ML20011E247

Analysis of Capsule U from Comm Ed Byron Unit 2 Reactor Vessel Radiation Surveillance Program.

ML20011E247
Person / Time
Site:	Byron
Issue date:	10/31/1989
From:	Lippincott E, Meyer T, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20011E244	List: ML20011E243 ML20011E247
References
WCAP-12431, NUDOCS 9002120303
Download: ML20011E247 (83)

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I e 1WCAP-12431:

W ANALYSIS OF CAPSULE U FROM THE' COMMONWEALTH EDISOH COMPANY.

BYRON UNIT 2 REACTOR VESSEL-RADIATION SURVEILLANCE PROGRAM 4

E. Terek E. P.-Lippincott L. Albertin October 1989 -

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• Work Performed Under Shop Order BFHP-106' Prepared by Westinghouse Electric Corporation:

for the Commonwealth Edison Company Approved by: b' T. A. Meyer, Mdnager k

Structural Materials and Reliability Technology n

I WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 1

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, PREFACE ,

I This report has'been technica11y' reviewed and verified. j i

l Reviewer Sections 1:through 5, 7, and 8 N. K. Ray .

W L Section 6 E. P. Lippincott- YhD( .!

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.Section Title. Page .l T*i > 1.0

SUMMARY

OF RESULTS. 11 I y

2.0: INTRODUCTION 2-1 f

3.0 BACKGROUND

~3-1 4.0 . DESCRIPTION OF PROGRAM- 4-1 5.0- TESTING OF SPECIMENS FROM CAPSULE U 5-l' 5.1 Overview- 5-1

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

. 5.3 Tension. Test Results 5-4 5.4- ' Compact Tension Tests

f. 5-5

[.; '6.0' = RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 , 3 6.3- Neutron Dosimetry 6-7 .

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7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1

8.0 REFERENCES

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LIST OF ILLUSTRATIONS Figure Title Page fl 4-1 Arrangement of surveillance capsules in the reactor. vessel 4-7 4-2 Capsule V diagram showing location of specimens, thermal 4-8 monitors and dosimeters 5-1 Charpy V-hotch impact properties for Byron Unit 2 reactor 5-13 vessel shell-forging MK24-3 (tangential orientation) 5-2 Charpy V-notch impact properties for Byron Unit 2 reactor 5-14

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vessel shell forging MK24-3 (axial orientation) j , 5-3. Charpy V-notch impact properties for Byron Unit 2 reactor 5-15 vessel weld metal m ,

5-4 Charpy V-notch impact properties for Byron Unit 2-reactor 5-16 j vessel weld heat affected zone metal 5-5 Charpy impact specimen fracture surfaces for Byron Unit 2 5 reactor vessel shell forging HK24-3 (tangentini orientation) 5-6 Charpy impact specimen fracture surfaces for Byron Unit 2 5-18 reactor vessel shell forging MK24-3 (axial orientation) j 5-7 Charpy impact specimen fracture surfaces for Byron Unit 2 5-19 1 reactor vessel weld metal i 5-B Charpy impact specimen fracture surfaces for Byron Unit 2 5-20 reactor vessel weld heat affected zone (HAZ) metal 5-9 Tensile properties for Byron Unit 2 reactor vessel shell 5-21 forging MK24-3 (tangential orientation) mo.,omeo in

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I  : Figure Title Page j 5-10 Tensile properties for Byron Unit 2 reacter vessel shell 5 7 forging MK24-3 (axial orientation) 5-11 Tensile properties for Byron Unit 2 reactor vessel weld 5-23 metal ,

5-12 Fractured tensile specimens from Byron Unit 2 reactor- 5-24 vessel shell forging MK24-3 (tangential orientation) r 5-13 Fractured tensile specimens from Byron Unit 2 reactor 5-25' vessel-she11' forging MK24-3 (axial orientation)

'5-14 Fractured tensile specimens from Byron Unit 2 reactor- 5-26' .~

vessel weld metal  : .;

5-15 Typical stress-strain curve for Commonwealth Edison 5-27 Company Byron Station Unit 2 shell forging MK24-3 tension specimens y

6-1 Plan view of a dual reactor vessel surveillance capsule 6-13 6-2 Core power distributions used in transport calculations 6-14' for Byron Unit 2 3990s/010490 10 jy

O LIST OF TABLES Table Title Page

• l

'7 '4-1 Chemical Composition and Heat Treatment of the Byron Unit 2 4-3  ;

Reactor vessel' Surveillance Materials 4-2 Chemical Composition of Byron Unit 2 Capsule U Irradiated' 4-4 Charpy Impact Specimens 4-3 Chemistry Results from the NBS Certified Reference Standards 4-5 4-4 Byron Unit 2 Reactor Vessel Toughness Data 4-6 1 Charpy V-Notch Impact Data for the Byron Unit 2 Shell 5-6 Forging MK24-3 Irradiated at 550'F, Fluence 3.96 x 1018n/cm 2

-a-(E > 1.0 MeV)

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' 5-2' .Charpy V-Notch Impact Data for the Byron Unit 2 Reactor 5-7 Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 3.96 x 10 18 n/cm2 (E > 1.0.MeV) j .

5-3 Instrumented Charpy Impact Test Results for Byron Unit 2 5-8 Shell Forging MK24-3 Irradiated at 550*F, Fluence ,

3.96 x 10 18 n/cm2 (E=> 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results. for Byron Unit 2 5-9 Weld Metal and HAZ Metal Irradiated at 550'F, Fluence t 3.96 x 10 18 n/cm2 (E > 1.0 MeV)

' 5-5 Effect of 550*F 1rradiation at 3.96 x 10 16 n/cm2 5-10 (E > 1.0 MeV) on Notch Toughness Properties of Byron Unit 2 Reactor Vessel Materials 5-6 Comparison of Byron Unit 2 30 ft-lb Transition Temperature 5-11 Shifts and Upper Shelf Energy Decreases with Regulatory mo.,om,o io Guide 1.99 Revision 2Predic$ ions

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-LIST OF_ TABLES-(Cont)  ;

Table Title Page ,

5-7.- . Tensile Properties for Byron Unit 2 Reactor Vessel Material 5-12 ,

Irradiated'at 550'F to 3,96 x 10 18 n/cm2 (E > 1.0 MeV) 6-l' Calculated-Fast Neutron Exposure Parameters at the - 6-15 Surveillance Capsule Center 6-2 Calculated Fast Neutron Exposure Parameters at the 6-56 E Pressure Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux 6-17 ,

(E > 1,0 NeV) within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-18 _,'

(E > 1,0 MeV)'within the Pressure Vessel Wall j

6-5 Relative Radial Distributions of Iron Displacement 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 Capsule U 6-21 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 Rates 6-25 ._

at the Surveillance Capsule Center .;

1 6-11 Adjusted Neutron Energy Spectrum at the Surveillance Capsule 6-26 l Center 1

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LIST OF TABLES (Cont)

I'? ~ Table Title s Page t9-

"'; 6-12: Comparisoniof Calculated and Measured Exposure Levels for 6-27 Capsule V 6-13 '.. Neutron. Exposure Projections at Key Locations on the 6-28 Pressure Vessel Clad / Base Metal Interface

