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Nonproprietary Analysis of Capsule V from Duke Power Co, McGuire Unit 2,Reactor Vessel Radiation Surveillance Program
	ML20141C857
Person / Time
Site:	Mcguire
Issue date:	01/31/1986
From:	Congedo T, Kaiser W, Yanichko S; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20141C848	List: ML20141C845; ML20141C857;
References
	TAC-59962, TAC-59963, TAC-62779, TAC-62780, WCAP-11029, NUDOCS 8604070328
	Download: ML20141C857 (82)
	v • d • e

WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION i

p ANALYSIS OF CAPSULE V FROM THE DUKE POWER COMPANY MCGUIRE UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko T. V. Congedo W. T. Kaiser January 1986 APPROVED:

M T. A. Meyer, M& nager Structural Materials and Reliability Technology Work performed under Shop Order No. DVFJ-106 Prepared by Westinghouse Electric Corporation for the Duke Power Company Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 8604070328 860402 ADom o 3767e:ld/012786 p

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5 PREFACE This report has been technically reviewed and verified.

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Reviewer

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i Sections I through 5, 7 and 8 C. C. Heinecke Section 6 S. L. Anderson JJ Thadann N Appendix A F. J. Hitt V / 2,/>*#

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TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 1 -1 2

INTRODUCTIGN 2 -1 3

BACKGROUND 3-1 4

DESCRIPTION OF PROGRAM 4 -1 5

TESTING 0F SPECIMENS FROM CAPSULE V 5-1 5 -1.

Overview 5 -1 5-2.

Charpy V-Notch Impact Test Results 5 -3 5-3.

Tension Test Results 5 -4 5 -4. Compact Tension Test Results 5-5 6

RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6 -1 6-1.

Introduction 6 -1 6-2.

Discrete Ordinates Analysis 6 -1 6-3.

Radiometric Monitors 6 -4 6-4.

Neutron Transport Analysis Results 6 -8 6-5.

Dosimetry Results 6 -9 7

SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7 -1 8

REFERENCES 8 -1 APPENDIX A HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION A -1 A-1.

Introduction A -1 A-2.

Fracture Toughness Properties A -2 A-3.

Criteria for Allowable Pressure -

Temperature Relationships A -2 A-4.

Heatup and Cooldown Limit Curves A-6

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4 LIST OF TABIES Table Title Page 4-1 Chemical Composition and Heat Treatment of The McGuire Unit 2 Reactor Vessel Surveillance Materials 4 -2 5-1 Charpy V-Notch Impact Data for the McGuire Unit 2 Intermediate Shell Forging 05 (Heat No. 526840),

18 2

Irradiated at 550*F, Fluence 3.06 x 10 n/cm (E > 1 Mev) 5-6 5-2 Charpy V-Notch Impact Data for the McGuire Unit 2 Reactor Vessel Weld Metal and HAZ Metal, Irradiated at 550*F, Fluence 3.06 x 1018 (E > 1 Mev) 5 -7 5-3 Instrumented Charpy Impact lest Results for McGuire Unit 2 Intermediate Shell Forging 05, Irradiated at 3.06 x 1018,fe,2 (E > 1 Mev) 5 -8 5-4 Instrumented Charpy Impact Test Results for McGuire Unit 2 Weld Metal and HAZ Metal, Irradiated at 18 3.06 x 10 njc,2 (E > 1 Mev) 5 -9 18 2

5-5 Effect of $50*F Irradiation at 3.06 x 10 n/cm (E > 1 Mev) on the Notch Toughness Properties of McGuire Unit 2 Reactor Vessel Materials 5-10 5-6 Tensile Properties for McGuire Unit 2 Reactor Vessel I0 2

Material Irradiated at 550*F to 3.05 x 10 n/cm (E > 1 Mev) 5-11 3767e:1d/012786 vi

LIST OF TABLES (Continued)

Table-Title Page 6-1 SAILOR 47 Neutron Energy Group Structure 6-11 6-2 Nuclear Constants for Radiometric Monitors Contained 6-12 in the McGuire Unit 2 Surveillance Capsules 6-3 Calculated Fast Neutron Exposure Parameters for the 6-13 Peak Location of the McGuire Unit 2 Reactor Vessel 6-4 Calculated Fast Neutron Exposure Parameters and Lead 6-14 Factors for the McGuire Unit 2 Surveillance Capsules 6-5 Calculated Neutron Energy Spectrum at the Center of 6-15 McGuire Unit 2 Surveillance Capsule V 6-6 Spectrum-Averaged Reaction Cross Sections at the 6-16 Center of McGuire Unit 2 Surveillance Capsule V 6-7 Irradiation History of McGuire Unit 2 Surveillance b-17 Capsule V 6-8 Comparison of Measured and Calculated Radiometric 6-18 Monitor Saturated Activities for McGuire Unit 2 Surveillance Capsule Y 6-9 Results of Fast Neutron Dosimetry for McGuire Unit 2 6-21 Surveillance Capsule V 6-10 Results of Thermal Neutron Dosimetry for McGuire Unit 2 6-22 Surveillance Capsule V 3767e:ld/010286 vil

LIST OF TA8LES (Continued)

Table Title g

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6-11 Summary of McGuire Unit 2 Fast Neutron Fluence Results 6-23 Based Upon Surveillance Capsule V l

A-1 McGuire Unit 2 Reactor Vessel Toughness Table A -7 i

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LIST OF lLLUSTRATIONS Figures Title Paje 4-1 Arrangement of Surveillance Capsules in McGuire Unit 2 Reactor Vessel (Updated Lead Factors for the Capsules Shown in Parentheses) 4 -4 4-2 Capsule V Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-5 5-1 Charpy V-Notch Inpact Data for McGuire Unit 2 Reactor Vessel Shell Forging 05 (Axial Orientation) 5-12 5-2 Charpy V-Notch Impact Data for McGuire Unit 2 Reactor Vessel Shell Forging 05 (Tangential Orientation) 5-13 5-3 Charpy V-Notch Impact Data for McGuire Unit 2 Reactor Vessel Weld Metal 5-14 5-4 Charpy V-Notch Impact Data for McGuire Unit 2 Reactor Vessel Weld HAZ Metal 5-15 5-5 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Shell Forging 05 (Axial Orientation) 5-16 5-6 Charpy Impact Specimen Fracture Surfaces for l

McGuire Unit 2 Reactor Vessel Shell Forging 05 (Tangential Orientation) 5-17 5-7 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Weld Metal 5-18 5-0 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 2 Reactor Vessel Weld HAZ Metal 5-19 5 -9 Comparison of Actual versus Predicted 30-ft-lb Transition Temperature Increases for the McGuire Unit 2 Reactor Vessel Material based on the Prediction Methods of Regulatory Guide 1.99, Revisier. 1 5-20 l

5-10 Tensile Properties for McGuire Unit 2 Reaccor Vessel Shell Forging 05 ( Axial Orientatio'))

5-21 3767e:1d/010286 ix

LIS1 0F ILLUSTRATIONS (Cont)

Fiqures Title Page l

5-11 Tensile Properties for McGuire Unit 2 Reactor Vessel Shell Forging 05 (Tangential Orientation) 5-22 i

5-12 Tensile Properties for McGuire Unit 2 Reactor Vessel j

Weld Metal 5-23 5-13 Fractured Tensile Specimens f rom McGuire Unit 2 i

Reactor Vessel Shell Forging 05 (Axial Orientation) 5-24 5-14 Fractured Tensile Specimens f rom McGuire Unit 2 Reactor Vessel Shell Forging 05 (Tangential Orientation) 5-25 5-15 Fractured Tensile Specimens f rom McGuire Unit 2 Reactor Vessel Weld Metal 5-26 5-16 Typical Stress-Strain Curve for Tension Specimens 5-27 6-1 McGuire Unit 2 Reactor Geometry 6-24 6-2 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-25 6-3 Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Reactor vessel - Surveillance Capsule Geometry 6-26 1

6-4 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Reactor vessel 6-27 6-5 Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Reactor Vessel 6-28 A-1 Predicted Adjustment of Reference Temperature, as a Function of Fluence, Copper, and Phosphorus Contents A -8 A-2 Fast Neutron Fluence (E > 1.0 Mev) as a Function of Full Power Service Life (EFPY)

A9 A-3 McGuire Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 8 EFPY A -10 A-4 McGuire Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 8 EFPY A-11 3767e:ld/012786 x

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SEC110N 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule V, the first capsule to be removed from the McGuire Unit 2 reactor pressure vessel, led to the following conclusions.

o The capsule received an average fast neutron fluence (E > 1.0 Mev) of I

3.06 x 10 n/cm.

o Irradiation of the reactor vessel intermediate shell forging 05 to 18 2

3.06 x 10 n/cm resulted in 30 and 50 ft-lb transition temperature increases of 10*F, for specimens oriented nornal to the principal working direction of the forging, and 30 and 50 f t-lb transition temperature increases of 65*F and 70*F, respectively for specimens oriented parallel to the forging principal working direction.

I8 2

o Weld metal irradiated to 3.06 x 10 n/cm resulted in 30 and 50 ft-lb transition temperature increases of 45'F.

o Weld HAZ metal showed a 30*F and 50*F transition temperature increase of 55'F and 60*F, respectively, af ter irradiation to 3,06 x 10 n/cm.

o Forging 05, weld metal, and HAZ metal all showed upper shelf energy levels well above 50 f t-lb af ter irradiation to 3.06 x 1018,fc,2 l

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SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule V, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the McGuire Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the McGuire Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Koyama and Davidsonbl. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E-185-73, "Recomended Practice for Survei. lance Tests for Nuclear Reactor Vessels"E2l. Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratosy, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens were performed.

This report summarizes the testing of and the postirradiation data ubtained from surveillance Capsule V removed from the McGuire Unit 2 reactor vessel and discusses the analysis of these data.

