ML20070W037

Analysis of Capsule U from Commonwealth Edison Co Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program

ML20070W037
Person / Time
Site:	Braidwood
Issue date:	03/31/1991
From:	Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20070W033	List: ML20070W032 ML20070W037
References
WCAP-12845, NUDOCS 9104120295
Download: ML20070W037 (165)

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WESTINGHOUSE CLASS 3 WCAP-12845 ANALYSIS OF CAPSVLE V FROM THE COMMONWEALTH EDISON COMPANY BRAIDWOOD UNIT 2 REACTOR VESSEL RADIATION SVRVEILLANCE PROGRAM E. Terek S. L, Anderson

., L. Albertin March 1991 4

Work Performed Under Shop Order BMVP-106 -

Prepared by Westinghouse Electric Corporation for the Connonwealth Edison Company

. Approved by:

T. A. Meyer, Manaher b -

Structural Reliability and

. Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division

" P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 C 1990 Westinghouse Electric Corp.

m_ -

1 PREFACE This report has been technically reviewed and verified, Reviewer i

1 Sections 1 through 5, 7, 8 J. M. Chicots N, b4-1'rb.

and Appendix B

/

Section 6 E. P. Lippincott . <-

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TABLE OF CONTENTS Seetion Iltle Pl91 1.0

SUMMARY

OF RESULTS 1-1 2.e INTRODUCTION 2-1

3.0 BACKGROUND

3-1

4.0 DESCRIPTION

OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE U 5-1 s 5.1 Overview 5-1 5,2 Charpy V-Notch Impact Test Results 5-4 5,3 Tension Test Results 5-7 5.4 Compact Tension Tests 5-8 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6.1 Introouction 6-1 6.2 Discrete Ordinates Analysis 6-2 6,3 Neutron 00simetry 6-7 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1

8.0 REFERENCES

8-1

d APPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS 11 1

LIST OF ILLUSTRATIONS -

Fiaure Iltle fjLqt L4-1 Arrangement of Surveillance Capsules in the Braidwood 4-9 Unit 2 Reactor Vessel 4-2 Capsule U Diagram Showing Location of Specimens, Thermal 4-10 Monitors and Dosimeters 5-1 Charpy V-Notch Impact Properties for Braidwood Unit 2 5-16 Reactor Vessel Shell Forging 500102-1/50C97-1 (Tangential Orientation)

, .c 5-2 Charpy V-Notch Impact Properties for Braidwood Unit 2 5-)

Reactor Vessel Shell Forging 500102-1/50C97-1 /

l(Axial Orientation) 5-3 Charpy V-Notch Impact Properties for Braidwood Unit 2 5-18 Reactor Vessel Weld Metal

, 5-4 -Charpy V-Notch Impact Properties for Braidwood Unit 2- 5-19L

-Reactor Vessel Weld Heat Affected Zone Metal 5-5: -Charpy Impact Specimen fracture Surfaces for Braidwood

• 5-20

. Unit'2 Reactor' Vessel Shell Forging 50D102-1/50C97-1 (Tangential Orientatiori) \

5-6 Charpy impact Specimen. Fracture Surfaces for Braidwood 5 Unit 2 Reactor- Vessel Shell Forging 150D102-1/50t97-1 e-

_(Axial Orientation) til 4

11 LIST OF ILLUSTRATIONS (Cont)

E199n lillt EARE 5-7 Charpy impact Specimen Fracture Surfaces for Braidwood 5-22 Unit 2 Reactor Vessel Weld Metal 5-8 Charpy impact Specimen Fracture Surfaces for Braidwood 5-23 Unit 2 Reactor Vessel Weld Heat Affected Zone Metal 5-9 Fracture Appearance of Specimen FL13 5-24 5-10 Fracture Appearance of Specimen FL4 5-25 5-11 Fracture Paths in Heat-Affected-Zone Charpy Specimens 5-26

. FH1 and FH2 5-12 Tensile Properties for Braidwood Unit 2 Reactor Vessel Shell 5-27 Forging 50D102-1/50C97-1 (Tangential Orientation) 5-13 Tentile Properties for Braidwood Unit 2 Reactor Vessel Shell 5-28 Forging 50D102-1/50C97-1 (Axial Orientation) 5-14 Tensile Properties for Reaidwood Unit 2 Reactor Vessel Weld 5-29

-Meta' 5-15 Fractured Tensile Specimens from Braidwood Unit 2 Reactor 5-30 Vessel Shell Forging 50D102-1/50C97-1 (tangential Orientation) 5-16 Fractured Tensile Specimens from Braidwood Unit 2 Reactor 5-31 Vessel Shell Forging 500102-1/50C97-1 (Axial Orientation) iv

. 1

LIST OF ILLUSTRATIONS (Cont)

Limtte lillt EA91 5-17 Fractured Tensile Specimens from Braidwood Unit 2 Reactor 6 32 Vessel Weld Metal Stress-Strain Curves for Tension Specimens FL1 and FL2 5-33

• 5-18 5-19 Streas-Strain Curves for Tension Specimens FL3 and Fil $-34 5-20 Stress-Strain Curves for Tension Specimens F12 and FT3 5-35 5-21 Stress-Etrain Curves for Tension Specimens FW1 and FW2 5-36 5-22 Stress-Strain Curve for Tension Specimen FW3 $-37 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 6-2 Core Power Distributions used in 1ransport Calculations 6-14 for Braidwood Unit 2

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LIST OF TABLES lahlt lilla hat 4-1 Chemical Composition of the Braidwood Unit 2 4-3 Reactor Vessel Intermediate Shell Forging 4-2 Chemical Composition of the Braidwood Unit 2 4-4 >

Reactor Vessel Lower Shell Forging 4-3 Chemical Composition of the Braidwood Unit 2 4-5 Reactor Vessel Weld Metal Used for the Upper to Lower Shell Closing Girth Seam 4-4 Heat Treatment History of the Braidwood Unit 4-6

. No. 2 Reactor Vessel Beltline Materiai 4-5 Chemical Composition of Braihood Unit 2 Capsule U 4-7 Irradiated Charpy impact Specimens 4-6 Chemistry Results from the NBS Certified Reference Standards 4-B 5-1 Charpy V-Notch Impact Data for the Braidwood Unit 2 5-9 Forging 500102-1/50C97-1 Irradiated at 550'F, Fluence 3.91 x 1018 n/cm2 (t > },o g,y)

,' 5-2 Charpy V-Notch Impact Data for the Braidwooo Unit 2 5-10 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550'F, Fluence 3.91 x 1018 n/r.m2 (E > 1.0 MeV) 5-3 Instrumented Charpy impact Test Results for the Braidwood 5-11 Unit 2 Shell Forging 500102-1/50097-1 Irradiated at 550'F, Fluence 3.91 x 10I8 n/cm2 (E > 1.0 kev) vi

LIST OF TAhlES (Cont)

Table 11tle Etqt 5-4 Instrumented Charpy Impact Test Results for the Braidwood b-12 Unit 2 Weld Metal and HAZ Metal Irradiated at 550'F, Fluence 3.91 x 1018 n/cm2 (E > 1.0 MeV) 5-5 Effect of 550'f irradiation to 3.91 x 1018 n/cm 2

$ 33 (E > 1.0 MeV) on Notch Toughness Properties of Braidwood Unit 2 Reactor Vessel Surveillance Materials A

5-6 Comparison of Braidwood Unit 2 Surveillance Material 5-14 30 ft-lb Transition Temperature Shif ts and Upper Shelf .'

Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions .-

5-7 Tensile Properties for Braidwood Unit 2 Reactor Vessel 5-15 Surveillance Material Irradiated at 550'F to 3.91 x 10I0 n/cm2 (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Parameters at the 6 15 Surveillance Capsula Center 6-2 Calculated Fast Neutron Exposure Parameters at the 6-16 ,

Pressure Vessel C1&d/ Base Metal Interface .

6-3 Relative Radial Distributions of Neutron flux 6-17 (E > 1.0 MeV) within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron flux 6-18 (E > 1.0 MeV) within the Pressure Vessel Wall vii l

i i

LIST OF TABLES (Cnnt) 4 lable Iltle 11qt 6-5 Relative Radial Distributions of Iron Displacemer,t Rate 6-19 (dpa) within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-20 6-7 Irradiation History of Neutron Sensors Contained in 6-21 Capsule U 6-8 Measured Sensor Activities and Reactions Rates 6-22 6-9 Summary of Neutron Dosimetry Results 6-24 6-10 Comparison of Measured and Ferret Calculated Reaction 6-25 Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-26 Capsule Center 6-12 Comparison of Calculated and Measured Exposure Levels 6-27 for Capsule U 6-13 Neutron Exposure Projections at Key locations on the 6-28 Pressure Vessel Clad / Base Metal Interface for Braidwood Unit 2 6-14 Neutron Exposure Values for use in the Generation of 6-29 Heatup/Cooldown Curves 6-15 Updated lead Factors for Braidwood Unit 2 Surveillance 6-30 Capsules viii

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SECil0N 1.0

$UMHARY OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule V, the first capsule to be removed from the Commonwealth Edison Company Braidwood Unit 2 reactor pressure vessel, led to the following conclusions:

o The capsule received an average fast neutron fluence (E > 1.0 Mev) of 3.91 x 1018 n/cm2 after 1.15 EFpY of plant operation.

o Irradiation of the reactor vessel lower shell forging 50D102-1/50C97-1 Charpy specimens to 3.91 x 10 18 n/cm2 (E > 1.0 MeV) resulted in no 30 and 50 f t-lb transition temperature increases for specimens oriented parallel to the major working direction

.. (tangential orientation). This results in a 30 ft-lb transition temperature of -10'F and a 50 f t-lb transition temperature of 15'F (tangential orientation),

o leradiation of the reactor vessel lower shell forging S00102-1/50C97-1 Charpy specimens to 3.91 x 1018 n/cm2 (E > 1.0 MeV) resulted in 30 and 50 ft-lb transition temperature increases of 5 and 10'F, respectively, for specimens oriented normal to the major working direction (axial orientation). This results in a 30 f t-lb transition temperature of -20'f and a 50 ft-lb transition temperature of 10'F (axial orientation).

o The weld metal Charpy specimens irradiated to 3.91 x 1018 n/cm 2 (E > 1.0 MeV) resulted in no 30 ft-lb transition temperature increase and a 50 ft-lb transition temperature increase of 5'F. This results in a 30 ft-lb transition temperature of -20'F cnd a 50 f t-lb transition temperature of 45'F for the weld metal.

P l-1 1

j i

) o Irradiation of the reactor vessel weld HAZ metal Charpy specitens to l 3.91 x 1018 n/cm2 (E > 1.0 MeV) resulted no 30 ft-lb and 50 f t-lb *l i transition temperature increases and no USE decrease. This results ,

~

l in a 30 f t-lb transition temperature of -135'r and a 50 f t-lb

! transition temperature of -105'r fer the weld HAT metal. ,i o The average upper shelf energy of lower shell forging I

500102-1/50097-1 (tangential orientation) resulted in no energy l

decrease af ter irradiation to 3.91 x 1018 n/cm2 (E > 1.0 MeV).

! This results in an upper shelf energy of 168 f t-lb (tangential

orientation).

I o The average upper shelf energy of lower shell forging 500102-1/50097-1 (axial orientation) resulted in a decrease in energy of 16 ft-lb after irradiation to 3.91 x 10 18 n/cm2 (E > 1.0 MeV).

This results in an upper shelf energy of 137 f t-lb (axial ..

orientation).

o The average upper shelf energy of the weld metal resuit:d in a decrease 9 f t-lb after irradiation to 3.91 x 10 18 n/cm2 (E > 1.0 MeV). This results in an upper shelf energy of 62 ft-lb.

o The surveillance capsule U test results do not indicate any significant changes in the RTNDT values of the reactor vessel surveillance materials, o The surveillance capsule materials exhibit a more than adequate upper ,

shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 f t-lb (

throughout the life (32 EFPY) of the vessel as required by 10CFR50, Appendix G.

1-2

i i

I i

o The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the Braidwood Unit 2 reactor vessel is as follows:

2 l Vessel inner radius * - 3.03 x 1019 n/cm l

Vessel 1/4 thickness - 1,66 x 1019 n/cm 2

2 i Vessel 3/4 thickness - 3.57 x 1018 n/cm

• Clad / base metal interface j o The above calculated 32 EFPY fluences are based on the original core and are expected to decrease with the innplementation of a low leakage fuel management program.

