 8 J. M. Chicots c / J';/ f lt t t's b Section 6 S, L. Anderson dcY (/%b2rMM i

_. _ _. - . . _ _ . . _- ~ _. _ _ . - _ - . _ _ _ . _ .

TABLE OF CONTENTS SEC112') Title Eagg 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 0F SPECIMENS FROM THE SURVEILLANCE 5-1 CAPSULE LOCATED AT 97*

5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-4 5.3 Tension Test Results 5-6 5.4 Hardness Test Results 5-7 5.5 Projected EOL Properties 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 6-5 7.0- SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1

8.0 REFERENCES

8-1 APPENDIX A - LOAD-TIME RECORDS FOR CHARPY-. SPECIMEN TESTS il

LIST OF TABLES (Cont)

Title Pace Iah_Le 4-1 Chemical Composition of the Unirradiated San Onofre 4-5 Unit 3 Reactor Vessel Surveillance Program Materials 4-2 Chemical Composition of Four San Onofre Unit 3 Charpy 4-6 Specimens Removed from Surveillance Capsule W-97*

4-3 Chemistry Results from the NBS Certified Reference 4-7 Standards 5-1 Charpy V-Notch Impact Data for the San Onofre Unit 3 5-8 Reactor Vessel Intermediate Shell Plate C-6802-1 Irradiated at 550*F, Fluence 0.8 x 10 19 n/cm2 (E>l.0 MeV) 5-2 Charpy V-Notch Impact Data for the San Onofre Unit 3 5-9 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 0.8 x 10 19 n/cm2 (E > 1.0 MeV) 5-3 Instrumented Charpy Impact Test Results for the San 5-10 Onofre Unit 3 Reactor Vessel Shell C-6802-1 Irradiated at 550*F, Fluence 0.8 x 10 19 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy impact Test Results for San Onofre 5-11 Unit 3 Surveillance Weld and HAZ Metal Irradiated at 550*F, Fluence 0.8 x 10 19 n/cm2 (E > 1.0 MeV) 5-5 Effect of 550*F Irradiation to 0.8 x 10 19 n/cm 2

5-12 (E > 1.0 MeV) on Notch Toughness Properties of the San Onofre Unit 3 Reactor Vessel Surveillance Materials l

iii

LIST OF TABLES (Cont)

Table Title Eagg 5-6 Comparison of the San Onofre Unit 3 Surveillance Material 5-13 30 f t-lb Transition Temperature Shifts and Uppe" Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-7 Tens 11e Properties for the San Onofre Unit 3 Reactor 5-14 Vessel Surveillance Material Irradiated at 550*F to 0.8 x 1019 n/cm2 (E > 1.0 MeV) 5-8 Room iemperature Rockwell B Hardness Values for the San 5-15 Onofre Unit 3 Surveillance Materials Irradiated at 550*F, Fluence 0.8 x 10 19 n/cm2 (E > 1.0 MeV) 5-9 Projected E0L (32 EFPY) RTNDT, RTPTS and USE for the 5-17 San Onofre Unit 3 Surveillance Material Contained in Capsule W-97 6-1 Calculated Fast Neutron Exposure Parameters at the 6-11 Surveillance Capsule Center 6-2 Calculated Fast Neutron Exposure Parameters at the 6-12 Pressure Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux 6-13 (E > 1.0 MeV) within the Pressure Vessel Wall iv

LIST OF TABLES (Cont) lable Title Pace 6-4 Relative Radial Distributions of Neutron Flux 6-14 (E > 0.1 MeV) within the Pressure Vessel Wall 6-5 Relative Radial Distributions of Iron Atom Displacement 6-15 Rate (dpa) within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-16 6-7 Irradiation History of i;autron Sensors Contained in 6-17 Capsule W-97 6-8 Measured Sensor Activities and Reactions Rates 6-18 6-9 Summary of Neutron Dosimetry Results 6-20 6-10 Comparison of Measured and Ferret Calculated Reaction 6-21 Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-22 Capsule Center 6-12 Comparison of Calculated and Measured Exposure levels 6-23 for Capsule W-97 v

l LIST OF TABLES (Cont)

Table Title Pagg 6-13 Neutron Exposure Projections at Key Locations on the 6-24 Pressure Vessel Clad / Base Metal Interface for San Onofre Unii 3 6-14 Neutron Exposure Values for use in the Generation of 6-25 Heatup/Cooldown Curves 6-15 Updated lead Factors for San Onofre Unit 3 6-26 Surveillance Capsules vi !

LIST OF ILLUSTRATIONS Fiaure 111.ls Eng 4-1 Typical Surveillance Capsule Assembly 4-8 4-2 Typical Locations of San Onofre Unit 3 Surveillance 4-9 Capsule Assemblies 4-3 Typical Tensile-Monitor Compartment Assembly 4-10 4-4 Typical Charpy impact Compartment Assembly 4-11 5-1 Appearance of the Thermal Monitors in the Various 5-18 Quartz Tubes 5-2 Charpy V-Notch Impact Properties for the San Onofre 5-19 Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for the San Onofre 5-20 Unit 3 Reactor Vessel Shell Plate C-6802-1 (Ls_,itudinal Orientation) 5-4 Charpy V-Notch Impact Properties for the San Onofre 5-21 Unit 3 Reactor Vessel Surveillance Weld Metal 5-5 Charpy V-Notch Impact Properties for the San Onofre 5-22 Unit 3 Reactor Vessel HAZ Metal 5-6 Charpy V-Notch Impact Specimen Fracture Surfaces for 5-23 the San Onofre Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (Transverse Orientation) vii 1

LIST OF ILLUSTRATIONS Ejaure Title Eagg 5-7 Charpy V-Notch Impact Specimen Fracture Surfaces for 5-24 the San Onofre Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (longitudinal Orientation) 5-8 Charpy impact Specimen Fracture Surfaces for San 5-25 Onofre Unit 3 Reactor Vessel Surveillance Weld Metal 5-9 Charpy impact Specimen Fracture Surfaces for the San 5-26 Onofre Unit 3 Reactor Vessel Surveillance HAZ Metal 5-10 Tensile Properties for the San Onofre Unit 3 Reactor 5-27 Vessel Intermediate Shell Plate C-6802-1 (Transverse Orientation) 5-11 Tensile Properties for the San Onofre Unit 3 Reactor 5-28 Vesse' Weld Metal 5-12 Tensile Properties for San Onofre Unit 3 Reactor 5-29 Vessel HAZ Metal 5-13 Fractured Tensile Specimens from the San Onofre Unit 3 5-30 Reactor Vessel latermediate Shell Plate C-6802-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens from the San Onofre Unit 3 5-31 Reactor Vessel Weld Metal 5-15 Fractured Tensile Specimens from the San Onofre Unit 3 5-32 Reactor Vessel HAZ Metal viii l

LIST OF ILLUSTRATIONS Fiaure Title Ease 5-16 True Stress-Strain Curves for the San Onofre 5-33 Unit 3 Reactor Vessel Shell Plate C-6802-1 Tension Specimens 2KE and 2KC 5-17 True Stress-Strain Curve for the San Onofre 5-34 Unit 3 Reactor Vessel Shell Plate C-6802-1 Tension Specimen 2K3 5-18 True Stress-Strain Curves for the San Onofre Unit 3 5-35 Reactor Vessel Weld Metal Tension Specimens 3JC and 3JE 5-19 True Stress-Strain Curve for the San Onofre Unit 3 5-36 Reactor Vessel Weld Specimen 3J4 5-20 True Stress-Strain Curve for the San Onofre Unit 3 5-37 Reactor Vessel HAZ Metal Tension Specimen 4JA 5-21 Engineering Stress-Strain Curves for the San Onofre 5-38 Unit 3 Reactor Vessel Shell Plate C-6802-1 Tension Specimens 2KE and 2KC 5-22 Engineering Stress-Strain Curve for the San Onofre 5-39 Unit 3 Reactor Vessel Shell Plate C-6802-1 Tension Specimen 2K3 5-23 Engineering Stress-Strain Curves for the San Onofre Unit 5-40 3 Reactor Vessel Weld Metal Tension Specimens 3JC and 3JE ix

LIST OF ILLUSTRNi10NS Fioure Title hge 5-24 Engineering Stress-Strain Curve for the San Onofre 5-41 Unit 3 Reactor Vessel Weld Metal Tension Specimen 3J4 5-25 Engineering Stress-Strain Curve for the San Onofre 5-42 Unit 3 Reactor Vessel HAZ Metal Tension Specimen 4JA x

SECTION 1.0

SUMMARY

OF RESULTS The analysis of the Southern California Edison Company San Onofre Unit 3 reactor vessel materials contained in the surveillance capsule removed from the 97* location, the first capsule to be removed from the San Onofre Unit 3 reactor pressure vessel, led to the following conclusions:

o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 0.8 x 10 I9 n/cm after 4.33 EFPY of plant operation, 2

o Irradiation of reactor vessel shell plate C-6802-1 Charpy specimens to 0.8 x 10 19 n/cm2 (E > 1.0 MeV) at 550*F resulted in a 30 and 50 ft-lb transition temperature increase of 55 and 67*F, respectively, for specimens oriented normal to the major working direction (transverse orientation),

o Irradiation of reactor vessel shell plate C-6802-1 Charpy specimens to 0.8 x 10 I9 n/cm2 (E > 1.0 MeV) at 550*F resulted in a 30 and 50 ft-lb transition temperature increase of 50 and 45'F, respectively, for specimens oriented parallel to the major working direction (longitudinal orientation).

I9 n/cm2 (E o The weld metal Charpy specimens irradiated to 0.8 x 10

> 1.0 MeV) at 550*F resulted in a 30 f t-lb transition temperature increase of 32*F and a 50 ft-lb transition temperature increase of 25'F.

o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 0.8 x 10 19 n/cm2 (E > 1.0 MeV) at 550*F resulted in a 30 ft-lb transition temperature increase of 45'F and a 50 ft-lb transition temperature increase of 70'F.

1-1

o The average upper shelf energy of shell plate C-6802-1 showed a decrease of 16 ft-lbs after irradiation to 0.8 x 1019 n/cm2 (E >

1.0 MeV) at 550'F for specimens oriented normal to the major working direction (transverse orientation).

o The average upper shelf energy of shell plate C-6802-1 showed a decrease of 22 ft-lbs after irradiation to 0.8 x 1019 n/cm2 (E >

1,0 MeV) at 550'F for specimens oriented parallel to the major working direction (longitudinal orientation).

o The average upper shelf energy of the weld metal showed a decrease of 12 ft-lbs after irradiation to 0.8 x 1019 n/cm2 (E > 1.0 MeV) at 550*F.

o The average upper shelf energy of the weld Heat-Affected-Zone (HAZ) metal showed a decrease of 11 f t-lbs after irradiation to 0.8 x 1019 n/cm2 (E > 1.0 MeV) at 550*F.

o The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the San Onofre Unit 3 reactor vessel is as follows:

Vessel inner radius * = 4.20 x 1019 n/cm 2 Vessel 1/4 thickness - 2.24 x 1019 n/cm 2 Vessel 3/4 thickness = 4.58 x 1018 n/cm 2

Clad / base metal interface ,

o Based on an E0L (32 EFPY) peak clad / base metal interface fluence of 4.20 x 1019 n/cm2 (E > 1.0 MeV), Regulatory Guide 1.99 Revision 2 and 10 CFR Part 50, the reactor vessel surveillance material, plate C-6802-1 and weld metal, are expected to meet the screening criteria of at least 50 ft-lb of upper shelf energy and an RTNDT of less than 270*F for plate material and axial welds and less than 300*F for circumferential welds through E0L (32 EFPY).

1-2 l

SECTION

2.0 INTRODUCTION

This report presents the results of the examination of the capsule located at 97', the first surveillance capsule to be removed from the San Onofre Unit 3 reactor vessel in the continuing surveillance program which monitors the effects of neutron irtadiation on the Southern California Edison Company San Orofre Unit 3 reactor pressure vessel materials under actual operating conditions, The surveillance program for the San Onofre Unit 3 reactor pressure vessel materials was designed by Combustion Engineering, Inc.[Il to the requirements of ASTM 185-73 " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". A complete description of the surveillance program has been reported by Combustion Engineering, Inc,Ill Westinghouse Power Systems personnel were contracted to aid in the preparatic, of procedures for removing the capsule located at 97* from the reactor and its shipment to the Westinghouse Science and Technology Center Meta 11ographic 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 the first surveillance capsule to be removed from the San Onofre Unit 3 reactor vessel and discusses the analysis of these data. The data are compared to the results of tests performed on unirradiated material from the reactor vessel.[23 2-1

1 l

SECT 10ti 3,0 PACKGROUND The Oility of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important f actor 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 SA533 Grade B Class 1 (base material of the San Onofre Unit 3 reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile failure,"

Appendix G to Section 111 of the ASME Boiler and Pressure Vessel Code. The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNOT)-

RTNOT is defined in the ASME Code as follows:

Determine a temperature THDT that is at or above the nil-ductility transition temperature by drop weight tests.

At a temperature not greater than TNDT+60*F, each specimen of the Charpy V-notch (C y ) test (NB-2321.2) shall exhibit at least 35 mils lateral expansion and not less than 50 ft-lb absorbed energy. When these requirements are met, TNDT is the reference temperature RTNDT' 3-1

i in the event that the above requirements are not met, conduct additional Cy tests in groups of three specimens to determine the temperature TCy at which they are met, in this case the reference temperature RTNDT = TCy

-6E N s, W & nce W a M % DT IS the higher of THDT and (TCy -60*F).

When a C y test has not been performed at TNDT + 60'F or when the Cy test at TNDT + 60*F does not exhibit a minimum of 50 ft-lb and 35 mils lateral expansion, a temperature representing a minimum of 50 ft-lb and 35 mils lateral expansion may be obtained from a full C y impact curve developed from the minimum data points of all the C y tests performed.

The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The Ki p 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 KIR 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.

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 vessel surveillance program such as the Program for Irradiation Surveillance of San Onofre Unit 3 Reactor Vessel Materiaisill, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTgg7 to adjust t'.: RT NDT for radiation embrittlement. This adjusted RTHDT (RTNDT initial + ARTNDT) is used to index the material to the KIR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

3-2

SECTION

4.0 DESCRIPTION

Or PROGRAM Six surveillance capsules for monitoring the effects of neutron radiation exposure on the San Onofre Unit 3 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up. A typical surveillance capsule assembly is shown in Figure 4-1. The six irradiation capsule assemblies are located at radial positions about the core, with their axial positions bisected by the midplane of the core. The capsule assemblies are contained in capsule holders positioned circumferentially about the core at locations which include regions of maximum flux as shown in figure 4-2. The six vessel surveillance capsule assemblies are located near the inside wall of the reactor vessel.

The first surveillance capsule, located at 97'F, was removed after 4.33 effective full power years of plant operation. This capsule contained Charpy V-notch and tensile specimens from intermediate shell plate C-6802-1, weld metal and weld Heat-Affected-Zone (HAZ) material.

The chemical composition of the San Onofre Unit 3 beltline materials is presented in Table 4-l[2] .

The chemical analysis reported in Table 4-1 was obtained from unirradiated material used in the San Onofre Unit 3 surveillance program [2]. In addition, a chemical analysis using Inductively Coupled Plasma Spectrometry (ICPS) was performed on irradiated weld metal specimens 378 and 367, intermediate shell plate C-6802-1 specimen 22B and HAZ metal specimen 46T and is reported in Table 4-2. The chemistry results from the NBS certified reference standards are reported in Table 4-3.

The San Onofre Unit 3 reactor vessel intermediate shell plate C-6802-1 was fabricated from SA 533 Grade B, Class 1 steel plate. The San Onofre Unit 3 surveillance materials are made from the same material. The steel plate was manufactured by Combustion Engineering and heat treated as follows:

4-1

l The heat treatment for the plate material consisted of austenitization at 1600'F 25'F for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; water auenched and tempered at l

1225'F i 25'F for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . After a 40-hour stress relief at ll50'F i 25'F the plates were furnace cooled to 600'F. The weldment received a final 41-hour and 45-minute stress relief at 1100 to 1150*F.