3767e:ld/0102d6 2-1

SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist f racture 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 SA 508 Class 2 (base material of the McGuire Unit 2 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 f ast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil ductility temperature (RT I

NDT

i RT is defined as the greater of either the drop weight nil-ductility NDT transition temperature (NOTT per ASTM E-208) or the temperature 60"F less than i

the 50 ft-lb (and 35-mil lateral expansion) temperature as determined f rom Charpy specimens oriented normal (transverse) to the major working direction of the material. The RT of a given material is used to index that NDT material to a reference stress intensity factor curve (K C"I**)

IR i

appears in Appendix G of the ASME Code. The K curve is a lower bound of IR dynamics, crack arrest, and static fracture toughness results obtained f rom several heats of pressure vessel steel. When a given material is indexed to the K curve, allowable stress intensity factors can be obtained for this gg material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

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RT and, in turn, the operating limits of nuclear power plants can be NOT adjusted to accouret for the ef fects of radiation on the reactor vessel material properties. The radiation embrittlement of changes in mechanical l

properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the McGuire Unit 2 Reactor Vessel Radiation Surveillance ProgramU l, 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 (ART

) due to irradiation is added to the original RT NOT to adjust the RT r radadon

Ment. N s aO us W RT MOT NOT (RT s use n ex em NOT NDT al to ne K,

g curve and, in tuivi, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

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SECTION 4 DESCRIPil0N OF PROGRAM Six survdillance capsules for monitoring the effects of neutron exposure on the McGuire Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. 1he six capsules were positioned in the reactor vessel between the neutron shielding pads and the vessel wall as showa in figure 4-1.

Theverticalcer\\terofthecapsulesis opposite the vertical center of the core, Capsule V was removed from the reactor af ter '1,.03 Ef f ec' ive Full Power Years t

(EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and Compact Tension (CT) specimens from submerged are weld metal fabricated with the same weld wire and flux as used in the reactor vessel core region girth weld, and Charpy V-notch, tensile, CT, and bend bar specimens from the intermediate shell forgirg 05. The capsule also contained Charpy V-notch specimens from weld Heat Affected Zone (HAZ) metal. All heat af fected zone specimens were obtained from the weld HAZ of forging 05. The chemistry and heat treatment of the program surveillance materials is, presented in table 4-1.

All test specimens were machined fri.m the 1/4-thickness location of the forging. Test specimens represent material taken at least one-forging thickness from the quenched end of the forging. Some base metal Charpy V-notch and tensile specimens were oriented with the longitudinal axis of the specimens normal to (axial orientation) and some parallel to (tangential orientation) the major working direction of the forging. The CT test specimens were machined so that the crack of the specimen would, propagate normal to (tangential specimens) and parallel to (axial specimens) the major working direction of the forging. All specimens were fatigue precracked per ASTM E399-72. The precracked bend bar was machined in the axial orientation.

Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens normal to (axial orientation)'the weld direction. Tensile specimens were oriented with the longitbeinal axis of the specimens normal to (axial orientation) the weld direction.

1 3767e:ld/012786 4 -1 t

1 ABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE MCGUIRE UNIT 2 REAC10R VESSEL SURVEILLANCE MATERIALS Chemical Composition (wt%)

Element Foroino 05 Weld Material C

0.18 0.20 O.055 0.07 Mn 0.69 0.63 1.81 1.66 P

0.012 0.010 0.016 0.005 S

0.017 0.012 0.015 0.004 Si 0.23 0.077

> (a) 0.29 0.089 Cr 0.43 0.37 0.03 0.03

>(b)

No 0.58 0.55 Ni 0.79 0.71 0.73 0.66 Cu 0.16 0.14 0.031 0.03 V

<0.002

<0.002 Sn 0.008

<0.002 8

<0.003

<0.003 Cb

<0.001 0.004 Ti

<0.002

<0.002 W

0.018 0.015 As

<0.002

<0.002 Zr

<0.002

<0.002 Sb 0.001 0.003 Pb 0.004 0.011 N2 0.019 0.007 Co 0.016 0.015 Heat Treatment History Material Temperature (*FL Time (Hr)

Coolant Forging 05 1688-1697 3.5 Water quenched (Ht. No. 526840) 1229-1238 7.5 Air cooled 1115-1165 22.0 Furnace cooled Weld Metal (C) 1115-1165 15.0 Furnace cooled l

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Analysis performed on irradiated Charpy specimen DT-18 l

b.

Analysis performed on irradiated Charpy specimen DW-30.

c.

Weld wire heat no. 895075 and Grau to Flux Lot No. P46.

i 3767e:1d/012786 4-2 I

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Capsule V contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15% cobalt (cadmium-shielded and unshielded).

In addition, cadmium shielded dosimeters of Np and U were contained in the capsule.

Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. The composition of the two alloys and their melting points are as follows.

2.5% Ag, 97.5% Pb Helting poir,t: 579'F (304*C) 1.75% Ag, 0.75% Sn 97.5% Pb Melting point: 590*F (312*C)

The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule V are shown in figure 4-2.

3767e:1d/0:2786 4-3

48/66/12216 8 REACTOR VESSEL O.

CORE BARREL Z (4.76) i NEUTRON PAD U (4.76)

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CoheACY CohPACT CohPACT C0hPAC" 800 BAR TENSILE TENSION TENSION CHARPY CHARPY OWPY TENS!CN TENSIO Dw6 Dw30 OH30 Dw27 OH27 f_W24 OH24 OL2 OWS Dwe DW7 Dw6 Dw5 Dw29 CH29 Dw26 OH26 Ow23 OH23 (1.8 CL7 CLS DL1 Dwe Dw29 CH28 Dw25 3425J Dw22 OH22 JL Cw - -e ll Il e--- A l

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_I CORE iE SPECIMEN NUWSERING CODE:

DL - INTEFtdEDIATE SELL FORGING 05 (TANGENTI AL)

DT - INTEREDI ATE SELL FORGING 05 ( AXI AL)

DW - WELD ETAL DH - NAT-AFFECTED-ZOE MATERI AL

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CHWFY CHAPPY TENSILE CHMFY OWFY CHWFY CHARPY OWFY ON TOS ON TD 41LE 0521 DH21 DWie DHIS DL6 DT30 CL30 0T27 CL27 DT24 CL24 0T21 CL21 OTIS CLl3 OTS i

OWE 0 CN2O OWf7 CN17 368 CL5 0T29 CL29 0T26 CL26 OT23 1 23 0720 CL20 OT'7 CLl7 OTS OT7 DT6 OTS DT5 owl' CNII DWI6 DHIS DL4 DT29 CL29 DT2S DL2S 0722 CL22 DTIS DLIS OT + 6 OLIS UT4 dL lb Cs===-1 W Al==. lSIC Ce Cv --.

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Ni Ni CENTEn nEo 0N OF VESSEL TO BOTTOW OF VESSEL Figure 4-2. Capsule V Diagram Showing T1 Location of Specimens, Thermal

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SECTION 5 TESTING OF SPECIMENS FROM CAPSULE V 5.1.

OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H, ASTM Specification E185-82[3], and Westinghouse Procedure RMF-8402, Revision 0.

Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9489E3 No discrepancies were found.

Examination of the two low-melting point 304*C (579'F) and 310*C (590*F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304*C (579'F).

The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-01sen Model 74,358.1 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 energy (E ).

From the load-time curve, the load of general yielding (Pgy), the D

time to general yielding (tgy), the maximum load (P ), and the time to g

maximum load (t ) can be determined. Under some test conditions, a sharp y

drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (P ), and the p

load at which fast fracture terminated is identified as the arrest load (P ).

g The energy at maximum load (E ) was determined by comparing the energy-time g

record and the load-time record. The energy at maximum load is roughly 3767e:1d/010286 5 -1

l equivalent to the energy required to initiate a crack in the specimen.

Therefore, the propagation energy for the crack (E ) is the dif f erence p

between the total energy to fracture (E and n e en ngy at maximum loat D

lhe yield stress (cy) is calculated from the three-point bend fonnula.

lhe flow stress is calculated from the average of the yield and maximum loads, also using the three-point bend formula.

Percent shear was determined f rom postf racture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-81 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc 45.

The upper pull rod was connected through a universal, joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensemeter gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67.

Elevated test temperatures were obtained with a three-zone electric resistance split-tube f urnace with a 9-inch hot zone. All tests were conducted in air.

Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.

Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dununy specimen and in each grip.

In the test configuration, with 4.., R ht loed on the specimen, a plot of specimen temperature versus upper ar.c lower grip.ind controller temperatures was 3767e:ld/010286 5-2 1

developed over the range room temperature to 550*F (288*C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures.

Experiments bdicated that this method is accurate to 12*F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield l

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 f racture area used to l

calculate the f racture stress (true stress at f racture) and percent reduction in area was computed using the final diameter measurement.

5-2.

CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials 18 2

contained in Capsule V irradiated at 3.06 x 10 n/cm are present.ed in tables 5-1 through 5-5 and figures 5-1 through 5-4.

The fractured surfaces of the impact specimens are shown in figures 5-5 through 5-8.

Irradiation of Charpy V-notch impact specimens from the reactor vessel 18 2

intermediate shell forging 05, to 3.06 x 10 n/cm as shown in figure 5-1 resulted in 30 and 50 ft-lb transition temperature increases of 70*F for specimens oriented nonnal to the principal working direction (axial orientation) of the forging. Specimens oriented parallel to the principal working direction (tangential orientation) of the forging as shown in figure 5-2 exhibited a transition temperature increase of 65*F and 70*F at the 30 and 50 ft-lb index temperatures, respectively. The upper shelf energy of the shell forging showed a 9 ft-lb decrease in the axial direction and a 22 ft-lb decrease in the tangential direction.

Weld metal specimens irradiated to 3.06 x 1018 2

n/cm resulted in a 30 ft-lb and 50 ft-lb transition temperature increase of 45'F, as shown in figure 5-3.

Irradiation caused the upper shelf energy of the weld metal to increase 5 ft-lb to a shelf energy level of 138 ft-lb.

3767e:ld/010286 5-3

A comparison of the 30 ft-lb transition temperature increase shown in figure 5-9 with the predicted increase based on U.S. Nuclear Regulatory Consnission Regulatory Guide 1.99, Revision 1 indicates that the weld metal and shell forging are slightly less sensitive than predicted it a fluence of 3.06 x 18 2

10 n/cm.

18 2

Weld HAZ specimens irradiated to 3.06 x 10 n/cm resulted in 30 ft-lb and 50 f t-lb transition temperature increases of 55'F and 60*F, respectively, as shown in figure 5-4.

The upper shelf energy of the HAZ metal decru.ed by 6 f t-lb due to the irradiation.