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SECTION 2.0 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 tu ef fer.t; of neutron irradiation on the Commonwealth Edison Compara Braldwood Unit 2 reactor pressure vessel materials under actual vperating conditions.

The suneillance program for the Braidwood Unit 2 reactor pressure vessel materi:1s was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-lll88

• Commonwealth Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Surveillance Program" by L. R. Singer III. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels"(6),

Westinghouse Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule "U" from the resctor and its shipment to the Westinghouse Science and Technology Center Hot Cell facility, where, the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

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

9 2-1

J I

SECTION 3.0 BACKGROUND

. The ability of the larp . teel pressure vessel containing the reactor core and itc 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 f ast neutron bombardment. The overall effects ci fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA508 Class 3 (base material of the Commonwealth Edison Company station Braidwood Unit 2 reactor pressure vessel lower shell forging) 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 til of the ASME Boiler and Pressure Vessel Code W .

The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)-

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 60*f less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (axial) to the major working direction of the material . 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 i

, G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Kjg curve, allowable -

stress intensity factors can be obtained for this material as a function of i temperature. Allowable operating limits can then be determined using these j allowable stress intensity factors.

3-1 l

i RTNDT and, in turn, the operating li.aii. of nuclear power plants can be ,

l adjusted to account for the effects of radiation on the reactor vessel material

. properties. The radiation embrittitment changes in mechanical properties of a ,

j given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Braidwood Unit 2 Reactor Vessel Radiation Surveillance .

program,Ill in which a surveillance capsule is periodically removed from the i operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTND7 (RTNDT initial +

6RTNDT) 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.

4 e

e 6

3-2

1 SECTION

4.0 DESCRIPTION

Of PROGRAM s

Six surveillance capsules for monitoring the effects of neutron exposure on the

]* Braidwood Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up. The six capsules were i positioned in the reactor vessel between the neutron shield pads and the vessel l wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. l Capsule U was removed after 1.15 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact tension T.T) specimens (Figure 4-2) from the lower shell forging 50D102-1/50097-1 and weld metal representative of the intermediate to lower shell beltline weld seam of the reactor vissel and Charpy V-notch specimens from weld heat-affected-zone (HAZ) material. All heat-affected zone specimens were obtained from within the HAZ of forging 50D102-1/50C97-1 of the representative weld.

The chemical composition and heat treatment of the surveillance material is presented in Tables 4-1 through 4-4. The chemical analysis reported in Ta',les 4-1 through 4-3 were obtained from unirradiated beltline material, in add'. ion, a chemical analysis using Inductively Coupled Plasma Spectrometry (ICPS) was performed on irradiated specimens from forging 500102-1/50017-1 and weld metal and is reperted in Table 4-5, also, reported in Table 4-6 are the chemistry results from the NBS certified reference standards.

All test specimens were machined from 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. Base metal Charpy V-notch impact and tension specimens were oriented with the lon'gitudinal axis of the specimen parallel to the major working direction of the forging (tangential orientation) and also normal to the major working direction (axial orientation). Charpy V-notch and tensile specimens from the weld metal were oriented such that the long dimension of the specimen was normal to the welding direction, i

4-1 i _ _ . _ _

l 1

The 1/2T CT test specimens in Capsule V are from the lower shell course forging 500107.-1/50C97-1 and were machined in both the axial and tangential orientations. Thus, the simulated crack in the specimen will propagate normal ,

and parallel to the major working direction of forging 50D102-1/50C97-1. The 1/2T CT Test specimens from the weld metal were machined with the notch ,

oriented in the direction of welding. Thus, the simulated crack in the specimen will propagate parallel to the weld direction. All CT specimens were fatigue precracked according to ASTM E399.

Capsule V contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium shielded dostmeters of neptunium (Np237) and uranium (U238) were placed in the capsule to measure the integrated flux at specific neutron energy levels.

Thermal monitors made from the two low-melting eutectic alloys and scaled in .-

Pyrex tubes were included in the capsule. The composition of the two alloys and their melting points are as follows:

• 2.5% Ag, 97.5% Pb Melting Point: 579'F (304*C) 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 590'F (310*C)

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

l l

4-2

TABLE 4-1  !

l CHEMICAL COMPOSITION OF THE BRAIDWOOD UNIT 2 REACTOR VESSEL INTERMEDIATE SHELL FORGING [b]

Chemical Compositon I (weight %) l Eletnent Upper Shell Forging 11 MK24 2 C .20l*l Mn 1.33 P .007-S .007

. SI .25 Ni .71

., Mo .53 Cr .08 Cu .03 Al .024 Co .012 Pb .0003 max.

W .005 max.

Ti .001 max

! Zr .005 max V .01 max.

Sn .003 As .006 Cb .005 max.

Ng .0097

. B Not Reported

a. Chemical Analyses by Japan Steel Works, Ltd.

b. Data reported here is the unirradiated chemistry results reported in WCAP-lll88 Ill 4-3

TABLE 4-2 CHEMICAL COMPOSITION OF THE BRAIDWOOD UNIT 2 REACTOR VESSEL LOWER SHELL FORGING ICI ,,

Chemical Compositon (weight %)

0 Lower Shell Forging 11 MK24 3 5

C .2 21*l .241*l Mn 1.30 1.38 P .006 .013 S .004 .009 Si .28 .30 .

Ni .75 .77 Mo .49 .56 ,.

C- .06 .095 Cu .06 .057 Al .025 .024 Co .011 .008 Pb .0003 max. < .001 W .005 max. < .01 Ti .005 inax. .004 Zr .005 max. < .002 V .01 max. < .002 Sn .007 .004 As .008 .007 -

Cb .005 max. < .002 .

N2 .0084 .009 -

B Not Reported < .001 .

a. Chemical Analyses by Japan Steel Works, Ltd.

b. Westinghouse Analyses from the Surveillance Program Test Plate.

c. Data reported here is the unitradiated chemistry results l reported in WCAP-lll88 Ill l

4-4 l

TABLE 4-3 j CHEMICAL COMPOSITION OF THE BRAID /OOD UNIT 2 REAC10R VESSEL WELD METAL USED FOR THE UPPER TO LOWER SHELL CLOSING GIRTH SEAM IC3 l

l l

Chemical Compositon Element Weld Filler Wire Hear Number 442011 Linde No. 80 flux. L..o, t Number 0344 l C .066W .069M Mn 1,44 1.45 P .015 .011 S .012 .013

. SI .48 .53 Ni .67 .64

. Mo .44 .40 Cr .10 .082 Cu .04 .040 Al .004 .007 Co .011 .004 Pb .0006 < .001 W .010 < .01 Ti .007 .003 Zr .003 < .002 V .006 < .002 Sn .005 .004 As .004 .004 Cb .004 < .002 N, .013 .012

. B .0007 < .001

a. Chemical Analysis of " Filler Wire Qualification Test" by Babcock and Wilcox, Company, Test No. WF 562

b. Westinghouse Analyses from the Su veillance Program Test Weldment,

c. Data reported here is the unirradiated chemistry results reported in WCAP-lll88 Ill 4-5

TABLE 4-4 l

HEAT TREATHENT HISTORY OF THE BRAIDWOOD UNIT NO. 2 REACTOR VESSEL Bil1LINE MATERI AL (13 g ,

~

Temperature Time 3*3 unterial ('F) (hr) Cooling Austenittring: 6 .7 53*3 Water quenchec Upper 1600 t 25 Shell Forging (871'C)

Tempered: 12.251*l Alt cooled 490963}*

• 49C904) 1226 t 25 +

(MK24 2) (663'C)

Stress Relief: 11.75fbl Furnace cooled 1150 t 50 (621'C)

Austenttizing: 6.5 (*) Water quenched 25 Lower 1600 (871 gC) .

Shell Forging Tempered: 1 2.2 5181 Air cooled 1225 t 25 500102l1.11 (663'C) g7 Stress Relief:

(MK24 3) 1 1 .7 5161 Fumace cooled 1150 t 50 (821*C) '

Upper Shell To Lower Strees Relief: 11.75fbl Fumace cooled Shell Cloei 1150 t 50 Girth Wold m (621*C)

Pese 4assit, mm unos so.

Las h 0844)

~

Surveillance Program Test Maternal Surveillance Program Post Wold- -l Test Forging Strees Roliet: 1 4 .2 5161181 Furnace-cooled 500102 1 1150 t 50

• 50C971 - (621'C) - -

Surveillance Poet Wold .

Program Tom Stress Relief: 12.5 lbilcl Fumace cooled Weidment 50 1150

'(621 gC)

a. Dess ottamed from Japan Steel Worus. Ltd. Matenal Test Reports.

b. Desa from annoook and w6icos, co.

c. The Strees Relef Heat Trootment recetved by tre Survoulance Test Forging and Wendment have toen mmutated.

4-6 l

.- .- . w .

j i

TA8LE 4-5 ,

l 1

CHEMICAL COMPOSITION OF BRAIDWOOD UNIT 2 CAPSULE U

, IRRADIATED CHARPY IMPACT SPECIMENS t Chemical Composition (wt.%)

l Specimen No. Cu Ni C Mn P S Si Cr . No V Co ,

i  !

FL-6 0.049 L 745 0.229 1.261 <.005 <.003 0.294 0.033 0.483 <.002 <.002  !

FW-1 0.032 0.704 0.070 1.628 0.011 0.467 0.090 0.466 0.006 <.002 l FW-7 0.034 0.754 1.687 0.013 0.009 0.450 0.092 0.503 0.007 <.002

FW-14 0.032 0.698 0.068 1.583 0.009 0.009 0.088 0.458 0.006 <.002

FW-2, 0.026 0.623 ,

FW-2 0.028 0.635  !

FW-3 0.031 0.679  !

FW-4 0.029 0.644  !

FW-5 0.032 0.699 4 i FW-6 0.034 0.765 ,

FV-8 0.031 0.673 0.038 FW-9 0.034 0.724 0.010  !

FW-10 0.035 0.747 FW-11 0.033 0.711  ;

FW-12 0.031 0.688 .

FW-13 0.035 0.750 0.010 FW-15 0.031 0.685 '

Analyses Method of Analysis j

Metals ICPS, Inductively Coupled Plasma Spectrometry Carbon EC-12, LECO Carbon Analyzer i Sulfur Combustion / titration j Silicor. Dissolution / gravimetric f

Iron (Matrix Element: Remainder by Difference)

• Second run to show duplication of results

, i 4-7 b

TABLE 4-6 4

CHEMISTRY RESULTS FROM THE NBS CERTlflED REFERENCE STANDARDS Material ID Low Alloy Steel: NBS Certified Reference Standards t

NBS 361 _ NBS 362 Certified Measured (a) Certified Measured (a)

Metals Concentration in Weight Percent l

Fe

• 95.60 (matrix) 95.30 (matrix)

Co 0.032 0.033 0.300 0.318 ,

Cu 0.042 0.043 0.500 0.514 l Cr 0.694 0.663 0.300 0.297 Mn 0.660 0.644 1.040 1.050 Mo 0.190 0.193 0.068 0.054 Ni 2.000 2.072 0.590 0.610 P 0.014 0.0144 0.041 0.0417 V 0.011 0.011 0.040 0.040 C 0.383 0.386 0.150 0.162/0,161 S 0.014 N. A. 0.036 0.0354 Si 0.222 0.208/0.219 0.390 0.383 Material ID Low Alloy Steel: NBS Certified Reference Standards NBS 363 NBS 364 Certified Measured (a) Certified Measured (a)

Metals Concentration in Weight Percent Fe

• 94.40 (matrix) 96 ' (matrix) to 0.048 0.051 0.10 0.149 Cu 0.100 0.102 0.249 0.252 Cr 1.310 1.315 0.063 u,058 NI 0.300 0.314 0.144 0.139 Mn 1.500 1.539 0.255 0.250 Mn 0.028 0.025 0.490 0.491 P 0.C29 0.0285 0.010 0.0096 V 0.310 N. A. 0.105 0.100 .

C 0.620 N. A. 0.87 N. A. .

S 0.0068 N. A. 0.0250 0.0247 Si 0.740 0.710 0.065 N. A.

1 .