'6-14 Neutron. Exposure Values for use in the Generation of 6 Heatup/Cooldown Curves

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l 6-15 Updated!Liad Factors for Byron Unit 2 Surveillance Capsules 6-30 b; -

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SECTION 1.0

SUMMARY

OF RESULTS

. The analysis of the reactor vessel material contained in surveillance Capsule U, the first capsule to be removed from the Commonwealth Edison Company Byron I Unit 2 reactor-pressure vessel, led to the following conclusions:

o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 3.96 x 10 13 n/cm2 after 1.15 EFPY of plant operation,

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o Irradiation of the reactor vessel lower shell forging MK24-3 Charpy specimens to 3.96 x 10 18 n/cm2 (E > 1.0 NeV) resulted in no 30 and 50 ft-lb transition temperature increases for specimens oriented parallel to the major working direction (tangential orientation) and a 25'F transition temperature increase for specimens oriented normal to the major working direction (axial orientation).

o The weld metal and weld HAZ metal Charpy specimens irradiated to

-, 3.95 x 10 18 n/cm2 (E > 1.0 MeV) resulted in 30 and 50 f t-lb transition temperature increases of 0 and 25'F, respectively. This results in a 30 f t-lb transition temperature of -65'F and a 50 f t-lb transition temperature of O'F for the weld metal and a 30 f t-lb transition temperature of -145'F and a 50 ft-lb transition temperature of -125'F for the HAZ metal.

o The average upper shelf energy of the shell forging MK24-3 showed no decrease in energy after irradiation to 3.96 x 10 18 n/cm2 (E >

1.0 MeV). The weld metal showed no decrease in upper shelf energy after irradiation to 3.96 x 10 18 n/cm2 (E > 1.0 MeV). Both I

materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an

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upper shelf energy of no less than 50 ft-lb throughout the life of the vessel as required by 10CFR50, Appendix G.

o The surveillance capsule test results do not indicate any significant changes in the RT NDT values projected for the reactor vessel, moucionolo 11 q

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t o The: calculated end'of-life (32 EFPY)- maximum neutron fluence; (E > .

1.0 MeV)-for-the Byron Unit 2 reactor vessel is as follows:

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Vessel inner radius - 3.03 x'10 19 n/cm 2

19 2 .

Vessel 1/4': thickness -11.66- x '10 n/cm 2

. Vessel 3/4 thickness a 3.57 x 1019 n/cm . ,

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SECTION

2.0 INTRODUCTION

• 1g This report presents the results of the examination of Capsule V, the first s capsule'to be removed from the reactor in the continuing surveillance program

', which monitors the effects of neutron irradiation on the Byron Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Byron Unit 2 reactor pressure vessel <

materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation '

mechanical properties of.the reactor vessel materials are presented by L. R.

Singer.(1) The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-79, " Standard Practice for conducting Surveillance Tests for light-water cooled Nuclear Power Reactor Vessels". Westinghouse Power Systems personne1'were contracted to aid in the preparation of procedures for removing capsule "V" from the reactor and its shipment to the Westinghouse Science and Technology Center

,- where the postirradiation mechanical testing of the Charpy V-notch impact and l

tensile surveillance specimens were performed at the remote metallographic facility.

This report summarized the testing of and the postirradiation data obtained from surveillance Capsule "U" removed from the Byron Unit 2 reactor vessel and discusses the analysis of these data.

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3990s/010490 10 2-1 l-

SECTION

3.0 BACKGROUND

• . The ability of the large steel pressure vessel containing the reactor core and-its primary coolant to resist fracture constitutes an important factor in

} ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 508 Class 2 (base materia'l of the _;

Qommonwealth Edison Company Station Byron 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.

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

. Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel [

Code. The method uses fracture mechanics concepts and is based on the l reference nil-ductility temperaturo (RTNDT)*

RT NDT 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 (transverse) 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 G of the ASME Code. The Kyp 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 Kyp curve, allowable stress intensity factors can be obtained for this l- material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

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L 'RTNDT and. in turn, the operating limits of nuclear power plants can be adjusted to account for_the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical-properties of a given reactor pressure vessel steel can be monitored by a -

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reactor surveillance program such as the Byron Unit 2 Reactor Vessel Radiation Surveillance Program,II) in which a surveillance capsule is periodically- l

. removed from tne 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 RT NDT f r radiation embrittlement. This adjusted RTNDT (RTNDT initial +

ARTNDT) is used to index the material to the KIR curve and, in turn,.to set operating limits for the nuclear power plant which take into account the l effects of irradiation on the reactor vessel materiais.'

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3-2 A

i-SECTION

4.0 DESCRIPTION

OF PROGRAM y

' *f Six surveillance capsules.for monitoring the effects of neutron exposure on the Byron Unit 2 reactor pressure vessel core region material were inserted in

'the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron shield pads and the vessel wall as shown in figure 4-1. The vertical center of the capsules is opposie J:e vertical center of the core.

Capsule U was removed after 1.15 effective full power years of plant ,

operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact  !

tension'(CT) specimens (figure 4-2) from the intermediate shell forging MK24-3 and submerged are weld metal representative of the intermediate to lower shell beltline weld seam of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) material. All heat-affected zone specimens were obtained from within the HA2 of forging MK24-3 of the representative weld. ,

The chemical composition, heat treatment and touchness data of the g survelliance material are presented in Tables 4-1 through 4-4. The chemical l

analyses reported in Table 4-1 were obtained from unirradiated material used l in the survelliance program. In addition, a chemical analysis using i Inductively Coupled Plasma Spectrometry (ICPS) was performed on irradiated {

specimens from forging HK24-3 and weld metal and is reported in Table 4-2.

The chemistry results from the NBS certified reference standards are reported in Table 4-3. Table 4-4 contains the toughness data for the reactor y vessel materials, I All test specimens were machined from the 1/4 thickness location of the f 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 longitudinal axis of the mo,,omeo io 41 l

3

specimen parallel to the major working' direction of the forging (tangential f orientation) and also normal to the major working direction (axial ,

orientation). Charpy V-notch and tensile specimens from the weld metal were L

or_iented with the' longitudinal axis of the specimens trensverse-to the welding- .

direction. The' CT specimens in the Capsule U were machined such that the ,

simulated crack in the specimen would propagate normal and parallel to the e: ~ major working direction for the forging specimen and parallel to the weld direction.

. Capsule V contained dosimeter wires of pure copper, iron, nickel, and

~a luminum-0.15% colbalt (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were

' contained in the capsule, t Thermal monitors made- from the two_ low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. The composition.of the two alloys and their melting points;are as follows:

2.5% Ag, 97.5% Pb ' Melting Point: _579'F (304'C) l'.75% Ag, 0.75% Sn, 97.5% Pb Melting Point: 590'F (310*C) -

1

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

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3990s/010490 10 4-2

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TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE' BYRON UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS' 0

' Chemical Cox)osition (wt%)

Element Lower Shell "orging MK24-3 Weld Metal'(a)-

C 0.1B ' O.09 Mn- 1.18 1.34 P 0.005 0.01 S 0.006 0.013 Si 0.23 0.55 Ni 0.65 0.65 Mo 0.43 0.45 Cr 0.06 0.08 Cu 0.07 0.03 A1 0.026 0.003 Co <0.01 (b) <0.01 (b)-

Pb <0.001 <0.001 W <0.02 <0.02 Ti <0.005 <0.005

-- 2r <0.002 <0.002 V <0.001 <0.001.