The fracture appearance of each irradiated Charpy specimen f rom the various inadiated materials is shown in figures 5-5 through 5-8.

Each of the vessel materials shows an increasing ductile or tougher appearance with increasing test temperature.

5-3.

TENSION TEST RESULTS The results of tension tests performed on material from the reactor vessel 18 intermediate shell forging 05 and weld metal irradiated to 3.06 x 10 n/cm are shown in table 5-6 and figures 5-10 through 5-12.

Forging 05 test results are shown in figures 5-10 and 5-11 and indicate that irradiation to 18 2

3.06 x 10 n/cm caused a less than 10 ksi increase in 0.2 percent yield strength and ultimate tensile strength. Weld metal tension test results presented in figure 5-12 show that the 0.2 percent yield strength increased by less than 10 ksi and ultimate tensile strength increased by less than 5 ksi with irradiation. The fractured tension specimens for the forging material are shown in figures 5-13 and 5-14, while the fractured tension specimens for the weld metal are shown in figure 5-15.

A typical stress strain curve for the tension tests are shown in figure 5-16.

3767e:1d/010286 5-4 1

5-4.

COMPACT TENSION TEST RESULTS The 1/2 T compact tension fracture mechanics specimens that were contained in Capsule V have been stored at the Westinghouse Research and Development Laboratory and will be tested and reported on at a later time.

3767e:ld/010286 5-5

TA8LE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE MCGUIRE UNIT 2 INTERMEDIA 1E SHELL FORGING 05 (HEAT NO. 526840)

IRRADIATED AT 550*F, FLUENCE 3.06 x 1018 n/cm2 (E > 1 Mev)

Temperature Impact Energy Lateral Expansion Shear Sample No.

(*F)

(*C)

(ft-lb)

(J)

(mils)

(mm)

_(11_

Tanaential Orientation DL19

-50

-46 13.0 17.5 12.5 0.32 2

DL24

-25

-32 8.0 11.0 7.5 0.19 3

j DL22

-10

-23 30.0 40.5 21.0 0.53 8

DL17

-10

-23 64.0 87.0 48.5 1.23 11 DL27 0

-18 56.0 76.0 43.0 1.09 29 DL20 0

-18 11.0 15.0 6.5 0.17 9

DL18 25

-4 55.0 74.5 42.0 1.07 28 DL30 50 10 83.0 112.5 63.0 1.60 43 DL28 82 28 99.0 134.0 59.5 1.51 81 DL23 100 38 120.0 162.5 78.5 1.99 98 DL21 150 66 127.0 172.0 83.0 2.11 100 DL16 200 93 114.0 154.5 78.0 1.98 100 DL25 250 121 145.0 196.5 86.0 2.18 100 DL29 300 149 134.0 181.5 83.0 2.11 100 DL26 375 191 152.0 206.0 85.5 2.17 100 Axial Orientation DT26

-50

-46 8.0 11.0 2.5 0.06 2

DI30 0

-18 27.0 36.5 21.5 0.55 10 DT23 25

-4 26.0 35.5 20.0 0.51 15 DT18 25

-4 19.0 26.0 17.5 0.44 9

D119 50 10 32.0 43.5 26.0 0.66 17 D125 50 10 42.0 57.0 32.0 0.81 23 DT22 82 28 41.0 55.5 32.0 0.81 28 DT17 85 29 50.0 68.0 36.5 0.93 30 DT20 100 38 45.0 61.0 41.0 1.04 39 DT16 125 52 54.0 73.0 51.0 1.30 57 DT28 150 66 77.0 104.5 59.0 1.50 79 DT27 200 93 82.0 111.0 68.0 1.73 100 DT29 200 93 84.0 114.0 65.0 1.65 100 DT21 250 121 82.0 111.0 70.5 1.79 100 DT24 375 191 92.0 124.5 69.0 1.75 100 3767e:Id/012786 5 -6

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE MCGUIRE UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 3.06 x 1018 n/cm2 (E > 1 Mev)

Temperature Impact Energy Lateral Expansion Shear Sample No.

(*F)

(*C)

(ft-lb)

(J)

(mils)

(mm)

_(11_

Weld Metal DW20

-100

-73 8.0 11.0 7.5 0.19 7

DW28

-50

-46 9.0 12.0 9.5 0.24 17 DW25

-10

-23 40.0 54.0 31.5 0.80 33 DW30

-10

-23 30.0 40.5 29.0 0.74 37 DW22 0

-18 42.0 57.0 28.5 0.72 38 DW26 25

-4 55.0 74.5 44.5 1.13 61 DW17 25

-4 39.0 53.0 32.0 0.81 56 DW18 50 10 63.0 85.5 46.5 1.18 64 DW19 82 28 96.0 130.0 68.5 1.74 89 DW27 150 66 112.0 152.0 87.5 2.22 97 DW16 200 93 125.0

'169.5 84.0 2.13 100 DW29 300 145 139.0 188.5 94.5 2.40 100 DW24 350

!,7 142.0 192.5 84.5 2.15 100 DW23 375 191 145.0 196.5 89.0 2.26 100 HAZ Metal DH30

-100

-73 4.0 5.5 6.0 0.15 2

DH16

-75

-59 32.0 43.5 21.0 0.53 11 DH27

-50

-46 30.0 40.5 26.5 0.67 15 DH17

-50

-46 31.0 42.0 24.5 0.62 28 DH24

-25

-32 38.0 51.5 23.0 0.58 44 DH18

-25

-32 40.0 54.0 26.0 0.66 33 DH19 0

-18 35.0 47.5 23.5 0.60 45 DH25 0

-18 35.0 47.5 24.5 0.62 39 DH22 25

-4 70.0 95.0 45.0 1.14 65 DH23 25

-4 59.0 80.0 43.0 1.09 83 Dh20 82 28 85.0 115.0 54.5 1.38 92 DH21 160 71 72.0 97.5 54.0 1.37 100 DH26 200 93 106.0 143.5 58.5 1.49 100 DH29 300 149 92.0 124.5 68.5 1.74 100 DH28 375 191 121.0 164.0 82.5 2.10 100 3767e:1d/010286 5-7

TA8LE 5-3 INSTRUMENTED CHARFY IMPACT TEST RESULTS FOR NCGUIRE UNIT 2 INTERMEDIATESHEg8n/cv(E>1Nev) t F08GING 05 1RRADIA1ED AT 3.06 x 10 Normalized Enerates Test Charpy Charpy Raximum Prop Yield Time Nazimum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress Number

[1El (ft-16)

(ft-lb/in2)

(kips) fusect (kins)

(usec)

(kins)

(kios)

(ksil (ksi)

Asial Orientation DT26

-50 8.0 64 46 18 3.95 110 4.05 140 4.05

.25 131 132 DT30 0

21.0 217 165 53 3.30 125 4.00 4f5 4.15

.15 109 121 DT23 25 26.0 209 192 17 3.70 120 4.35 440 4.30

.20 123 1 34 8118 25 19.0 153 122 31 3.60 105 4.00 300 4.05

.25 118 126 ST19 50 32.0 258 226 32 3.55 1 30 4.40 515 4.35

.35 117 131 DT25 50 42.0 338 253 86 3.50 110 4.30 560 4.30

.35 116 129 DT22 82 41.0 330 214 116 3.35 105 4.20 490 4.15

.15 111 125 DT11 85 50.0 403 259 143 3.40 105 4.30 575 4.25

.10 112 121 0T20 100 45.0 362 219 144 3.35 115 4.15 515 4.00 1.00 111 125 i

DT16 125 54.0 435 223 212 3.00 125 3.80 585 3.55 1.00 99 112 i

DT28 150 17.0 620 288 332 3.15 115 4.15 670 3.35 1.60 105 121 i

DT27 200 82.0 660 209 451 2.60 105 3.65 560 85 103 DT29 200 84.0 676 209 467 2.60 105 3.60 565 85 102 DT21 250 82.0 660 225 435 2.90 105 3.85 560 95 111 0124 315 92.0 141 255 486 2.55 110 3.65 670 84 103 on S

Taneential Orientation DL19

-50 13.0 105 60 44 3.45 1 30 3.65 210 3.50

.25 113 til 1

DL24

-25 8.0 64 36 29 3.40 100 3.50 130 3.45

.20 112 114 DL22

-10 30.0 242 221 21 3.75 105 4.50 475 4.50

.40 124 137 DL11

-10 64.0 515 295 221 3.65 110 4.55 625 4.40

.30 121 1 36 DL21 0

56.0 451 331 120 3,25 110 4.25 750 4.15

.25 1 00 124 OL20 0

11.0 89 10 19 3.95 140 4.10 215 4.10

.20 131 133 4

DL18 25 55.0 443 342 101 3.50 105 4.50 130 4.40

.15 116 132 i

DL30 50 83.0 668 317 352 3.45 105 4.45 685 3.15

.25 115 131 DL28 82 99.0 191 329 468 3.25 105 4.35 725 3.15 1.35 107 126 4

DL23 100 120.0 966 308 659 3.30 105 4.30 685 110 126 DL21 150 127.0 1023 331 692 3.15 130 4.20 170 105 122 DL16 200 114.0 918 282 636 2.65 135 3.80 130 88 107 i

DL25 250 145.0 1168 308 860 3.30 105 4.30 685 110 126 1

DL29 300 134.0 1079 302 177 2.80 105 3.95 140 92 111 DL26 315 152.0 1224 304 920 2.90 135 3.85 180 96 112 i

]

TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR MCGUIRE UNIT 2 WELD METAL AND HAZ METAL IRRADIATED AT 3.06 x 1018 n/cm2 (E > 1 Nev)

Normalized Enereies Test Cha rpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress Number

[_* F1 (ft-lb)

(ft-1b/in21 (kips)

(usec)

(kipst_

(usec)

(kios)

(kips)

(ksi)