• Matrix element calculated as difference for material balance.

l Tentative value, certiffed 100% of value.

l N. A. - Not analyzed L

l (a) Method of analysis -- Inductively Coupled Plasma Spectrometry (ICPS) for I all elements except C, S and Si.

4-8

O'

~ REACTOR YESSEL

/ CORE BARREL j NEUTRON PAD 3 (301.5') 2 -

CAPSULE U (58.6')

/ y c c%, 56.6' \p 58.6' N

/ 61'

~

l 270' 90' 1

(241 ') Y (238. 5 ') X  %.-

W (121.5')

REACTOR Vt!SSEL 180' f PLAN VIEW l

~W l [ -)

\'

CAPSULE CORE ll!lllllll;l -

.- . CORE

, Q MIDPLANC s

l  : s Nb

NEUTRON PAD F CORE BARREL ELEVATION VIEW Figure 4-1. Arrangement of Surveillance Capsules in the Braidwood Unit 2 P.eacter Vessel 4-9

1. .% op***

CDE _ _ ' " . _"' " - , _"*

U U h%4 hNI4 ,Y $ k hN!h N

hh j l g $g pgg m ,-n ,,,o .. ... , ,, nn  % ,,,

LEGEND: FL LOWER SHELL FORGING 60D1021/50C971 (TANGENTIAL)

FT . LOWER SHELL FORGING 5001021/50C971 (AXIAL)

FW . WELD METAL FH HEAT AFFECTEDZONE MATERIAL e

m

'-O h

s

'M 40 pg n.. n ., ni, n ,, n, n, n, ,o n, 'u r.

h - - -

em n, w g, n- n. n,, ,,,, n, n, m ,o n, n8 ru m n, ni m

[ ._

[ ewi e,,, g, no nu no nie m n, n. n. 7 7 m Si AginnTURE CARD Also Ava'iblile 00 Aperture CDMI figure 4-2 Capsule U Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters C1l04tgogc;5.o) 4-10

. ~ . _ , , ,

J

1 1

SECTION 5.0 TESTING Of SPECIMENS FROM CAPSULE U 5.1 Overvi_ew The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Power Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and HI23, ASTM Specification E185-82[6), and Westinghouse Procedure MHL 8402, Revision 1 as I modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1. I Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-lll88(Il. 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 E23-88l73 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J 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 (ED ). From the load-time curve (Appendix B), the load of general yielding (Pgy), the time to general yielding (tcy), the maximum load (Pg), and the time to maximum load (tg) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pr), and the load at which fast fracture terminated is identified as the arrest load (PA)*

5-1

The energy at maximum load (Eg) was determined by comparir.g 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 )D and the energy at maximum load. .

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

oy = Pay * {L/[B*(W-a)2*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 & = 45' and p = 0.010', Equation 1 is valid with with C = 1.21. Therefore (for L =

4W),

oy = Pgy * {L/[B*(W-a)2*l.21)) = (3.3P gy)/[B(W-a)2) W (2) c 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 Pay (3)

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

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods 11 compliance with ASTM Specification A370-89I83, i

The lateral expansion war measured using a dial gage rig similar to that shown in the same specification.

5-2

j l Tension tests were performed on a 20,000 pound Instron, split-console test

machine (Model 1115) per ASTM Specification EB-89bl93 and [21-79 l f, (1988)(103, 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 irtprove axiality of loading. The tests l were conducted at a constant crosshead speed of 0,05 inches per minute throughout Ae test.

! Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch, The extensometer is rated as Class B-2 perASTMEB3-85Illl.

1 Elevated test temperatures were obtained with a three-zone electric resistance

-, split-tube furnace with a 9-inch hot zone. All tests were conducted in air.

'. Because of th0 difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.

ChmC.-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip, in the test configuration, with a slight load on the specimen, a plot of specimen 4 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 actuel testing the grip temperatures were used to obtained desired specimen temperatures. Experiments indicated that this me nod is accurate to 2*F, l[ The yield load, ultimate load, fracture load, total elongation, and uniform l, 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

i. computed using the final diameter measurement.

5-3 l-

-__a m c----u- .=e.g -W y-mvs=- < - - - , . , ww-.P 4,== -wva----er ---*.9 --ypq-ev -.t--up-' ++----,m7.,- + .g-,,g-.y 2-39 6eim 4---- , . - t.g---w.,-mvg .-,--- w.-

l i

[ 5.2 Charov V-Notch Impact Test Results i .

The results of Charpy V-notch impact tests performed on the various materials '

contained in Capsule V irradiated to 3.91 x 10 l8 n/cm2 (E > 1.0 MeV) are l

! presented in Tables 5-1 through 5-4 and are compared with unirradiated ,

resultsII3 as shown in Figures 5-1 through 5-4. 1he transition temperature increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-5.

Irradiation of the reactor vessel lower shell for,ging 500102-1/50097-1 Charpy specimens to 3.91 x 10 18 n/cm2 (E > 1.0 MeV) at 550'F (Figure $~1) resulted no 30 and 50 ft-lb transition temperature increases for specimens oriented perpendicular to the major working direction (tangential orientation). This resulted in a 30 ft-lb transition temperature of -10'F 1 and a 50 f t-1b transition temperature of 15'F for specimens oriented perpendicular to the major working direction (tangential orientation). .

The average upper shelf energy (USE) of the lower shell forging .

SM102-1/50C97-1 Charpy specimens (tangential orientation) resulted in no energy decrease after irradiation to 3.91 x 10 18 n/cm2 (E > 1.0 MeV) at 550'F, This resulted in an average USE cf 168 f t-lb (Figure 5-1).

Irradiation of the reactor vessel lower shell forging 500102-1/50C97-1 Charpy specimens to 3.91 x 10 I8 n/cm2(E>1.0MeV)at550'F(Figure 5-2) resulted in a 30 f t-lb transion temperature increase of 5'F and a 50 f t-lb transition temperature increase of 10'F for specimens oriented parallel to the major working direction (axial orientation). This resulted in a 30 ft-lb '

transition temperature of -20'F and a 50 f t-lb transition temperature of 10'F for specimens oriented perpendicular to the major working direction l

(axial orientatioit).

5-4

\

The averago upper shelf energy (USE) of the lower . hell farsing 50D102-1/50C97-1 Charpy speetmens (axial orientation) resulted in a decreaso of 16 ft-lb in energy after irradiatinn to 3.91 x 10 l8 n/cm2 (E > 1.0 ,HeV) at 550'F. This resulted in an average USE of 137 f t-1b (Figure 5-2).

Irradiation of the reactor vessel core region weld metal Charpy specimens to 3.91 x 1018 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-3) resulted in no 30 f t-lb transition temperature increase and a 50 f t-lb transition temperature increase of 5'F. This resulted in a 30 ft-1b transition temperature of

-20'F and a 50 f t-lb transition t'emperature of 45'r The average upper shelf energy (USE) of the reactor vessel core region weld metal resulted in a decrease of 9 ft-lb in energy af ter irradiation to 3.91 x 1018 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in $a average USE of 62 ft-lb.

Irradiation of the reactor vessel weM metal Heat-Affe:ted Zone (HAZ) specimens to 3.9\ x 10 IU n/cm2 (E > 1.0 MeV) at 550*F (Figure 5. 4) resulted in no 30 and 50 ft-lb transition temperature increases. This resulted in a 30 f t-lb ,

transition temperature of -135'F and a 50 ft-lb transition temperature of

-105'F The average upper saelf energy (USE) of the reactor vessel HAI metal resulted in an increase of 45 f t-lb af ter irradiation to 3.91 x 10 18 n/cm2 (E > 1.0 MeV) at 550*F, however this is not unexpected due to the large teatter of

data points. 'his resulted in an average USE of 200 ft-lb.

The fracture appearance of each 1" radiated Chrrpy specimen from the various

,' materials is shown in Figures 5 5 through 5-0 and show t.n increasingly ductile l

or tougher appearance with. increasirig test temperature.

l l

5-5

. _ . . . - . . - . = - . - , .. - - . - . - - - -. - ._ -

A comparison of the 30 ft-lb transition temperature increases for the various Braidwood Unit 2 surveillance materials w',th predicted increases using the -

methods of NRC Regulatory Guide 1.99, Redston 2(33 is presented in Table 5-6. This comparison indicates that thv thnsition temperature increases and the USL decreases resulting from irradiation tu 3.91 x 10 18 n/cm2 (E > 1.0 ,

MeV) are less than the Guide predictions.

Unusual energy .nd fracture behavior was shown by tangl.ntial base metal i

specimens (U3 and FLl4. The impact energy value of ','6.0 f t-lb at -35'l for specimen h13 i close to the impact energy value of 71.0 f t-lb for specimen FL11 wh th was tested at +20'F, yet our test vecords indicate that specimen FL13 was tested at the prescribed temperature. The fracture apeetrances of r ciws FL13 and F111 are also similar (Figure 5-5), and ation of a irregular f acture path and brine 111ng of the notched f ace of

r u .a FL13 (Tigure 5-9) suggest that the high energy vam is correct and that the specimen was tested properly. The impact energy va'rc cf 19.0 ft-lb ,

for specimen FL4 tested at -25'F seem; to be low. It should be w *e like 30 't-lb. Th' fracture appearance of this specimen (Figure 5-5; at ggests a low ..

irpact enemy v0 m, and the more brittle fracture path of this sptcL'on seen in Figure H O is ni a in line with a low fracture toughness behayW HAI metal Charpy spet/ mn FH2 showed an unusually high imtact energy s./;ue (.'43 ft-lb) at 250 'F. I k specimens fracture path was examined and compared to the fracture path of another HAZ specimen (FHl) which showed a considrably lower impact energy value (162 ft-lb) at 225'F. The results are shown in Figure 5-11. The comparison showed that fracture in specimen FH2 was more irreguira *M in specimen FHl. Fractura in specimen FH2 appeared to have staaed in 1 iA y tough microstructure of the Heat-Affect-Zone, with some fracture actually nit iating toi. side the notch root. The specimen, 4n effect, '

'aehaved if ke a blunt atch ty, specimen, thus accounting for the high energy nine during fracture.

The ) pad-time records for the .9dividual instromented Charpy specimens are contained in Appendix A.

5-6

6.3 Tension Test Results The results of tension tests performed on shell forging 500102-1/50097-1 (tangential and axial orientation) and the weld oetal irradiated to 3.91 x 1018 n/cm2 (E > 1.0 MeV) are shown in Table 5-7 and are compared with unirradiated results{ll as shown in Figures 5-12 through 5-14. The tension test results for forging 500102-1/50C97-1 are shown in Figures 5-12 and 5-13 and indicated that irradiation to 3.91 x 1018 n/cm2 (E > 1.0 MeV) caused a less than 6 ksi increase in the 0.2 percent offset yield strength and ultimate tensile strength. The weld metal tension tests results are shown in Figure

, 5-14 and show that the ultimate tensile strength and the 0.2 percent offset yield strength increased by less than 5 ksi with irradiation to 3.91 x 1018 n/cm2 (E > 1.0 MeV). The small increases in 0.2% yield strength and tensile strength exhibited by the forging material and weld metal indicate that these materials are not highly sensitive to irradiation to 3.91 x 1016 n/cm2 (E >

., 1.0 HeV), as is also ir. die:tc? the Charpy impact *.est results. The fractured tension specimen > for the forging material are shown in Figures 5-15 and 5-16, while the fractured specimens for the weld metal are shown in Figure 5-17. The engineering stress-strain curves for the tension tests are shown in Figures 5-18 through 5-22.

9 0

6 5-7 1

I (3 ___.

5.4 [xanart._.Tensinn Te:::

Per the surveillance capsule testing program with the Commonwealth Edison ,

Company, 1/2 T-compact tension fracture mechanics specimens will not be tested and will be storri at the Westinghouse Science and Technology Center Hot Cell. .