Sn <0.002 <0.005 As <0.005 0.005 Cb <0.003 <0.003 N 0.007 0.006 2

B <0.005 0.005 Heat Treatment History Material Temperature (*F) Time (Hr) Coolant Lower Shell (Forging MK24-3) Austenitizing 1575-1625 9.0 Water quenched Tempered 1200-1250 10.25 Air cooled Stress Relief 1100-1200 12.75 Furnace cooled WeldMetal(8) Stress Relief 1100-1200 12.75 Furnace cooled (a) This weldment was fabricated by The Babcock and Wilcox Co., using 5/32 inch weld filler wire, heat number 442002 and Linde 80 flux, lot number 8064 and is identical to that used in the actual fabrication of the 7~

reactor vessel intermediate to lower shell girth weld.

(b) Westinghouse analysis from surveillance program test plate.

n ouioins so 4-3

m TABLE 4-2 CHEMICAL COMPOSITION FOR BYRON UNIT 2 CAPSULE U IRRADIATED CHARPY IMPACT SPECIMENS Chemical Composition (wt.%)I*I .

Weld Metal:

Specimen Nc. Cu Ni C Mn P- ~S "Si Cr - No V Co YW-6 0.024 0.740 0.080 1.401- 0.008 0.013 0.496 0.085' O.397 <0.005 <0.010-YW-15 0.024 0.786 0.078 -1.509 0.016 0.013 0.513 0.093 0.452 <0.005 <0.010 YW-1 0.022 0.704 YW-2 0.020 0.681 .

YW-3 0.021 0.706 YW-4 0.020 0.697 YW-5 0.019 0.668 YW-7 0.022 0.759-YW-8 0.021 0.714 4- YW-9 0.020 0.678

2. YW-10 0.020 0.695 YW-11 0.019 0.689 YW-12 0.021 0.744 YW-13 0.022 0.738 YW-14 0.022 0.771 Forging MK24-3 Specimen No. Cu Ni C Mn P S Si Cr No V Co-YT-1 0.022 0.689 0.068 1.353 0.014 0.017 0.493 0.083 0.405- <0.005 <0.010 l (a) Method of analysis -- Inductively Coupled Plasma Spectrometry (ICPS) for all elements except C,:S, and Si.

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l~ .h TABLE 4 CHEMISTRY RESULTS FROM THE NBS~

CERTIFIED REFERENCE STANDARDS Material ID Low Alloy Steel: NBS Certified Reference Standards

'I NBS 361 NBS 362 Certified. Measured (a) Certified Measured ~(a)

' Metals- Concentration in Weight Percerit Fe 95.6 (matrix) 95.3 (matrix)

Mn 0.66 0.675 1.04 1.077 Cr- 0.694 0.7 06 0.30 0 311 Ni 2.00 above calib 0.59 0.6 37 Mo 0.19 0.208 0.068 o.061 Co 0.032 0.033 0.30 0.3 54 Cu 0.042 0.0 45 0.50 0.529 P 0.014 o.0189 0.041 o . o 47 2 V 0.011 o.01.11 0.040 -o.0398 C 0.383 0 386 0.160= 0.161-S 0.014' N.A. 0.036 0.0 406

, Material ID Low Alloy Steel:

. NBS Certified Reference Standards NBS 363 NBS 364

, Certified Measured (a) Certified Measured (b)

Metals Concentration in Weight Percent Fe

• 94.4 (matrix) 96,7 (matrix)'

Mn 1.50 1.518 0.255 0.252 Cr 1.31 -1.339 0.063 0.06o Ni 0.30 0 323 0.144 0.149 Mo 0.028 0.024 0.49 0. 47 5 Co 0.048 0.0 50 0.15 0.17 2 Cu 0.010 0.103 0.249 0.258 P. 0.029 0.0339 0.01 0.0147 V 0.31 0.276 0.105 0.105 0.62 N.A. 0.87 N.A.

- S 0.0068 N.A. 0.0250 0.0251

• Matrix element calculated as difference for material balance.

T'entative value, certified + 100% of value.

NA - Not analyzed; NR, Not requested (a) Method of analysis -- Inductively Coupled Plasma Spectrometry (ICPS) for all elements except C, S anc Si.

me.<omso io 4-5

TABLE 4-4 BYRON UNIT 2 REACTOR VESSEL TOUGHNESS DATA 50 ft/lb  : Upper Shelf Energy.

! Material Cu P Ni TNDT 35 mil RTNDT NMWD(a) MWD (b) l Component Heat No. Spec. No. (%) (%) (%). _ (*F) Temp.(*F) (*F) (ft-lb)- _(ft-lb). ,}.

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Closure Head Dome C4375-2 A533 B C1.1 .12 .013 .65 -40 <20 -40. 114 ---

Closure Head Ring 48C1300-1-1 A508 C1. 3 .05 .007 .69 -30 <30 108 ----

Closure Head Flange 2029-V-1 A508 C1. 2 ---

.011 .71 0 <60 0 157 ---

Vessel Flange 124L556VA1 A508 Cl. 2 ---

.008 . 70 30 <90 30 129 ---

Ialet Nozzle 51-2979 A508 C1. 2 .07 .010 .86. -10 <50 -10 130 ' ---

Inlet Nozzle 51-2979 A508 C1. 2 .07 .009 .86 <40 -20 121 ---

Inlet Nozzle 42-5105 A508 C1. 2 .07 .008 .84 0 <60 0 122 ---

Inlet Nozzle 42-5105 A508 C1. 2 .07 .011 .84 0 <60 0 121 ---

Outlet Nozzle 11-5052 A508 C1. 2 -.09 .010 .85 -10 <50- -10 108 ---

u Outlet Nozzle 11-5052 A508 C1. 2 .08 .007 .81 -10 <50 -10 121 ---

E Outlet Nozzle 4-2953 A508 C1. 2 . 09 .010 .78 -20 <40 -20 133 -- -

Outlet Nozzle 4-2956 A508 C1. 2 .09 .009 .81 -10 <50 -10 121 ---

Nozzle Shell 4P-6107 A508 C1. 2 ---

.014 .74. 10 <70 10 155 ---

Upper Shell 490329/ 'A508 C1. 3 .01 .007 . 70 -20 <40 -20 149 149 49C297/1-1 Lower Shell 490330/ A508.C1. 3 .05 .008 .73 -20 <40 -20 127 159 49C298/1-1 Bottom Head Ring 48D1566/1-1 A508 C1. 3 .07 .007 .67 -30 <30 -30 126' ---

Bottom Head Dome C3053-1 A5338 C1. 1 .06 .004 .64 -30 40 -20 121 ---

Upper Shell to WF447 (c) SAW .059 .009 .62 10 <70 10 80. ---

Lower Shell Girth Weld Weld HAZ --- ---

-60 <0 -60 143 ---

(a) Normal to major working direction.