(ksil Weld Metal DW20

-100 8.0 64 36 28 4.15 125 4.15

.70 DW28

-50 9.0 72 32 40 3.50 130 3.40

.70 DW30

-10 30.0 242 151 91 3.60 110 4.15 360 4.10 1.95 120 129 DW25

-10 40.0 322 232 90 3.60 115 4.40 515 4.35 1.45 120 133 DW22 0

42.0 338 190 149 3.25 140 3.80 515 3.95 1.80 107 117 DW17 25 39.0 314 151 163 3.50 110 4.05 365 4.00 2.55 116 126 I

DW26 25 55.0 443 250 193 3.50 100 4.25 555 4.25 2.50 116 128 DW18 50 63.0 507 301 206 3.45 115 4.30 680 4.15 2.45 113 128 DW19 82 96.0 773 326 447 3.30 110 4.25 735 3.95 3.40 109 125 DW27 150 112.0 902 312 590 3.15 110 4.05 735 103 118 DW16 200 125.0 1007 316 690 2.65 110 3.75 825 87 1 06 DW29 300 139.0 1119 310 810 2.60 110 3.70 800 86 104 l

DW24 350 142.0 1143 291 852 2.45 70 3.75 740 80 102 l

DW23 375 145.0 1168 266 902 1.90 90 3.45 760 62 88 n

i ND HAZ Metal DH30

-100 4.0 32 34

-2 4.30 115 4.30

.15 DH16

-75 32.0 258 212 46 4.05 110 4.75 440 4.70

.40 135 146 DH27

-50 30.0 242 194 47 3.70 105 4.40 435 4.35

.43 122 1 34 DH17

-50 31.0 250 164 86 4.10 130 4.55 365 4.40 1.75 136 143 DH24

-25 38.0 306 208 98 3.90 100 4.60 435 4.60 1.50 130 141 DHl8

-25 40.0 322 205 117 3.60 100 4.30 455 4.25 1.35 120 131 DH19 0

35.0 282 166 116 3.25 130 3.95 430 3.95 1.15 107 119 DH25 0

35.0 282 161 120 4.05 425 4.50 375 4.50 2.15 135 142 DH22 25 70.0 564 279 2 84 1.70 105 4.55 590 4.05 2.45 123 137 DH23 25 59.0 475 221 2 54 3.80 110 4.25 485 4.15 3.55 125 133 DH2O 82 85.0 684 253 432 3.50 110 4.25 560 2.90 2.15 115 128 DH21 160 72.0 580 213 366 3.35 115 4.10 510 111 123 DH26 200 106.0 854 278 576 2.80 110 3.90 705 92 110 DH29 300 92.0 741 197 544 2.50 105 3.45 565 82 99 DH28 375 121.0 974 262 712 2.40 105 3.50 720 79 97

TABLE 5-5 18 n/cm2 (E > 1 Mev)

EFFECT OF 550*F 1RRADIAT10N AT 3.06 a 10 ON NOTCH TOUGHNESS PROPERTIES OF MCGutRE UNIT 2 REACTOR VESSEL MATERIALS Average Average 35 mil Average Average Energy 50 ft-lb Temperature (*F1 Lateral Expansion Temperature (*F) 30 ft-lb Tenocrature (*F1 Absorbtion at Full Shear (f t-lb)

Material Untrradiated Irradiated AT Untrradiated Irradiated AT Untrradiated Irradiated AT Untrradiated Irradiated Aft Ib Ferging 05 25 95 10 25 15 50

-25 45 10 94 85

-9

( Axial)

Farging 05

-60 10 10

-10 5

15

-75

-10 65 156 134

-22 (Tcngential)

Weld Metal

-25 20 45

-35 10 45

-50

-5 45 133 138

+5 MA2 Metal

-55 5

60

-55 0

55

-90

-35 55 104 98

-6 Y

^

5 I

3767e:1d/010286

TA8LE 5-6 TENSILE PROPERTIES FOR MCGUIRE UNIT 2 REACTOR VESSEL MATERIAL IRRADIATED AT 550*F TO 3.06 x 10 n/cm2 (E > 1.0 Mev)

I8 Test 0.25 Yield Ultimate Fracture Fracture Fracture Uniform Total.

Reduction Specimen Temperature Strength Strength Load Stress Strength Elongation Elongation in Area j

Material Number

(*F)

(ksil (ksil (kid)

(ksil (ksil (1)

(5)

(1) j Forging 05 DT4 75 65.7 89.6 3.07 167.0 62.5 11.3 23.0 63 l

(Axial)

DT6 175 10.3 87.6 3.00 142.0 61.1 10.5 23.0 57 DT5 550 63.2 89.6 3.45 152.0 10.3 9.8 18.9 54 Forging 05 DL5 15 72.8 92.7 2.87 111.4 58.5 12.0 26.0 66 (Tangential)

DL4 150 69.3 89.6 2.80 175.6 57.0 11.3 24.3 68 DL6 550 55.0 85.6 2.90 173.2 59.1 9.0 20.3 66 Weld Metal DW4 75 77.4 89.0 2.70 152.8 55.0 11.3 25.2 64 DWS 250 73.8 81.5 2.45 179.3 49.9 7.5 19.8 72 DW6 550 68.2 83.5 2.60 116.6 53.0 10.1 22.5 10 T=

i 1

48/66/2216-10 12216-10 (OC)

-100 50 0

50 100 150 200 120 2 2 100 W2 e e

80 e

g

/

r 4

60 3

y 2

/

M 40 y

O 20

2.,,7 2

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0 50 F 5

20 r

0.5 0

0 120 160 100 3

(

80 UNIRRADIATED g,

120 0

IRRADIATED (550 F) g 60 3.06 X 1018 n/cm - 80 3 2

l 40 e,4 e 70 F W

8--- 70 F 40 0

4-0 0

200

-100 0

100 200 300 400 TEMPER ATURE (OF)

Figure 5-1. Charpy V-Notch Impact Data for McGuire Unit 2 Reactor Vessel Shell Forging 05 (Axial Orientation) 5-12

48/66/2216 11 12216-11 (OC)

-100 50 0

50 100 150 200 120 I

I I

I p.- +2.-.I 2

2 y

80 5

g 60 o

Z*

40 j*

ee 20

/

0 Ga 100 2.5 g

8 8

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40 O##

1.0 5 0

g 75 F 5

20 e !e 0.5 0

0 180 240 160 O

n O,

200 O

140 o

UNIRRADIATED

,e 120 4

160 O

l#

Og 0

IRRADIATED (550 F) 80 o!

3.06 X 1018 n/cm2 120 2 O

y 60 OOe

~

80 0

70 F 40 0

20

/

40 65 F ef e n

1 O

0 200 100 0

100 200 300 400 TEMPER ATUR E (OF)

Figure 5-2. Charpy V-Notch Impact Data for McGuire Unit 2 Reactor Vessel Shell Forging 05 (Tangential Orientation) 5-13

48/66/2216 12 12210 12 (OC)

-100 50 0

50 100 150 200 120 l

I l

l 1

I i

100

,es re-ee 80 7

3 60

[

/

x 40 p

20 2

,/

0 100 2.5 t1 80

- 2.0 e

-E.

60

/

1.5 a-E E

40

_ op 1.0 3

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/

05 0

Y 0

160 N

140 0

0, e #-

120

/.'

160 UNIRRADIATED

$ 100 g-80 3.06 X 1018 IRR ADIATED (5500F) -

120 n/cm2 0

$ 60 80 3 z

O 45 F 0

40 0

45 F 20

[

40 l

l l

l 0

200 100 0

100 200 300 400 TE MPER ATUR E (OF)

Figure 5-3. Charpy V Notch impact Data for McGuire Unit 2 Reactor Vessel Weld Metal 5-14

48/66/2316-13 12216-13 (OC)

-100

-50 0

50 100 150 200 120 l

l l

2 2

100

-M e

e 80

[

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40

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W" 20 g

e,e, 2

0 100 2.5 g

80 8-2.0 d

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n o e

k 60 e

1.5 I

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7 w

40 S

1.0 -

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550F 3

20 OI 0.5 e

0 0

140 120 g

e-160

- 100 UNIRRADIATED O

e 8

120 80 o

g O

e 60 e

IRRADIATED (550 F) 80 0

O 9 2

40

" " * ^ =

40 gop 20 0

O 200 100 0

100 200 300 400 TEMPERATURE (O )

F Figure 5-4. Charpy V-Notch Impact Data for McGuire Unit 2 Reactor Vessel Weld HAZ Metal 5-15