4 e

9 5-8

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE BRAIDWOOD VNIT 2 FORGING 500102-1/50097-1 1RRADIATED AT 550'f,

, Ft.UENCE 3.91_x 1018n/cm2 (E > 1.0 MeV)

Sample No. 0$ $$ bf $ '

kf Tangential Orientation FL12 - 75 -5 o.O 7.5 4.0 0.10) 0 FL13 - 35 - 76.0 103.0 46.0 1.17 70 FL4 - 25 - 19.0 26.0 14.0 0.36} 15 FL9 0 -

33.0 44.5 23.0 0.58 25 FL5 15 - '47.0 63.5 35.0 0.89 45 FL11 20 - 74,0 100.5 52.0 1.32 70 FL10 40 89.0 120.5 58.0 1,47 75 FL6 75 138.0 187.0 80 0 2.03 90 FL7 100 140.0 190.0 83.0 2.11 90 '

FL3 125 136.0 184.5 84.0 2.13 95-FL2 150 175.0 237.5 93.0 2 :=6 100 FL15 200 172.0 233.0 86.0 2.18 100 FL1 250 1 184.0 249.0 81.0 2.06 100 .*

FL8 300 1 173.0 234.5 84.0 2.13 100 Axial Orientation FT3 - 75 - 8.0 11.0 4.0 0.10 5 FT14 - 50 -

10.0 13.5 6.0 0.15 10 FT6 --20 -

-40.0 54.0 24.0 0.6) 25 FT4 0 -

35.0 47.5 24.0 0.61 30 FT8 20 - 82.0 111.0 57.0 1.45 70-

, FT1 25 -

63.0 85.5 40.0 1.02 55-FT2 40 71.0 96.5 50.0 1.27 65 FT11 60 74.0- 100.5 55.0 1.40 70 FT15 80 100.0 135.5 68.0 1.73 80 3 FT13 105 116.0 157.5 78.0 1.98 90

. FT10 105 128.0- 174.0 82.0 2.08 -100 FT5 200 125.0 169.5 81.0 2.06 .100 FT9 200 139.0 188.5 83.0 2.11 130-

. '. FT12 FT7-250 300 1

1 149.0 145.0 202.0 19.6.5 86.0 82.0 2.18 2.08 100 100 l

5-9 L

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

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE BRAIDWOOD UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL-IRRADIATED AT ,

550'F, FLUENCE 3.91 x'1018 n/cm2 (E > 1.0 Hev)

Temperature Impact Energy Lateral Expansion Shear Sample No. ('F) ('0) (ft-lbl 2), (sils) (mm) A Weld Vetal FW11 -110 - 18.0 24.5 12.0 0.30 10 FW8 - 80 -

7.0 0.5 5.0 0.13 5

, FW7 - 70 .

18.0 24.5 16.0 0.41 15 FW5 - 45 -

25.0 34.0 19.0 0.48 30 FW1 - 20 -

31.0 42.0 23.0 0.58 25 FW3 0 -

33.0 44.5 28.0 0.71 30 FW12 20 -

44.0 59.5 35.0 0.89 45 FW14 35 43.0 58.5 36.0 0.91 45 FW9 60 46.0 62.5 38.0 0.97 50 FW6 90 59.0 80.0 54.0 1.37 100 .

FW4 120 63.0 85.5 60.0 1.52 '100 FW10 150 59.0 80.0 54.0 1.37 100 FW2 200 60.0 81.5 49.0 1.24 100 .-

FW1: 250 1 68,0 92.0 83.0 1.60 100 FW1; 300 1 63.0 85.5 61.0 1.55h 100 EAZ Wetal FH3 -170 -112 7.0 9.5 10.0 0.25 5 FH8 -150 -101 19.0 26.0 17.0 0.43 15 FH4 -125 - 87 82.0 111.0 47.0 1.19 75 FH6 -115 - 82 29.0 39.5 15.0 0.38 20

. FH15' 71 68.0 92.0 38,0 0.97) 60 FH7 59 76.0 103.0 41.0 1.04) 65 FH9 - 25 - 32 97.0 131.5 58.0 1.47 85 FH11 5 - 15 111.0 150.5 59.0 1.50 95 4 FH14 30 -

1 174.0 236.0 85.0 2.16 100 FH12 60 ) 155.0 210.0 80.0 2.03 95 -

FH10 100 56.0 76.0 43.0 1.09 60 FH13 150 204.0 276.5 82.0 2.08 100 FH5 200 191.0 259.0 83.0 2.11 100 -

FH1 225 162.0~ 219.5 55.0 1.40 100 FH2 250 243.0 329.0 69.0 1.75 100 e

5-10

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TABLE 5-5 EFFECT OF 550*F 1RRADIATION TO 3.91 x 1018 ,fe ,2 (E > 1.0 Mev)

OM NOTCH TOUGHNESS PROPERTIES OF BRAIDWOOD UNIT 2 REA" TOR YESSEL StRVEILLANCE MATERIALS Average 30 ft-lb Average 35 mil Average 50 ft-lb Average Energy Transition lateral Expansion Transitica Absorption at l Temperature (*F) Temperature (*F) Temperature (*F) Full Shear (ft-1b)

Material Unirradiated Irradiated h? Unirradiated Irradiated AT unirradtated trradiated AT Untrradiated Irradiated a(ft-ib)  !

Forging - 10 - 10 0 10 10 0 15 15 0 158 176 +B 500102-1/50C97-1 (Tangential)

.I Forging - 25 - 20 5 -5 5 10 0 to 10 153 137 -16  ;

50D102-1/50C97-1  !

(Axial) l Weld Metal - 20 - 20 0 5 15 10 40 45 5 71 62 -9 e

HAZ Metal -135 -135 0 -80 -75 5 -105 -105 0 155 200 +45 l I

• These values reflect scatter in the data and not real increases. Thus, the values will be reported as 168 f t-lb for the plate (tangential orientation) and 155 f t-1b for the HAZ metal.

5-13

TABLE 5-6 COMPARISON OF BRAIDWOOD UNIT 2 SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATU:

• SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE I.99 REVISION 2 PREDICTIONS 30 ft-lb Transition Temo. Shift Upper Shelf Enerav Decrease R.G. 1.99 Rev. 2 Capsule U R.G. 1.99 Rev. 2 Capsule U Fluence (Predicted)(a) (Predicted)

Material 10 18 n/cm 2

(.F) (*F) (%) (%)

Forging 50D102-1/50097-1 3.91 25.2 0 15.5 0 (Tangential)-

15.5 10.5 1

Forging 50D102-1/50C97-1 3.91 25.2 5 i (Axial)

Weld Metal 3.91 33.3 0 15.5 12.7 l

i l HAL Metal 3.91 --

0 --

0 f a) Mean wt. % values of Cu and Ni from Reference 'I and Table 4-5 were used to calculate the chemistry I factors for the forging and weld metal.

i i

3  ;

l '

i  :

c

! 5-14 8

f A 4 e g

TABLE 5-7 TENSILE PROPERTIES FOR BRAIDWOOD UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIAL IRRADIATED AT 550*F TO 3.91 x 10 18 n/cm2 (E > 1.0 MeV) &

Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp. Strength Strength Load Stree. Strength Elonnation Blongation in Area.

Waterial. Number (*F) (kei) (kei) (kini (keil Jk=1) (%) (%) (%)

Forging FL1 75 70.3 91.7 2.70 203.4 55.0 11.4 25.2 69 500:02-1/ FL2 300 es.2 es.5 2.50 15e.s 50.9 9.0 21.9 em 50097-1 FL3 550 e0.8 87.6 2.70 220.0 55.0 9.9 22.2 e@

(Tangent.

Orient.)

Forging FT1 75 70.8- 94.7 2.88 284.7 58.7 11.1 24.2 69 50D102-1/ FT2 300 65.2 84.5 2.70 211.5 55.0 9.0 21.9 67 50C97-1 FT3 550 62.6 90.7 3.30 250.6 65.2 9.9 19.7 64 *

(Axial Orient.'s Weld FW1 75 74.9 99.e 3.10 131.4 63.2 9.9 21.0 81 Weld FW2 300 88.8 82.5 2.95 166.9 60.1 8.4

• 18.9 59 Weld FWB 550 87.7 as.e 3.15 172.5 64.2 7.5 17.3 55 5-15

__. = . - . ..

1 (SC)

- 100 - 50 0 50 100 150 50 250 1 I g 2 3 12 Q I .

E 80 - 2 I '

/

A2 kM -

/

5 40 -

/ -

9 g _ 2 _

2 , i i i 0 1 ,.

100 a _ Au i ,_ t 2.5 a

=

i i i

~ ~

i E

80 - -

2.0

~

60 -

2 ** -

1. 5 E

.\ -

L0 b d E -

R5

.3 ' ' ' ' ' '

0 0 200 i i 280 i i i i i i .

180 -

2-o -

240 p .a_ .a. _ .a s..

160 -

o o o

~

140 .,

.O O g 120 - -

160 8 o --- Unirradiated o ._,

_ 100 -

Irradiated at 550 F - 120 -

E c

80 -

18

" 2 60 -

to 3.91 x 10 n/cm - 80

^

40 -

( E ) I MeV) . .

~

o -

20 ,

0- T- ' ' ' ' i

- 0

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

Figure 5-1. Charpy V-Notch Impact Properties for Braidwood Unit.2 Reactor Vessel Shell Forging 50D102-1/50C97-1 (Tangential Orientation) 5-16

( C)

-100 - 50 0 50 100 150 200 250 i i i i i3 i3 i i 100 - -

- i e g 80 -

e/ 2 -

'g M -

• Eg 2 B' 2 [

0 l@ l ' ' I 100 2.5 g I ' I o . ' Mi- A2 I -

I g 80 -

\2'42 2.0 60 - - 1,5li ct. e d5 40 -

_,p' 10*F

1. 0 -

m - -

a5 s

4" 0 0

200 280

.. = i i i i i i i i 180 -

240 34

~?.

200 3 o /

• g 120 -

/ -

160

~ 100 -

• /

---Unirradiated o -

120 80 O 60 - Irradiated at 550 F -

80 10 F

1. 18 2 40 -

to 3.91 x 10 n/cm 20 -

f^o 5 F ( E > 1 MeV) .

40

- 200 -100 0 100 200. 300 40] 500 Temperature ( F) i Figure 5-2. Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Shell Forging 500102-1/50C97-1 (Axial Orientation) 5-17

(OC)

-100 - 50 0 50 100 150 200 250 ,

100

• 2 -

3  %.3 6g g 80 -

/ 2 .

fm 60 -

. ., l l

5 40 -

2 o .

i 20 - o T

~

0

..100- 2.5 3 2 li 80 -

o -

2.0 g 60 -

- # 9* *-d-- -

1. 5 g

1. 0 =

] 40 10 F

.5 20 - -

0.5 0 0

/

100

'~

90 -

120 80 -

o o

o/--j- [a

-. 100 70 -

m 6 of '

? 60 2 .. -

80

~

/

5 50 -

5F

@40 -

o 60  ;

g o o --- Unirrad!ated o ~

o w -30 -

40 Irradiated at 550*F .

20 -

' 18 2 l o o to 3.91 x 10 n/cm - 20 i 10 -

. ( E > 1 MeV) .

0

,l  ! ' ' ' ' '

0

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

Figure 5-3. Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Weld Metal 5-18

('C)' I

-100 - 50 0 50 100 150 200 950 I ' 3b ' ' ' I i 1 100

• f ~r3 3r--

E 80 -

, o/

of o

k c

60 -

2,P / .

Si 40 -

/

o Y?

100 -

2.5 2

5 80 -

• os 4-- t% *- T- d- -

2.0 E o8 'o

- 60 -

4 '

1.5 li

. / e

$40 5F go o

1. 0 -

r4 20 - -

0.5

' 2,

., 0 0 250 ,

320 225 -

g _

E

_ 175 -

. o 240 o .oo 1h -

o o -

M

- o

~ 125 - /e o

o j

160 _

@100

= -

120 ~

O . --- Unirradiated o -

75 - *o o *oo . Irradiated at 550 F - 80

. 50 -

o g to 3,91 x 10 18 n/cm2 -

O 40 25 ( E > I MeV) .

0 ' I i I i I 0

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

Figure 5-4. Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Weld Heat Affected Zone Metal 5-19

f l

FL13 FL4- FL9 FL5 FL12l W E

, , n;g ,. e ._. .

3' k , ?Y f, _', ' A't I .h'? l'~ ' A 'I l wemmmmum -

':FL11 FL10 =FL6: FL14- ~ FL7

< gFL3: FL2 - FL15 FL1 -7i .FL8 -

l Figure 5-5. Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel Shell Forging 500102-1/50C97-1 (Tangential Orientation) 5-20

i i

m FT3' FT14- -FTC FT4 FT8 g en :.. - M~ ' : tW ' (Wt kkh;\,  ?