Because the lower and intermediate shell courses of the reactor vessel experience the highest fluences in the entire vessel, selection of the candidate surveillance base materials was restricted to the six plates in the intermediate and lower shell courses. Selection criteria followed the general guidelines of ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels".

The San Onofre Unit 3 surveillance materials consist of intermediate shell plate C-6802-1 which was made of SA 533 Grade B Class I plate, Mil B-4 submerged arc-weld with Linde 124 flux and Heat-Affected-Zone (HAZ) metal from plate C-6802-1.

All test specimens were machined from the 1/4 thickness location of the plate materi al . Test specimens represent material taken at least one plate thickness from the quenched end.

The longitudinal base metal Charpy V-notch test specimens were oriented with the longitudinal axis of the specimen parallel to the major roiling direction of the plate whereas transverse base metal specimens were oriented with the longitudinal ar,is perpendicular to the major rolling direction and parallel to the surface of the plate. The axis of the notch of the Charpy specimens was machined perpendicular to the major surfaces of the plate.

The weld metal Charpy impact test specimens were oriented transversely with their major axis perpendicular to the direction of the weld and parallel to the surface of the plate. The notch in each of the test specimens was oriented perpendicular to the surface of the weld.

4-2

The Heat-Affected-Zone (HAZ) metal Charpy V-notch test specimens were oriented transversely with the longitudinal axis of the specimen perpendicular to the direction of the weld and parallel to the surface of the plate. The notch in each of the test specimens was oriented perpendicular to the surface of the weld and the root of the notch was centered on a line parallel to and 1/32 inch from the weld fusion line on the Heat-Af fected-Zone side.

The longitudinal base metal tensile test specimens were oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate whereas transverse base metal specimens were oriented with the longitudinal a.is perpendicular to the major rolling direction and parallel to the surface of the plate.

The weld metal tensile test specimens were oriented transversely with their major axis perpendicular to the direction of the weld and parallel to the surf ace of the plate.

The Heat-Affected-Zone (HAZ) metal tensile test specimens were oriented transversely with the longitudinal axis of the specimen perpendicular to the direction of the weld and parallel to the surface of the plate. The test specimens were positioned so that the HAZ was centered as closely as possible in the gage length of the specimen.

The capsule contained fission threshold detectors (U-238) and threshold detectors of nickel (Ni), titanium (Ti), iron (Fe), sulfur (S), copper (Cu) with known cobalt (Co) content and Cobalt (Co) to monitor the thermal neutron exposure. The flux monitors were located in the three tensile-monitor compartment assemblies as shown in Figure 4-3.

4-3

Thermal monitors made from low-melting eutectic alloys sealed in quartz tubes were included in the capsule and were also located in the three tensile-monitor compartment assemblies as shown in Figure 4-3. The composition of the alloys and their melting points are as follows:

Melting Temperature Chemical Comoosition (*C) (*F) 80% Au, 20% Sn 280 536 90% Pb, 5% Sn, 5% Ag 292 558 97.5% Pb, 2.5% Ag 304 580 97.5% Pb, 0.75% Sn,1.75% Ag 310 590 Contained in Figure 4-4 is A typical Charpy impact compartment assembly.

4-4

TABLE 4-1 Chemical Composition of the Unirradiated San Onofre Unit 3 Reactor Vessel Surveillance Program Materials

Weicht Percent Plate Wold Element C-38Q.?d C-6802-2 & 3 Si 0.23 0.39 S 0.014 0.009 P 0.008 0.004 Mn 1,38 1,54 C 0.24 0.12 Cr 0.07 0.05 Ni 0.57 0.08 Mo 0.58 0.55 V 0.005 0.005 Cb <,01 < 01 B 0.0005 0.001 Co 0.010 0.015 Cu 0.05 0.03 Al 0.033 0.006 W <.01 0.01 Ti <.01 <.01 As 0.009 < 001 Sn 0.005 0.002 Zr 0.002 <.001 N 0.010 0.006 Sb ---

0.0012 Pb ---

<.001

Chemical Composition as Reported in Reference 2.

4-5

TABLE 4-2 Chemical Compostion of Four San Onofre Unit 3 Charpy Specimens Removed from Surveillance Capsule W-97' Chemical Composition (wt.%)

Standard Reference Metal Weld Metal Specimen No. 37B 367 46T 228 Material WELD WELD HAZ BASE

........................................................................... l Fe MATRIX ELEMENT: Remainder by Difference Mn 1.430 1.464 1.295 1.307 Cr 0.063 0.061 0.102 0.101 Ni 0.108 0.094 0.567 0.575 Mo 0.563 0.544 0.527 0.531 Co 0.010- 0.012 <0.010 <0.010 Cu 0.034 0.032 0.057 0.058

-P 0.014 0.011 0.012 0.009 V 0.015 0.014 0.007 0.014 C 0.116 0.111 0.221 0.237 S 0.0075 0.0075 0.0136 0.0135 Si 0.411 0.355 0.244 0.048 I

l l

l Analyses Method of Analysis Metals ICPS, Inductively Coupled Plasma Spectrometry Carbon EC-12, LECO Carbon Analyzer l Sulfur Combustion / titration Silicon Dissolution / gravimetric l

I 4-6

TABLE 4-3 Chemistry Results from the NBS Certified Reference Standards Material 10 Low Alloy Steel: NBS Certified Reference Standards NBS 361 NBS 362 Certified Measured Certified Measured

............................o............................................

Met al s Concentration in Weicht Percent Fe

95.60 (matrix) 95.30 (matrix)

Mn 0.66 0.671 1,04 1.075 Cr 0.694 0.701 0,30 0.310 Ni 2.00 2.12 0.59 0.615 Mo 0.19 0.195 0.068 0.065 Co 0.032 0.015 0.30 0.347 0.042 0.043 0.50 0,519 Cu P 0.014 0.0141 0.041 0.0395 V 0.011 0.014 0.040 0.0355 C 0.383 0.3844 0.160 0.1620 S 0.014 N.A. 0.036 0.0362 Si 0.222 0.228 0.39 N.A.

Material 10 Low Alloy Steel: NBS Certified Reference Standards NBS 363 NBS 364 ____

Certified Measured Certified Measured Metals Concentration in Weicht Percent Fe

94.4 (matrix) 96,7 (matrix)

Mn 1.50 1.556 0.255 0.251 Cr 1.31 1.366 0.063 0.061 Ni 0.30 0.321 0.144 0.142 Mo 0.028 0.029 0.49 0.495 to 0.048 0.040 0.15 0.160 Cu 0.10 0.102 0.249 0.253 P 0.029 0.0316 0.01 0.0094 V 0.31 0.251 0.105 0.106 C 0.62 N.A. 0.87 N.A.

S 0.0068 N.A. 0.0250 0.0249 Si 0.74 N.A. 0.065 N.A.

.r..................................................... .................

Matrix element calculated as difference for material balance. ( )

N. A. - Not analyzed 4-7

1l -

Lock Assembly l

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sp -

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Compadment s:: s

.> Charpy Impact Compartments D'# -

Tensile -Monitor Compartment N s ,-

\i s- s

%d

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> Charpy Impact Compartments

\s::' ,

Tensile -Monitor Compartment ,

5 Figure 4-1. Typical Surveillance Capsule Assembly 4-8 )

0 180 Outlet Nozzle

[~~~~~~~l

/ \

Vessel / - :

0 ~.

104 P-^ ,

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Reactor Vessel fj 970 ~ , p ,

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0 Elevation j Enlarged Pian View View l Figure 4-2. Typical Locations of San Onofre Unit 3 Surveillance Capsule Assemblies EdffIM$[a

=-

Wedge Coupling - End Cap

y Flux Spe:trum Monitor Cadmium Shielded Flux Monitor Housing-

$b s, H Stainless Steel Tubing Stainless Steel Tubing 7"]

Threshold Detector - '

Cadmium Shield

. 9 . Threshold Detector Flux Spectrum Monitor %

% a uadz Tuung L Temperatur e Monitor L g '

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( g, Low Melting Alloy

\,

Tensile Specimen :m Split Spacer = .

'lll k

Tensile Specimen Housing =

NRectangular Tubing

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[ -Wedge Coupling - End Cap l

Charpy impact Specimens i

Spacer ,,

s N .-

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%p

-Redangular Tubing k -Wedge Coupling - End Cap N

Figure 4-4. Typical Charpy impact Compartment Assembly 4-11

SECTION 5.0 TESTING Of SPECIMENS FROM THE SURVEILLAf4CE CAPSULE LOCATED AT 97*

5.1 Overview The post-irradiation mechanical testirg of the Charpy V-notch and tensile specimens, contained in the surveillance capsule removed f rom the 97' location, was performed at the Westinghouse Science and lechnology Center Hot Cell Laboratory with consultation by Westinghouse Power Systems personnel.

Testing was performed in accordance with 10CIR50, Appendices G and Hl33, ASTM

$pecification E185-82l43 and Westingnouse Procedure MHL 8402, Rdvision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.

Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list supplied by the Southern California Edison Company. No discrepancies were found.

The surveillance capsule had four low melting cutectic alloy thermal monitors which were removed from Blocks C214, C242, and C273. The thermal monitors were encapsulated in quartz tubes of 1, 1.25, 1,50, and 1.75 inches long and had melting temperatures of 536' , 550', 580' and 590* f, respectively, Examination of the four low-melting, 536', 558',

580' and 590*f eutectic alloys indicated melting of the 536'f monitor, but not of the three remaining thermal monitors in blocks C214 and C242. Based on this examination, the maximum temperature to which these blocks were exposed was less than 558'f but more than 536'f. At block C273 both the 536'f and the 558'f melting point alloys melted, indicating that the temperature of this block reached the range from 558'f to 580*f. Figure 5-1 shows the condition of the thermal monitors.

The Charpy impact tests were performed per ASTM Specification E23-88I63and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 35BJ machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology Model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy 5-1

energy (ED ). from the load-time curve, the load of general yielding (Pgy),

the time to general yielding (tcy), 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 fradre load (Pr), and the load at which fast fracture terminated is identified as the arrest load (PA )-

The energy at maximum load (Eg) 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 (E p ) is the difference between the total energy to fracture (E )Dand the energy at maximum load.

The yield stress-(oy) is calculated from the three-point bend formula having the following expression:

oy Pgy * (L / (B * (W-a)2 f C)) (1) where the constant C is dependent on the notch flank angle (d), 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 4 45' and p = 0.010", Equation 1 is valid with with C = 1.21. Therefore (for L =

4W),

oy Pg y*(L/(B*(W-a)2*l.21)]=[3.3P cyW)/[B(W-a)2] (2)

For the Charpy specimens, B = 0.394 in., W = 0.394 in, and a = 0.079 in.

Equation 2 then reduces to:

oy - 33.3 x Pgy (3) i

where oy is in units of psi and Pgy is in units of lbs. The flow l stress was calculated from the average of the yield and maximum loads, also l using the three-point bend formula.

l 5-2 l

Percent shear was determined from post fracture pnotographs using the ratio-of-areas methods in compliance with ASTM Specification A370-89[6). The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-89b{73 and [21-79(1988)[83, 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.

Extension measurements were made with a linear variable displacement transducer (l.VDT) cxtensometer. The extenrometer knife edges were spring-loaded to the specimen and oper ated through specimen f ailure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-85I93 E"evated 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 eac5 riid 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 obtained desired specimen temperatures. Experiments indicated that this method is accurate to 2'T, 5-3

?

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.

Rockwell B hardness values were obtained with a RAMS Rockwell Hardness Tester Model No. 30-R.

5.2 Charny V-Notch impact Test pesults The results of Charpy V-notch impact tests performed on the various materials contained in the San Onofre Unit ? surveillance capsule located at 97*

irradiated at 550*F to 0.8 x 10l9 n/cm2 (E > 1.0 MeV) are presented in Tables 5-1 through 5-4 and are compared with unirradiated resultsI23 as shown ,

in Figures 5-2 through 5-5. The transition temperature increases and upper shelf energy decreases for the various materials contained in the San Onofre Unit 3 surveillance capsule located at 97' are summarized in Table 5-5.

4 Irradiation of the reactor vessel intermediate shell plate C-6802-1 Charpy specimens to 0.8 x 10 19 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 55'F and a 50 ft-lb transition temperature increase of 67'F for specimens oriented with the longitudinal axis perpendicular to the major rolling direction and parallel to the surface of the plate (transverse orientation). This results in a 30 ft-lb transition temperature of 105'F and a 50 f t-lb transition temperature-of 160*F.

The average upper shelf energy (USE) of the intermediate shell plate C-6802-1 Charpy specimens (transverse orientation) resulted in a decrease of 16 ft-lbs after irradiation to 0.8 x 10I9 n/cm2 (E > 1.0 MeV) at 550'F. This results in an average USE of 76 ft-lbs (Figure 5-2).

5-4

Irradiation of the reactor vessel intermediate shell plate C-6802-1 Charpy specimens to 0.8 x 10 19 n/cm2 (E > 1.0 MeV) at 550'r (figure 5-3) resulted in a 30 f t-lb transition temperature increase of 50*r and a 50 ft-lb transition temperature increase of 45'F for specimens oriented with the longitudinal axis parallel to the major rolling direction (longitudinal orientation). This results in a 30 ft-lb transition temperature of 130'f I and a 50 ft-lb transition temperature of 170'F.

1 The average upper shelf energy (VSE) of the intermediate shell plate C-6802-1 Charpy specimens (longitudinal orientation) resulted in a decrease of 22 ft-lbs after irradiation to 0.8 x 10I9 n/cm2 (E > 1.0 MeV) at 550*F. This results in an average VSE of 70 ft-lbs (Figure 5-3).

Irradiation of the reactor vessel surveillance weld metal Charpy specimens to

.0.8 x 1019 n/cm2 (E > 1.0 MeV) at $50*F (figure 5-4) resulted in a 30 f t-lb transition temperature increase of 32*F and a 50 ft-lb transition temperature increase of 25'f. This resulted in a 30 ft-lb transition temperature of 5'f and a 50 ft-lb transition temperature of 45'f.

The average upper shelf energy (VSE) of the reactor vessel surveillance weld metal resulted in a energy decrease of 12 ft-lbs after irradiation to 0.8 x 10l9 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average USE of 70 ft-lb (figure 5-4).

Irradiation of the reactor vessel weld metal Heat-Affected-Zone (HAZ) specimens to 0.8 x 1019 n/cm2 (E > 1.0 MeV) at 550'F (figure 5-5) resulted in a 30 ft-lb transition temperature increase of 45'f and a 50 ft-lb transition temperature increase of 70*f. This results in a 30 ft-lb transition temperature of 35'F and a 50 ft-ib transition temperature of 120*F.

The average upper shelf energy (VSE) of the reactor vessel HAZ metal resulted in a decrease of 11 ft-lbs after irradiation to 0.8 x 10 I9 n/cm2 (E > 1.0 MeV) at 550'F. - This resulted in an average USE of 74 f t-lb (figure 5-4).

5-5

The fracture appearance of each irradiated Charpy specimen from the various materials are shown in figures 5-6 through 5-9 and show an increasingly ductile or tougher appearance with increasing test temperature, A comparison of the 30 ft-lb transition temperature increases and the upper shelf energy decreases for the various San Onofre Unit 3 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2[10] is presented in Table 5-8. This comparison indicates that the 30 ft-lb transition temperature increases resulting from irradiation to 0.8 x 1019 n/cm2 (E > 1.0 MeV) are higher than the NRC Regulatory Guide 1.99, Revision 2 predictions and the average USE decreases resulting from irradiation to 0.8 x 1019 n/cm2 (E > 1.0 MeV) are in close agreement with Regulatory Guide 1.99, Revision 2 predictions.