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Figure 5-7. Cnarpy Irnpact Specimen Fracture Surf aces for McGuire Unit 2 i
Reactor Vessel Wold Metal t
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Figure 5 9. Comparison of Actual Versus Predicted 30 Ft Lb Transition Temperature increases for the McGuire Unit 2 Reactor Vessel Material Based on the Prediction Methods of Regulatory Guide 1.99, Revision 1 1
5-20 1
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g 30 k g g __
TOTAL ELONGATION
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t 10 UNIFORM ELONGATION l
l I
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100 200 300 400 500 600 TEM PER ATURE (OF)
Figure 510. Tensile Properties for McGuire Unit 2 Reactor Vessel Shell Forging 05 (Axial Orientation) 5-21
48/66/2216-16 12216 16 (DC) 0 50 100 150 200 250 300 120 l
l l
l l
l l-800 110 100 ULTIM ATE TENSILE STRENGTH - 700
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o 30 k g_
O TOTAL ELONGATION 20 as 1
e n
10
~
g UNIFORM ELuNGATION O
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O 100 200 300 400 500 000 TEMPER ATURE (OF)
Figure 5-11. Tensile Properties for Mcuuire Unit 2 Reactor Vessel Shell Forging 05 (Tangential Orieritation) 5-22
48/66/2216-17 13216-17 (OC)
O 50 100 150 200 250 300 120 l
l l
l l
l l-800 110 100 700 ULTIMATE TENSlLE STRENGTH
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600 80 y
9 E
f 00 g O
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=
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50 h
3 40 i:o g
30 TOTAL ELONGATION O
20
=A F
10 S
g j
j UNIFORM ELONGATION O
O 100 200 300 400 500 600 TEMPER ATURE (OF)
Figure 5-12. Tensile Properties for McGuire Unit 2 Reactor Vessel Weld Metal 5-23
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0 10THS INCHES 1
F;gure 5-13. Fractured Tensile Specimens From McGuire Unit 2 Reactor Vessel Shell Forging 05 (Axial Orientation) 1 5-24
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5-25
0 SPECIMEN DW-4 75 F y,
, 7-.,,
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0 10THS INCHES 1
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48A'9/221618
'12216 18 120 108 96 84 w
72 E*
\\
48 36 SPECIMEN DW 6 0
24 TEST TEMP 550 F-12 -
1 O
^
I I
I I
I I
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0 0.03 0.06 0.09 0.12 0.15 0.18 0.21 0.24 0.27 0.30 STR AIN (in/in)
Figure 5-16. Typical Stress-Strain Curve for Tension Specimens 5-27 l
i
SECTION 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6 -1.
INTRODUCTION Knowledge of the neutron environment within the reactor pressure vessel / 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 nettron 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.
This section describes a discrete ordinates S transport analysis performed n
for the McGuire Unit 2 reactor to determine the fast (E > 1.0 MeV) neutron flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules. The analytical data were then used to
' develop lead factors for use in relating neutron exposure of the reactor vessel to that of the surveillance capsules. Based on the use of spectrum-averaged reaction cross sections derived from this calculation and the McGuire Unit 2 power history, the analysis of the neutron dosimetry contained in Capsule V is presented.
6-2.
DISCRETE ORDINATES ANALYSIS A plan view of the McGuire Unit 2 reacter geometry at the core midplane is shown in figure 6-1.
Since the reactor exhibits 1/8th core symmetry, only a 3767e:ld/111185 6-1
45-to 90-degree sector is depicted. Six irradiation capsules attached to the neutron pad are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at 58.5 degrees (V,Y) and 56 degrees (U,W,X,Z) f rom the cardinal axes as shown in figure 6-1.
A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in figure 6-2.
The stainless steel specimen containers are approximately 1-inch square and approximately 56 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot-high reactor core.
From a neutron transport standpoint, the surveillance capsule struttures are significant. They have a marked ef fect on both the distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pad and the reactor vessel. In order to properip determine the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. This requires at least a two-dimensional calculation.
In the analysis of the neutron environment within the McGuire Unit 2 reactor geometry, predictions of neutron flux distributions and energy spectra were made with the 001 two-dimensional discrete ordinates transport code. The radial and azimuthal distributions were obtained from an R,0 calculation l
wherein the geometry shown in figures 6-1 and 6-2 was represented in the analytical model. In addition to the R,0 calculation, a second calculation in R,Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest.
In the R,Z analysis, the reactor core was treated as an equivalent volume cylinder. The surveillance capsules were not included in the R,Z model.
Both the R,e and R,Z analyses employed 47 neutron energy groups and a P 3
expansion of the scattering cross sections. The cross sections used in the analyses were obtained from the SAILOR cross section library [6] which was 3767e:1d/030686 6-2
developed specifically for light water reactor applications. The neutron energy group structure used in the analysis is listed in table 6-1.
A key inpat parameter in the analysis of the integrated neutron exposure of the reactor vessel is the core power distribution. For this analysis, core power distributions representative of time-averaged conditions derived from statistical studies of Icng-term operation of Westinghouse 4-loop plants were employed. These input distributions include rod-by-rod spatial variations for all peripheral fuel assemblies.
This generic, design basis, core power distribution is intended to provide a vehicle for the long-term (end-of-life) projection of reactor vessel exposure. Since plant-specific core power distributions reflect only past operation, their use for projection into the future may not be justified. The use of generic data which reflects long-term operation of similar reactor cores may provide a more suitable approach.
Benchmark testing of these generic core power distributions and the SAILOR cross sections against surveillance capsule data obtained from two, three,
and four-loop Westinghouse plants indicate that this analytical approach yields conservative results, with calculations exceeding measurements from 10 to 25 percent.
One further point of interest regarding these analyses is that the design basis assumes an out-in fuel loading pattern (fresh fuel on the periphery).
Future comitment to low-leakage core loading patterns could significantly reduce the calculated neutron flux levels presented in section 6-4.
In addition, surveillance capsule lead factors could be changed, thereby influencing the withdrawal schedule of the remaining surveillance capsules.
Having the results of the R,0 and R,Z calculations, three-dimensional variations of neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor.
3767e:ld/llll85 6-3
$(R,Z,0,E ) = $(R,0,E ) x F(Z,E )
(6-1) g where
$(R,Z,0,E ) = neutron flux at point R,Z,0 within energy group g g
$(R,0,E )
= neutron flux at point R,0 within energy group g g
obtained from the R,0 calculation F(Z,E )
= relative axial distribution of neutron flux within energy g
group g obtained from the R,Z calculation This analysis is consistent with established ASTM standards.
6-3.
RADIOMETRIC MONITORS The passive radiometric monitors included in the McGuire Unit 2 surveillance program are listed in table 6-2.
The first five reactions in table 6-2 are used as fast neutron monitors to relate fast (E > 1.0 MeV) neutron fluence to measured material property changes.
In order to address the potential for burnout of the product nuclides generated by fast neutron reactions, it is necessary to also determine the magnitude of the thermal and resonance region neutron' fluxes at the monitor location. Therefore, bare and cadmium-shielded cobalt-aluminum monitors are also included.
The relative locations of the various radiometric monitors within the surveil?ance capsule are shown in figure 4-2.
The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the axial center of the capsule. All monitors are located radially at the center of the capsule and azimuthally within ! 0.23 degrees of the capsule center.
3767e:1d/llll85 6-4
The use of passive monitors such as those listed in table 6-2 does not yield a direct measure of the energy-dependent neutron flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated ef f ect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived f rom the activation measurements only if the irradiation parameters are well known.
In particular, the following variables are important.
l o The operating history of the reactor o The energy response of the monitor o The neutron energy spectrum at the monitor location o The physical characteristics of the monitor The analysis of the passive monitors and the subsequent derivation of the average neutron flux requires two operations. First, the disintegration rate of product nuclide per unit mass of monitor must be determined. Second, in order to define a suitable spectrum-averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.
The specific activity of each of the monitors is determined using established ASTM p roc ed u re s. [13,14,15,16,17,18,19,20,21 ] Following sample preparation, the activity of each monitor is detennined by means of a lithium-drif ted germanium, Ge(L1), gamma ray spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration.