(b) Major working direction (c) Welded using 5/32 inch weld wire Heat No. 442002 and Linde 80 Flux Lot. No. 8064.

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- (238.5 *) X- W IIII'0*I 1

REACTOR

' VESSEL .j

[* I

.. 180' , i 1

5 PLAN VIEW 1 L ,

h h [ VESS$.L WALL -  ;

) { k CAPSULE Neutron Pads are located CORE l

'lllllllll[j !

" [  ;

at 45 degree angles from each axis, l !

; s

)' . CORE MIME -q

< l l N l N NEUTRON PAD 1 I

6 s k I /

" CORE BARREL {

f l f l

. 1 ELEVATION VIEW Figure 4-1. Arrangement of Surveillance Capsules in the Reactor Vessel l

ma.-no,ese ,o 4,

i

c - ---. - -

~#

m b i

I i

SPECIMEN NUpeERING LEGEND. YL - LOWER SHELL FORGesG MK24 30ml(TANGENTIAL OMENTATION)

YT - LOWER SHELL FORGSIG MK24 3 (AxlAL ORIENTATION)

YW - WELD METAL YH - HEAT AFFECTED ZONE MATERIAL

a. Forgeg Heat Number 49D3301/49C296-1 or .

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M M o' M TO TOP OF VESSEL 19 EE':ffW. .F 1p e Figure 4-2. Capsule U Diagram Showing 1.ocation of Specimens, Thermal Monitors and Dosimeters a

. O h W W $ p

_ - . - .~

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

a  :

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SECTION 5.0 -

TESTING OF SPECIMENS FROM CAPSULE U L

5.1 Overview

. The post-irradiation mechanical testing of. the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center with 3 consultation by Westinghouse Power Systems personnel. Testing was performed in acecrdance with 10CFR50, Appendices'G and H,(2) ASTM Specification E185,.

and Westinghouse Procedure RMF-402, Revision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.

Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-10395(1) No discrepancies were found.

Examination of the two low-melting point 304*C (570'F) and 310'C (590'F) t 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 Cherpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure.8103, Revision 1 on 'a Tinius-01sen Modol 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, the load of general yielding (PGY), the time to general yielding (tGY), the maximum load (Pg ), and the time to maximum load (ty) 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 (Pp), and the load at which fast fracture terminated is identified as the arrest load (PA)*

mwomeo io 5-1

+ r g i

s

-) f ' .

The energy at maximum. load (E g ) was deterinined b'y comparing the energy-time record and the load-time record. -The energy at maximum load is roughly equivalent'to the energy required to-initiate a. crack in the specimen.

Therefore, the propagation energy for the crack (E p

) is:the difference between the total energy to fracture (E )Dand the energy at maximum load. ,

The yield stress (cy) is calculated from the three point bend formula. .

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

Percent shear was determined from post-fractura photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The

, lateral expansion was measured using a dial gage rig similar-to that shown in the same cpecification.

Tension tests were performed on a 20,000 pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-83 and E21-79, and RMF

. Proc'edure 8102, Revision 1. All pull rods, grips, and pins were made of .

^

Inconel 718 hardened to R: 45. The upper pull rod was connected through a universal joint to' improve axiality of loading. The tests were conducted at a -

constant crosshead speed-of 0.05 inches per minute throughout the test.

Deflection measuremen~ts were made with a linear variable. displacement

-transducer (LVOT) 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 0-2 per-. ASTM E83-67.

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.

Becauseofthedifficultyinremotelyattachingathermocoupledirectlytothe '

specimen, the following procedure was used to monitor specimen temperature. '

Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In the test configuration, with a slight load on the specimen, a plot of specimen neo.-ioinsto 5-2

p g, 1 A. ' '

f temperature versus upper and lower grip and controller temperatures was i developed over the range of room temperature to 550'F (288'C). The upper grip was used to control the furnace temperature. Du' ring the actual testing the l grip temperatures were used to obtained desired specimen temperatures.

Experiments indicated that this method is accurate to +2*F. -

I

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

strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to ,

calculate the fracture stress (true stress at fracture) and percent reduction- ,

in area was computed using the final diameter measurement.

5.2 Charpy V-Notch Impact Test Results The results of Charpy V-notch impact tests performed on the various materials y contained in Capsule U irradiated to 3.96 x 10 18 n/cm2 (E > 1.0 MeV) are presented in tables 5-1 through 5-4 and are compared with unirradiated

?.; results(1) ae s%wn in Figures 5-1 through 5-4. The transition temperature increases t . nr shelf energy decreases for the Capsule U materials are summarized , e e 5-5.

Irradiation of the vessel lower shell forging MK24-3 material (tangential orientation) specimens to 3.96'x 10 18 n/cm2 (Figure 5-1) resulted in no increase-in the 30 and 50 ft-lb transition temperatures and no upper shelf energy decrease. Irradiation of the vessel lower shell forging MK24-3 l material (axial orientation) specimens to 3.96 x 10 18 n/cm2 (Figure 5-2) l resulted in a 25'F increase in the 30 and 50 f t-lb transition temperatures and no upper shelf energy decrease.

I l Weld metal irradiated to 3.'96'x 1018 n/cm2 (Figure 5-3) resulted in no 30 1

and 50 ft-lb transition tempe.aturat ncrease i and no upper shelf energy L decrease.

l L l 1

l t

a.wioi... io 1

1

.$_3 l x

i k

18 i Weld HA2 metal irradiated to 3.96 x 10 n/cm2 (Figure 5-4) resulted in a 30 and 50 f t-lb transition temperature increase of 25'F and no upper shelf energy decrease, respectively.J A large scatter in data was observed which is typical of many HAZ Charpy tests for other surveillance programs. .

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasingly ductile .'

or tougher appearance with increasing test temperature, ,

A comparison of the 30 ft-lb transition temperature increases for the various Byron Unit 2 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2(3) is presented in Teble 5-6. This.

. comparison indicates that the transition temperature increases resulting from irradiation to 3.96 x 10 18 n/cm2 are less than the Guide precictions.

5.3 Tension Test Results The results of tension tests performed on shell forging HK24-3 (tangential ar.d -

18 2 '

axial orientation) and the weld metal irradiated to 3.96 x 10 n/cm are shown in Table 5-7 and are compared with unirradiated resultsII) as shown in ,

Figures 5-9, 5-10 and 5-11. Forging MK24-3 test results are shown in Figures 10 2 5-9 and 5-10 and indicated that irradiation to 3.96 x 10 n/cm caused a less than 5 ksi increase in the 0.2 per. cent offset yield strength and ultimate tensile strength. Weld metal tension tests results shown in Figure 5-11, show w that the ultimate tensile strength and the 0.2 percent offset yield strength increased by less then 1 ksi with irradiation. The small increases in 0.2%

yield strength and tensile strength exhibited by the forging material and weld metal indicate that there materials are not highly sensitive to radiation at 3.96 x 10 18 n/cm2 , as it, also indicated by the Charpy impact test rasults. The fractured tention specimens for the forging material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14. A typical stress-strain curve for the tension tests is shown in Figure 5-15.