5.3 Tension Test Results The results of tension tests performed on the reactor vessel intermediate shell plate C-6802-1, weld metal and weld HAZ metal irradiated to 0.8 x 10 19 n/cm2 (E > 1.0 MeV) at 550*F are shown in Table 5-7 and are compared with unirradiated results(2) as shown in Figures 5-10, 5-11 and 5-12.

Irradiation of the reactor vessel intermediate shell plate C-6802-1 tensile specimens to 0.8 x 10 19 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-10) resulted in an increa:e of 2 to 8 ksi in 0.2 percent offset yield strength and an increase of 0 to 4 ksi in ultimate tensile strength for specimens oriented with the longitudinal axis perpendicular to the major working direction of the plate (transverse orientation).

Irradiation of the reactor vessel weld metal tensile specimens to 0.8 x 10 19 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-11) resulted in an increase of 0 to 14 ksi for the 0.2 percent offset yield strength and an increase of 0 to 22 l

ksi for the ultimate tensile strength.

5-6

1rradiation of the reactor vessel weld HA2 metal tensile specimens to 0.8 x 1019 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-12) resulted in an increase of 2 to 7 ksi for the 0.2 percent offset yield strength and an increase of 0 to 3 ksi for the ultimate tensile strength.

l The fractured tension specimens for the reactor vessel intermediate shell plate C-6802-1 tensile specimens are shown in Figure 5-13.

The fractured tension specimens for the weld metal are shown in Figure 5-14.

The fractured tension specimens for the weld HAZ metal are shown in Figure 5-15.

True stress-strain curves for the tension specimens are shown in Figures 5-16 through 5-20.

Engineering stress-strain curves for the tension specimens are shown in Figures 5-21 through 5-25.

5.4 Hardness Test Results The results of Rockwell B hardness tests are presented in Table 5-8.

5.5 Pro.ietted E0L Properties The projected EOL (32 EFPY) RTNDT, RTPTS and USE for the materials contained in the capsule removed from the 97* location are presented in Table 5-9. Table 5-9 was generated using the projected E0L (32 EFPY) peak vessel fluence of 4.2 x 1019 n/cm2 (E > 1.0 MeV).

5-7

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE SAN ON0fRE UNil 3 REACTOR VESSEL INTERMEDIATE SHELL PLATE C-6802-1 IRRADIATED AT 550'F, FLUENCE 0.8 x 10 19 n/cm2 (E > 1.0 MeV)

Temperature Impact Energy Lateral Expansion Shear Sample No. (*F) ('C) (ft-lb) (J) (alls) (mm) (fl_

Longitudinal Orientation 147 50 6.0 8.0 5.0 0.13 5 153 75 22.0 30.0 15.0 0.38 15 15K 100 29.0 39.5 24.0 0.61 25 113 125 35.0 47.5 32.0 0.81 30 14Y 150 32.0 43.5 26.0 0.00 30 15B 165 45.0 61.0 43.0 1.09 45 14A 175 30.0 40.5 27.0 0.69 45 ,

14B 200 46.0 62.5 46.0 1.17 55 11T 225 75.0 101.5 66.0 1.68 80 12K 250 45.0 61.0 48.0 1.22 80 12L 250 48.0 65.0 46,0 1.17 95 14M 275 90.0 122.0 76.0 1.93 100 Transverse Orientation 25U 0 -1 19.0 26.0 18.0 0.46 10 23K 25 -

16.0 21.5 11.0 0.28 10 21A 50 9.0 12.0 12.0 0.31 10 23L 75 21.0 28.5 20.0 0.51 15 22B 100 25.0 34.0 25.0 0.64 20 25J 115 55.0 74.5 48.0 1.22 45 25L 130 50.0 68.0 44.0 1.12 50 247 150 33.0 44.5 33.0 0.84 50 223 165 36.0 49.0 37.0 0.94 65 250 200 92.0 124.5 73.0 1.85 100 245 225 62.0 84.0 59.0 1.50 100 23M 250 73.0 99.0 77.0 1.96 100 5-8

~r - . , - , - - . , --,. wwe

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE SAf4 Ot10FRE UNIT 3 REACTOR VESSEL WELD METAL AND HAI METAL 1RRADIATED AT 550'f, FLUENCE 0.8 x 10 19 n/cm2 (E > 1.0 MeV)

Sneple No. b (k $1 ,(m 1 $I') (

Weld Metal 3 - .0 . . '.!

31D 0 -1 5.0 7.0 6.0 0.15 5 37B 10 -1 42.0 57.0 34.0 0.86 60 367 25 - 42.0 57.0 44.0 1.12 65 331 50 10 44.0 59.5 44.0 1.12 70 37U 60 16 61.0 82.5 57.0 1.45 95 371 80 65.0 88.0 59.0 1.50 95 365 105 67.0 91.0 60.0 1.52 100 34P 150 75.0 101.5 73.0 1.85 100 33D 190 77.0 104.5 76.0 1.93 100 377 225 107 63.0 85.5 58.0 1.47 100 HAZ Metal 441 -75 -59 9.0 12.0 8.0 0.20 5 42L -50 -46 18.0 24.5 16.0 0.41 15 450 -20 -29 3.0 4.0 4.0 0.10 5 43P O -18 26.0 35.5 23.0 0.58 35 464 25 -4 35.0 47.5 37.0 0.94 45 410 60 16 33.0 44.5 33.0 0.84 50 47E 95 35 43.0 58.5 44,0 1.12 75 46T 125 52 41.0 55.5 38,0 0.97 70 45M 145 63 45.0 61.0 44.0 1.12 80 437- 165 74 63.0 85.5 61.0 1.S5 100 45P 200 93 84.0 114.0 68.0 1.73) 100 434 250 121 75.0 101.5 65.0 1.05) 100 5-9

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O hk WT ebTNeNNO mbe c 00 N mom,owmMeowom OTOOOVom 0

Y T t,s M 00 to e v b4 N N N M si M o M to t- @ MmN t c. t N T T N N b T m a yo Z .r. N U

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nnnnn 8888*n88888 -----nnn do :>

HHV U v T U r i. b r=4 G -

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(fJ Z M M M M M M M M M e4 M M N N t1 N N N N N O N N N e e 5-10

TARI E 5-4 IfiSTRUMENTED EHARPY IMPACT TEST RES' TS FOR THE SAfl OfiOFRE U!IIT 3 SURVEILLAfiCE WELD At1D HAZ METAL IRRADIATED AT 550*F, FLUENCE 0.8 x 10 I9 n/cm2 (E > 1.0 MeV)

Normalleed Enerstice Test Charpy Charpy Waximum Prop Yield Time Maximum Time to Fracture Arrent Yield Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Strees Number (*F) (f t-lb) (ft-Ib/in2) (kipe) (usee) (kire) __funec) (kipa) (kipe) (hel)

Weld Metal 363 - 50 23.0 185 53 132 4.40 0 4.55 80 4.55 0.55 146 i 33L - 25 27.0 217 146 71 4.10 36 4.55 250 4.55 0.35 136 31D 0 5.0 40 COMPLTER MALFUNCTION +-

37B 10 42.0 338 146 192 4.45 65 4.75 240 4.70 2.25 147 367 25 42.0 338 134 204 4.40 45 4.75 7.20 4.70 1.95 146 331 50 44.0 354 155 199 3.50 45 4.00 316 3.85 1.40 115 37U 60 61.0 491 159 332 3.80 70 4.25 305 4.15 2.70 126 371 80 65.0 523 168 355 4.30 120 4.45 295 4.20 3.60 142 W 365 105 67.0 540 167 373 3.95 135 4.25 315 -- -. 131 1 34P 150 75.0 604 192 412 3.65 2.90 125 60 4.05 3.55 390 470

-. -. 121

~

33D 100 77.0 620 199 421 -. -. 96 377 225 63.0 507 151 357 3.75 140 4.00 305 -. -. 124 HAZ Metal l

! 441 - 75 9.0 73 COMPLTER MALFL*1CTION **

j 42L - 50 18.0 145 7G 70 4.35 40 4.55 120 4.55 0.01 144 45C - 20 3.0 24 COMPLTER MM'CTION ==

43P O 26.0 209 85 125 4.40 70 4.55 140 4.35 1.10 145 464 25 35.0 282 158 123 3.95 10 4.60 275 4.60 0.70 131 1 41C 60 33.0 266 58 <208 3.30 40 3.55 120 3.55 2.65 109

! 47E 95 43.0 346 110 236 3.20 66 3.55 245 3.50 0.25 106 46T 125 41.0 330 151 180 3.55 65 4.05 305 3.75 2.15 118 l 237 3.90 130 4.05 240 3.55 1.75 129 45M 145 45.0 362 125 437 165 63.0 007 138 369 3.35 65 3.65 300 -. -. 111 45P 200 84.0 ts76 199 477 3.70 60 4.05 395 -. -. 122 434 250 75.0 604 183 421 3.25 90 3.70 405 -. -. 107

~

-Fully ductile fracturer no arrest load

- When test was rtsn data did not transfer to cmter d:sk i

l l _ - - - -_ - - ________________ _____- _____ _.

i 1lj

)

b t

f 5 s 2 2 1 i 2 1 1 A( - - - - 5

) h; b d r i

- t n

v f t

ia s 7

0 7

4 1 b r

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a r

r 2

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R n E

T w

t A %

M h s

GN t 7 5 5 0 s

a A A c 4 2 7 s

L t L

) I d s e e MES T L

b

)

t a

i 0 G 5 0 t

p y

0 S 1 ** d 5 T 4 2 1

- ( a 1 I 1 a

L t r b

> E S

f e r E r I S O u E E ' t a

(

V a d h 2 e g r e t R t

/

m c

n O

T C

A a

r

e. T vi e d a

a 3 5 0 0 f ec r

E A r 2 2 2 5 R r 1 s i t 0 3 n n 1

T e n e

5 w I r 5 8 ?t a t

E 0 E T a R 3 5 7 0 d 7 L

P O F A 4 5 3 3 1 A T 0 e -

T 5 h 5 M 0 d t O e I

1 n )

t b; T

A A . o r ia 3 5 5 0 r S t I

D l ts *

(

d a 3 5 2 5 o A E i n r 1 1 -

H m a h E T p e r

r t R 5 s I I

F 3 E u d O t a d e

F e l t

S g a r e t 0 E a r ep t i 5 I

e m ta f 5 T e t e d ev R v a T a 5 0 2 0 F E A t r e 0 1 2 O r r 1 - r o i u

T R c C F h t

E F S e

F S tt E E 4 m

+ o Gt T 5 0 2 5 f r

0 A 5 5 3 4 1 d a

H d e C e r T ) t O b ia 5 0 5 5 e M l F 0 3 3 u

-

d 1 1 l 1 t ( a a 0 f r v e r 0 r I -

3 u b t t a a d g r e s a t a A

r e

v T

mi d a

a r

0 5

0 0

7 2

0 1

d a

m r f t a n d U

is 1 1

- - l

2 ) 2 a ) E 0 e n 0 n n G 3 s fo 8 i o i A

0 r 0 d i a l D

l - e t - u t t a t a C v a C t a e t v i s t i t M e A r e n n e g

> n M ~

e t a ie t -

ie d t a r la o i 2

a l T r l r A _

M P ( O P ( O W H TZ ll1Iill ll l l1I'l 1

TABLE 5-6 COMPARISON OF SAN ONOFRE UNIT 3 SURVEILLANCE MATERIAL 30 FT-LB TRAflSITI0ff TEMPERATUR EllIFTS AND UPPER SilELF Ef4ERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDI 30 ft-lb Transition Temp. Shift Upper Shelf Enerav Decrease R.G. 1.99 Rev. 2 Measured R.G. 1.99 Rev. 2 F (Predicted) Measured (Predicted) 10jgence /*F) (%) (%)

' Material Capsule - n/cm 2 (*F) 55.0 18 17.0 Plate C-6802-1 W-97 0.8 31 (Transverse Orientation) 50.0 18 24.0 W-97 0.8 31 Plate C-6802-1 (Longitudinal Orientation) 32.0 18 15.0 cn Weld Metal W-97 0.8 27 l "'

1 45.0 -- 13.0

~

ilAZ Metal W-97 0.8 ---

TABLE 5-7 TENSILE PROPERTIES FOR SAN ON0FRE UNIT 3 REACTOR VESSEL SURVEILLANCE MATERIAL l9 n/cm2 (E > 1.0 MeV)

IRRADIATED AT 550*F TO 0.8 x 10 Uniform Total Reduction Test O.2% YieldStrength Ultimate Fracture Ioad Fracture Strese Fracture Strength Elongation Elongation in Area Sample Temp. Strength (%) (%) (%)

(kei) (kio) (kei) (kei)

Material Number (*F) (kei) 208.6 63.2 13.0 23.3 70 2KE 74 71.3 91.7 3.10 Plate C6802-1 215.6 60.1 10.5 23.0 72 2KC 200 67.7 86.6 2.95 (Transverse) e0.1 86.6 3.00 202.0 61.1 10.5 22.2 70 2K3 550 185.1 58.1 12.0 24.5 69 74 67.2 87.6 2.85 HAZ 4JD 2.63 148.5 53.5 12.0 25.0 64 4JK 175 62.1 77.9 68 2.75 172.4 56.0 12.0 24.5 4JA 550 60.1 84.5 168.9 53.0 11.4 20.4 69 3JC 74 51.4 75.4 2.60 Weld 93.7 3.45 232.3 70.3 10.5 20.3 70 3JE 250 62.6 73.3 13.5 22.7 64 550 54.5 94.7 3.60 203.7 3J4 Y

f

TABLE 5-8 ROOM TEMPERATURE ROCKWELL B HARDNESS VALUES FOR lHE SAN ON0f RE UNIT 3 SURVElLLANCE MATERIALS 1RRADIAlED AT 550'f, FLUENCE 0.8 X 10 I9 n/cm2 (E > 1.0 MeV)

Specimen Identification Rockwell_B Hardness Averanc (1) Base Metal (Lono.)

15K 89 89 87 88.3 12K 87 87 85 86.3 12L 85 87 8'i 85.7 153 86 87 Bi 86.3 14Y 85 84 85 84.7 ISB 88 88 87 87.7 IIT 87 89 87 87.7 14E 09 91 90 90.0 14M 87 89 88 88.0 147 90 91 92 91.0 14A 89 90 89 89.3 113 89 89 90 89.3 (2) Base Metal (Trans.)

245 84 89 88 87.0 21A 83 89 81 84.3 223 73 84 87 81.3 23K 88 86 81 85.0 23M 84 89 88 87.0 23L 86 89 86 87.0 22B 77 87 89 84.3 25V 90 91 91 90.7 247 87 91 90 89.3 25J 83 89 87 86.3 25C 81 85 88 84.7 25L 86 88 89 87.7 5-15

TABLE 5-8 (cont)

ROOM TEMPERATURE ROCKWELL D HARDNESS VALUES FOR THE SAN ONOFRE UNIT 3 SURVEILLANCE MATERIALS IRRADIATED AT 550'f.

19 FLUENCE 0.8 X 10 n/cm2 (E > 1.0 MeV) joecimen identification Rockwell B Hardness Average l (3) Weld Metal _ l 367 88 91 95 91.3 34P 88 83 84 85.0 33L 89 87 91 89.0 377 91 94 93 92.7 37U 92 92 93 92.3 37B 93 95 94 94.0 371 91 94 94 93.7-310- 81 81 82 81.3 330 90 ' 85 87 87.3 .