For the samples removed f rom McGuire Unit 2, the overall 2a deviation in the measured data is determined to be plus or minus 10 percent. The neutron energy spectrum at the monitor location is determined analytically using the method described in paragraph 6-2.
Having the measured activity of the monitors and the neutron energy spectrum at the monitor locations of interest, the calculatiori of the neutron flux 3767e:ld/llll85 6-5
proceeds as follows. The reaction product activity in the monitor is expressed as n
A = N, F Y o(E) $(E) dE (1-e-j) e d
(6-2)
E max
),)
where A
= induced product activity (dps per gram)
N
= number of target element atoms per gram 9
F
= weight fraction of the target nuclide in the target material Y
= number of product atoms produced per reaction o(E)
- energy dependent reaction cross section
$(E)
- energy dependent neutron flux at the monitor location with the reactor at full (reference) power P)
= average core power level during irradiation period j P,,,
= maxi m m or reference core power level A
= decay constant of the product nuclide t)
= length of irradiation period j t
= decay time following irradiation period j d
n
= total number of irradiation periods Because the neutron flux distributions are calculated using multigroup transport methods and, further, because the main interest is in the fast (E > 1.0 MeV) neutron flux, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the following relation.
I t
F l
a(E) $(E) dE = a $f JE l
where 4
l r
i 3767e:ld/llll85 6-6
47 a(E) $(E) dE "g *g o
a=1
/=
18 1 MeV
+g g=1
' 18 f=31MeV,(E)dE=[+9 g=1 1
g = group number f rom Table 6-1 Thus, equation (6-2) is rewritten i
i n
(1-e-At)),-Atd A = N, F Y a $f
=x 33 or, solving for the fast (E > 1.0 Mev) neutron flux,
}
3767e:1d/111185 6-7
-.--.----,m,
$f=
n
-At ),-Atd (6-3)
N, F Y a (1-e j
aax j.i The total fast (E > 1.0 MeV) neutron fluence is then given by n
'f " 'f t)
(6-4) p j-1 max where n
total ef fective full power seconds of reactor P,, tj = operation up to the time of capsule removal 3,j An assessment of the thermal neutron flux levels within the survelliance capsules is obtained f rom the bare and cadmium-covered Co ' (n,Y) Co data by means of cadmium ratios and the use of a 37-barn, 2,200 m/sec cross section. Thus, 0-1 R
E 3
bare D
(6-5)
$7h
~~N F n~~
P o
Ys v g _,-At ) g-Atd j
j'=7 max where D is defined as R
/Rg g.
l j
An assessment of the potential for product nuclide burnout has determined that l
such an effect is negligible for McGuire Unit 2 Surveillance Capsule V.
6-4.
NEUTRON TRANSPORT ANALYSIS RESULTS l
Results of the discrete ordinates transport calculations for the McGuire Unit 2 reactor are summarized in this section.
In figure 6-3, the calculated maximum fast (E > 1.0 MeV) neutron flux levels at the radius of the surveillance capsule center, the reactor vessel inner radius, the reactor l
l 3767e:1d/111185 6-8
vessel 1/4 thickness location, and the reactor vessel 3/4 thickness location are presented as a function of azimuthal angle. The local influence of the surveillance capsules on the fast neutron flux distribution is clearly evident.
In figure 6-4. the radial distribution of maximum f ast (E > 1.0 MeV) neutron flux through the thickness of the reactor vessel is shown. The relative axial variation of fast neutron flux within the reactor vessel is given in figure 6-5.
Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6-3 or 6-4 by the appropriate values from figure 6-5.
Table 6-3 provides the calculated fast neutron exposure parameters for the McGuire Unit 2 reactor vessel.
Table 6-4 provides the calculated fast neutron exposure parameters and updated lead factors for all of the McGuire Unit 2 surveillance capsules. The lead factor is defined as the ratio of the fast (E > 1.0 MeV) neutron flux at the docimeter block location (capsule center) to the maximum fast neutron flux at the reactor vessel inner radius.
In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The calculated neutron energy spectrum at the center of the McGuire Unit 2 surveillance capsule V is listed in table 6-5.
The calculated spectrum-averaged cross sections for each of the fast neutron reactions are given in table 6-6.
6-5.
DOSIMETRY RESULTS The irradiation history of the McGuire Unit 2 reactor up to the time of removal of Capsule V is listed in table 6-7.
Comparisons of measured and calculated saturated activity of the radiometric monitors contained in Capsule V based on the irradiation history shown in table 6-7 are listed in table 6-8.
The fast (E > 1.0 MeV) neutron flux and fluence levels derived for Capsule V using the spectrum averaged cross-sections listed in table 6-6 are presented in table 6--3.
Table 6-10 summarizes the thermal neutron flux obtained f rom the cobalt-aluminum monitors.
i 3767e:ld/121985 6-9 l
l
An examination of table 6-9 shows that the average fast (E > 1.0 MeV) neutron 10 flux derived f rom the five threshold reactions ranges f rom 8.33 x 10 II 2
n/cm -set to 1.09 x 10 n/cm -sec, a total span of less than 31 II 2
percent. The calculated flux value of 1.16 x 10 n/cm -sec exceeds all of the measured values, with calculation to experimental ratios ranging from 1.06 to 1.39.
A summary of measured and calculated current fast neutron exposures for Capsule V and for key reactor vessel locations is presented in table 6-11.
The measured value is given based on the average of all five threshold reactions listed in table 6-9.
End-of-life (EOL) reactor vessel fast neutron fluence projections are also included in table 6-11.
Based on the data given in table 6-9, the best estimate fast neutron exposure of Capsule V is 18 4 = 3.06 x 10 n/cm2 (E > 1 MeV) at 1.03 EFPY.
i 1
3767e:1d/121985 6-10
TA8LE 6-1 SAILOR 47 NEUTRON ENERGY GROUP STRUCTURE Group Group Energy Lower Energy Energy Lower Energy Group (MeV)
GrouD (MeV) 1 14.19(a) 25 0.183 8
2 12.21 26 0.111 3
10.00 27 0.0674 4
8.61 28 0.0409 5
7.41 29 0.0318 6
6.07 30 0.0261 7
4.97 31 0.0242 8
3.68 32 0.0219 9
3.01 33 0.0150
-3 10 2.73 34 7.10x10
-3 11 2.47 35 3.36x10
-3 12 2.37 36 1.59x10
~4 13 2.35 37 4.54x10
~4 14 2.23 38 2.14x10 15 1.92 39 1.01x10'4
-5 16 1.65 40 '
3.73x10 17 1.35 41 1.07x10-
-6 18 1.00 42 5.04x10 19 0.821 43 1.86x10~
~I 20 0.743 44 8.76x10 21 0.608 45 4.14x10^
22 0.498 46 1.00x10-23 0.369 47 0.00 24 0.298 l
a) The upper energy of group 1 is 17.33 MeV.
3767e:1d/111185 6-11
TABLE 6-2 NUCLEAR CONSTANTS FOR RADIOMETRIC MONITORS CONTAINED IN THE MCGUIRE UNIT 2 SURVEILLANCE CAPSULES Reaction Target Fission of Weight Product Yield Monitor Material Interest Fraction Half-life (5)
Iron wire Fe (n,p) Mn$*
0.0585 314 dy g
Nickel wire NiS8 (n.p) Co 0.6777 71.4 dy 8
63 (n,a) Co60 Copper wire Cu 0.6917 5.27 yr Uranium-238(a) in U 0 (n, ) s 1.0 30.2 yr 6.0 38 Neptunium-237(a) in Np0 Np (n,f) Cs 1.0 30.2 yr 6.5 2
Cobalt-aluminum (a) wire CoS9 (n,y) Co60 0.0015 5.27 yr Cobalt-aluminum wire Cob 9 (n,y) Co60 0.0015 5.27 yr i
t I
l l
i a) Denotes that the monitor is cadmium-shielded l
3767e:1d/111185 6-12
TABLE 6-3 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS FOR 1HE PEAK LOCATION OF THE MCGUIRE UNIT 2 REACTOR VESSEL Iron Radial Location Fast Neutron Flux Displacement 2
Within the (n/cm -sec)
Rate Reactor Vessel (E > 1.0 MeV)
(E > 0.1 MeV)
(dpa/sec) 10 10
-II Inner Surface 2.86 x 10 5.65 x 10 4.40 x 10 (R = 86.500 inches) 10 10
-II 1/4 Thickness 1.59 x 10 4.98 x 10 2.77 x 10 (R = 88.657 inches) 9 10
-12 3/4 Thickness 3.22 x 10 2.30 x 10 9.08 x 10 (R = 92.970 inches) 9 10
-12 Outer Surface 1.28 x 10 1.20 x 10 4.46 x 10 (R = 95.126 inches) i 3767e:1d/111485 6-13 i
TA8LE 6-4 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AND LEAD FACTORS FOR THE MCGUIRE UNIT 2 SURVEILLANCE CAPSULES Iron i
Azimuthal Fast Neutron Flux Displacement I
Capsule Location ")
(n/cm -sec)
Rate Lead I.D.
LDecrees)
(E > 1.0 MeV) (E > 0.1 MeV)
(dpa/seci Factor II
-10 U
56*
1.36 x 10 6.21 x 10" 2.77 x 10 4.76 W
124*
1.36 x 10" 6.21 x 10" 2.77 x 10 4.76
-10 X
236*
1.36 x 10" 6.21 x 10 "
2.77 x 10 4.76
-10 Z
304*
1.36 x 10" 6.21 x 10" 2.77 x 10 4.76
-10 V
58.5*
1.16 x 10" 5.14 x 10" 2.32 x 10 4.06 l
-10 Y
238.5*
1.16 x 10" 5.14 x 10" 2.32 x 10 4.06
-10 i
a) The radius of the surveillance center is 81.620 inches.
b) The lead factor is the ratio of the fast (E > 1.0 MeV) neutron flux at the center of the surveillance capsule to that at the peak location on the reactor vessel inner surface.
l 3767e:1d/111185 6-14
TABLE 6-5 CALCULATED NEUTRON ENERGY SPECTRUM AT THE CENTER OF MCGUIRE UNIT 2 SURVEILLANCE CAPSULE V Energy Neutron Flux Energy Neutron Flux 2
Group (nic_m -sec)
Group (n/cm -sec) 2 I
1 2.16 x 10 25 6.48 x 10 10 I
2 7.90 x 10 26 6.73 x 1010 8
3 2.72 x 10 27 5.42 x 10 10 8
4 4.95 x 10 28 3.80 x 10 10 8
5 8.18 x 10 29 1.15 x 1010 9
6 1.80 x 10 30 6.17 x 10' 7
2.47 x 10' 0
31 1.66 x 10 9
8 4.97 x 10 32 1.06 x 1010 9
4.52 x 10 33 1.92 x 1010 10 3.79 x 10' 10 34 2.84 x 10 11 4.55 x 10' 10 35 4.81 x 10 12 2.28 x 10' 10 36 4.31 x 10 13 6.99 x 10 37 6.03 x 10 14 3.51 x 10' 1
38 3.29 x 10 15 9.41 x 10' 1
39 3.65 x 10 10 16 1.27 x 10 40 4.93 x 1010 17 1.98 x 10 41 5.82 x 10 10 10 18 4.41 x 10 42 3.22 x 1010 0
19 3.37 x 10 43 3.68 x 10 10 10 20 1.60 x 10 44 2.29 x 10 10 21 5.82 x 10 45 1.78 x 1010 0
22 4.55 x 10 46 2.52 x 10 10 7.3 5.69 x 10 47 3.13 x 10 10 0
24 5.51 x 10 3767e:1d/111185 6-15 a
TABLE 6-6 SPECTRuft-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF MCGUIRE UNIT 2 SURVEILLANCE CAPSULE V Spectrum-Averaged Cross Section(a)
Reaction of Interest (barns)
Fe (n.p) Mn s
0.0590 Ni (n.p) Co 0.0816 Cu63 (n a) Co60 0.000529 238 (n,f) Cs137 U
0.3185 Np (n,f) Cs 3.277 i
i a(E) $(E) dE a) i=
I1MeV a(E) dE 1
I l
3767e:1d/111185 6-16
-.~
TABLE 6-7 1RRADIAT10N HISTORY OF MCGUIRE UNIT 2 SURVEILLANCE CAPSULE V P)
P,,x P)/P Irradiation Time Decay Time max Month Year (MWt) (MWt)
(Days)
_ (Days)
.5 1983 52 3411 0.015 24 838 l
6 1983 391 3411 0.115 30 808 7
1983 1
3411 0.000 31 777 8
1983 960 3411 0.281 31 746 9