3900s/01049ti 10 5-4

4 . ' .

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(.

+

a . - 5;4' Compact Tension Tests. ,1 l

3..

A Per the surveillance capsule testing program with the Commonwealth Edison 4

~

. Company,1/2 T-compact tension fracture mech"anics- specimens will .not be tested -

. and will-be stored at the hot cells- at the Westinghoujeg &T S Csnter,

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q 't TABLE 5-1  :

CHARPY Y-NOTCH IMPACT DATA FOR THE BYRDN UNIT 2 SHELL FORGING MK24-3 IRRADIATED AT 550'F. ,

FLUENCE 3.96 x 10 18 nf,,2 (E > 1.0 MeV)

\ t To

• erature Impact Energy Lateral Expansion Shear -

Sample No, 1*fd, Ut-lb) H}. IAME). 13R). 1 Arial Orientation -

YT15 -50 - 9.0 12.0 7.0 0.18 5 YT4 -26 -

59.0 80.0 45.0 1.14 30 -

YT14 -25 - -

14.0 19.0 10.0 0.25 10 i YT1 -25 - 25.0 34.0 20.0 0.51 15 l YT10 -10 - 32.0 43.5 25.0 0.64 25 ,

YT9 -10 - 35.0 47.5 29.0 0.74, 25 i YT3 0 - 62.0 84.5 46.0 1.17 30 l YTS 0 - 62.0 84.0 45.0 1.14 30 YT13 90 - 60.0 81.5 48.0 1.22 30 4 YT2 50 84.0- 114.0 56.0 1.42 40 YT7 72 112.0 152.0 68.0. 1.73 70

• YT11 150 160.0 217.0 85.0 2.16 100 '

YT5 250 165.0 223.5 78.0 1.98 100 YT12 350 161.0 218.5 79.0 2.01 100 YT6 400 154.0 209.0 79.0 2.01 100 .' ;

fannentialOrient.t. USA -

YL1 -80 - 5.0 7. 4.0 0.10 2 -

YL12 -50 -

34.0 46.0 26.0 0.66 15 YLS -50 - 10.0 13.5 7.0 0.18 5 YL6 -25 -

50.0 68.0 39.0 0.99 35 YL2 -25 -

21.0 28.5 18.0 0.46 15 YL3 -10 - 54.0 73.0 40.0 1.02 45 YL5 -10 - 41.0 55.5 33.0 0.84 35 YL14 0 - 68.0 92.0 49.0 1.24 55 YL10 20 - 58.0 78.5 42,0 1.07 55 YLS 50 115.0 156.0 79.0 2.01 80 YL13 72 125.0 169.5 81.0 2.06 80 YL15 150 160.0 (217.0 89.0 2.26 100 YL11 250 176.0 (238.5 88.0 2.24 100 ,,

YL4 350 228.0 309.0 74.0 1.88 100 .

YL7 400 202.0 274.0 73.0 1.85 100

..so.. io 5-6

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

~

18

- 550'F FLUENCE 3.96 x 10 n/cm2 (E > 1.0 WaV) 9 Yesperature Impact Energy Latern3 hpwwiu Shear Sample No. I'Z), .CA), fft-lbi LD, Mid Isp1 .IEL Wald lle o,a1 YW12 -180 -118 14.0 i h9.0 %.0 0,%3) 1G YW5 -150 -101 18.0 d34.5 23,0 0.03) 15 YW2 -135 - 87 19.0 96.0 12.0 0.30) 15 YW9 - 75 - 59 33.0 44.5 27.0 (0.69) 3D YW3 - 75 l - 18.0 S4.5 15.0 (9 38) 15 YW4 - so f 59 51 GPERAYOR EBROR - - -

YWe - so v-51 s8.0 38.0 32.0 0 Es s5 YW1 - 25 l-32 38.0 51.5 22.0 0.41 35 YW14 - 25 - 32 35.0 47.5 30.0 0 76 30

. YIl0 25 - 4 56.0 76.0 $0.0 3.27 55

- YW7 35 - 4 56.0 76.0 48.0 1.22, 55 YW8 72 22 72.0 97.5 60.0 h.63) 95

• YW11 150 66 73.0 99.0 64.0 1.03 100 YW15 360 121 79.0 107.0 69.0 1.75 100 YW13 350 177 82.0 111.0 71.0 1.70 100 RAE Metal .,

. YR13 -200 -129 10.0 13.5 5.0 (th13 5 YIl -175 -115 59.0 80.0 38.0 (0.97 40 YR3 -175 -115 23.0 31.0 10.0  !;0.25 15 YR10 -150 -101 32.0 43.5 26.0 1 25 YE6 -150 -101 28.0 38.0 15.0 {0.66 0.38 20 YR2 -135 - 87 52.0 70.5 29.0 (0.74 35 YR4 -135 - 87 100.0 148.0 60.0 (1.52 70 YR12 -125 - 87 37.0 50.0 26.0 (0.66 30 YB5 -100 - 73 78.0 106.0 46.0 (1.17 60 YR8 -100 - 73 52.0 70.5 28.0 0.71 50 YB11 - 50 - 46 115.0 156.0 65.0 1.65 80 YR14 72 22 159.0 215.5 83.0 2.11 100 YB15 205 93 162.0 219.5 80.0 2.03 100 YB9 300 149 193.0 261.5 74.0 1.88 100 f_ YB7 300 149 DID NOT BREAK - - 100 i

m.o..ioi io 5-7

i TABLE 5-3

{

INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE BYRON UNIT 2 SHELL FORGING MK24-3 IRRADIATED AT 550*F. FLUENCE 3.96 x 1018 ,7c ,2 (E > 1.0 MeV) i Normalised Beermies Test Charpy Charpy Maxiome Prop Yield Time Maximus Ties to Fractere Arrest Yield Fles Sample Temp Emergy Ed/A Em/A 2 Ep/A Load to Yield Lead Monisse Load Load Strees 94rees Number f*F1 H t-lb) fft-lb/is ) ggg,,) g,,,,) ,ggg,,) g,,,,) ggg,,) ggg,,) gg,g) gg,g)

Amial Orientation Y715 -50 9.0 72 34 38 2.90 95 3.90 130 2.90 0.20 95 112 YT14 -25 14.0 113 59 54 3.15 135 3.45 206 3.45 -

104 109

. YT1 -25 25.0 201 157 44 3.30 85 4.10 370 4.06 -

108 122 YT4 -25 59.0 475 301 174 2.90 85 4.35 875 4.30 0.15 96 120 YTIO -10 32.0 258 203 54 2.95 80 4.10 475 4.10 0.16 98 117 YT9 -10 35.0 282 229 53 3.15 80 4.35 506 4.30 0.25 106 125 YT3 0 62.0 499 292 208 2.85 75 4.25 SEE 4.20 0.15 96 118 YT8 0 62.0 499 305 194 3.10 120 4.80 000 4.25 0.20 103 128 T