365 92 94 93 93.0 ,

331 90 90 90 90.0 363 87 89 87 87.7 (4) HAZ Metal 437 89 91 88 89.3 42L 88 86 84 86.0 41C 87 89 86 87.3 45M 89 89 89 89.0 434 92 92 89 91,0 450 88 87 85 86.7 46T 86 88 84 86.0 464 88 89 88 88.3 45P 89 89 87 '88.3 47E 86 87 86 86.3 43P 94 95 93 94.0 441 89 90 90 89.7

5-16 e -giea. ; . wy. .-.., we p +--.3 m.y .m,ye.3.9g yypy -%n ni, 3 yw ..gg,yw,my y,g f.,.,,,gi -u. yp 9 pqy+. cp,,wgw egy,,.,g g ,g- .g ,wep--. py-%~re-- e

TABLE 5-9 PROJECTED EOL (32 EfPY) RTND7, RTp73 AND USE FOR THE SAN ONOFRE UNIT 3 SVRVEILLANCE MATERIAL CONTAINED IN CAPSVLE W-97 1 ,

I R.G. 1.99(l*2) R.G. 1.99(3)

REV. 2 REV. ?

CURRENT PTS (l) PROPOSED PTS (l) 32 EFPY 32 EFPY l RVLE RTPTS RVLE RTPTS CALCVLATED CALCULATED l SURVEILLANCE AT 32 EfPY AT 32 EfPY RTNDi USE MATERIAL ('F) ('f) (*f) (ft-lb) 4 PLATE 127 120 120 71 C-6802-1 (116)

, (Transverse Orientation) i WELD 23 62 46 63 METAL (39)

(1) The fluence value used for the above calculations was 4.2 x 1019 n/cm 2 (E > 1.0 MeV). This is the projected EOL (32 EFPY) peak vessel fluence.

(2) Values in () calculated using 1/4 T (32 EfPY) peak vessel fluence. _

(3) Calculated using.1/4 T (32 EfPY) peak vessel fluence.

5-17

pr ,r, s

\ '

g .

tt .,

, n .

. , , W:3 ;4

/M -:.. a -).

f 4

BLOCK C214 8

-' i

a ..

\

59 j [' y,sJ e.

.w .

a a.

BLOCK C242 i

^"

[J '"

9- $N6

~

gue p&,s3%v p$,1;#gN M j,

. .i Y

.st wW 4 tp$I Q(f%@. f"%)f4

j Wg%

,; a

i ' 3 o"

( k N' 4

4M.MAQiv. ; @?M g,

.e

,1;[f d.

?

";a L .... a n..t un%

block C273 Figure 5-1. Appearance of the Thermal Monitors in the Various Quartz Tubes.

5-18

('C)

-100 -50 0 50 100 150 200 250 i i i i i i i yi 100 - %- -

O o -

g 80

~

60 e op .

5 40 - o -

2 20 - a 2 -

3

- 2. 5 100 i i i i i i i i 5

80 - o

o w -

LO

- 8 0 * -

1. 5 g 60 2

_: 40 -

ageF d_ .. -

1. 0 -

3 20 0.5 2

160 i i i i i i i i 2W 140 _

120 o 160

- o e 100 -

_o 120 5 80 _

g 16 Ft-lb _

& C 0

g 60 2 -

80 3[2, 40 -

3- -

A 55'F - 40 20 1 ' ' i I O 0

- 200 -100 0 100 200 300 400 500 Temperature ('F)

A Unirradiated Data from FSAR o Unirradiated Oata from Surveillance Program o IrradiateC at 550*F to 0.8 x 10 19 n/cm 2

Figure 5-2. Charpy V-Notch Impact Properties for the San Onofre Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (Transverse Orientation) 5-19

(oC)

-105 - 50 0 50 100 150 200 250 i i i ' ' '

i 2'

100 o

g 80 - . -

g 60 -

p 6i 40 - 2 -

20 2 y /* _

0 100 , i i i i i i i 2. 5 e

2 S0 -

e 20

= o I o.. M. i'!

i 20 0.5 25 ' ' ' '

0 0 160 i i i i i i 1 i

~

140 -

120 160 3 100 -

o

~

b go _

o 22 Ft-lb 5 B

b 60 80 5 _

45*F -

. g 50'F -* ** -

40 20 -

2 i ' ' ' '

0 0

-200 -100 0 100 200 300 400 500 Temperature ('F) o Unirradiated Data (Typical figures 5-3 through 5-5)

Irradiated at 550'F to 0.8 x 10 19 n/cm 2

(Typical Figures 5-3 through 5-5)

Figtre 5-3. Charpy V-Notch Impact Properties for the San Onofre Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (Longitudinal Orientation) 5-20

1 l

1

(*C) l

-100 - 50 0 50 100 150 200 250 l I 6 I I i i i l 3

100 -

g 80

, f 4 60 - -

l b fG -

OO 0 -

20 - 0 0 -

1

'^ ' ' ' ' i 0

100 i ' i ' ' i ' ' 2. 5 2

3 80 - o "

c -

2.0 E n 8

- 60 -

1.57 40 -

['__ 37 e p -

1.0 3 0.5 h 20 '

0 lo I I I I 0

160 i i i i i i i i

~

140 120 160 o

12 ft-lb 3 103 -

o s o o

.v n1 -

120

- 80 -

,. o ;

~

h60 -

25* F - u-I -

80 3 ,

40 -

o 40 32* F -

20 -

I o/[, ' ' ' ' I 0

0

-200 -100 0 100 200 300 400 500 Temperature ( 'F)

Figure 5-4. Charpy V-Notch Impact Properties for San Onofre Unit 3 Reactor vessel Surveillance Weld Metal 5-21

('C)

-100 -50 0 50 100 150 200 250 i i i i i i i !

'3 100 - -

80 -

o. ,

g 60 - o -

540 -

28

~

2 N -

2 ,

V -

l 0 '\ 'l I i ! I 100 i i i ! , , , i 2.5 e 80 o

o -

2.0 G

~g 60 -

o -

1. 5 )

d 40 2 o -_p'p -

1. 0 ~

3 20 - -

0.5 0 0 160 i i i i i ; i

~

140 -

120 - -

160 2 100 -

o il ft-Ib

~

80 - o h

e 60 -

o 80

" 40 --

2 70*F o **

N

45 F -

40 2,

20 0 I ' ' I !

0

- 200 -100 0 100 200 300 400 500 Temperature ( F)

Figure 5-5. Charpy V-Notch Impact Properties for the San Onofre Unit 3 Reactor Vessel Surveillance HAZ Metal 5-22

t

! c e

9i y~

g g DisM m

,- nan c...,.

e'

, l

} l c,

=

w t i{

.a 25l' 23K CIA 23L

( ,.e , , , - , :q -

1 8

8

, -4 i ;Jl! .

i 't (

4. - i .

., .,-.,. i

. 5 ,

an 6,w.skG

.n w

m sLSQ%pf s e v _m.

Q x. o. ,,j' i \

\, !

I _..

l <

-(

> ',}

n n.

r w i

1, g@ .t. 4 S

, a

'I h - .,

e

, s-

%,ggpy; -W ysamp i.g ,.

o,;.sv,

> w. . ..a: , , vr f , - ~ . - >i tmp; , c, ,,

j \ , , ;* , .

. , 4,t r4 wq

'y 7

p A ;y . ;

se,

.a. i . .,.,;i.(

- .. > s. . ,., 5 t

.sse ,

& [

'M A * ' ' M& '

223 2LC 2n 2w figure 5-6. Charpy V-Notch Impact Specimen f racture Surfaces for the San Onof re Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (Transverse Orientation)

E-23

- . - . _. .~ . - - - . - . - . - . . . _ _ . - . - . - . . -

1 I

3 i

i

)Asok% -

q';*p

, s

> r

-'*-W

< - .' t

+.

a4t %% wh l i t

m :;>, ~,

2 l

.z; ,;

I t

147 153 15K 113 l

vm m

c

); y %T 1- '

i- ')

L, r . r m rv & .a%,,, .j 3 ,, . ;3 use ,:!

4 g K.

.4 .. i :

y f < .

<- (

n n. m "

14Y 15B 14A 14E

, , p ,,

}-

'"e i kbi.,th. t, .,9 ?

, ;"r ;

ww 1

(? t. y 4 .3 11T 12E 12L 14M Figure 5-7. Charpy V-Notch Impact Specimen fracture Surfaces for the San Onofre Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (Longitudinal Orientation) ,

5-24

%-iz+- , . .s..ymy -

.__.,,,,-m,..

,,y y ,, , , --,,-,,,wm--w-w-- ---.-y,.

7 _ , - ..

{liveM ..

,7 e /d. c L._ m a, .4 Ud

. >.A..4 g . m.n. ~

cw%>e m f.:- -;** y ;

.,,..y.-

y ,~~g <%

.,4

- 1 /, ;L-

,q-

4w

^ -

t ua x _ _,

L> Jb*4 e..

363 33L 31D 3713

.. W ' . .

- , x a"

r !. R I 4'

( ,. 4 i

r a.:@v ;

g3 %- '

sa%2h w >+ a-;. . .

73 s- j%

hems ., E a l 1' _.

307 331 37U 371 g

)

ic v.- 1 j men

4,L g g n-mg.-

[6 ,, , , ,

f i2%#3pl '

., , %%Mk

'g i 3 :. . i hgg- 4 i 3

i. .

' i. ( -)

av.- -- & _._ -

305 34P 33D 377 Figure 5-8. Charpy Impact Specimen Fracture Surfaces for San Onofre Unit 3 Reactor Vessel Surveillance Weld Metal 5-25

i

}

) i '

,, / r~q l

a ggm vrN%W m nwpmnj e+v

\

Cic, M

\ g' j i.

, 3 .

+W-

441 42L 45C 43P l

g , - - -

. d j E g 4 ,, ,

.; 4

y T-?;!:i]

.O Wm N r

. ,-. e:,' l,hs

, -4 s

t 1-'"' -

404 410 ..

47E 46T

~

,,s

.w -

j y( Ql

ap 7 M{+ y# 3 , w , ~

~~, ,%

J kyy.!'.l& n $mq-EAAlILEJ -

.m ,

, , 3,: 4 . .,x

, it .

r i % . ..-.b e -

45M 437 45P 434 Figure 5-9, Charpy Impact Specimen Fracture Surfaces for San Onofre Unit 3 l Reactor Vessel Surveillance HAZ Metal 5-26

( C) 0 50 100- 150 200 250 300 1M i i i i i i ' - %0 110 Ultimate Tensile Strength -

700

'100 -

p

90 -

__ /

2 g- 600g m % -

g -

500 ~ ,

400 l

50 o 0. 2 % Yield Strength g i i i i l' -

E

, Code: .,

l Open Psmts - Unirradiated R 19 2 Closed Points -Irradlated at 500 F to 0.8 x 10 n/cm l b, i i i i i i i-

^

70 b _

0 -

c;. -

2'

~

Reduction in Area

. 50

~ 40 - -

X 2 Total Elongation 2 b

~

ag _

a 7

h -

2

- ~

_ ~

10 -

0 Uni, form Elongation ,

0 100 200 300 400 500 600 Temperature ( F)

Figure 5-10. Tensile Properties for the San Onofre Unit 3 Reactor Vessel Intermediate Shell Plate C-6802-1 (Transverse Orientation) 5-27

- ___=-_-____-__-_-__--_--_-_________--_________________--___:

( C) 0 50 100 150 200 250 300 120 ' ' ' ' ' i l -

800 110 Ultimate Tensile Strength -

700 100 -

e

.: 90 -

g ^- 600 80 -

~

5 ~

@ 70 o-2 l M

60 -

g; -

8- 400 50 A

o 0. 2 % Yleid Strength -

300 I I ,

40 Code:

Open Points - Unirradiated 19 2 Closed Points -Irradiated at 550 F to 0.8 x 10 n/cm 80 i i i i I i i 60 Reduction in Area o e

m -

E 40 5 30 -

2 Total Elongation -

5 20 -

b- i$ -

. m 10 -

j a-0 -

Uniform Elongation -

I i l I I 0 100 200 300 400 500 600 Temperature ( F)

Figure 5-11, Tensile Properties for the San Onofre Unit 3 Reactor Vessel Weld Metal 5-28

l

(*C) 0 50 100 150 200 250 300-i i- i i i i '- 800' 110 -

100 -- -

700

- 90 - Ultimate Tensile Strength 80 -

?

E0 7 -

5003

[

60 -

o 400 3,P A 50 -

2 0,2 % Yield Strength

' l '

300 40 I I i

Code:

Open. Points - Unirradiated Closed Points - Irradiated at 550*F to 0,8 x 1019.n/cm 2 80 i i i i i i i 70- -

Reduction in Area -

= ~

60 - -

g 50 - -

[ 40 - -

5 30 - Total Elongation -

~

20 -

e Q j-10 -

F n- p'-

0 -

2[ Uniform Elongation -

1 I I i 1-0 '100 200 300 400 500 600 Temperature ( F)

Figure 5-12. Tensile Properties for San Onofre Unit 3 Reactor Vessel HAZ Metal 5-29

y.3 .

e wg - ,>6 -

,x_ + ,.,c Q uirk ., u. a m

.:h+

Q. ,

, -a

,, w. ,::.

Specimen 2KE 74*F c'+

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> se

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?"9? Y ._

Wusi . ,;,- .WQ _

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s

%N w TWSME?b2.%~

e - e,-m \ , , . ,

.>co,. -gg.y,,y,,u3,

._q

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l -~m ; l -- s p i ::

-6 T.

.i 4 -

3 "

j *No. C 0058 D .

m; -: +\

3 - 4 1 c M y.. ;

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<; C4,,-:. =. $ hh; ce Specimen 3JC 74'F

+

vm ;

t d f,,

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)

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Specimen 3J4 550*F Figure 5-14. Fractured Tensile Specimens from the San Onofre Unit 3 Reactor Vessel Weld Metal i 5-31

T 1h s

?

h C305R <3-

] ,

.; , ,, a . [ = 0 I6; Yi BU 4 " 2):!dM Q 1:

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p a ,.

gc im --

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'Q Specimen 4JA 550*F Figure 5-15. Fractured Tensile Specimens from San Onofre Unit 3 Reactor Vessel HAZ Metal i

5-32

120

~

100-G-

i 80-w

% 60-w w

@ 40- ,

l F- l SPEC 2KE l 20-550 F 0 . . . i

-0 0.05 0.1 0.15 0.2 0.25 TRUE STRAIN, IN/IN 100 90-

_ 80-G x 70- ,_,

w 60-m 5 50-w 40-D 30-20-SPEC 2KC 10] - 200 F 0' . , , ,

0 0.05 0.1 -0.15 0.2 0.25 TRUE STRAIN, IN/IN Figure 5-16. True Stress-Strain Curves for the San Onofre Unit 3 Reactor Vessel Shell Plate C-6802 I Tension Specimens 2KE and 2KC 5-33

100,-

90-80-G5 x 70-N w

60-

.$ 50-

.w 40- :

a K 30-20~ SPEC 2K3

.19 550 F G , , , ,

0 0.05 0.1 0.15 0.2 0.25 TRUE STRAIN, IN!!N 1 t

figure 5-17. True Stress-Strain Curve for the San Onofre Unit 3 Reactor Vessel Shell Plate C-6802-1 Tension Specimen 2K3 5-34

i 120 100-G x . - '

m -,

w 1 w

% 60-m g-

@ 40-SPEC 3JC

.20_

74 F G , ,

0 0.05 0.1 0.'15 0'2 0.25 TRUE STRAIN, IN/lN i.