;; ? .

~

lN

.g .

g_

FT1 FT2 FT11' FT15 FT13-

_C:y . 4 - .. -

&, gam;vg

- FT10 FT5 , -_

m FT9. FT12 : ._ FT7 4

I Figure 5-6. Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel Shell Forging 500102-1/50C97-1 (Axial Orientation) 5-21

- . . _ _ . - . . . _ . . . _ . -. . _.. _ _ _ _ _ _ . _ ___ _. _ ___ _ __ _ _ _ _ _ ~ . . _ . . _ . . . _ _ _ . _ _ . . .

m. n a ., ,n.

{$7[q #I $N/ J('l~ I I;y ," j'"! i M; ."~Giv L *

'FW11 FW8' FW7 FW5 Fil W:m ,

%Q Wh ,

.?? 'i d

h ..

't. . c

1L' , ' h' ..

$,*{.

FW3 FW12 FW14- FWB, FW6 ' .

0

I j ly .- 4 .. s,

vi

i .i  ;

. , q kt ,

.f- .- f:!.

,,gfp

{

.FW4 FW10f 'FW2; 'FW15 FW13 Figure 5-7, Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel Weld Metal 5-22

,P' < *""*

.%;%?M :s iY I

j mi -

y.'

, ]$Yji (g;,

3, cu.

1. b

f. f[

.: FH3 - TH8 FH4 Fil6 FH15-namen_ W^ M

.FH7 FH9 FH11 FH14: FH12~  ;

J FH10 FH13 FH5 ' ,

FH1' Fl!2 Figure 5-8. Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 j Reactor Vessel Weld Heat Affected Zone Metal 5-23 l

1

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

1 ,. .

~ T'op i

+

Bottom e

.t

~--.~.-+..._n . ;e;;-

r.v .

1^ 7 W -. . ,

. q. jig kJ

! M .#

l Side 1

".10,. , ;3 t o. , -

,,, . tw :n n;

+ ,.

_m N'k4 7!,tj j ,, . .' , g ;? { ,; : . b, Ndo ,. . , -m , ' -

m, fj'* hW,y . l '< ,

9 W ': .' '_u..

9- , ., - 1lr. :. . . ., ,

f.=arvwa'

' t,4:olt*=a+ ,

'*o,: se.g.q. .r4v n y, ,3 .

. ,., ny a kl% . ' s Side 2 l

l .

I t

l l

Figure 5-9. Fracture Appearance of Specimen FL13 6-24

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

l

\

%h6 '..'.,.t'--

. D.+ t 6ce [..'.'..,.

l

+[.y

, +-

g.:

y y, weppAa " * * +-- % v

. ett w $ > [ Ng.-- I,.., 4 )g Top am --u

.p . .

s Wm9 >

..p- . v r v-l gTg1; '

QQ:s:QiCI.i@

r .y 97.*

• l

,. .c

,e <*

->!:%.y. .. u .V ~ e s O

, ^

s ,' .*

N %.* k E :gf p, '= 4 'y e '+

Bott]m I

- ~; m g4 .4: p;, u--

v..

.m m- . n . ..

M.I!

4g<

s d[ t . . *:, E.

,i/Q  ;: .

w~ , ..y, .

w, , . .

wayeatyw e s +m ,

, e -;

gyt.5> ,

Side 1 p

4 i , a

(

~n:m a__ -

,+

.w -

~J  : vgind~ ,

, () ~

' %- *T4

-. Y .. s

&6. 3MNi.:

. 7-p

, q; 7.e,

,r. ag.eps vs I -

I l

Side 2 l

(

Figure 5-10. Fracture Appearance of Specimen FL4 5-25

l 1

i i

l .

A. e 9:

l

, Fracture of Charpy Specimen FH1 ,

l

~

n ' ,

, Fracture of Charpy Specimen FH2 Figure 5-11. Fracture Paths in Heat-Affected-Zone Charpy Specimens FH1 and FH2 5-26 I

( 'S C) 0 -

50 -100 150 200 250 300 120- , , , , , , .

800 110 -

100 - ~

= Ultimate Tensile Strength 22 90 -

- 2 ,

1o -

600 3

O! 80 -

a E 2  :

E 70 -

500 E 60 50

[ (- 400

0. 2 % Yield Strength 399 40 '

Code:

Open Points - Unirradiated 18 2 Closed Points - Irradiated at 550*F ( 3.91 x 10 n/cm )

80 , , , , , ,

i 70 -

B C -

%~

~

60 -

Reduction in Area -

o

} x50 -

g 40 -

Total Elongation EM c

a )----

~

7 +

20 - -

Uniform Eloncation'~

10 -

2 f i

'2 0 ' ' ' ' ' '

0 100 200- 300 400 500 600 Temperature ( F)

Figure 5-12. Tensile Properties for Braidwood Unit 2 Reactor Vessel Shell i Forging 50D102-1/50C97-1 (Tangential Orientation) )

i 1

5-27 l

1

( C) 0 50 100 150 200 250 300 120 , , , , , , ,

800 110 -

100 -

Ultimate Tensile Strength

~

E

% ~

2 . g 0l 80 -

i N. 5 E 2

~

^ 70 ~

~

500 E y

60 . +~ -

400

0. 2 % Yield Strength 50 -

300 40 , ,

Code:

Open Points - Unirradiated .

2 Closed Points - Irradiated at 550 F ( 3.91 x 1018n/cm )

80 , , , , , , , ,.

70 - o m _

• ; D' 60 -

Reduction in Area

@ 50. _

$ 40 -

Total Elongation l30 -

e a

a 20 -

e a _

a -

Uniformjlongation -

10 _

0_,

2'A , , .\2 m

~

100 200 300 400 500 @

Temperature ( F)

Figure 5-13. Tensile Properties for Braidwood Unit 2 Reactor Vessel Shell Forging 50D102-1/50C97-1 (Axial Orientation) l 5-28 l

l

^

(

• C) 0 50 _100 150 200 250 300

. IN , , , , , , ,

800

, 110 -

~ '

Ultimate Tensile Strength 2 . 1a

1. 2 m 80 -

" a

= 6 5 70 .-

500 s 2 i v

3

0. 2 % Yield Strength 50 -

300 40 Code:

. Open Points - Unirradiated 2

Closed Points - Irradiated at 550 F (3.91 x 1018n/cm) 80 70 --

Reduction in Area -

, 2 m - e 5 m -

m E%

$'40 - -

is g 30 -

Total Elongation

~

2 20 -- Y -

.- e

- UnHorm Elongation __

10 _

27-

, , , , , , R2 0

0 100- 200 300 400 500 600 Temperature ( F)

Figure 5-14. Tensile Properties for Braidwood Unit 2 Reactor Vessel Weld Metal 5-29 l

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

i I

y  ;.-

e

[f.

7 a l

-,.2- .

g .,' ,

[$

t, Specimen FL1 75'F c . p-

^ ~ _ , %2 s y -$hbW i

J. _. ... ,

,t e. ,.

k;c-.

ga 1

(

Specimen FL2 300*F

.';flg W 1n ', phi +1; ..,-

. 3 . 7 ,

-

- . NM ;;[* 1

' U.7 y!t I E , I

.nw t . , w. m !y *

. ;.vqfJpt .,, /* +-7 -

m;c vr ,

,'; % e .o ,

, - a. .

W'L ? ~

! Specimen FL3 550'P Figure 5-15. Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel l

Shell Forging 50D102-1/50C97-1 (Tangential Orientation) 5-30 i

I

\__-_----_.--._--

l L

l l

l t

l 4

l

\

. -W:..' p ' '~. , : )

9> . c4. t -

i , %g._-

. ~-

l I

\

i h&W l I

I s i%n . Tw-. ,

i

i

,..;;;g .it(Pr >

-j =

l Specimen FT1 75'F l

,  : ,9 , ,i .

L l

3;. mw,v m.m . w; j' .

s. .,

a 3.

Specimec FT2 300*F (4

4 44 t  !

j , , ,MM e

k.

v (6'  ;

(~ . g .- ,y g: , 'im:

\

u

~ . ,:n Specimen FT3 550'F Figure 5-16. Fractured Tensile Specimens from Braidwood Onit 2 Reactor Vessel i

i Shell Forging 500102-1/50C97-1 (Axial Orientation) j l

l 1

1 i

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

I l 1

l

\

l l l 1 -

l l-l l -

P~ i- T/ .

n.-- -s .-

C" -

..I" (a i

- i, g s

.( -

l - -

Specimen FW1 75'F M

) ~ 'f f j 1 .>

. ; a sw .. ~

w

, 5-g 1 r"- ,

t < i Specimea FW2 300*F l \ . ;-  ;~

g. x.

s; O! l 1

m.

4

.-y l}. s . j, y

, =

c 2 -

k ( P-is 3

f, . .

Specimen FW3 550'F l

Figure 5-17. Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel Weld Metal 5-32 l.

100 -,

90- --

~~

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x 604

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, 0 0.05 0.1 0.15 0.2 0.2 S 0.3 STRAIN, N/N 100 - --

90-80- /' .

\

70-W x 60- s p 50-

\ '

W p 40--

M -

39_

20-SPEC FL2 10-300 F 0 i . < .

0 0.05 0.1 0.15 0.2 0.25 STRAIN, M/M Figure 5-18. Stress-Strain Curves for Tension Specimens FL1 and FL2 5-33

l 9Q- '

80- /'

['p 70-x 60 V ,

. I p 50- '

c d 40 .- y 30-20- SPEC FL3 10' 550 F  !

0-

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~

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609

-[; 50-h 40-30- -

2.0- SPEC FT1 10-75 F

0. -

--e ---

0 'O.05 0.1 0.15 0.2 0.25 577tAIN, lH/N Figure 5-19. Stress-Strain ci.rves for Tension Specimens FL3 and FTl 5-34

_ . _ . . , , m.e, . .

P 100- -

90-

80- -

N.

70-M x

60 I ,

y 50 '

h 40-  !

30-20-SPEC FT2 10-300 F 0

0 0.05 0.' 1 0.i 5 0.'2 0.25 STRAIN, IN/IN 100 '

90- - - -

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70-x 60-

\

d 50 d

$ 40-

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30-20-10- SPEC FT3

' 0 550 F '

O 0.b2 0 b4 0.b6 0.b 8 0.'1 0.I 2 0. I 4 0. I 6 0. I 8 0.2 STRAIN, lN/lN Figure 5-20. Stress-Strain Curves fo Tensioi. Specimens FT2 and FT3 5-35  ;

100 , ,

90- -

80- ,

70- .

M x 60-N w

50-

$ 40-

^

30- t 20- SPEC FW1 10- 75 F 0 , - , ,

0 0.05 0.1 0.15 0.2 0.25 STRAIN, IN/lN 100 _

90- ,

80-70- 1 M

x 60-y 50-N 40-Ui 30- ~

20- -

SPEC FW2  :

10- ,

300 F l 0 0.b2 0.b4 0.b6 0.b8 0.' 1 0.I 2 0. I 4 0.k 6 0. k 8 0.2 STRAh4, IN/lN figure 5-21. Stress-Strain Curves for Tension Specimens FW1 and FW2 5-36

100 . _ _

90-804 '

70- {

G x 60-p 50-

'h 40-s M 30-  ;

.. 20-10- SPEC FW3 0 550 F 0 0. 2 0.64 0 b6 0 bB 0.' 1 0.i 2 0.i 4 0.i B 0.18 STRAIN. IN/iN ,

4 Figure 5-22. Stress-Strain Curve for Tension specimen FW3 5-37 M

=

d l

o 4-1 4

i .

4 I

l 4

1-4 1

1 1

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l &

t i

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.;,-.-, . . , , . + , - . . . . , , - . . -----.---,n.--.,..,,--.~ . .- - , , - - - . , - . --.--~~.--. - ~ . - - - - - - - . - - , , - - - ----- - - . - . - - _ - - - - -

.- -. . . - . - - - _ _ _ _ _ - . - - _ _ - = _ _ - . - - -

SECTION 6.0 RADIATION ANALYSIS AND NCUTRON DOSlHETRY

, , 6.1 Introducti2D

. 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 rc'ationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specinens. The former requirement is normally met by employing a combination of 'igorous 5 analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured ,

materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more l 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 6-1

(dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. T.% energy dependant dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritit 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 vessel wall has already been promulgated in Revision .