YT13 20 60.0 483 302 181 2.85 90 4.45 680 4.25 0.45 94 120 i YT2 50 84.0 676 284 393 2.65 120 4.25 000 4.00 1.15 88 114 YT7 72 112.0 902 343 559 2.90 120 4.45 775 - -

98 122 YT11 150 160.0 1288 323 965 2.65 75 4.10 785 - -

87 111 YT5 250 165.0 1329 324 1005 2.60 70 4.00 780 - -

88 100 YT12 350 161.0 1269 308 988 2.65 105 3.80 780 - -

. 87 107 YT6 400 154.0 1240 293 947 2.05 55 3.00 755 - -

68 93 Tammential Orientation YL1 -80 5.0 40 31 10 3.00 105 3.75 130 3.75 -

98 111 YL9 -50 10.0 81 32 49 2.75 125 3.80 145 3.90 -

90 108 E12 -50 34.0 274 229 45 3.40 90 4.50 500 4.45 -

112 130 YL2 -25 21.0 169 140 30 3.40 90 4.10 335 4.10 -

113 124 YL6 -25 50.0 403 332 70 3.00 80 4.35 730 4.35 0.15 99 121 YL5 -10 41.0 330 256 74 3.00 85 4.30 500 4.30 -

101 121 E3 -10 54.0 435 294 140 3.00 80 4.70 620 4.65 0.20 99 127 YL14 0 68.0 548 312 235 3.20 130 4.60 700 4.35 0.30 108 129 E10 20 58.0 467 328 139 2.95 65 4.00 715 4.55 0.55 98 125 YL8 50 115.0 926 379 547 2.80 95 4.50 850 3.30 1.25 92 120 E13 72 125.0 1007 379 627 3.10 115 4.50 850 3.05 -

107 126 YL15 ISO 160.0 1288 344 944 2.50 55 4.05 820 - -

82 108 YL11 250 176.0 1417 342 1976 2.45 70 3.95 825 81 108 E4 350 228.0 1836 265 1571 2.35 120 3.85 700 - -

78 103 E7 400 202.0 1629 - - - - - - - - - -

TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE BYRON UNIT 2 WELD ' METAL AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 3.% x 1018 ,fc ,2 (E > 1.0 MeV)

Normalised Emeraies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flos j Sample Temp Energy Ed/A Em/A 2 Ep/A Load to Yield Load Maximum Load Load Stress 84rees i gg;,,) g,,,c) ggg,,) g,,,,) (gg,,) ggg,,) gg,;) gg,g)

Number M (ft-lb) (ft-lb/ia )

Weld Metal  !

YW12 -180 14.0 113 80 33 3.25 90 4.45 200 4.30 -

10T 127  !

YW5 -150 18.0 145 79 66 3.45 70 4.55 185 4.35 -

113 132 .

! YW2 -125 19.0 153 92 61 3.45 80 4.40 206 4.30 -

113 129 I YW3 - 75 18.0 145 114 31 3.50 165 3.95 345 3.95 -

116 124 YW9 - 75 33.0 266 229 37 3.25 85 4.35 495 4.35 0.15 108 126 YW4 - 60 OPERATOR EE80R - - - - - -

YW6 - 60 28.0 225 186 39 2.80 165 4.25 495 4.15 -

92 117 m YW14 - 25 35.0 282 218 83 2.90 75 4.06 500 3.90 -

97 115 i

, E YW1 - 25 38.0 306 221 85 3.15 80 4.10 500- 4.00 -

104 120  !

YW7 25 56.0 451 232 219 3.85 55 4.30 520 3.95 1.10 95 119 i YW10 25 56.0 451 211 240 2.85 80 3.95 506 3.35 1.00 94 113 '

YW8 72 72.0 580 256 324 2.85 106 4.15 000 3.25 1.95 95 118

, YW11 150 73.0 588 208 380 2.90 110 3.90 525 - - . 95 112  ;

1 YW15 250 79.0 636 181 455 2.80 75 3.90 470 - -

SS 105 1 YW13 350 82.0 660 254 406 2.85 ISO 3.70 715 - -

95 100 i EAE Metal YH13 -200 10.0 81 59 21 4.00 155 4.60 210 4.55 -

132 142  :

l i YH3 -175 23.0 185 132 53 3.00 70 4.75 275 4.70 -

100 128 YH1 -175 59.0 475 368 10T 3.30 106 5.06 706 4.90 0.20 100 138 YH6 -150 28.0 225 187 39 3.25 50 4.85 380 4.85 -

107- 134 ,

YH10 -150 32.0 258 191 66 3.50 70 4.90 385 4.85 -

115 138 i YH12 -125 37.0 298 92 208 1.95 85 3.70 275 - -

64 93 YH2 -125 52.0 419 360 58 3.20 120 5.25 895 4.95 -

105 140 YH4 -125 100.0 878 337 541 3.60 115 4.90 000 3.70 0.35 119 141 I YH8 -100 52.0 419 340 79 3.06 125 3.90 706 4.90 0.35 101 131 YE5 -100 78.0 628 339 289 3.75 90 4.90 665' 4.55 -

124 143 YH11 - 50 115.0 926 378 548 3.40 90 4.80 765 3.65 1.10 113 136 YH14 72 159.0 1280 329 952 3.15 105 4.70 690 - -

105 130 YHIS 200 162.0 1304 268 1036 2.00 45 3.40 775 - -

66 89  ;

YH9 300 193.0 1554 278 1276 2.20 85 3.25 825 - -

72 90 YH7 300 DID NRY BREAK - - - - -

t

. .y4 ., . ,. .se < , , ,, .% +-y.3 ..r _..__m_ ____.,__m

F-- ~

.7 TABLE 5-5 EFFECT OF 550*F IRRADIATION AT 3.96 x 10'8 vi/cm# (E > 1.0 Mev)

ON NOTCH TOUGSMESS PROrERTIES OF SYRCN LMIT 2 RE ACTOR VESSEL MATERI ALS Average 35 ott Average Energy Average 30 ft-Ib Lateral Expansion Average 50 ft-ib Absorption at Temperature ("F) Temperature (

• F ) Temperature ( ~ F) Full Shear (ft-1b)

Matertal Unirradtated Irradiated AT unteradiated treedtated AT thit erad$ ated arcadiated AT untreadiated teradiated attt-tb)

Forging -30 -30 0 -25 '-25 0 -25 -25 0 170 170 .O m24-3 (Tangentia1)

Forging -50 -25 25 -35 -10 25 -30 -5 25 154 154 0 m24-3 g ( Axtal) s

>.D O

Weld Metal -65 -+. 5 0 -25 -25 0 0 0 0 67 67 0 HAZ Matal -170 -145 25 -135 -110 25 -150 -125 25 131 131 0 3990s/010490:10

W . .. .n,-.,

4 .i TABLE 5-6

' COMPARISON OF BYRON UNIT 2 30 FT-LB TRANSITION TEMPERATURE SHIFTS ,

AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS 30-ft-lb Transition Temp. Shift Upper Shelf Energy Decrease ,

Fluence R.G. 1.99 Rev. 2 Capsule U R.G. 1.99 Rev. 2 Capsule U (Predicted)