120-e 100-G

. 80-m m

w-fm 60-

-w '

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120 7

100-li)

m. 80-w i p 60-m w

@ ' 40-F-

SPEC 3J4 2&

550 F 0 . . . i

-- O 0.05 0.1~ 0.15 0.2 0.25 TRUE STRAIN, IN/IN l

l Figure 5-19. True Stress-Strain Curve for the San Onofre Unit 3 Reactor Vessel Weld Metal Tension Specimen 3J4 l- 5-36 t

l l

l

120 1 100- ,_

a_

w 80-p 60- f f

v>

uJ

@ 40-F-

SPEC 4JA 20- 550 F 0 0.'05 0'1 0.'15 0.2 0.'25 0.3 TRUE STRAIN, IN/IN Figure 5-20. True Stress-Strain Curve for the San Onofre Unit 3 Reactor Vessel HAZ Metal Tension Specimen 4JA 5-37

100 0- - 3 m

x 60q Lj 50-E 40-Di 30-20- SPEC 2KE 10- 74 F 0- . . .

0.25 O 0.05 0.1 0.15 0.'2 STRAIN, IN/lN 90 80-70-g 60-x

,50-m y40-E 30-20-SPEC 2KC 10~l 200 F

'l 0i . > > i 0 0.05 0.1 0.15 0.2 0.25 STRAIN, IN/lN Figure 5-21. Engineering Stress-Strain Curves for the San Onofre Unit 3 Reactor Vessel Shell Plate C-6802-1 Tension Specimens 2KE and 2KC 5-38

90 80-70q m 60 x i

,509 m

O 40 m 30-20~ SPEC 2K3 10-550 F 0 i i i i 0 0.05 0.1 0.15 0.2 0.25 STRAIN, IN/lN Figure 5-22. Engineering Stress-Strain Curve for the San Onofre Unit 3 Reactor Vessel She'll Plate C-6802-1 Tension Specimen 2K3 5-39

80 '

704 l

60 '

$ 50 d vi 40

$ 30 -

tn 20-SPEC 3JC 10-74 F 0

0 0.b5 0.' 1 0. 1' 5 0.'2 0.25 STRAIN, IN/lN 100 90-80-70' en x 60-y 501 E 40 d G

30- >

20-SPEC 3JE 10 ] 250 F 0! . . . .

0 0.05 0.1 0.15 0.2 0.25 STRAlN, IN/lN Figure 5-23. Engineering Stress-Strain Curves for the San Onofre Unit 3 Reactor Vessel Weld Metal Tension Specimens 3JC and 3JE 5-40

Y 100 90-80- .

70-

_v) x 60-p 501 E 40 G

30 !

~

SPEC 3J4 10- 550 F 0 . . . <

0 0.05 0.1 0.15 0.2 0.25 STRAIN, IN/lN Figure 5-24. Engineering Stress-Strain Curve for the San Onofre Unit 3 Reactor Vessel Weld Metal Tension Specimen 3J4 !

1 5-41

I 90 80-70-g60- [

M

,50-m 40q m 30-201 J SPEC 4JA 10 l 550 F 0 i i i 0 0.05 0.1 0.15 0.2 0.25 STRAIN, IN/lN Figure 5-25. Engineering Stress-Strain Curve for the San Onofre Unit 3 Reactor Vessel HAZ Metal Tension Specimen 4JA 5-42 l l

SECTION 6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of 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 material property changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve cata to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall .

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853,

" Analysis and Interpretation of Light Water Reactor Surveillance Results,"

recommends reporting displacements per iron atom (dpa) along with fluence 6-1

(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," The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure ves:;el wall has already been promulgated in Revision 2 to the Regulatory Guide l 1.99, " Radiation Damage to Reactor Vessel Materials." j This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in the surveillance capsule irradiated at the location attached to the vessel wall at 97' (W-97).- Fast neutron exposure parameters in terms of neutron fluence (E >

1.0 MeV), neutron fluence (E > 0.1 Mov), 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. 1 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. Six irradiation capsules attached to the vessel wall are included in the reactor design to constitute the reactor vessel surveillance program. The

! capsules are located at azimuthal angles of 83*, 97*, 104*, 263*, 277*, and

!' 284* relative to the core cardinal- axes as shown in Figure 4-1.

A view of a surveillance capsule assembly is shown in Figure 4-2. A total of 7 stainless steel specimen containers hold the charpy and tensile monitors. The assembly is positioned axially centered on the core midplane, thus spanning the central part of the active fuel zone.

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 core l barrel 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.

I In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two transport calculations were carried out. j I

These represent two fuel loading geometries for the San Onofre Unit 3 plant as follows:

1. Cycles 1 through 3: A calculation was performed with a power distribution averaged over these three cycles which all have fresh fuel loaded on the periphery,

2. Cycle 4: A calculation was performed with the power distribution averaged over this single cycle which loaded some burned fuel on the periphery. This cycle was assumed to be typical of future cycles and was used to extrapolate exposures to 32 EFPY, These calculations were combined to produce average relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters (p(E > 1.0 Mev), p(E > 0.1 Mev), and dpa) through the vessel wall integrated over time. 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/p(E > 1.0 MeV), within the pressure vessel geometry. The relative radial gradient information was required to l

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. In addition differences between the two geometries in calculated dosimeter reaction rates at the center of the surveillance capsule were used to correct the flux history to provide corrected measured reaction rates.

6-3

_ _ ~

The transport calculations for the reactor model summarized in Figure 4-2 were carried out in R, O geometry using the DOT two-dimensional discrete ordinates codellll and the SAILOR cross-section libraryll23 The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications, in these analyses anisotopic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an S8 order of angular quadrature.

Selected results from the neutron transport analyses performed for the San Onofre Unit 3 reactor are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall.

In Table 6-1, the calculated exposure parameters [p (E > 1.0 MeV), ((E

> 0.1 MeV), and dpa].are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with

analysis. The design basis data derived from the forward calculation are 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 cycle 1 plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad / base metal L.terface; and, thus, . represent the maximum exposure levels of the vessel wall itself.

Radial gradient information for neutron flux (E > 1.0 MeV), neutron flux (E >

0.1 MeV), and iron 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, 6-4

For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by:

= p(221.45, 45') F (226.93, 45')

41/4T(45')

where: = Projected neutron flux at the 1/4T position on

@l/4T(45*)

the 45' azimuth

& (221.45,45') = Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth.

F (226.93, 45*) = Relative radial distribution function from Table 6-3.

Similar expressions apply for exposure parameters in terms of & (E > 0.1 MeV) and dpa/sec.

The D0T calculations were carried out for a typical octant of the reactor. For the fuel power distributions calculated for cycles 1 through 4, only negligible differences are noted between the different octants, so the use of a single octant to represent the entire reactor geometry is accurate. The model did, however, include wall surveillance capsules at both 7 degrees and 14 degrees with respect to the cardinal axes and these capsules are not included in all octants. This will result in a small correction to the calculated fluence to the reactor vessel at angles near 7 and 14 degrees for octants without the capsules.

6.3 Neutron Dosimetry The passive neutron sensors included in the San Onofre Unit 3 surveillance program are listed in Table 6-6. (The sulfur monitor has been omitted because the delay between irradiation and receipt at the laboratory precluded analysis. 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 [d (E > 1.0 Mev), p (E > 0.1 MeV),dpa].

6-5

The relative locations of the' neutron sensors within the capsules are shown in Figure 4-1. The sets of bare and cadmium shielded monitors were placed in holes drilled in flux monitor housing blocks at three axial levels within the capsules. The monitors' are all located at the same radial distance from the Core.

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 mon: tors may be derived from the' activation measurements only if the irradiation parameters are well known.

In particular, the following variables are of interest:

1 o The specific activity of each monitor. !

o The operating history of the reactor.

o The energy response of the monitor.

o The neutron energy spectrum at the monitor location.

o The physical characteristics of the monitor.

The specific activity of each of the neutron monitors was determined using established' ASTM procedures (13 through 25]. Following sample preparation l and weighing, the activity of each monitor was determined by means of a l lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the San Onofre Unit 3 reactor during cycle I was obtained from NUREG-0020, l " Lie.ensed Operating Reactors Status Summary Report" for the applicable period, i

The irradiation history applicable to capsule W-97 is given in Table 6-7.

Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7.

i 6-6 l

In the analysis of the uranium monitors, it was found that the cadmium cover material had become intimately mixed with the uranium and was recovered as a powder. After counting, the mixture was heated in a furnace to attempt to volatilize the cadmium and enable an accurate uranium weight to be obtained.

Unfortunately, the mixture reacted with the crucible and no direct weight could be measured. The~ uranium was then dissolved and a weight obtained. This weight was used to obtain the specific activities in Table 6-8. However, these values were found to be significantly different when ratioed to the other .

l monitors when compared to data from a similar plant. It was concluded that the l l

uranium weight was not reliable. This conclusion is also supported by the variability in weights and ratio of uranium to remaining (not volatilized) fission products. Unfortunately no "as-built" weights were available. The bare uranium values are presumably more reliable but also were not used because more than 50% of the fissions are estimated to have occurred in the U-235 impurity in the depleted uranium. j It was also found that the initial analysis of the Co-60 was not consistent.

It was decided to measure the amount of Co-59 in each of the Co-Al wires and the content was found to average 0.7% which is very different from the specified value of 0.17%. Reaction rates for Co were based on the measured

concentration for each wire.

1 i

I Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code (26]. The FERRET approach used the measured reaction rate data and the calculated neutron i energy spectrum at the the center of the surveillance capsule as input and l proceeded to adjust a priori (calculated) group fluxes to produce a best fit (ir 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 L the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly

- related to the flux p by some response matrix A:

! 6-7 l

(s,a) (s) (a) f =2 A. p 9 19 9 where i indexes the measured values belonging to i single data set s, 9 designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, R -I o p i g ig g relates a set of measured reaction rates Rj to a single spectrum p g by the multigroup cross section ojg. (In this case, FERRET also adjusts the cross-sections.) The lognormal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.

In the FERRET analysis of the dosimetry data, the continuous quantities (i.e.,

fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-Il code (27). This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed to the group scheme used in FERRET.

The cross-sections were also collapsed into the 53 energy-group structure using

-SAND 11 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:

6-8 -

M gg,=Rf+R g R,P g gg, where _ Rg specifies an overall fractional normalization uncertainty (i.e.,

complete correlation) fer the corresponding set of values. The fractional uncertainties Rg specify additional random uncertainties for group g that are correlated with a correlation matrix:

Pgg, = (1 - 0) 6gg, + 0 exp ( )

The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 7 (0 specifies the strength of the latter term). '

for the a priori calculated fluxes, a short-range correlation of 7 - 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E.

Maerker[28]. Maerker's results are closely duplicated when 7 - 6. For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.

Results of the FERRET evaluation of the capsule W-97 dosimetry are given in Table 6-9. The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 8.00 x 10 18 n/cm2 (E > 1.0 MeV) with an associated uncertainty of 11%. Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the adjusted spectrum are provided in Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.

6-9

A summary of the measured and calculated neutron exposure of capsule W-97 is presented in Table 6-12. The comparison between calculation and measurement indicates that the calculation is biased low by 20-24% for all fast neutron exposure parameters listed. Vessel exposure extrapolations have been normalized to include this bias.

Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (4.33 EFPY) exposure derived from the capsule W-97 measurements, projections are also provided for an exposure period of 20 EFPY and to end of vessel design life (32 EFPY). The calculated exposure rates based on the Cycle 4 fuel loading were used to perform projections beyond the end of cycle 4.

In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the San Onofre Unit 3 reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14. In order to access RTNDT vs. fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations p' (1/4T) = 4 (Surface) {dpa ( face))

3 d' (3/4T) - 4 (Surface) { ja uf e 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 San Onofre Unit 3 surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules. 6-10

TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER IRRADIATION CYCLE TIME & (E > 1.0 MeV) 6 (E.> 0.1 MeV) dpa/sec (EFPY) (n/cm2 -sec) (n/cm2 -sec) __ 7* Capsule. Cycle 1-3 2.807 5.17 x 10 10 9.58 x 10 10 7.31 x 10-II Cycle 4 1.520 3.73 x 10 10 6.87 x-10 10 5.28 x.10-Il

 ?
 ~~

Cycle 1-4 4.328 Average -4.66 x 10 10 8.63 x 10 10 6.60 x 10-II 14' Capsule ! Cycle 1-3 2.807 3.72 x 10 10 6.86 x 1010 5.29 x 10-11 7 Cycle 4 1.520 2.61 x 10 10 4.79 x 10 10 3.73 x 10-II I l Cycle 1-4 4.328 1 10 ! Average 3.33 x 10 6.13 x 10 10 4.74 x 10-11 l l l l

TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE-METAL INTERFACE 0' 15' 30' 45' p( E > 1. 0Me.yl 2 (n/cm -sec) Cycle 1-3 4.20 x 10 10 2.66 x 1010 2.37 x 10 10 1.78 x 10 10 Cycle 4 3.22 x 1010 1.90 x 10 10 1.61 x 1010 1.22 x 1010 Cycle 1-4 Average 3.86 x'1010 2,39 x 10 10 2.10 x 10 10 1.58 x 10 10 6(E > 0.1Mevi 2 (n/cm -sec) Cycle 1-3 9.03 x 1010 5.82 x 1010 5.07 x 10 10 3.79 x 1010 Cycle 4 6.88 x 1010 4.13 x 1010 3.43 x 1010 2.57 x 1010

  -Cycle 1-4 Average     8.28 x 1010     5.23 x 1010       4.49 x 10 10     3.36 x 10 10 doa/sec Cycle 1-3           6.20 x 10-11    3.97 x 10-11      3.51 x 10-II     2.66 x 10-Il l.

Cycle 4 4.75 x 10-Il 2.84-x 10-11 2.39 x 10-11 1.82 x 10-11 l Cycle 1-4 Average 5.69 x 10-Il 3.57 x 10-11 3.11 x 10-11 2.37 x 10-Il l l l- (a) Radies of clad / base metal interface at 221.45 cm. 6-12

m - TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS Of NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL Radius fraction (cm) of Vessel O' 15' 30' 45' 221.45(l) 0.000 1.0000 1.0000 1.0000 1.0000 221.73 0.013 0.9825 0.9808 0.9824 0.9821 222.38 0.042 0.9289 0.9240 0.9283 0.9287 223.21 0.080 0.8527 0.8457 0.8517 0.8528 224.13 0.122 0.7667 0.7595 0.7655 0.7678 225.05 0.164 0.6842 0.6779 0.6828 0.6861 225.97 0.206 0.6074 0.6026 0.6062 0.6100 226.93 0.250 0.5345 0.5313 0.5336 0.5376 228.03 0.300 0.4607 0.4589 0.4o00 0.4643 229.22 0.355 0.3907 0.3901 0.3899 0.3948 230.41 0.409 0.3302 0.3309 0.3296 0.3345 231.61 0.464 0.2784 0.2801 0.2781 0.2829 232.80 0.510 0.2343 0.2366 0.2341 0.2387 233,99 0.572 0.1967 0.1994 0.1968 0,2011 235.19 0.627 0.1647 0.1678 0.1650 0.1692 s 236.38 0.681 0.1375 0.1408 0.1381 0.1420 237.57 0.736 0.1145 0.1178 0.1153 0.1190 238.77 0,791 0.0948 0.0982 0.0960 0.0994 239.96 0.845 0.0779 0.0814 0.0794 0.0828 241.15 0.899 0.0631 0.0668 0.0652 0.0686 242.25 0.949 0.0508 0.0548 0.0535 0.0572 243.06 0.986 0.0421 0.0464 0.0456 0.0495 243.36(2) 1.000 0.0384 3.0431 0.0424 0.0464 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-13

TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS Of NEUTRON FLUX (E > 0.1 MeV) WITHIN THE PRESSURE VESSEL WALL Radius Fraction (cm) of Vessel O' 15' 30' 45' 221.45(l) 0.000 1.000 1.000 1.000 1.000 221.73 0.013 1.008 1.008 1.008 1.009 222.38 0.042 1.000 1.005 1.005 1.009 223.21 0.080 0.989 0.989 0.988 0.996 224.13 0.122 0.960 0.961 0.959 0.970 225.05 0.164 0.924 0.927 0.923 0.937 225.97 0.206 0.884 0.890 0.884 0.901 226.93 0.250 0.841 0.850 0.841 0.861 228.03 0.300 0.791 0.803 0.792 0.814 229.22 0.355 0.737 0.751 0.739 0.763 230.41 0.409 0.683 0.700 0.686 0.713 231.61 0.464 0,630 0.649 0.635 0.663 232.80 0.518 0,579 0.599 0.585 0.614 233.99 0.572 0.528 0.550 0.536 0.566 235.19 0.627 0.479 0.503 0.489 0.519 236.38 0.681 0.432 0.456 0.444 0.474 237.57 0.736 0.385 0.411 0.400 0.430 238.77 0.791 0.340 0.366 0.357 0.387 239.96 0.845 0.296 0.323 0.315 0.345 241.15 0.899 .. 0.251 0.279 0.274 0.304 242.25 0.949 0.209 0.238 0.235 0.266 243.06 0.986 0.175 0.206 0.206 0.238 243.36(2) 1.000 0.161 0.192 0.193 0.226 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-14 l

l

TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL 1 Radius fraction (cm) of Vessel 0' 15* 30' 45' 221.45(l) 0.000 1.000 1.000 1.000 1.000 221.45 0.000 1.000 1.000 1.000 1.000 221.73 0.013 0.986 0.985 0.986 0.986 222.38 0.042 0.942 0.938 0.941 0.941 223.21 0.080 0.879 0.874 0.877 0.879 224.13 0.122 0.808 0.803 0.865 0.808 l 225.05 0.164 0.738 0.735 0.736 0.740 225.97- 0.206 0.673 0.672 0.671 0.676 226.93 0.250 0.611 0.611 0.609 0.614 228.03 0.300 0.546 0.548 0.544 0.551 229.22 0.355 0.483 0.487 0.482 0.489 4 230.41 0.409 0.427 0.433 0.426 0.434 231.61 0.464 0.377 0.384 0.377 0.386 l 232.80 0.518 0.333 0.341 0.333 0.342 233.99 0.572 0.293 0.302 0.294 0.304 l j 235.19 0.627 0.258 0.267 0.260 0.269 l 236.38 0.681 0.226 0.236 0.228 0.238

237.57 0.736 0.196 0.207 0.200 0.210 l 238.77 0.791 0.170 0.180 0.174 0.184 239.96 0.845 0.145 0.156 0.151 0.161 l

l 241.15 0.899 0.121 0.133 0.129 0.139 242.25 0.949 0.100 0.112 0.109 0.120 l 243.06 0.986 0.0835 0.0960 0.0946 0.106 243.36(2) 1.000 0.0764 0.0896 0.0885 0.100 l i l NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius i

6-15

TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Weight Response Product Yield Monitor of Fraction Range Half l ife (%) Material Interest C.6917 E > 5 MeV 5.272 yrs Copper Cu63(n.o)Co60 Ti46(n,p)Sc 46 0.0778 E > 4 MeV 83.85 days Titanium Fe54(n,p)Mn54 0.0582 E > 2 MeV 312.2 days Iron NiS8(n p)CoS8 0.6830 E > 2 MeV 70.90 days Nickel U238(n,f)CsI37 1.0 E > 1 MeV 30.17 yrs 5.99 Uranium-238* U238(n,f)CsI37 1.0 E > 1 MeV 30.12 yrs 5.99 Uranium-238 CoS9(n,d)Co60 0.0070** 0.4ev>E> 0.015 MeV 5.272 yrs Cobal t-Al uminum* CoS9(n,0)Co60 0.0070** E > 0.015 MeV 5.272 yrs Cobalt-Aluminum

Denotes that monitor is cadmium shielded.

 ** Average value of 6 measurements. Measured cobalt concentration value for each wire was used to get activity per unit mass of cobalt.

TABLE 6-7 1RRADIAT10N HISTORY Of NEUTRON SrNSORS CONTAINED IN CAPSULE W-97 Irradiation Pj Irradiation Pj 1rradiation Pj Period (MWg ) Period (MWt ) Period (MWt ) 9 /1983 118958 12 /1985 0 3 /1988 2505525 10 /1983 728980 1 /1986 1137179 4 /1988 2323149 1049800 2 /1986 951789 5 /1988 0 11 /1983 12 /1983 1651300 3 /1986 1060286 6 /1988 0 415741 4 /1986 2313945 7 /1988 0 1 /1984 2 /1984 0 $ /1986 2503081 8 /1988 517640 3 /1984 1415800 6 /1986 2422034 9 /1988 2435963 4 /1984 2185194 7 /1986 2129940 10 /1988 2477912 5 /1984 1157057 8 /1986 2221739 11 /1988 2433112 6 /1984 755354 9 /1986 2001976 12 /1988 2469278 7 /1984 638154 10 /1986 610986 1 /1989 2300423 8 /1984 1643144 11 /1986 2032603 2 /1989 2238355 9 /1984 2405823 12 /1986 1873034 3 /1989 2509434 10 /1984 2169606 1 /1987 100514 4 /1989 1454279 11 /1984 0 2 /1987 0 5 /1989 2516439 12 /1984 1961635 3 /1987 1431146 6 /1989 2252528 1 /1985 2106963 4 /1987 2438732 7 /1989 1783842 2 /1985 0 5 /1987 2459096 8 /1989 2480029 3 /1985 752146 6 /1987 2043680 9 /1989 2424722 4 /1985 1798082 7 /1987 2468870 10 /1989 2465612 5 /1985 2340249 8 /1987 2505525 11 /1989 2420731 6 /1985 1583225 9 /1987 2431401 12 /1989 2492573 NOTE: Reference Power = 3390 MWt 6-17

i TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated Reaction Monitor and Activity Activity Rate Axial location (dis /see-om) (dis /sec-am) (RPS/ NUCLEUS) [y_53 (n.o) Co-60 Top 2.13 x 10 5 5.89 x 10 5 Middle 2.45 x 105 6.78 x 105 Bottom 2.11 x 10 5 5.84 x 105 Average 2.23 x 10 5 6.17 x 105 9.41 x 10~17 Fe-54(n.0) Mn-54 Top 2.01 x 10 6 4.78 x 106 Middle 2.00 x 106 4.75 x 106 Bottom 1.86 x 10 6 4.42 x 106 Average 1.96 x 10 6 4.65 x 106 7.43 x 10-15 Ni-58 (n.0) Co-58 Top 4.83 x 10 6 7.11 x 10 7 Middle 4.61 x 10 6 6.79 x 10 7 Bottom 4.37 x 10 6 6.44 x 10 7 Average 4.60 x 10 6 6.78 x 10 7 9.68 x 10-15

  .Ti-46 (n.0) Sc-46 Top                                       1,49 x 10 5             1.52 x 10 6 Middle                                    1.41 x 10 5             1.44 x 106 Bottom                                    1.32 x 10 5             1.35 x 10 6 Average                                   1.41 x 10 5             1.44 x 10 6                                            1.38 x 10-15 6-18

TABLE 6 8 MEASVRED SENSOR ACTIVITIES AND REACTION RATES - cont'd Measured Saturated Reaction Monitor and Activity Activity Rate Axial location (dis /sec-om) (dis /sec-cil (RPS/ NUCLEUS) V-238 (n.fi Cs-137 (Cd) Top 5.36 x 105 5.85 x 10 6 Middle 4.47 x 10 5 4.88 x 10 6 Bottom 4.96 x 10 5 5.41 x 10 6 Average 4.93 x 10 5 5,33 x }o 6 3.55 x 10-I4 V-238 (n.fi Cs-137 (Bare) Top 8.60 x 10 5 9.40 x 10 6 Middle 7.74 x 10 5 8.45 x 10 6 Bottom 6.73 x 10 5 7.35 x 10 6 Average 7.69 x 10 5 8.40 x 10 6 5.54 x 10-I4 Co-59 (n.7) Co-60 Top 1.42 x 10 8 3.93 x 10 8 Middle 1.42 x 10 8 3.93 x 10 8 Bottom 1.02 x 10 3 2.82 x 10 8 Average 1.29 x 10 8 3.56 x 10 8 4.88 x 10-12 Co-59 (n.7) Co-60 (Cd) Top 1.70 x 10 7 4.70 x 10 7 Middle 1.84 x 10 7 5.09 x 10 7 Bottom 1.57 x 10 7 4.34 x 10 7 Average 1.70 x 10 7 4.71 x 10 7 6.68 x 10-13 6-19

TABLE 6-9

SUMMARY

OF NEUTRON DOSIMETRY RESULTS TIME AVERAGED EXPOSURE RATES 2

 & (E > 1.0 MeV) {n/cm -sec)                            5.86 x 1010                3 33g 2

4 (E > 0.1 MeV) {n/cm -sec) 1.14 x 10 11 18% dpa/sec 8.52 x 10-11 11% 2 d (E < 0.414 eV) {n/cm -sec} 1.47 x 10 ll 23% JNTEGRATED CAPSULE EXPOSURE 4 (E > 1.0 MeV) (n/cm ) 2 8.00 x 10 18 33g 2

   + (E > 0.1 MeV) (n/cm )                               1.56 x 1019                   18%

dpa 1.16 x 10-2 1 33g 2 t (E < 0.414 eV) (n/cm ) 2.01 x 10I9 23% NOTE: Total Irradiation Time 4.33 EFPY 6-20

TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVE!LLANCE CAPSULE CENTER Adjusted Reaction Measured Lalculation (jf Cu-63 (n.o) Co-60 9.41x10-17 9.40x10-17 1.01 Fe-54 (n,p) Hn-54 7.43x10-15 7.47x10-15 1.00 Ni-58 (n.p) Co-58 (Cd) 9.68x10-15 9.54x10-15 o,99 Ti-46 (n,p) Sc-46 1.38x10'15 1.40x10-15 1.01 Co-59 (n,1) Co-60 (Cd) 6.69x10-13 6.61x10-13 0.99 Co-59 (n,7) Co-60 4.8Bx10-12 4.56x10-12 0.93 6-21

TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SVRVElLLANCE CAPSVLE CENTER Energy AdjusgedFlux Energy Adjus}edflux Group (Mev) (n/cm sec) Group (Mev) (n/cm -sec) I 1.73x10 1 9.00x10 6 28 9.12x10-3 3.20x10 9 2 1.49x101 2.15x10 7 29 5.53x10-3 4.19x10 9 3 1.35x10 I 9.46x10 7 30 3.36x10-3 1.33x10 9 4 1.16x10 I 2.34x10 9 31 2.84x10-3 1.30x10 9 5 1.00x10I 5.51x10 3 32 2.40x10-3 1.30x10 9 6 8.61x10 0 9.78x10 8 33 2.04x10-3 3.78x109 7 7.41x10 0 2.43x10 9 34 1.23x10-3 3.62x10 9 8 6.07x10 0 3.45x10 9 35 7.49x10-4 3.52x10 9 9 4.97x10 0 6.15x10 9 36 4.54x10'4 3.52x109 10 3.68x100 5.95x109 37 2.75x10-4 3.94x10 9 11 2.87x10 0 9.86x10 9 38 1.67x10-4 5.30x10 9 12 2.23x10 0 9.21x109 39 1.0lx10'4 4.29x10 9 13 1.74x10 0 9.27x10 9 40 6.14x10-5 4.14x10 9 14 1.35x10 0 7.05x10 9 41 3.73x10-5 3.84x109 15 1.lix10 0 9.75x10 9 42 2.26x10-5 3.53x10 9 16 8.21x10-l 8.99x10 9 43 1.37x10-5 3.28x10 9 17 6.39x10-1 8.07x10 9 44 8.32x10-6 3.0lx10 9 18 4.98x10-I 5.84x10 9 45 5.04x10-6 2.68x10 9 19 3.88x10-1 6.6Bx10 9 46 3.06x10-6 2.43x109 20 3.02x10'I 9.85x109 47 1.86x10-6 2.18x109 2i 1.83x10-I 8.38x10 9 48 1.13x10-6 1.62x10 9 22 1.llx10-1 6.87x10 9 49 6.83x10-7 1.66x10 9 23 6.74x10-2 4.75x10 9 50 4.14x10-7 4.12x10 9 24 4.09x10-2 2.68x10 9 51 2.51x10-7 1.17x10 10 25 2.55x10-2 3.66x10 9 52 1.52x10-7 2.54x1010 9 9.24x10-8 1.06x10ll 26 1.99x10-2 1.73x10 53 27 1.50x10-2 2.17x109 NOTE: Tabulated energy levels represent the upper ener9y of each group. 6-22

TABLE 6-12 COMPARIS0ll 0F CALCULATED Af4D MEASURED EXPOSURE LEVELS FOR CAPSVLE W-97 C al cul at ed Masured [M f(E > 1.0 MeV) (n/cm ) 2 6.36 x 10 18 8.00 x 10 18 0.00 f(E > 0.1 MeV) {n/cm ) 2 1.18 x 10 19 1.56 x 1019 0.76 dpa 9.01 x 10-3 1.16 x 10-2 0.78 6-23

l TABLE 6-13 , NEUTR0ti EXPOSURE PROJECTIONS AT KEY LOCAT10t45 ON THE PRESSURE VESSEL CLAD / CASE METAL ll4TERFACE FOR SAN ONOFRE V!41T 3 A71MUTHAL ANfdl , 0.(a) 35 39 45' 4.33 EFPY 4(E>1.0 MeV) 6.63 x 10 18 4.11 x 10 18 3.61 x 1018 2.72 x 1018 2 (n/cm ) f(E>0.1 MeV) 1.49 x 1019 9.44 x 10 18 8.11 x 1018 6.07 x 1018 2 (n/cm ) dpa 0.0100 0.00628 0.00547 0.00416 20.0 EFPY f(E>1.0 MeV) 2.66 x 10 19 1.59 x 10 19 1.36 x 1019 1.03 x 10 19 2 (n/cm ) 4(E>0.1 MeV) 5.99 x 1019 3.64 x 10l9 3.05 x 1019 2.29 x 10I9 2 (n/cm ) dpa 0.0463 0.0290 0.0253 0.0192 32.0 EFPY f(E>l.0 MeV) 4.20 x 1019 2.50 x 10 l9 2.13 x 1019 1,61 x 10 19 2 (n/cm ) f(E>0.1 MeV) 9.44 x 10 19 5.71 x 10 19 4.77 x 1019 3.57 x 1019 2 (n/cm ) dpa 0.0634 0.0382 0.0324 0.0246 (a) Maximum point on the pressure vessel 6-24

       ,                                             TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/E00LDOWN EURVES 20 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE                                               dpa SLOPE 2

(n/cm ) (equivalent n/cm2 ) j Surface '1/4 T 3/4 T Surface 1/4 T 3/4 T 0* 2.66E+19 1.42E+19 2.91E+18 2.66E+19 1.63E+19 5.02E+18 15* 1.59E+19 8.46E+18 1.79E+18 1.59E+19 9.73E+18 3.18E+18 30* 1.36E+19 7.27E+18 1.50E+18 1.36E+19 8.30E+18 2.64E+18 45* 1.03E+19 5.53E+18 1.17E+18 1.03E+19 6.31E+18 2.09E+18 32 EFPY 4 f " C el FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 i* (n/cm ) (equivalent n/cm2 ) Surface 1/4 T- 3/4 T Surface 1/4 T 3/4 T 0* 4.20E+19 2.24E+19 4.58E+18 4.20E+19 2.56E+19 7.91E+18 15* 2.50E+19 1.33E+19 2.81E+18 2.50E+19 1.53E+19 4.99E+18 30* 2.13E+19 1.14E+19 2.35E+18 2.13E+19 1.30E+19 4.13E+18 45' l.61E+19 8.64E+18 1.83E+18 1.61E+19 9.87E+18 3.26E+18 i l l (a) Maximum point on the pressure vessel l

TABLE 6-15 I UPDATED LEAD FACTORS FOR SAN ON0fR2 UNIT 3 SURVEILLANCE CAPSVLES Capsul t lead Factor (a) 83 1.21 97 1.21 104 0.86 263 1.21 277 1.21 284 0.86 (a) Plant specific evaluation for fuel cycles 1 through 4. If future fuel loading patterns similar to Cycle 4 are used, the lead factors for capsules at 830 , 2630 , and 2770will decrease towards an asymptotic value of 1.16 and the capsules at 1040 and 2840 will decrease towards 0.81. 6-26

I SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE-The following removal schedule meets ASTM E185-82 and is recommended for future b capsules to be removed from the San Onofre Unit 3 reactor vessel-l Capsule Capsule Estimated Location Lead Removal Fluence

(deg.) Factor EFPY (b) (n/cm) 2 I

l 97.0 1.21 4.33 (a) 0.8 x 1019 (actual) 83.0 1.21 6.0 1.0 x 1019 l 263.0 1.21 15.0 2.4 x 10I9 277.0 1.21 32.0 5.1 x 1019 104.0. 0.86 Standby ---- 284.0 0.86 Standby ---- (a) Plant Specific Evaluation _ (b) Effective full power years from plant startup. l l l l 7-1 L

 ~- . , . , , .         . . . . - . - - , , - . , . , . . -                       -     ,   . . . . , . . - . . . . - - . - . -   . - . . - . . -

SECTION 8.0

                                         ,                           REFERENCES

1. Combustion Engineering Report No. S-NLM-002, " Program foe Irradiation Surveillance of San Onofre Reactor Vessel Materials San Onsfre Nuclear Generating Station, Units 2 and 3", October 30, 1974

2. Combustion Engineering Report No. TR-S-MCM-004, " Southern California Edison San Onofre Unit 3, Evaluation of Baseline Specimens, Reactor Vessel 1 Materials Irradiation Surveillance Program", November 30, 1979

3. Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," V.S. Nuclear Regulatory Commission, Washington, l D.C.