1983 1783 3411 0.523 30 716 10 1983 1891 3411 0.554 31 685 11 1983 2330 3411 0.683 30 655 12 1983 2201 3411 0.645 31 624 1
1984 579 3411 0.170 31 593 2
1984 2742 3411 0.804 29 564 3
1984 3101 3411 0.909 31 533 4
1984 3197 3411 0.937 30 503 5
1984 2865 3411 0.840 31 472 6
1984 3309 3411 0.970 30 442 7
1984 2095 3411 0.614 31 411 8
1984 826 3411 0.242 31 380 9
1984 3156 3411 0.925 30 350 10 1984 3038 3411 0.891 31 319 11 1984 2532 3411 0.742 30 289 12 1984 2311 3411 0.678 31 258 1
1985 3247 3411 0.952 25 233 CAPSULE V REMOVED Note: 1) Decay time is referenced to 9/16/85
~~2) Total irradiation time is 3.24 x 107 effective full power seconds (EFPS) or 1.03 effective full power years (EFPY)~~
~~3) Pj is the average core power level during the irradiation period 3767e:1d/111185 6-17~~
TABLE 6-8 COMPADISON OF MEASURED AND CALCULATED RADIOMETRIC MONITOR SATURATED ACTIVITIES FOR MCGUIRE UNIT 2 SURVEILLANCE CAPSULE V Radiometric Monitor Monitor Saturated Activity and
( Di sinteg rat ions /Second -Gram)*
Axial locaticq(a)
Measured Calculated C/E fe (n.P) Mn 6
Top 3.22 x 10 6
Middle 3.36 x 10 6
Bottom 3.28 x 10 Average 3.29 x 106 4.48 x 106 1.36 Ni (n.9) Co Top 4.91 x 10 Middle 4.43 x 10 7
Bottom 5.01 x 10 Average 4.78 x 107 6.68 x 107 1.40 Cu (n.o) Co Top 3.38 x 10 Middle 3.55 x 10 Bottom 3.31 x 10 Average 3.41 x 10 4.07 x 10 1.19
~~Basis is one gram of wire.~~
3767e:1d/111185 6-18 l
TABLE 6-8 (continued)
COMPARISON OF MEASURED AND CALCULATED RADIOMETRIC MONITOR SATURATED ACTIVIllES FOR MCGUIRE UNIT 2 SURVEILLANCE CAPSULE V Radiometric Monitor Monitor Saturated Activity and (Disintegration 5/Second-Gram)*
Axial Location (a)
Measured Calculated C/E t
U238 (n.f) CsI3I (b) 6 Middle 5.95 x 10 Corrected (c) 5.30 x 106 5.62 x 106 1.06 Mp237-(n.f) CsI3I (bl Middle 5.23 x 107 6.25 x 107 1.20 b
Co ' (n.v) Co60 (b)
I Top 4.44 x 10 Middle 4.16 x 10 Bottom 4.18 x 10 Average 4.26 x 107 5.43 x 107 1.27 CoS9 (n.y) Co60 r
7 Top 8.86 x 10 Middle 7.34 x 10 Bottom 7.80 x 10 Average 8.00 x 10 7.62 x 10 0.95 l
3767e:1d/111185 6-19 1
TABLE 6-8 (continued)
COMPARISON OF MEASURED AND CALCULATED RADIDMETRIC MOWITOR SATURATED ACTIVITIES F0E MCGulRE UNIT 2 SURVEltLANCE CAPSULE V
)
a) Refer to Figure 4-2 for the locations of the various radicmetric monitors, b) This radiometric monitor was cadmium shielded.
c) The measured value has been moltiplied by 0.09'to correct for the effect of 322 ppe U and the build-in of Pu ',
i r
r
,I
't 1
f 3767e:1d/111185 6-20 l
TABLE 6-9 RESULTS OF FAST NEUTRON 00SIMETRY FOR MCGUIRE UNIT 2 SURVEILLANCE CAPSULE V Current Radiometric Monitor Fast (E > 1.0 MeV)
Fast (E > 1.0 MeV)
Saturated Activity (*
Neutron Flux Neutron Fluence 2
Reaction (dos /am)
(n/cm -sec)
(n/cm 1 of Interest Measured Calculated Measured Calculated Measured Calculated b4 6
6 10 18 Fe (n.p) Mn 3.29x10 4.48x10 8.54x10 2.77x10 8
10 18 Ni (n.p) Co 4.78x10 6.68x10 8.33x10 2.70x10 Cu63 (n,m) Co60 5
5 10 18 3.41x10 4.07x10 9.76x10 3.16x10 238 (n,f) CsI3 '
5.30x10 5.62x10 1.09x10" 3.53x10 6
6 18 U
l I
I 10 18 Np (n,f)'Cs 5.23x10 6.25x10 9.67x10 3.13x10 10 18 18 Average 9.44x10 1.16x10" 3.06x10 3.76x10 i
a) Refer to Table 6-9.
b) Total irradiation time for surveillance capsule V is 3.24x107 effective full power seconds (EFPS).
I l
3767e:1d/111285 6-21 l
TABLE 6-10 RESULTS OF THERMAL NEUTRON 00SIMETRY FOR MCGUIRE UNIT 2 SURVEILLANCE CAPSULE V Saturated Activity Axial (dos /em)
+
2 Location Bare Cd-covered (n/cm _3,c)
Il Top 8.86 x 10 4.44 x 10 1.25 x 10 7
7 10 Middle 7.34 x 10 4.16 x 10 9.03 x 10 7
7 ll Bottom 7.80 x 10 4.18 x 10 1.03 x 10 7
7 II Average 8.00 x 10 4.26 x 10 1.06 x 10 2
(
1 e
j i
I 3767e:1d/012786 6-22
TABLE 6-11
~~SUMMARY~~
OF MCGUIRE UNIT 2 FAST NEUTRON FLUENCE RESULTS BASED UPON SURVEILLANCE CAPSULE V End of Life Current fast (E > 1.0 MeV)
Fast (E > 1.0 MeV)
Neutron Fluence (a)
Neutron FluenceID) 2 (n/cm )
(n/cm )
IC)
Calculated Measure.l(c)
Calculate 0 location Measured 18 18 Capsule V 3.06x10 3.76x10 lI lI I9 I9 Vessel IR 7.54x10 9.28x10 2.35x10 2.89x10 II ll I9 I9 Vessel 1/4T 4.19x10 5.16x10 1.31x10 1.61x10 16 II 18 18 Vessel 3/4T 8.47x10 1.04x10 2.64x10 3.25x10 a) Current fluences are based on operation at 3411 MWt for 1.03 EFPY b) EOL fluences are based on operation at 3411 MWt for 32 EFPY.
c) The measured results of surveillance Capsule V were extrapolated to the reactor vessel locations using the following calculated lead factors:
Inner Radius - 4.06 1/4 Thickness - 7.30 3/4 Thickness - 36.13 3767e:1d/012786 6-23
11756 1 CORE BARREL 6=650 0 = 58. S '
/
6 = 56 *
/
NEUTRON PAD 0 = 450 e
/
/ /
~
/ /
/
7
//[
/
? l
/
/'
//
I
//
/
V t
l Y
a
~~X Figure 6-1.~~
McGuire Unit 2 Reactor Geometry l
3767e:1d/111185 6-24
-,,--.,,-ee,--
i f
' ' ' ' rCHARPY SPECIMEN
/ l
/////
/////7 A
NN % N N N % % N N % % N N N N N K NEUTRON PAD
\\ N N N \\ \\ \\ \\ \\\\ \\ \\\\ \\ \\\\ \\ \\
i l
f Figure 6-2.
Plan View of a Dual Reactor Vessel Surveillance Capsule 3767e:ld/lll185 6-25
11756 3 100.0
)
i 70.0 50.0
?
E 20.0 x
U SURVEILLANCE CAPSULE y 10.0 R = 20731 m N
Z ER 5
5.0
>a lE PRESSURE VESSEL IR I
2.0 f
1/4t LOCAT 1.0 d
0.7 Z 0.5 3/41 LOCATION 0.2 -
l l
l 0.1 45 50 55 60 65 70 75 80 85 90 AZIMUTHAL ANGLE (DEGREE)
Figure 6-3.
Calculated Azimuthal O t'it' lon of Maximum Fast (E > 1.0 MeV) Neutrco nk hin the Reactor Vessel - Surveillance capw : Geometry 3767e:1d/111185 6-26
11756 2 i
100.0 2
70.0 219.71 50.0
~
225.19
~~o 20.0 e~~
X U 10.0 m
=
to 7.0 5.0 1
I j
2.0 l
A 1.0 m
N 0.7
~
~
2 J
0.5 0.2 1R 1/4t 3/4t OR 0.1 215 220 225 230 235 240 245 RADIUS (cm)
Figure 6-4.
Calculated Radial Distribution of Maximum Fast (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel 3767e:1d/111185 6-27
11756-5 1.000 0.700 0.500 0.200 0.100 y 0.070 d
2 0.050 e
H B
2
$ 0.020 P<
dx 0.010 0.007 l
0.005 CORE MIDPLANE 0.002 TO VESSEL
' CLOSURE HEAD 0.001 300
-200 100 0
100 200 300 DISTANCE FROM CORE MIDPLANE (cm) l Figure 6-5.
Relative Axial Variation of Fast (E > 1.0 Mev)
Neutron Flux Within the Reactor Vessel
(
3767e:1d/111185 6-28
SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule is reconnended for future capsules to be removed from the McGuire Unit 2 reactor vessel.
Lead Removal Estimated Fluence Capsule
,Factot Time [a]
(n/cm x 10 'l I
V 4.06 Removed (1.03) 0.306 (Actual)
X 4.76 4
1.72[b]
U 4.76 7
3.01[c]
Edl Y
4.06 15 5.50 W
4.76 Standby Z
4.76 Standby a.
EFPY from plant startup b.
Approximates vessel end of life 1/4 thickness wall location fluence c.
Approximates vessel end of life inner wall location fluence d.
Approximates vessel inner wall location fluence for plant life extension to 60 EFPY. Should be adjusted accordingly if plant life is extended to another time period.
l 4
i
\\
3767e:ld/030686 7 -1
SECTION 8 REFERENCES 1.
~~Koyama, K., Davidson, J.~~
A., " Duke Power Company William 8. McGuire Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9489, May,1979.
2.
ASTM E185-73 " Practice for Surveillance Tests for Nuclear Reactor" in ASIM Standards, Part 10 (1973), Amet ican Society for Testing and Materials, Philadelphia, Pa., 1973.
3.
ASTM E185-82 " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," in ASTM Standards, Section 3. Volume 03.01. (1984), American Society for Testing and Materials, Philadelphia, PA 1984.
4.
Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, April 1977.
l 5.
Soltesz, R. G., Disney, R.
K., Jedruch, J., and Ziegler, S.
L., " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5 - Two-Dimensional Discrete Ordinates Transport Technique," WANL-PR(LL)-034, Vol 5, August 1970.
6.
"0RNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".
7.
Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology (to be published).
i 8.
ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance," in ASTM Standards,
)
4 Section 12, American Society for Testing and Materials, P511adelphia, Pa.,
1984.
l 3767e:ld/030686 8-1 l
9.
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.
10. 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.
11. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
12. 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, Philadelphia, Pa., 1984.
13. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,
~~1984.~~
14. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
15. 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.
16. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards, Section 12, America Society for Testing and Materials, Philadelphia, Pa., 1984.
3767e:ld/010286 8-2 l
l
17. 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,~~
18. ASTM Designation ES23-82, " 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.
19. 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,
20. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.
21. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
3767e:ld/010286 8-3 l
APPENDIX A HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION A -1.
INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART RT is designated as the higher of either NDT.
NDT the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.
Rf increases as the material is exposed to fast-neutron radiation.
MDT Therefore, to find the most limiting RT at any time period in the NDI reactor's life, ART due to the radiation exposure associated with that NDT time period must be added to the original unirradiated RT The extent of NDT.
the shift in RT is enhanced by certain chemical elements (such as copper NDT and phosphorus) present in reactor vessel steels. Design curves which show the effect of fluence and copper and phosphorus contents on ARI r
NDI reactor vessel steels are shown in figure A-1.
Given the copper and phosphorus contents of the most limiting material, the radiation-induced ART can be estimated from figure A-1.
The most NDT limiting material occurs in the reactor vessel intermediate shell forging 05 Heat No. 526840 listed in table A-1.