2 to the Regulatory Guide 1.99, " Radiation Embrittlement of Reactor vessel Materials."

This section provides the results of the neutron dosimetry evaluations performed in conjunction with tb analysis of test specimens contained in surveillance capsule V. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 MeV), and iron atom displacements (dpa) are established for the capsule irradiation history.

1he analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure -

parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided. -'

6.2 Discrete Ordinates Aqalyi h A plan view of the reactor geometry at the core midplane is shown in figure >

4-1. Six irradiction capsule: attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillar.se program. Tre capsules are located at azimuthal angles of 58.5, 61.0, 121.5, 238.5, 241.0, and 301.5 relative to the core cardinal axes as shown in Figure 4-1.

A plan view: of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1. The stainless steel specimen containers are 1.182 by , ,

1-inch 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.

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 neutron pad 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 distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameten {p(E > 1.0 MeV), ((E > 0.1 MeV), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/p(E >

> 1.0 MeV), within the pressure vessel geometry. The relative radial gradient information was required to permit tne projection of measured exposure

~ parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations.

The second set of calculations consisted of a series of adjoint analyscs relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute nredictions of neutron exposure at the locations of interest for the cycle 1 irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects of varying neutron yield 6-3

1 per fission and fission spectrum introduced uy the build-up of plutonium as ihe burnup of individual fuel assemblies increased.

~

The absolute cycle specific data from the adjoint evaluations together with relative neutron erurgy spectra and radial distribution information from the forward calculation provided the means to: ,

1. Evaluate neutron dosimetry obt61ned from surveillance capsule locations.

2. Extrapolate dostmetry results to by locations at the inner radius and through the thickness of the pressure vessel wall.

3. Enable a direct comparison of analytical prediction with measurement.

4. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves. <

The forward transport calculation for the reactor model summarized in figures 4-1 and 6-1 was carried out in R, O geometry using the DOT two-dimensional discrete ordinates codell23 and the SAILOR cross-section ibrhryll33. The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an 58 order of angular quadrature.

The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse loop ,

plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core ,

periphery. Furthermore, for tiw peripheral fuel assemblies, a 20 ,

uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2a 6-4

,u--y<,,. -

--*ce- ~-,,c-w. in-w,,,-, ,r-, - ww n - , ce, ,, rv w es-m we en, y w-mw , evmay,,r w ey ,w+ C-y e=, -

gw r,- = wye e ey

level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.

, A'el adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library.

. Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well ai, the geometric center of each surveillance capsule. Again, these calculations were run in R, O geometry to provide neutron source distribution importance functions for the exposure paramet,er of interest; in this case, d (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:

R (r, 0) = fr [0 [E !(r, 0, E) S (r, 9, E) r dr de dE where: R (r 0) = p (E > 1.0 MeV) at radius r and azimuthal angle 0 I (r, O. E) Adjoint importance function at radius, r, azimuthal s angle 0, and neutron source energy E.

S (r, 0. E) = Neutron source strength at cota location r, O and energy E.

Although the adjoint importance functions used in the Braidwood Unit 2 analysis were based on a response function defined by the threshold neutron flux (E >

1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratiu of

[ dpa/p (E > 1.0 MeV) is insensitive to changing core source distributions.

In the application of these adjoint important functions to the Braidwood Unit 2 reactor, therefore, the iron displacement rates (dpa) and the neutron flux (E >

0.1 MeV) were computed on a cycle specific basis by using dpa/p (E > 1.0 MeV) and 4 (E > 0.1 MeV)/p (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific p (E > 1.0 MeV) solutions from the individual adjoint evaluations.

l 6-5

l The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design report for the first operating cycle of Braidwood Unit 2[I43. The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of the pressure vessel and surveillance capsules are summarized in figure 6-2. For comparison purposes, the core power distribution (design basis) used in the .

reference forward calculation is also illustrated in figure 6-2.

Selected results from the neutron transport analyses performed for the Braidwood Unit 2 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 tr 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), 6(E

> 0.1 MeV), and dpa) are given at the geometric renter 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 6.bsolute comparison of v.casurement with analysis. The design basis data derived from the forward calculation are provided as a poiht of reference against which piant 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 distribution, it is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel well itself. ,

Radial gradier.t information for neutron flux (E > 1.0 MeV), neutron flux (E > '

O.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 calctlation, are presented on a relative basis for each exposure parameter at severci azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.

6-6

for example, the neutron flux ([ > 1.0 MeV) at the 1/41 position on the 45' azimuth is given by:

fl/4T(45') = d(220.27, 45') f (225.75, 45')

where:

41/4T(45') = Projected neutron flux at the 1/41 position on

. the 45' azimuth d (220.27,45') = Projected or calculated neutron flux at the vessel inner radius on the 45" azimuth, f (225.75, 45') = Relative radial distribeion function from Table 6-3.

Similar expressions apply for exposure parameters in terms of 6 (E > 0.1 MeV) and dpa/sec.

. The DOT calculations were carried out for a typical octant of the reactor.

However, for the neutron pad arrangement in Braidwood Unit 2, the pad extent for all octants is-not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron pad extending from 32.5 - 45.0 degrees which produces the maximum vessel flux. Other octants have neutron pads spanning larger azimuthal sectors which provide more shielding. For the octant with the 12.5 degree pad, the maximum flux to the vessel occurs near 25 degrees and the values in the tables for the 25 degree angle are vessel maximum values.

Exposure values for 0,15, and 45 can Se used for all octants; values in the tables for 25 and 35 degrees are maxiinum values and only apply to octants with a 12.5 degree neutron pad, 6.3 Neutron Dosimetry

[ The passive neutron sensors included in the Braldwood Unit 2 surveillance program are listed in Taolo 6-6. Also given in Table 6-6 are the primary nuclear reactions.and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest (p (E > 1.0 MeV), p (E > 0.1 HeV), dpa).

6-7 i

1 The relative locations of the neutron sensors within the capsules are shown in

Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors were ,

l accommodated within the dosimeter block located near the center of the capsule.

The ust of passin monitors such as those listed in Table 6-6 uoes not yield a direct measure of the energy dependent flux level at the point of interest.

Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors rnay be derived from the activation treasurements only if the irradiation parameters are well known.

l In particular, the following variables are of interest:

o The specific activity of each monitor, o The operating history of the reactor. .

o The energy respcnse of the monitor.

l 0 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 determ8 ned using l established ASTM procedures D 5 through 28), following sarrple preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the Braidwood Unit 2 reactor during cycle I was cotained from NUREG-0020,

" Licensed Operating Reactors Status Summary Report" for the applicable period.

The Irradiation history applicable to capsule V is given in Table 6-7.

Measured and saturated reaction product specific activities as well as measured full power reection rates are listed in Table 6-8. Reaction rate values were ,

derived using the pertinent data from iables 5-6 and 6-7.

6-8 l.

Values of key fast neutron exposure parameters were derived from the measured l

I reaction rates using the FERRET least squares adjustment code (29), The

~

FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit

. (in a least squares sense) to the reaction rate data. The exposure parameters 4

along with associated uncertainties where then obtained from the adjusted spectra.

In the FERRET evaluations, a log normal leastesquares algorithm weights both the a priori values and the measured dath in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly )

related to the flux & by some response matrix A:

(s,a) (s) (c) f I A &

9 19 9 s

where i indexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be simultaneously adjusted. Fct example, R =I o (

i g ig g relates a set of measured reaction rates Rj to a single spectrum pg by the multigroup cross section ogg. In this case, FERRET also adjusts the cross-sections. The lognormal approach automatically accrunts for the physical constraint of positive fluxes, even with the large assignea uncertainties.

9 4

6-9

,_ _ , . . . . . . , _ - , ..,-..c,-,_, _ . , _ , , , . , - , . _ , , , , , , . . . - _ . . _,.._.,._..y.m. . , . ~ . . - - . . , - _ . - . . , . . . , -

.,-...m--

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 (30). This procedure was carried out by ,

first expanding the a priori spectrum into the SAND-!! 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 group:) 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.

i 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, e a simple parameterized form is used; i

2 Hgg, = R7+R R,P gg, g g where Rg specifies an overall fractional normalization uncertainty (i.e.,

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

~

2 Pgg, = (1 - 0) 6gg, + 0 exp [ 'L) *

, 6-10

_ - ~ _ _ . . _ _ . _ _ . , . _ _ _ _ _ ~ _ . _ _ _ _ . _ . _ _ _ _ . . - _ _ _ _ _ _ _ ___ . _ _ . _

The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 4 (0 specifies the strengthofthelatterterm).

1 :

> For the a priori calculated fluxes, a short-range correlation of a = 6

. groups was used. This choice implies that neighboring groups are strongly correlated when r is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E.

Maerker[31). Maerker's results are closely duplicated when a = 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 V dosimetry are given in Table 6-9. The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 3.91 x 10 18 n/cm2 (E > 1.0 MeV) with an associated 4 uncertainty of 18%. Also reported are capsule exposures in terms of fluence i (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, enellent results i

' were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tebulated in Tabic 6-11 for the FERRET 53 energy group structure.

A summary of the measured and calculated neutron exposure of capsule U is presented in Table 6-12. The agreement between calculation and measurement falls within i 12% for all fast neutron exposure parameters listed. The thermal neutron exposure calculated for cycle 1 underpredicted the measured value by 59 percent.

Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (1,15 EFPY) exposure derived from the capsule U measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EfPY). The calculated design basis exposure rates given in Tabic 6-2 were used to perform projections beyond the end of cycle 1.

6-11

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

l In the calculation of exposure gradients for the Braidwood Unit 2 reactor i coolant system, exposure projections to 16 EFPY and 32 EfPY were employed.

Dr.ta 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 RTHDT .

4 vs. fluence trend curves, dpa equivalent fast neutron fluence levels for the '

l 1/4T and 3/4T positions were defined by the relations e' (1/4T) =

4(Surface)([p[a { ))

a-d' (3/4T) =

f (Surface) { y ))

! Using this approach results in the dpa equivalent fluence values listed in

lable 6-14. ,

in Table 6-15 updated lead factors are listed for each of the Braidwood Unit 2 surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules.

2

+

t 6

6-12

. - . , , - , , . . .- m__ - ---m ._ . --- ,,--, ,.~.-re_.-,i.n.m-. --,v.,, , ,m.._-,,-..,.___.w.y.. _ _ _ , , . . _ . _ , .

..m. . , _ , - - _ - - . __ . - - , _ . , ,-uv.

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

e (TYPICAL)

Co

- 58.58 - 41.08

..x s

- 81.825 IN.

y '

m'%'Sth:h hxAh Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13

4 7

1 4 .

t l

1 0.74 0.70 0.76 0.59 Cycle 1 1.01 1.04 0.96 0.77 Design Basis 0.99 1.02 0.97 0.95 0.84 0.57 1.02 1.10 1.00 1.05 1.10 0.71 .