4. ASTM E185-82, " Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)."

S. ASTM E23-88, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials." , 6. ASTM A370-89, " Standard Test Methods and Definitions for Mechanical , Testing of Steel Products "

7. ASTM E8-89, " Standard Test Methods of Tension Testing of Metallic Materi al s . "

l

8. ASTM E21-79 (1988), " Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."

9. ASTM E83-85, " Standard Practice for Verification and Classification of i

Extensometers." j 10. Regulatory Guide 1.99, Proposed Revision 2, " Radiation Damage to Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, February,1986. 8-1 L

11. 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.

12. *0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors'.

13. ASTM Designation E482-82, " 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, 1984.

14. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984,

15. 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,1984.

16. ASTM. Designation E706-81a, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.

i i i 8-2 ,

i

17. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results', in ASTM Standards, Section 12, American Society for Testing and Materials, l Philadelphia, PA,1984. j i

18. ASTM Designation E261-77, " Standard Method for Dete mining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

19. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12.

American Society for lesting and Materials, Philadelphia, PA, 1984.

20. ASTM Designation E263-82, " 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, 1984.

21. ASTM Designation E264-82, " Standard Method for Determining fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984,

22. ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux -

Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984. .

23. ASTM Designation E523-82, " Standard Method for Determining fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.

24. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.

8-3

a

25. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance", in AS1H Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984,

26. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979, t

,                  27. 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.1-IV. Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.

28. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981.

8-4 . l

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                                                                                                        .ZZ.2cr.:w':===:=,,0 0.0000 2.T. . ... .... . . .. . . . . . .. .. . . .. .. .. .. . . .... . ... . .. . 'i.6000E+03                       2.4000E+ 3 E+00        0.0000E+00                      9.0000E+02 Time, cicroseconds i

I Figure /-10. Charpy impact test load-time record for Specimen 12K.

                       ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ^   ~ ~ ~ ~ ~  ~ ~ ~ ~ ~ ~ ~ ~

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                                                         ' Time, microseconds t

I 6

                      , Figure A-11. Charpy impact test load-time record for Specimen 12L.

E+03 3.7500 rr'? T^^ E+03 [ '-N

                                                 \%
                       >P99 c~                                         N O                                                                                                                               {

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2.5000 %. E= 90.0 ft-lbs 4

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                                                 /                           '

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0.0000 c.,..................... c ...... ... ....... . . . ..... ......... ... .. ..... ... . 3.0000E+02 1.6000E+03 2.4000E+03 E+00 0.0000E+00 Time, microseconds Figure A-12. Charpy impact test load-time record for Specimen 14M.

                       ~       ~ ~ ~ ~ '          '        ' ' ' " "' ~~~~

530% E+03 Pr,3 Pf 3.7500 E+03

E- 2

5 0 - 2.5000

         "                                                                                                       0 E+03 1.2500 E+03 E=140ft-lbs

0. 0000 .'.e.....u,i..;l.re.ce:2 0,y.su: _z.L.<ne aQ2 2.4000E+03 w._.cm- .=

hs. 0.0000E+00 3.0000~t02 1.6000EiO3 E+00 Time, microseconds f Figure A-13. Charpy impset test load-time record for Specimen 25U. m .

                                                                                                             ~ ~ ~
                                  "~     ~ "~~~         "~ ~ """          " ~ " ~ " " " " ~
                       ~

5.0000 E+03 Rf 3.7500 Pg3~ E+03 e-- Q e., c

 -                                                                                                                 3
 ^     -   2.5000
       "                                                                                                           K E+03 1.2500 E+03           .

E= 1E.0 ft-lbs 0.0000 K W ..'.o,.v_.s s . o_. . a_.m _ x _ 7,__, u ,m_ m _

                                                                                                     ,,,c_ _

E+00 0.0000E+00 3.0000E+02 i.6000E+03 2.4000E+03 Time, microseconds l l l Figure A-14. Charpy impact test load-time record for Specimen 23K. l __

s i E+03 ff Ps3 l 3.7500 l E+03 ! c-- o

8. c g 2.5000 3 E+03 L i

1.2500 , E+03 e

                       . Pa .
                              /
                                        ,                                                                                          E= 21.0 ft-lbs l

0.0000 p. l Gn..... ...:.~.:l..~2.h. h.D.e.hme-rs. sea -v .w--:. c. < - E+00 0.0000E+00 3.0000E+02' 1.6000E+03 2.4000E+03 Time, microseconds Figure A-15. Charpy impact test load-time record for Specimen 23L.

5.0000 E+03 80 3.7500 s~ 'Pf E+03 Pg'd e O - 9 C* : E - 2.5000 er e E+03 1.2500 . E+03 ph,'., E= 25. it-lbs F .

0. 0000 ..i..'..' 4..r (d$h...........:..:.r l

l ..'Dc":aZ::=a~-: uL; -%c: ^~ h=-=~=~ ; -

E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds l l Figure A-16. Charpy impact test load-time record for Specimen 22B. L

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......'...-.-l-.'- - -- - . -
E.0000E+0? 1.6000E+03 E.4C 0E+03 E+00 0.0000E+00 Time, microseconds Figure A--18. Charpy impact test load-time record for Specimen 25L.
-n- ;a. a. . n. o-E+03 3.7500 _Pr ' Pg3 EiO3 e-o - - i e I E 2.5000 ,
E+03 .- Pa-t i 1.2500 's E+03 -
~
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~_.
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0.0000 lid.Er...... ..... ... ........ .... . .. 7. ~. ~' ~.?:f:- 3.0000E+0) 1.6000E+03 2.4000E+0? E+00 ' 0.0000E+00 Time, microseconds Figure A-19. Charpy impact test load-time record for Specimen 247.
t E+03 3.7500 E+03 ,fFf
~ > N +Es3 e
c. =
4 2 c - 2.5000 a-E+03 3 Pa. . . , i.2500 ~ i_ ,. E+03 ;, E= 33.0 ft-lbs _- 2.
, ~,6 ~' % _
2.'.~~::~.c.tw:.e=,cu.o. .e: e..,. V.~. . ... . .. . .. . .... . . ............l. 0.0000 ..w.-m c.,,-. E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-20. Charpy impact test load-time record for Specimen 223.
E+03
,,. -Ro. . ,, ,- t a./,500 , -s \
E+03 .Pqq-- . l e-O
> u -
C i n. RJ =
~
p 2.5000 w,N. E= 92.0 ft-lbs 5
- e E+03 _
y'~s
. ~~ ',i / 5 1.2500 -.,
_~% E+03 .Ed' _,
,__.~.___-
0.0000 ,/ ~ E+00 5.6056'EiU6"""'$f.66565762 ~ "" i'.T6UUEiUI"""~ 2.id'66'ET65 """" Time, microseconds Figure A-21. Charpy impact test load-time record for Specicen 25C.
- ~
0.000v,, r , n -- ttuo 3.7500 _- f'R-^s__ E+03 :P94 -- . r-O '
>r P 4
e - , 4
" - 2.5000 " 5' E+03 i. .~
N E= 62.0 ft-lbs s 1.2500 ~~~ _.h E+03 _, ~ __
./_ _. ~ ^^- - ,._ _ ~ ~ ~
0.0000 !c::'_/ Em. .. . . . . . . .. . .. . .... 1.6000E+03 2.4000E+03 E+00 ' 0.0000E+00 3.0000E+02 Time, microseconds Figure A-22. Charpy impact test load-time record for Specimen 245.
1 l i i E+03 i 3.7500 E+03 _fv1n. Pdi ~". e _._w _' C,, - -
> a - ,
E .
" ?
a- 2.5000 - E+03 ; M i ', E= 73.0 ft-lbs
~
_ N. _____-_- i.2500 ' ! E+03 E.n ! n.0000 l ':.I. . . . . . .. . . . . . . . . . . .. . . ... .. .... ....... ............ . . . ~ ~~2 : E400 ' O.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-23. Charpy impact test load-time record for Specimen 23M.
_ ..s .. ..... . ............ . ........ . . . . . . . . . . . ..
5. v. ovv_
E+03 Bf F33 3.7500 E+03 e-O
> P a
i b 2
* - .s -- ;
o- c.ov00 t, E+03 3 i.2500 ' E+03 : P.a ' ., E= 23.0 ft-lbs-1%3e,e.c,. mo.s_,a_-3
0. 0000 !gi.e.......' -Vvc . _. _. ..
E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 i Time, microseconds I Figure A-24. Charpy impact test load-time record for Specimen 363. '
-- ~ - - - -
e---- - _ _ _ _ _ - _ _ _ _
s.vuv. E+03 .5 + Pgy--
,,0 3/ 0 i
E403 l e. O
> P l . o-l u - .. t i ~
o- 2. Guntvie m s-
._ i i
l E+03 L l 1.2500 E+03 , E= 27.0 ft-lbs- t
' - E r. .. :a~ . . s. ; ,i , , , . -e.................,...tx.....:L. .Tv.
a
. . .r. m n-a n.on9 v ' ,.., ,_ n .n ,, .................vt...,u., ....w...,v..,,,v_.,.,nN,. . cn www - :-
E+00 0.0000E+00 d.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds ! r f
' Figure A-25. Charpy impact test load-time record for Specimen 33L.
ll ll i
,: / ,r 1
1!i:l, Ii,i ,:i, I i r6
..,t -
3 B 0 7 c- + 3 b E l n
- 0 e .0 t 0 m 4 i f c e
r-t
. 'Ig2 S p
m U- r
- o = f % d r
3 o 0 c
. %. E 4 e r
e
^0 0 0 m 6 s i d t o d aTins c
e s o l t a o
._- r s c e ~ _. . 2 .
i t 0
= + t .E e c
m a 0 p 0 i
.- m .- . 0 T i . .0 y
8 p
.? gl.
r
. a ~j -
h C
... _ . _~ .
0 .
- 0 6 .e ._- .+ 2 .E - .0 A p , .F . h .
0 e u 0 r
.,. .;sE g/. . 0 - .;a .
u g i _r {k. O ' F 1tI - a rJ 0 rj. 0 e, . 0 e. 0 o< 00 nw 3 53 wac0 3 00 n0 70 n=m_.
s. + .+ . +
. + .+ E 0E JE 3E - .. c - <1 ROE - we >.eC li
.t h
E+03 ... O P f , 1 Ps.k. s
'3.7500 n c.'
u- -- e . o ! F
c. ,
u -
; 2.5000 6
r. [
t+0a / Pa j 1.2500 - E= 42.0 ft-lbs E+03 1 '. *
-T. - ?,-
r
^ ~
0.0000 . . . . . . . . . . . . . . . . . . . . . . .. . ... . .?.T.T.. . :.h.r:~.d. .c .= ,c:,,=--r . E+00 ' O.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03
. Time, microseconds t
Figure'A-27. Charpy impact test load-time record for Specimen 367. 1 _ . _ . ._
.*- t 5.0000' E+03 Pn 3.7500 .-J~ ~P f E+03 Pyj A
r. e-O W '. > c. , M . = ~ 1 e 2.5000 E+03 Pa? E= 44.0 ft-lbs 1.2500 .,. E+03 ____in; .,. t
,/- v. , <. .
1
- v'- n _, ,
_/' Yn __,
.. .. . . .:=. . :':2:~'~=~= ~-
0.0000 c.:........................... . ..... . . .. . .... . . . . . . .. ... . . . . 1.6000E403 2.4~0 e v E + 0 3=r7 " 0.0000E+00 3.0000E+02 E+00 Time, microseconds Figure A-28. Charpy impact test load-time record for Specimen 331. I
5.0000 E+03
,.j-"P5 P.f 3.7500 Pfy s -
E+03 7, e o y E . u - Ps a G '2.5000 ~.--
~
7 E+03 U
'\ i._ E= 61.0 ft-lbs s
u 1.2500 E+03 ,/
~'. V'u~- .y _ / -~ / ' Em ' ' % ,_ ~ -w 0A0000 CI......................... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . = - - . . . . . . . . . . . - - WCOTn'*er. . .
E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-29. Charpy impact test load-time record for Specimen 37U.
5.0000[ i E+03
.Pc. 6Fk ,.
Pt 3.7500 E+03 a F;n t-O .- _.
> p O.
t * ,- 63 s o i e 2.5000 , i E+03
.\ E= 65.0 ft-lbs x-1.2500 - /.-
E+03 ,' i_
,9 ' ,- -n.__-
0.00()0 / ' E+00 5.iOUE5Ud~~~a'd60iE+I2" """ " 5205UEid5 "" ~ 2I4000E+03 Time, microseconds Figure A-30. Charpy impact test load-time record for Specimen 371.
i l 1 i [ - l o.vvvu , E+03- i Py y,gy I'JA, r _
,s.1 ,5 0 v,. . .
t+ v,.o, s e o P c. c,3 - _
" I ; 2.5000 .,
6 E+03 T 5 Ns - ,., . t= t / . o + ' ,i t's s.
' 1. C500, E+03 ,..-
EK =-, A.0000 !/
~ ^ ~ - . . . . . . . . . . . . ~ . .l. . - . . . . . . . . . . . . . . _ . - - 'c.:
E+00 0.0000E+00 3.0000E+02 1.6000E+03 ME+03 Time, microseconds Figure A-31. Charpy impact test load-time record for Specimen 365.
8. . , .
t. E+03 l . i k
.p. . _~~'-
3.7500 .~~~ ~ E + 0:. cu
. .- r _ _
e O
> P s c- g ~ ~
a t.:
2.5000 P G ' E= 75.0 ft-lbs E+03 ,
;m ~~~~__ ,- _~ .- ^~ .
i.2500i ' ~~- ! E+03 ..- - i-fi .'e..._
._w ~~
f c~". 4000E+03 0.0000 . -c~.~. . . . . . . . . .
'i 6000E+03 E+00 0.0000E+00 . 8.0000E+('E Time, microseconds Figure A-32. Charpy impact test load-time record for Specimen 34P.
i
~~' ~
f1 . . . . _ 1 I
i E t.itG C4 I
< l:
1 l l j . r. i .
'n '. c O + M LU l ; gj M a .D ' i !, c2 ~
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o O O { , a
. - i._ g I' I t% ,1 L I o il 6 . **4 L13 'l -
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I .
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23 0 4
+ p
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i
o e ri
, y~ , x p.
w
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i
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cia cji oo OO
... . M LO TD [1
c. c, rs o . .e . +
L AI C' L' J, (~[, g .f , (t3 ::'*J Ld Load, lb A-33
l l4l{l;Ii Il 1\ 1
.b- T .; :l;:
i.
.l-;!-l ;*!i1i 7 . 3 7 0 3
_ .* n