IO 2 Material 10 n/cm (*F) (*F) (*F) (*F)

Forging NK24-3 (Tang.) 3.% 15.0 0 15 0 Forging MK24-3 (Axial) 3.% 15.0 0 15 0 Weld Metal 3.% 30.0 0 15 0 b

a) Cu and Ni values from table 4-1 were used to determine R.G. 1.99 predictions. ,

e 3990s9tO*9010

[

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

- TABLE 5-7 L

TENSILE PROPERTIES FOR BYRON UNIT 2 REACTOR VESSEL MATERIAL IRRADIATED AT 550*F TO 3.% x 1018 ,7c ,2 (E > 1.0 MeV)

Uniform Total Bedsetion Test 0.25 Yield Ultimate Fracture Fracture Fracture la Aswa Sample Temp. Strength Strength Load Strees Strength Blongation Blongation (ksi) (his) (kei) (kei) (5) (5) f5)

Material Number f*F) (ksi)

Forging 183.8 56.0 13.5- 28.1 SS BK24-3 YT1 78 87.7 88.8 2.75 80.8 79.5 2.15 154.1 43.8 10.5 23.2 69 (Axial 172 300 22.4 83-50.1 85.8 2.70 209.2 55.0 10.5 Orient.) YT3 550 Forging 182.7 51.9 14.2

• 30.8 72 MK24-3 YL1 78 88.8 89.8 2.55 u,

80.5 2.50 172.4 50.9 11.3 24.9 70 a (Tangential YL2 300 81.8 30.5 82 59.1 85.8 2.85 142.8 54.0 10.5 Orient.) YL3 550 71.3 88.8 3.15 144.3 84.2 10.5 21.0 58 'j Wald YW1 78 80 300 83.7 79.5 2.95 148.8 80.1 9.8 18.9 Metal YW2 81.1 9.0 18.0 88 YW3 550 83.7 80.5 3.00 192.1 me,-seiew se

curvo 757609-A

('C)

-150 -100 -50 0 50 100 150 200 i i i i 1 ' 3 i 100 -

'2: '2 :  :  :  : -

p E -

2 g 60 -

, 52 -

2 '

~

E -

2 .

0 ' ' ' ' ' ' '

'100 i i i , , , m i i 2.5 e

80 -

, ,- 2. 0

- 60 - e -

1. 5 'il 840 - *

1. 0 5 g20 -

0. 5 0 t i i i i J l 0

'. 220 i i i i i i 200 -

[228 Ft 'Lb)" .- 280

.. o Unirradiated 180 18 2

. Irradiated at 3.96x 10 nlcm ae o 240 1@ -- o.

,_. le -

1. 1 a *

~g 120 -

160 :::;

~

~ 1@ -

80 - 120 3 8 o e -

% 1

,o

. - g m - e 0 i i i i i i i 0

~

- 300 - 200 -100 0 100 200 300 @0 Temperature (*F)  :

Figure 5-1. Charpy V-Notch Impact Properties for Byron Unit 2 Reactor '

Vessel Shell Forging MK24-3 (Tangential Orientation) ano..iom io 5-13

L l Curve 757611-A (oC) 1

-150 -100 -50 0 50 100 150 200 l

l I I I I I i I >

- o- /g :  :  :  : -

. }@

3 E -

e

.]

! g 60 -

5e -

.l

.l

.2 - 2 l

0 I ' ' ' 3 '2 ' ' ' ' '

100 2.5 i i i i g_ i i .i g 80 -

4 . . .-- - - - -s , - , - - .- - , -

2. 0 i

E t

60 -

,5 -

1. 5 e l

,N 5

$40 -

25 'F

1. 0 i g 20 -

,f '

0. 5

' ' i 1 I I O

0 220 i i i i i i i i g _

22 -

1 80 - -

2@  :

160 -

200- o p'

140

,/

8 - Unirradiat _ 3g _

120

~

E 100 - <

b 80 - I Irradiated (550*F) 120 5 '

i .

18 2 60 -

/* 3 n/cm -

80

/= 25'F . 96x 10 40 -

• 25'F -

40 i [,, i 20 ,

O i i i i i 0

- 300 - 200 -100 0 100 200 300 400 .

Temperature (*F)

Figure 5-2. Charpy V-Notch Impact Properties for Byron Unit 2 Reactor Vessel Shell Forging MK24-3 (Axial Orientation) me.-ioiesi in 5-14

, curve 757608-A l ('C)

-150 -100 -50 0 50 100 150 200

' ' ' ' '3 '2 ' '

1% -  !=  ?= = = -

. E E -

u g - -

2  %

2

,'. . Si e _

. 8 ,

20 2 . .

t ' ' I I '

0 100 i i i i 2.5 i i i i 5E e

_........4..e 10 60 -

. g- - -

1. 5 _s

d. 40 j$ - -

1.0E g20 -

I

, 2 i i i i 1

0. 5 0

0

'. 220 i i , i i i i i 280 200 -

180 - o Unirradiated -

2@

18 2 160 -

. Irradiated at 3.96 x 10 Icm

~

_a 140 -

g 120 - -

160 =;

~

~ 100 -

h

. 120 e 80 -

_- ,--- a- - * - - , - - - -

w m_.e g -

2' -

2 2

40 -

2

. _ g ,

20 .. .

. O i i i i i i I 0

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

Figure 5-3. Charpy V-Notch Impact Properties for Byron Unit 2 Reactor Vessel Weld Metal woo.-ioisonia 5-15

curva 757610-A (SC) 150 -100 -50 0 50 100 150 200  ;

i i l

• J I I I IW - - -

y 5 - . o -

i gm - .

o g 5e .

20 -

' ' I i I '

0 i 100 i i i i i y , i 2.5

_ .l 2 80 -

.... ..... 4.... -

2. O  ;

3 a , e ..

60 - .

1. 5 'il  ;

d. .- e a

40 -

25' F

1. 0 e i

/sr.

20 - -

0. 5 3 ^ ' ' ' ' ' ' 0 0

220 i , , i i i i , .;

g _

280 180 - '

20 -

160 -

200 .

o

_ 140 / f, ' 8 Unirradiated , i e

j g 120 -

160 =;

~

~x 100 -

/ o >

, Irradiated at L 60 . , -

80 25o F 18 2 l- g _

/,- 3.96x10 n/cm 25'F -

40 .

l 20 - o ,

L 0 #'  ! ' I I I ' 0

- 300 - 200 -100 0 100 200 300 400 -

Temperature (*F)

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

neo. ioin io 5-16 .

-- ,, ,-y w ,,-.--,w---,-,-w,-av.-- e O

l

. - - ~, - . 7 7 m m .

m .7 ;. e --. y

g. ., 7,

- - . y%y ~

< r,j - -

. } j y;;; , L* 9 .

]hM y,

u.w n .a . . .n . ~ a , . .

YL1 YL9 YL12 YL2 YL6 +

m., ,, - , .e , ,,

p e r-y g.g 7,,,, r,, .

p r . .  :; . g. 3 . . . .a e

Yl*

Yl

^

h* sj v

• ; ._ 'k ,

,; ,, , ?. %,

. li .

s im , m < . _ m e .. - m. .. . m.