1.13 1.09 1.07 1.05 0.98 1.01

._ 1.05 0.87 0.87 1.07 1.00 1.05 e 1.14 - 1.13 1.13 1.14 1.08 1.09 1.06 0.88 1.10 1.04 1.18 1.14 1.14 1.20 0.90 1.04 1.12 0.92 Figure 6-2 -Core Power Distributions Used in Transport Calculations

- for Braidwood Unit 2 1

6-14 ye r - + wi s- ,v,4s e- =<-r - v-----vw~ -e- e er e- #e*-iew-- de w *'ww w s ewr e"'r = ra + w -ev--ww----e -'Wwvm*-+--***w---evne-e=----------- -

l TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER DESIGN BASIS CYCLE 1 21.0* ALi' 21d'. II.di'

( (E > 1.0 MeV) 1.13 x 1011 1.21 x 1011 8.84 x 1010 9.51 x 1010 2

(n/cm.sce)

& (E > 0.1 HeV) 5.07 x 1013 5.44 x 1011 3.97 x 1011 4.28 x 1011 2

(n/cm-sec)

dpa/sec 2.21 x 10-10 2.37 x 10-10 1.73 x 10-10 1.86 x 10-10

+

9 i 9 4

6-lE

TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL !.NTERFACE ,

DESIGN BASIS O' 15' 25' 35' 45' 10

((E > 1.064V) 1.78 x 10 10 2.66 x 10 10 3.01 x 10 10 2.45xIMO 2.81 x 10 2

(n/cm -sec) d(E > 0.lMeV) 3.70 x 10 10 5.60 x 10 10 8.22 x 10 10 6.96 x 10 10 7.04 x 10 10 2

(n/cm -sec) ,

dpa/sec 2.77 x 10'II 4.12 x 10'II 5.04 x 10'II 4.15 x 10'Il 4.48 x 10'II CYCLE 1 SPECIFIC 0' 15' 25' 35' of

((E > 1.0MeV) 1.32 x 10 10 2,06 x 10 10 2.38xly0 1.98 x 10 10 2,31 x 10 10 2

(n/cm -sec) 10 10 10 10

((E > 0.lMeV) 2.74 x 10 4.34 x 10 6.50 x 10 5.62 x 10 5.79 x 10 10 ,

2 '

(n/cm -sec) dpa/sec 2.05 x 10'II 3.19 x 10'II 3.99 x 10'II 3.35 x 10'II 3.68 x 10'II 6-16

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

1 TABLE 6-3

RELATIVE RADIAL DISTRIBUTIONS Of NEUTRON FLUX (E > 1.0 MeV)

WITHIN THE PRES $URE VESSEL WALL Radius

.(em)_ O' 15' 25' ,

35' 45' 220.27(l) 1.00 1.00 1.00 1.00 1.00 220.C4 0.976 0.979 0.980 0.977 0.979 221,66 0.888 0.891 0.893 0.891 0.889 222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.5b0 0.554 0.552 0.543 226.95 0.462 0.460 0.465 0.463 0.452

.228.28 0.386 0.384 0.388 0.386 0.375 229.60 0.321 0.319 0.324 0.321 0.311

, 230.92 0.267 0.265 0.271 0.267 0.257 232.25 0.221 0.219 0.223 0.221 0.211

, D3.57 0.183 0.181 0.185 0.183 0.174 234.89 0.151 0.149 0,153 0.151 0.142 236.22 0.124 0.122 0.126 0.124 0.116 237.54 0.102 0.100 0.104 0.102 0.0945 238.86 0.0823 0.0817 0.0846 0.0835- 0.0762 240.19 0.0671 0.0660 0.0689 0.0679 0.0608 l 241.51 0.0538 0.0522 0.0550 0.0545 0.0471 242.17(2) 0.0506 0.0488 0.0518 0.0521 0 0438 NOTES: 1) Base Metal Innse Radius

2) Base Metal Outer Radius 6-17

,_ -, . , . , . . _ , . , , - ---.,s. , , - - . . .-- - , ,....,,.~ _ ._,.,,,r. , _ . . . - , , , - ~ . - - , . - - . . . .

l 4

1 TABLE 6-4 i

RELATIVE RADIAL DISTRlDUTIONS OF NEUTRON 7 LUX (E > 0.1 MeV)

WITHIN THE PRESSURE VESSEL WALL ,

1 Radius y, 0' ,_ 15' 25' 35' 45' 220.27(I) 1.00 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 1.00 221.66 1.00 1.00 1.00 0.999 0.995 222.99 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901 225.63 0.874 0.865 0.874 . 850 0.842 226.95 0.818 0.808 0.818 0.792 0.782 228.28 0.761 0.750 0.716 0.734 0.721 e l 229.60 0.705 0.693 0.704 0.677 0.662 230.92 0.649 0.637 0.649 0.621 0.605 '

232.25 0.594 0.582 0.594 0.567 0.549 233.57 0.540 0.529 0.542 0.515 0.495 234.89 0.487 0.478 0.490 0.46T 0.443 23b.22 0.436 0.428 0.440 0.416 0.392 237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 0.279 0.248

-241.51 '

0.244 0.238 0.249 0.233 0.201 242.17(2) O.233 0.226 0.237 0.223 0.188' ,

e NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-18

l TABLE 6-5 l* RELATIVE RADIAL DISTRIBUTIONS Of IRON DISF < CEMENT RATE (dpa)

WITHIN THE PRESSURE VESSEL WALL Radius

,. (cm) 0* 15' 25' 35' 45' 220.27(l) 1.00 1.00 1.00 1.00 1.00 220.64 0.984 0.981 0.984 0.983 0.984 221.66 0.912 0.909 0.917 0.921 0.915 222.99 0.815 0.812 0.826 0.833 0.821 224.31 0.722 0.719 0.737 0.747 0.730 225.63 C.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572 4 228.28 0.497 0.493 0.519 0.533 0.506 229.60 0.439 0.435 0.462 0.475 0.447 230.92 0.387 0.383 0.410 0.423 0.394 232.25 0.341 0.338 0.364 0.376 0.347 233.57 0.300 0.297 0.322 0.334 0.305 234.89 0.263 0.261 0.285 0.295 0.766 236.22 0.230 0.228 0.250 0.260 0.231 237.54 0.199 0.198 0.218 0.227 0,199 238.86 0.171 0.170 0.189 0.196 0.169 240.19 0.145 0.144 0.161 0.167 0.140 241.51 0.121 0.119 0.135 0.139 0.113 242.17(2) 0.116 0.113 0.128 0.134 0.106 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-19

f!!  !  ! i!l((I!'r1l5![1 t . isf!  !!; 1I l!, ,ii! .

n .

o d 9 0 i l ) 9 5 s e % (

s i 5 6 .

i Y F

s s e s y y s s s s r

f r a a r r r t i y d d y y y y c L u - 2 2 0 2 2 2 2 d f 7 9 1 1 7 7 o

r a l 2 2 2 2 1 0 0 0 P  !

! 5 3 7 3 3 5 5

. S R

. O T V I e N M V O V e M V V V V e 5 M e e e e M 1 X e M M M M 0 5 U 's 8 1 L n e 7 0 0 4 0 0 0 F o g p n 4 1 1 0 0 E

> 0 N s a > .

O R

e R > > > > > v >

R e 6 T E E E E E 4 E

- U 6 E 0 ,

N E

L R B O A F n 1 o 7 2 0 5 5 S t t i 1 8 3 1 1 R e h t 9 5 8 0 0 E g g c 6 0 6 0 0 0 0 T r i a E a e r 0 0 0 1 1 0 0 M T W F A .

R d A e P d l

R e A i E h L 7 s C 7 3 U 0 4 8 3 I 0 0 m N 6 5 S I s 6 6 u n

o t

s o n o s C o o i C M C C ) C C m i e ) ) ) ) ) ) d r p p t f c

a o et a.

n n n

, f.n f.

(

n 0, n

3 n

a c .

e n ( ( ( ( 7 ( ( s R I 3 4 8 8 3 9 9 i .

. 6 5 S 3 2 S S u e i 2 p o o r C F N U N C C o .

t i

n .

o m m m

.

• u u 7 n n t

• 3 i i a 8 2 m m h 3 - u u t 2 m l l r a l - u A A s o i r l m

u i

n t t

t e

t r e e i u l l o i e p n k n t a a n n t p o c a p b b e o a o r i r e o o D MM C I N U N C C *

! ! ,f  !  ;: ' ;1

TABLE 6-7 1RRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U Irradiation Pj Pj trradiation Decay Period (HWt ) Pref. Time (days) Time (days)-

5/88 691 .203 7 869 6/88 741 .217 30 839 7/88 1572 .461 31 808 8/88 2576 .755 31 777 9/88 1368 .401 30 747 10/88 670 .196 31 716 11/88 2248 .659 30 686 12/88 2636 .773 31 655 1/89 3016 .884 31 624 2/89 1078 .316 28 596 3/89 408 .170 31 565 s 4/89 3164 .9i8 30 535 5/89 2252 .660 31 504 6/89 2381 .698 30 474 7/89 2496 .732 31 443 8/89 2821 .827 31 412 9/89 2651 .777 30 382 10/89 2960 .868 31 351 11/89 3172 .930 30 321 12/89 3323 .974 31 290 1/90 2636 .773 31 259 2/90- 1862 .546 28 231 3/90 1505 .441 15 216 NOTE: Reference Power = 3411 MWt 6-21 -

.,--+,el

, . - -r. .,,,n -

-c-eve.-- - ,-v r , rw--, m m w r. ~ - - , - - - . ~ r, .-- -- - ,,,-,,w~ .--u-,-- e.,,..- .,,-, ,,r-s--

_ . . . _ . _ _ . _ _ _ . . _ _ _ _ _ . . _ _ , _ . _ . . _ _ . ~ . _ _ . _ . _ _ _ _ _ _ _ _ ._.

r i

I j TABLE 6-8 i MEASURED SENSOR AC".VITIES AND REACTION RATES i

l- Measured Saturated Reaction 1

Monitor and Activity Activity Rate j Axial location Idis/see-am) (dis /see-am) ,(,RPS/ NUCLEUS 1 .

I  !

Cu 63 (nia) C0-60 i

Top 5.41 x 10 4 4.30 x 105 1 Middle 4.78 x 10 4 3.80 x 105

, Bottom 4.81 x 10 4 3.82 x 105 l Averige 5.00 x 10 4 3.97 x 105 6.06 x 10-17 Fe-54(n.p) Mn-54 Top 1.31 x 106 4.03 x 106 ,,

i Middle 1.16 x 106 3.57 x 106 '

Bottom 1,14 x 10 6 3.51 x 106 Average 1.20 x 106 3.70 x 106- 5.90 x 10-15 Ni-58 (n.p) 00-58  ;

Top 4.99 x 100 5.77 x 107 Middle 4.46 x 10 6 5.15 x 10 7 ~

Bottom 4.44 x'106 5.13 x 107 Average 4.63 x 105 5.35 x 107 7.63 x 10~I6 ,

- U-238 (n, f)' Cs-137 (Cd) r Middle 1.34 x 105 5.25 x 106 3.46 x 10"14

6-22 l

l.

. _ . . . _ _ . . . . - _ , . _ , _ . . ._. .._. ,. _ - . , _ . .-.~...-._,__..m. .._c__ _.

_ . . _ _ . , . . _. r..

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

TABLE 6-8 MEASURED SENSOR I,CTIVITIES AND REACTION RATES - con','d Measured $atu aated Reaction Monitor and Activity Aci ,vity Rate

. Axial L_qat.ign .uii s / sec-am) 11L /see-ard (RPS/NUC1EQSJ Np-237(n,f) Cs-137 (Cd) i Midale 1,39 x 10 6 5.44 x 10 7 3.30 x 30-13 ,

Co-59 (n,6) C0-60.

Top 1.08 x 10 7 8.58 x 10 7

. Middle 1.07 x 10 7 8.50 x 10 7 Bottom 1.06 x 10 7 8.42 x 10 7 Average 1.07 x 10 7 8.50 x 10 7 5.54 x 10 12 C0-59 (n,0) 00-60 (Cd)

Top 5.42 x 106 4.30 x 30 7 Middle 5.57 x 106 4.42 x 10 7 Bottom 5.56 x 106 4,42 x 10 7 Average 5.52 x 10 6 5.81 x 10 7 2.86 x 10-12 4

9 6-23

,m,, ,, a--,, . , . + - -- , , - - , - - . . . , - y.e.. ,, - - - - , - , ,, ..,-,,m, -..-..,,4 -e

TABLE 6-9 SUMHARY OF NEUTRON DOSIMETRY RESULTS TlHE AVERAGED t1POSURE RhK $

2 4 (E > 1.0 MeV) {n/cm -sec) 1.08 x 10Il 1 8%

p (E > 0.1 MeV) {n/cm2-sec) 4.76 x 1011 i 15%

dpa/sec 2.08 x 10-10 11%

p (E < 0.414 eV) (n/cm2.3ec) },31 x 19 11 1 21%

INTEGDATED CAPSULE EXPOSURE ,

t (E > 1.0 MeV) {n/cm )

2 3,93 x jolB 1 8% a t (E > 0.1 HeV) {n/cm )