Fast neutron fluence (E > 1 Mev) at the vessel inner surface, the 1/4 T (wall thickness), and 3/4 T (wall thickness) vessel locations are given as a function of full-power service life in figure A-2.
The data for all other ferritic materials in the reactor coolant pressure boundary are examined to ensure that no other component will be limiting with respect to RT NOT' 3767e:ld/012786 A -1
A-2.
FRACTURE 100GHNESS PROPERTIES The preirradiation fracture-toughness properties of the McGuire Unit 2 reactor vessel materials are presented in table A-1.
The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review PlanUl. The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the McGuire Unit 2 Vessel Material Surveillance Program.
A-3.
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERA 10RE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup g
or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. K is btained from the IR reference fracture toughness curve, defined in Appendix G to the ASME Code.E 3 1he K curve is given by the following equation.
IR KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)]
(A-1) where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code [2] as follows.
CKg& Kit < KIR (A-2) i l
l 3767e:1d/012986 A-2
where IM = stress intensity factor caused by membrane (pressure) stress K
= stress intensity factor caused by the thermal gradients It KIR = function of temperature relative to the RINDT f the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K is determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference f racture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K r ne It, reference flaw are computed. From equation A-2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.
3767e:1d/012786 A-3
During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel 10.
This condition, of course, is not true for the steady-state situation.
It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K
at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K IR exceeds Kg, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4 T IR crack during heatup is lower than the K f r the 1/4 T crack during IR steady-state conditions at the same coolant temperature.
During heatup, especially at the end of the transient, conditions may exist so that the ef fects of compressive thermal stresses and lower K
's do not offset each IR other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed
)
in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
3767e:1d/012786 A-4
i The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 T deep outside
_ surface flaw is assumed. Unlike the situation at the vessel inside surface, I
the thermal gradients established at the outside surface during heatup produce l
stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
}
l Following the generation of pressure-temperature curves for both the I
steady-state and finite heatup rate situations, the final limit curves are
{
produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three j
values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is i
possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Then, composite curves for the heatup rate data and the cooldown 4
rate data are adjusted for possible errors in the pressure and temperature j
sensing instruments by the values indicated on figures A-3 and A-4.
Finally, the new 10CFR50[3] rule which addresses the metal temperature of
}
the closure head flange and vessel flange regions is considered. This 10CFR50 l
rule states that the metal temperature of the closure flange regions must 1
z exceed the material RT by at least 120*F for normal operation when the NOT pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for McGuire Unit 2).
Table A-1 indicates that the limiting RT I
N01 l'F occurs in the closure head flange of McGuire Unit 2, and the minimum allowable temperature of this region is 121*F at pressures greater than 621 psig.
l 3767e:ld/012786 A -5
f A -4. HEATUP AND COOLDOWN LIMIl CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in section A-3.
The derivation of the limit curves is presented in the NRC Regulatory Standard i
Review Plan.
Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program. Charpy test specimens from Capsule V indicate that the surveillance weld metal and limiting core region intermediate shell forging 05 exhibited shifts in RT f 45*F and 70*F, respectively. These NDT 18 2
shifts at a fluence of 3.06x10 n/cm are within the appropriate design curve (figure A-1) prediction. As a result, the heatup and cooldown curves are based on the ART given in figure A-1 for the most limiting beltline NDT material which is the intermediate shell forging 05. The resultant heatup and cooldown limit curves for normal operation of the reactor vessel are presented in figures A-3 and A-4 and represent an operational time period of 8 EFPY.
The heatup limit curves in Figure A-3 is not impacted by the new 10CRF50 rule. However, the cooldown curves are impacted by the 10CFR50 rule as shown in Figure A-4.
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in figure A-3.
This is in addition to other criteria which must be met before the reactor is made critical.
The leak test limit curve shown in figure A-3 represents minimum temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of references 2 and 4.
Figures A-3 and A-4 define limits for ensuring prevention of nonductile
~~failure, i~~
3767e:ld/012786 A -6
TABLE A-1 t
NCGUIRE UNIT 2 BEACTOS VESSEL 100GMNESS TABLE Material Unser Shelf Enerov y
RTNO fede m
Specification Heat Cu P
N1 Tg(f)
(*F){8)
(ft-Ib)
(ft-Ib)
Component Number Number (5)
(5)
(5)
.011
.63
-31 12 132 Closure head done A5338, C1. 1 55154-1 closure head ring A500 C1. 2 007055
.006
.86 16 16 156 Closure head flange A500 C1. 2 526916
.012
.82
-13 1
155 Vessel flange A508 C1. 2 218572
.016
.82
-4
-4 174 Inlet nozzle A508 C1. 2 526341-1
.04
.006
.76
-13
-13 141 Inlet nozzle A500 C1. 2 526395-1
.05 009
.73
-31
-31 114.5 Inlet nozzle A508 C1. 2 526537
.06
.009
.76
-22
-22 129 Inlet nozzle A500 C1. 2 526537
.06 009
.78
-40
-40 132 Outlet nozzle A500 C1. 2 526341
.04
.007
.77
-13
-7 124 Outlet nozzle A508 C1. 2 525789
.05
.011
.83
-40
-24 103 Outlet nozzle A508 C1. 2 525789
.05
.010
.84
-49
-16 116 Outlet nozzle A508 C1. 2 526395-2
.03
.010
.14
-40
-30 121
~
i Nozzle shell A500 C1. 2 411085
.006
.89
-4 1
151.5 (c)
Inter. shell A508 C1. 2 526840
.16
.012
.85
-4
- 4(b) 147 96(b) 141(b)
Lower shell A500 C1. 2 411337-11
.15
.004
.88
-30
-30(b) 152(C) 109 Bottom head ring.
A508 C1. 2 527428
.06
.013
.77
-4 15 Bottom head segment A5338 C1. 1 55126-2
.007
.59
-49
-2 136 Bottom head segment A5338, C1. 1 55126-2 007
.59
-40
-40 131 Bottom head segment A5338, C1. 1 55292-2
.006
.58
-13
-13 142 7
Bottom head segment A5338, C1. 1 55292-2
.006
.58
-13
-13 132 i
N Bottom head done A5338, C1. 1 55292-3
.006
.58
-40
-40 127 Intermediate to lower shell weld (d)
.05
.010
.70
-76
-68(b) 127(b)
Weld HAZ
-76
-76(b) 125.5(b) led 0 - Major Working 01rection INed0 - Normal to Major Working 01rection a) Estimated per MUSEG-0000 USnBC Standard Soview Plan, Branch Tech. Position - NTE8 5-2.
b) Based on actual data.
c) 1005 shear not reached, upper shelf energy is greater than listed.
d) Submerged arc weld (weld wire heat 895075 and Grau to Flum Lot No. P46),
i i
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l l
l l
l l
0 2
4 6
8 10 12 14 16 18 20 22 24 26 28 30 32 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)
Figuiu A-2. Fast Neutron Fluence (E>1 Mev) as a Function of Full Power Service Life (EFPY)
A-9
16412.1 2600 Z
INSERVICE LEAK TEST
- MINIMUM TEMPERATURE \\ l 2
1 I
2200 Z l
I
__ CURVE APPLICABLE FOR l
! ll
- THE SERVICE PERIOD UP TO 2000 - 8 EFPY AND CONTAINS Z MARGINS OF IO'F ANO l
I 60 PSIG FOR POSSIBLE l
1800 - INSTRUMENT ERRORS gl E
~
l 1600 w
l 1400 HEATUP RATES TO u) 60*F/HR MATERIALS BASIS Z
CONTROLLING MATERIAL A
1200 Z
-REACTOR VESSEL INTERMEDIATE SHELL 05 y
1000 COPPER CONTENT O.16 u
RT INITIAL-4*F 800 NOT RT AFTER 8 EFPY MT 600 l/4T 85'F 2
3/4T.36'F
~
400 CRITICALITY LIMIT Z
/
200 Z I I III II II! IIIII II fI II fIIII IIII If f O
50 100 150 200 250 300 350 400 INDICATED TEMPERATURE (*F)
Figure A-3 McGuire Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 8 EFPY A-10
i 16412.3 2600
~
I 2400 2 CURVE APPLICABLE FOR THE SERVICE PERIOD UP TO 8 EFPY AND 2200 CONTAINS MARGINS OF IO'F AND 60 PSIG FOR POSSIBLE Z
INSTRUMENT ERROR 2
2000 Z
O 1900 m
E 5
~
B600 W
1400 1200 0
2 MATERIAL BASIS N
1000 Z CONTROLLING MATERIAL 6
Z
-REACTOR VESSEL 000 COOLDOWN RATE INTERMEDIATE SHELL 05
~~F/HP COPPER CONTENT O.I6~~
}
RT
~~INITIAL,~~
-4*F 600 WT Z
60 RT AFTER 8 EFPY 400 100 l/4T.05'F 3/4T.36*F 200
=IIIIIIIIII IIIIIIIIIIIIllIIIIIIIIII 0
O 50 100 150 200 250 300 350 o
INDICATED TEMPERATURE (*F) t i
i Figure A 4 McGuire Unit No. 2 Reactor Coolant System Cooldown Limitations Applicable for the First 8 EFPY A-11
APPENDIX A REFERENCES l
1.
" Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants. LWR Edition, NUREG-0800,1981.
2.
ASME Boiler and Pressure Vessel Code,Section III, Division 1 -
Appendices, " Rules for Construction of Nuclear Vessels," Appendix G,
" Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York,1983.
3.
Code of Federal Regulations,10CFR50, Appendix G. " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,
Amended May 17, 1983 (48 Federal Register 24010).
4.
" Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.
3767e:ld/012786 A-12
-_.

v t e Letter Sequence Other
TAC:59962, (Open) TAC:59963, (Open) TAC:62779, (Open) TAC:62780, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML20100H434, ML20100H458, ML20141C857
MONTHYEAR1986-10-21 [Table View]

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