l. -E e

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- i
t. 0 c
(, 4 e p
's - 2 S . ,Ip r T_ :
C. 2
. o - f d
iu ~ r
.~~ 3 o ~ .0 c
e
- . 4 r E - . 00 e m . .. 0 i t - 6 s - - . 1 d d . n a . . . o o ~
c l
. e s t ,.- o s . r e c t .i . ..2 t . 0 m c ~ 4 , a . E e p . 0 m m .. 0 - .i i - - T y - 0 p 9 r a ~ . h C ._ 0 4 0 3 .. .+ A . ". .E - o,?. - st 0 e 0 .. ~. 0 r
u 6~ . 0 g N~ .
. . i i, it Pt ~ /. . 0 F . 0 () . . .- 0 o.
n- r,v e 0
a. 0
. T_ 6_ X. nc i,._ 5 ,o 00 7, D. E r). nc Q 00 . ,.+ 0_ .+ . a.
l t
.+
OE
+
_n. F_. 2- V- ? cu tOva' wT y, E
,Illl lilt
0. 0Vvv E+03 ff
.Pt.y 3.7500 E+03 e
o E- 4
? -
o : m ~ e ,c.500v. L E+03 1.2500 E+03 E= 18.0 fi-lbs 1 ym cr>: .a:-s r .~s
~ * .s I -'
j. o
. i hi..~ .v.W:.s.::.:: . . : .x. . .Uv h CL..-
(gv m 00(-)()-
'-i.'m'.1 3.0000E+02 1.6000E+03 1.4000E+03 E+00 ~ 0.0000E+00 Time, microseconds Figure A-35. Charpy impact test load-time record for Specimen 42L.
D . Y,OUV E+03' Pm fM 4* 3.7500 E+03 e-o
> W , , .a ,
3
" 2.5000 P
E+03 1.2500 Pas'- t +0,s .
!'sf . . f= 2E.0 ft-Ibs . s .
I O____ < v ,
. s ,.
Q,QQQQ l/ 's " Y^J w.,mwn, , _ . , _ C +00 6500005756 -"E'.'h66d5Id i ~"~~ i.'UUU'5h"id~^.__T__'[b55'EIdi~_~' Time, microseconds Figure A-36. Charpy. impact test load-time record for Specimen 43P. I l
-' d' WMb-lah 'M dt'e_.Pl h & %n Wh*W M N wh-e'reer' Olk 6
w t,O 't s. i J (:. . k M b v
j: O
us c' +
,o Itd C I L' +
e t (O O O N to
p O > v
~ ; 4. !. res . o to.4 l it j T w .
b
' re; O }.
o V l 1+ h
td O o
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e T 4 4
w C T 0 N
. i- O O l v v=<
In O 4 4 M v
.O es 4 N El O 4
+ = 0 LM 0 d
(- O a
.aet 0
. .p$ .o O b D*'l O* A u.
[/. iO
.y 4 _
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* +
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LJ =< E O
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e . '.S
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m
N
m .lt o is.
a
>,' ty3
O.
o O O O O O $ am O OO OM OM Om NO CO NO go + OO + - g1 41 41 -w OW Load, Ib A-37
E+03 3.7500 E+03 y Ff P9'd C ^ y V 4 a
> ." Ets- 1 = -
e 2.5000 C
'^
E+03 h 1,
1. <500 .
E= ,a,a . 0 t c l .os E+03
, , ~ . ..
i -
's __.
I 0.0000 [ d r..... ............:. I'g:r.e ,_v_c _ _ _ _,_'___ _ 0E403 , 0.000'f;+00 8.0000E+02 *i.6000E+03 2.406 E+00 Time, microseconds Charpy impact test load-time record for Specimen 41C. Figure A-33. !
-~
I we- (% W l
,tO Ea3 O t~
m + v
.CJ ;W 0 - .o I ,o .fj- . - o
4 ,Y Q
.i e p o t cJ n, . v2 g
9- _ {. - W O II w
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o O
. m
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b
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o
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/ \ \
lo
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i !
Et> m to co , ,. s.\:h. :o
.te e -. m o o o o o o o o o om oo 8m oc om o, u mo ca. +o oo +
s. o. .+
ca u; -w ow d,1 rn w Load, Ib A-39
o, 6 , . - . t f **J G C* b y is i +
.O , Lt1 C - c. O s ' *: U . e4 4, c. U ..+ ~r , G 4
I 6 ,j V2
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. l; ra i O
c. O
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g
.C .H e4 '
t** (n i:w C' y d . $ iro
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[ } I, e
+ ..'sm ; . C' c) '., ; C' y ct? ct. * , .+ s 1 4 2'S' \' m I;.. U'.
s* ,c, p Laa ,' 4 s c.
; ye To *.,,,! .C,' . .hs A, 'n l v' ,
g ce C) C' o, c c) C* *> C# f *~.8 LO I C' C' 7, r**J (CJ f * > "C' Ce
,7, c, rs c. LO c. C 'd C) ,+ , .+ . + =+
c'J LAJ *-* Lt.1 C' Lal
.c. u. r i L2.t Load, Ib A-40
E+03
,.f4 3.7500 19g3 ' ,~
E+03 Pf E [
~
E p 2.5000 4 5 M E+03 P'a E= 45.0 ft-lbs 1.2500 N _ E+03 - __4.-
,- .m-p ~ ' ~
O.0000 <::[.I.n..... .. ......... E+00 0.0000E+00 .... ... .... . ..C.T.T. h1. 3.0000E+02 6000E + 03~~~~-'.~.::carr.hn
' 2.4000E+03 i Time, microseconds Figure A-41. Charpy impact test load-time record for Specimen 45M.
v r="A N t
' r*"i 8
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. tri :+ v lW
EJ. .. O C iO v q
li .:- a
+ y j; e; a
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OW
*+
LO W MW NW Load, lb A-42
i I I v O C1. M O e A V1 + 4
,,C3 ', LD - .O C:
O O
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O i1 A *H W 'O
to b O )
. +
LM
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/

f. ,(
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) \.N 6
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l ',. A :o t O v b
. 33 3~. .[ tn \'\., . . V 'i l., , D .m -A 4 O O O O O O O O O O OM OM C r9 OM OO OO NO OO .+ N. O + O. O + *+ . +
u,w ro w co w -w Ow Load, Ib A-43
5.0000' .. E+03 3.7500 E+03 , ('Pil '-s_., 863 e o .- s j
>, W ~. I =
a. -
3 A
^ 1 ~
2.5000 4 E+03 N i E= 75.0 ft-lbs
,~-
1.2500 ,' _
,/ %
E+03 .
~' #~" . -~r , _,7 . .. .. . .. . . . ..l . . . .. :. . . . :=====<
0. 0 0 0 0 ................-.......... .
[d.T.................. ..... 3.0000E+02 1.6000E+03 2.4000E+03 E+00 0.0000E+00 Time, microseconds Figure A-44. Charpy impact test load-time record for Specimen 434.
I 4 i l l I ENCLOSVRE B I APPLICABILITY OF HEAT AFFECTED ZONE MATERIAL IN CALCULATING l REACTOR VESSEL Nil-0UCTILITY REFERENCE TEMPERATURE (RTm) i
*--n . gm -t --- -awa==--se - . ,e --=ve--e-+,-r
ENCLOSURE B APPLICABILITY OF HEAT AffECTED ZONE MATERIAL IN CALCULATING NIL-DUCTILITY REFERENCE TEMPERATURE (RT ,) SAN ON0fRE UNIT 3 l l Backoround Information The original Unit 3 limiting beltline material based on the 1979 Combustion Engineering (CE) analysis of unirradiated surveillance capsule specimens was the heat-affected zone (HAZ) metal. On June 12, 1989, Southern California Edison (SCE) submitted Proposed Change Number (PCN) NPF-15 292 revising Technical Specification (TS) 3/4.4.8, " Pressure / Temperature Limits,* to be effective up to 8 Effective full Power Years (EfPY) of operation. In the submittal, SCE noted the HAZ was no longer the limiting material based on the Battelle materials test report, and that the limiting material is now intermediate shell plate C 6802-1. Discussion l From the original CE HAZ metal V-Notch Charpy impact energy test data, three data points a)peared to be anomalous in comparison to the other data (see figure 1). Taose anomalous data points led CE to use conservative bounding values (low values) for the HAZ metal Charpy Upper Shelf Energy (USE). As a result, the calculated RT overly conservative. m in the CE analysis was believed to be high and The Battelle materials test performed in 1989 re evaluated the conservatism of the CE analysis. Battelle fabricated new specimens from archive HAZ metal, which was the limiting mtterial at that time, and tested these samples. The results of the Battelle material test indicated much higher Charpy USE values for the HAZ metal than determined by CE (see figure 2). We believe the inconsistencies were due to the use by CE of only 2 data points per test temperature and the three anomalous data points. Other factors contributing to the inconsistencies could have been from the non standardized techniques in the preparation of the samples, the location of the machined notch relative to the HAZ metal, and an off-centered impact for some specimens. The 1991 Westinghouse data from the first Unit 3 irradiated surveillance capsule specimens did not consider the HAZ metal data in the calculation of RT, . This is consistent because HAZ metal is not limiting. Also, HAZ data shows significant scatter and, therefore, may not be representative of the strength of the full thickness of the HAZ material. Results of SCE Evaluation The Westinghouse Char)y USE curve for the HAZ metal is consistent with the CE Charpy USE curves wit 1 the exception of the three anomalous CE data points. The Battelle Charpy USE test results indicated values up to 59 ft-lbs higher than the Westinghouse values.
u 2-ENCLOSURE B Table 1, Figure 1, and Figure 2, provide specific information on llAZ metal
, and the irradiated specimen test data Charpy RT,33 impact Table 1 properties, predicted initial values RT, were based on CE unirradiated specimen data.
These values will be updated based on the irradiated surveillance capsule test results. As shown in Table 1, the Battelle unirradiated specimen initial RT,3, is -40'F and the Westinghouse irradiated specimen test data indicated an shift of 45'F. We believe this 40*F RT is appropriate because of the RT,3, care taken by Battelle in preparing (machining,3,, cutting, and notching relative to the ilAZ) their specimens. The Battelle data in Figure 2 shows very little scatter compared to CE's data in Figure 1 which shows significant scatter, including three data points considered to be anomalous. Flowever, as indicated in the transmittal letter, we used the CE unirradiated specimen data in conjunction with the Westinghouse irradiated specimen data to shif t for the llAZ metal. We believe this is appropriate determine because both the RT,3,illance capsule specimens were prepared by CE. surve IIAZ metal was determined to be not limiting, therefore, ilAZ metal is not a factor in determining RT,33
TABLE I SAN ON0FRE NUCLEAR GENERATING STATION UNIT 3 Predicted Values of RT,3, for Reactor Vessel Boltline Materials at 8 EFPY NOTE: The predicted values in this Table were based on CE unirradiated specimen data. The predicted values need to be updated based on the surveillance capsule irradiated specimen test results. Plate or Cu Ni intl. Adj RT Description Block No. % % RT,3, RT,3, ShWt Margin Inter. Shell C 6802-1 0.05 0.57 +40 92.4 27.3 25.1 Inter. Shell C 6802-2 0.04 0.55 +0 44.6 22.9 21.7 Inter. Shell C 6802 3 0.05 0.57 +20 72.4 27.3 25.1 Lower Shell C 6802-4 0.05 0.57 +55 7.4 27.3 25.1 Lower shell C 6802-5 0.05 0.57 +10 62.4 27.3 25.1 Lower shell C-6802 6 0.06 0.6? +20 71.7 32.5 19.2 Long. Weld 2 203A 0.05 1.00 -40 71.2 59.8 51.4 Long. Weld 2 203B 0.05 1.00 40 71.2 59.8 51.4 Long4 Weld 2-203C 0.05 1.00 -40 71.2 59.8 tl.4 Long. Weld 3 203A 0.04 0.16 -10 58.1 36.1 32.0 Long. Weld 3 203B 0.04 0.16 10 58.1 36.1 32.0 Long. Weld 3-203C 0.04 0.16 10 58.1 36.1 32.0 Circum, Weld 9-203 0.05 0.04 -50 2.4 27.3 25.1 HAZ (Battelle Data) 0.05 0.04 40 21.7 32.5 29.2 HAZ (CE Data) - -
+70 2
Predicted Fluence at 1/4 T - 6.51E18 n/cm WESTINGHOUSE DATA Plate or Predicted Test Data Description Block No. RT,3, Shift RT,3, Shift Inter. Shell Transverse C-6802 1 31 55 Inter. Shell Longitudinal C 6802-1 31 50 Weld Metal 27 32 HAZ Metal - 45
'C 120 - 0 100 - r * - Eum 0
g 80 = 0,- - o .3 l
% w <
a /- M
40 - -
20 - - - - 2 ,4 oY' 0-N. i 200 100 0 100 200 300 400 500 100 2.5 80 - - -- --- C - - ---
- 2.0 E
E 60 -- - - -- 0 -- - - ef & c --- --
-1.5 _
ci. E b 40 - 2
-1,0 b 30*F shif t 20 - y -0.5 /,,
0 ! -0 200 100 0 100 200 300 400 500 160
- 200 140 --
120 - =160 11 ft Ib F 100 - - unirrediated
~ ~- - ~ ~~
f s k ,_ oh -120 *q w 80 -- - g- .-p - 60 - - 80 [ 2 " 70' , C.[ g 5 40 - 2
\ - - -
v 'I@* A -45'F Irr' dlated at 550'F - 40 20 -
\-
to0.8x1019 n/cm 2 _
o ! ! -
- 0 200 100 0 100 200 300 400 500
(*F) FIGURE 1: Charpy V Notch impact Properties for the San Onofre Unit 3 Reactor Vessel H AZ Metal O Unitradiated G lrradiated at 550*F to 0.8 x 1019 n/cm 2 CE
@ Untrradiated Anomalous Data Points
100 50 0 50 I DI 100 150 200 250
' I ' '
120 ' ' ' 100 -
- - c, " , 0 0 . 2 Q
o 80 - - O
$ 60 - ,# .c .
D 40 ) e
/
0 T f, 200 100 50 100 200 300 400 500 100 2.5 80 ------- Q 9 7 . O
'-#-"--ga~ = 2.0 = 0 -
5. 60
~ ,-1.5
m 40 , e -
o e e - -- -1.0 5 - 20 0
, -0.5 0^ i e 200 100 0 0 100 200 300 400 500 160 g O O - 200 140- --
unitradiated - 120 - \*- -O - . A - 160 E;- 100_ J 59 f t.lb
-120 5 80 - i sY 60 --
135'F Shift - --- - < 04 > 80 *g Lu 40 - - g irradiated at 550' 107 F Shilt -- Z-. to 0.8x10" n/cm { - 40 0 200 0 100 0 100 200 300 400 500 ('F) FIGURE 2: Charpy V Notch Impact Properties for the San Onofro Unit 3 Reactor Vessel HAZ Metal O Unitradiated 2
# Irradiated at 550'F to 0.8 x 10 n/cm Dattelle}}

ML20073H818

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ML20073H818

Text

SUMMARY

2.0 INTRODUCTION

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8.0 REFERENCES

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2.0 INTRODUCTION

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SUMMARY

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