2 1,72 x 10 l? 1 15%

dpa 7.53 x 10-3 1 11%

t (E < 0.414 ev) {n/cm2 }

4.02 x 1018 i 21%

NOTE: Total Irradiation Time = 1.15 EFPY S

4 o

6-24 I

TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED

- . REACi!0N RATES AT THE SURVEILLANCE CAPSULE CENTER Adjusted Reaction Measured Calculation [jf Cu-63 (n,n) 00-60 6.06x10-17 5.93x10-17 0.98 Fe-54 (n.p) Ma-54 5.90x10-15 5.79x10-15 0.98 Ni-58 (n.p) Co-58 7.63x10-15 7.72x10-15 1.01 v 238 (n,f) C0-137 (Cd) 3.46x10-14 3.31x10-I4 0.96 Np-237 (n,f) Cs-137 (Cd) 3.30x10-13 3.38x10-13 1.02 Co-59 (n,0) Co-60 (Cd) 2.86x10-12 2.87x10-12 0.99 Co-59 (n,0) 00-60 5.E4x10'l2 5.50x10-12 1 00 i

s 6

P e

U 4

b 6-25

. a _ _ __

l TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM Ai THE SURVEILLANCE CAPSU'.C CENTER Erargy Adjusged Flux Energy AdjusgedFlux Group (HeV) (n/cm -sec) Group (MeV) (n/cm -sec)

• 1 1.73x101 P 15x10 6 28 9.12x10-3 2.18x1010 2- 1.49x10 I 1.90x107 29 5.53x10"3 2.83x1010 (

3 1.35x101 7.34x107 30 3.3r,10-3 8.83x109 ,

4 1.16x10 1 1.64x10 8 31 2.84x10-3 8.44x109 5 1.00x10I 3.60x10 8 32 2.40x10'3 8.13x109 6 8.31x100 6.14x10 3 33 2.04x10-3 2.29x1010 7 7.41x10 0 1.41x109 34 1.23x10-3 2.lix1010 8 6.07x10 6 2.02x109 35 7.49x10-4 1.95x1010 0 4.54x10'4 1.86x1010 9 4.?7x10 4.26x109 36 10 3.58x10 0 5.60x109 37 2.75x10-4 2.00x1010 11 2.87x10 0 1.19x1010 38 1,67x10-4 2.14x1010 12 2.23x10 0 1.65x10 10 39 1.0lx10-4 2.16x10 10 ,

13 1.74x10 0 2.32x1010 40 6.14x10-5 2.15x10 10 14 1.35x10 0 2.59x10 10 41 3.73x10-5 2.10x1010 15 1.lix10 0 4.75x1010 42 2.26x10-5 2.04x10 10 16 8.21x10~1 5.43xlC 10 43 1.37x10-5 1.98x1010 17 S.39x10-1 5.64x10 10 44 8.32x10-6 1.89x10 10 18 4.98x10'l 4.09x1010 45 5.04x10-6 1.74x1010 19 3.88x10-l 5.75x1010 46 3.06x10-6 1.63x1010

, 20 3.02x10-1 5.91x1010 47 1.86x10-6 1.50x1010 01- 1.83x10-l 5.86x1010 48 1.13x10-6 1.lix10 10 -

f. 1.lix10*l 4.69x1010 49 6.83x10'7 1.43x10 10

'3 6.74x10'2 3.26x1010 50 4.14x10-7 1.90x1010 '

24 4.09x10-2 1.85x1010 51 2.31x10-7 1.90x10 9 '

25 2.55x10-2 2.42x1010 52 1.52x10-7 1.81x10 9 k

26 1.99x10-2 1.19x1010 53 9.24x10-8 5.46x1010 27 1.50x10-2 1.51x1010 NOTE: Tabulated energy levels represent the upper energy of each group.

6-26

TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED

, EXPOSURE LEVELS FOR CAPSULE V ,

Calculated Meaten ed UB 2

f(E'> 1.0 MeV) (n/cm ) 3.44 x 1018 3.91 x 1018 0.88 2

f(E > 0.1 MeV) {n/cm ) 1.55 x 1019 1.72 x 1019 0.90 dpa 6.74 x 10-3 7.53 x 10-3 0.90 2

f(E < 0.414 eV) {n/cm ) 1.65 x 1018 4.02 x 1018 0.41 s.

4 9 I 6-27 I

._._ _ ~ _ . ._

TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD /8ASE METAL INTERFACE FOR BRAIDWOOD UNIT 2 A7IMUTHAL ANGLf, O' 15' 25'(*} 35' 45' 1.15 EFPY f(E>l.0 MeV) 5.43 x 10 8.48 x 10 9.80 x 10 8.15 x 10 9.51 x 10 2

(n/cm )

0 18 I I f(E>0.1 MeV) 1.10 x 10 1.74 x 10 2.61 x 10 2.26 x 10 2.33 x 10 '

2 (n/cm )

dpa 8.30 x 10 1.30 x 10' 1.62 x 10' l.35 x 10' l.49 x 10' 16.0 EFPY f(E>1.0 MeV) 8.88 x 10 1.33 x 10 ' l.51 x 10 ' l.23 x 10 ' 1,41 x 10 '

2 .-

(n/cm )

f(E>0.1 MeV) 1.84 x 10 ' 2.80 x 10 4.11 x 10 3.49 x 10 3.53 x 10 '

2 (n/cm )

dpa 1,38x16 2.06x16 2.52x16 2,08x16 2.25x16 32,0 EFPY f(E>l.0 MeV) 1.79 x 10 ' 2,68 x 10 3,03 x 10 2.47 x 10 2.83 x 10 2

(n/cm )

f(E>0.1 MeV) 3.71 x 10 ' 5.62 x 10 8.26 x 10 ' 7.00 x 10 7.08 x 10 ,

2 (n/cm )

l dpa 2.78x16 4.14 x 16 5.07 x 16 4.17x16 4.51 x 16 i

(a) Maximum point on the pressure vessel 6-28

TABLE 6-14 NEUTR0t1 EXPOSURE. VALUES FOR USE lti T:1E GENERATI0f! 0F llEATUP/C00LDOWN CURVES 16 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dea SLOPE (n/cd) (equivalent n/cm 2)

Surface IL4 T 3/4 T Surface 1/4 T 3/4 T 0* 8.88 x 10 18 4.82 x 10 18 1.03 x 10 18 8.88 x 10 18 5.60 x 10 18 1.95 x 10 18 15* 1.33 x 10 19 7.20 x 10 18 1.51 x 10 18 1.33 x 10 19 8.34 x 10 18 2.88 x 10 18 25*(a) 1.51 x 10 19 8.24 x 10 18 1.78 x 10 18 1.51 x 10 19 9.80 x 10 18 3.59 x 10 18 35* 1.23 x 1019 6.69 x 10 18 1.43 x 10 18 1.23 x 10 l9 8.15 x 10 18 3.05 x 10 18 45* 1.41 x 10 I9 7.54 x 10 18 1.52 x 10 18 1.41 x 10 l9 9.02 x 10 18 3.09 x 10 18 32 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dpa SLOPE 2 (equivalent n/cm2)

(n/cm )

Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 0* 1.79 x 10 l9 9.72 x 10 18 2.07 x 10 18 1.79 x 10I9 1.13 x 10 19 3.92 x 10 18 15* 2.68 x 10 19 1.45 x 10 I9 3.05 x 10 l8 2.68 x 10 19 1.69 x 10 19 5.81 x 10 18 ,

25*(a) 3.03 x 10 l9 1.66 x 10 19 3.57 x 10 18 3.03 x 10 19 1.97 x 10 19 7.21 x 10 18 35* 2.47 x 10 l9 1.35 x 10 19 2.86 x 10 18 2.47 x 10 I9 1.64 x 10 l9 6.12 x 10 18 45* 2.83 x 10 l9 1.52 x 10 l9 3.06 x 10 18 2 83 x 10 l9 1.81 x 10 l9 6.20 x 10 18 (a) Maximum point on the pressure vessel 6-29

.. -- . . . - . _= - -

TABLE 6-15 UPDATED LEAD FACTORS FOR BRAIDWOOD UNIT 2 SURVEILLANCE CAPSULES Cap.iglg lead Factor V 4.00(a):

X 4.02 W 4.02 Z 4.02

\ 3.75 Y 3 75 (a) Plant specific evaluation 1

l.

O e

6-30

SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future

. capsules to be removed from the Braidwood Unit I reactor vessel:

Capsule Estimated Location Lead Fluence 2

Capsule (deg.) Factc* Removal Time (b) (n/cm)

U 58.5 4.00 1.15 (Removed)(a) 3.91 x 1018 (Actual)

X 238.5 4.02 4.5 1.7 x 1019 (c)

V 61.0 3.75 9.0 3.2 x 10l9 (d) e Y 241.0 3.75 15.0 5.3 x 1019 W 121.5 4.02 Standby ---

2 301.5 4.02 Standby ---

(a) - Plant Specific Evaluation (b) Effective full Power Years (EFPY) from plant startup.

'c) Approximate fluence at 1/4 thickness of reactor vessel wall at end of life (32 EFPY).

-(d) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY).

~.

7-1 .

4 -.-. u a n - .--- - .,- ,s'. L - - ~..

4 4

e r

S 9

e 9

1 a r

i

SECTION

8.0 REFERENCES

1. Singer, L,R., " Commonwealth Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-11188, December 1956.

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

3. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988.

4. Section 111 of the .GME Boiler and Pressure Vessel Code, Appendix G,

" Protection Against Nonductile Failure,"

i 5. ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels."

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

7 ASTM E23-88, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials."

8. ASTM A370-89, "Standaro Test Method: and Definitions for Mechanical Testing of Steel Products."

9. ASTM E8-89b, " Standard Test Methods of Tension Testing of Metallic

~

Materials."

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

l 8-1

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

11. ASTM E83-85, " Standard Practice for Verification and Classification of Extensometers."

12. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Hodification, Updating and input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport ,

Technique", WANL-PR(LL)-034, Vol . 5, August 1970.

13. "0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors". -

14. J. V. Alexander, et. al., " Core Physics Parameters and Plant Operations Dat: for the Braidwood Generating Station Unit 2 Cycle 1", WCAP-ll475, June 1987. (Proprietary)

15. ASTM Designation E482-82, " Standard Guide for Application of Neutron i Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,

• 1984,

16. 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, 17 ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Termt of Dispiacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, ,

Philadelphia, PA,1984.

18. 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.

8-2

19. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillant.e Results", in ASTM Standards, Section 12, American Society for Testir,? and Materials, Philadelphia, PA,1984.

. 20. ASTM Designation FS61-77, " Standard Method for Determining Neutrcn Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984,

21. 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.

22. ASTM Designation E263-82, " Standard Method for Determining f ast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, 3 American Society for Testing ano Materials, Philadelphia, PA,1984.

23. 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.

24. 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.

25. 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,

26. 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.

1 1

8-3 l

\

1

27. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptuntum-237", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.

28. 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.

29. F. A. Schmittroth, FERRET Data Analysis Cort, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

30. W. N. McElroy, S. Berg and T. Crocket, A Computer- Automated Iterative Method of Neutron Flux Soectra Determined by Foil Activation, AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.

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

O 8-4

0 APPENDIX A 4

Load-Time Records for Charpy Specimen Tests 4

e G

S 1

l A-0 l l

1 l

i

! t .i[ ,* j'  ! ,

t!  : ;I! t i. [I!ti  ?!I .!I ?; t

=

D A D O A L O E L R T U S E

T C R A R R A

_- F =

p W = A g

P P

D A -

O L

- / I I i i I I I d

r.

o

= c.

M e U r M e

. I m X i t .

A -

M E d a

- M o g I T

l P d

= e

, g l Il l 1 iI i l I i I I I z

i l

= - a e

d I

1 1

A e

r u.

g -

i F

W x a

t m .

L A .

R E D .

N A E O .

G L

= D c yLE n l l jI 1 li I y

_ gl f

P Y I c =

r

--o4cs_

Y-

.. - - ~ _- . . - -

l l

1 l

l l

l 1

g.

3 M

C L.LJ l O l C l o

(

• I. ICN 1

'i t.

M N,=

O y

+

I La W d,c C l o O h l, t.k i m .m a T U a &

. g ,.-4 h

C O Cf4 0 be M

4 O O k *H U T eH k

< C'.) 8 O O = 0 l + G 1 LLJ G k*

' o S O v Q $

R

.m h o .

4 k $1

) j

.s. c.

>:e N

+ <

.' L6J g

I l

'9

i. y

r3 g O

'/ * *9 Ga.

'- . c,,

l O O C o c.

C C C o O

u>M CM LO to oo

• OM r'" O oC N. C, LO. +O + . +

O. .e C -

M LJ e LAJ o LJ N L.J f.0 LAJ sq l ' P*o'I A-2 1

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