YL5 YL3 YL14 YL10 YL8

,. y W* -

g- .i _N l

4 f_ . . l, . y, p. .

, ,' f4i - .. :  ; g. s r ..

.a a .

,'Y

j. ___

• La ,

=

V;+

9 YL13 YL15 YL11 YL4 YL7 Figure 5-5. Charpy Impact Specimen Fracture Surfaces for Byron Unit 2 Reactor Vessel Shell Forging MK24-3 (Tangential Orientation) me. iomi so 5-17 RM-21315

I l

4

)

as - %P M +>y_,* .,-t', 4 li % n 8-T' *  ?- W # ) t # '" !^ j ^

, y ,

y ~",

l& ,

iLed ,1,i. ~

# ,,m .

,a e-YT15 YT14 YT1 YT4 YT10 i

i

" ' ' *~

• n - ': <r- '

i $: ,

bhY h  :

[ E 5 ,

9

{

YT9 YT3 YT8 YT13 YT2 l

V 4

anumme,3[m:

Pih 3  ?

.Q 'l m' .

YT7 YT11 YT5 YT12 YTS Figure 5-6. Charpy Impact Specimen Fracture Surfaces for Byron Unit 2 Reactor Vessel Shell Forging MK24-3 (Axial Orientation) me.-ioiess io 5-18 m-2n14 1 x L- . , - _ . . _ -- _ . _ _ , _ _ . . _ . . . . . ~ _ . - - - _ - - . -_

l i'

p., . .orag ;>m.m>,gry.s  : , p . e., y, p.y , .y , , , . , e_ c y ,, 7

, , , , , .n., .

,f .

, . n .A _ .. ->

x.  ;

W12 W5 U2 W3 W9 )

i L

l p ,s ,t m 3 .n.y .- - y,,. - -. x,,  ;

, t p@.: .

/-7 '

q ' , ,.. . t 7

p

1

-W. 7 4

l. 3 4

g c i

• A 0 J v ws '

g e .s .

W6 W14 W1 W7 f

Y W10 W8 Wil W15 W13 Figure 5-7. Charpy impact Specimen Fracture Surfaces for Byron Unit 2 Reactor Vessel Weld Metal me..ioises sa 5-19 RM-21316 l- . . _ - . .

i l

l m- - .- . . , , . . cy. ,., . y7. .. , , , 7 .,

Uk k? m;\ '

jff . . ; {' ,; 'Y '

..hA. . .{$W"'

t y ,...

.~ .

4[ ,

L

= mn

'N .

e..

.- 2 l .m c.,.-., ,

Ab. i YH13 YH3 YH1 YH6 YH10

)

l l

g; , , 7 n y r. ?,-.n ,, nr ,,..r. c.m , . .

p v.g., . ~ ..,,r.v. ,n.

y.  ;.v . z. .. .

, ~T[ .

f_ c W g. ,

YH12 YH2 YH4 YH8 YH5 i

e  % 1 L-

~

e  ; .:.

} (a==.t; s, .,$s

4.'

y'%

s < -

I l

1. y

. ,g

.. M M g. L , - ,

g ,

YH11 YH14 YH15 YH9 Figure 5-8. Charpy Impact Specimen Fracture Surfaces for Byron Unit 2 Reactor Vessel Weld Heat Affected Zone (HAZ) Metal an..ioins ie 5-20 RM-21317

i l

Curve 757612-A

'C

- 50 0 50 100 150 200 250 300

. 120 i i i i i i i i _

g 110 -  !

700 100 -

• 90 Tensile Strength 80 -

2 E -

E~

G 70 60 2

g, 2'O. 2 % Yleid Strength m - -

400

% - i { ,

.e i i i i i i l E I Code:

Open Points - Unirradiated 18 2

", Closed Points - Irradiated at 3.%x 10 n/cm j E i i i i i i i i l 70 -  :  ? g2 -

2

_ 60 Reduction in Area fg - -

x E 40 Total Elongation -

D g 30 -

2e _.

g 20 -

10 - I 2 ^ -

I I I nH rp SongaUon' 2 - ,

. 0

-100 0 100 200 300 400 500 600

. Temperature ( F)

Figure 5-9. Tensile Properties for Byron Unit 2 Reactor Vessel Shell forging HK24-3 (Tangential Orientation) neo. ioine io 5-21

1 i

l l

I curve 757613-A

'C .

- 50 0 50 100 150 200 250 300 l

, 120 I I I I I I - -

I I 800 ~

110 -

,j 700 100 _

j90 - Tensile Strength j 5 80 -

2 & 20 h 2

g 70 -

2 _

m- 1

0. 2 % Yield Strength  !

5  : -

@ l

[ E g i i i i i i i- ,

Code:

Open Points - Unirradiated  :

I 2 Closed Points - Irradiated at 3.96x 10 n/cm ,

80 i- i i i i i i i m _ Reduction in Area _

J-70 _

o M -

E

-g -

E 40 2

Total Elongation g 30 -

h -

g,2 20 -

Uniform Elongation 10 -

6- 5 s-2 -

I I I I I I i 0 .

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

Figure 5-10. Tensile Properties for Byron Unit 2 Reactor Vessel Shell Forging MK24-3 (Axial orientation) me.-ioissi in 5-22

curve 757614- A

'C

- 50 0 50 100 150 200 250 300 120 i i i , , , , , _

110 -

. 100 - -

700

'.5 90 -

g  % -- 600g

,Tenslie Strength g 2 g - -

E 500 -

g 70 A

,0.2% Yleid Strength 60 -

2' -

400 50 40 i i i i i i i- 300 Code:

Open Points - Unirradiated 18 2 Closed Points - Irradiated at 3.96x 10 n/cm S I I I I I I I l 70 - -

m Reduction in Area 9

_ 60 fg -

( 40 -

3 30 - Total Elongation _

20 -

b y 2 , _

10 _

_ Uniform Elongalon v2 2F 7;i 0 l- 1 I e i i

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

Figure 5-11. Tensile Properties for Byron Unit 2 Reactor Vessel Weld Metal neo. ioine io 5-23

s hgp %dh i.

ps.an >,."- : . *fpM M.mp w % ? M,%,id%

,*g 4Wl $'f T*- <Yh'f;*,6??ww'Eh

W '

qy' I '~ W W ~

w$*t,g vh .AyL"Y.

i V:s,N.G$!r,t,a:,.;

., 4 gMjNdbitPM

.y

';t' G b fl  % y;v,.:y 8%y;pt,q a ,)v.[y; s e%k&;gy 7-

*z\d y.

LAJf12

., g i

'fh,, &%y$p i c nh. s,h:

GQ' .

& w?  %: oa

.a

.~m.A(mynmlW AAa jUfl;Q.y;, w:' A_' w,.(.L;ft 3 'g l'gd;> 1 Specimen YL1 78'F i

I 1

. .w . , . .. <.7e.is ,

g ., -> A <-

_w, V

, ,y., f . .  ?)>. s

,'v m

. , . . ,.,%_. ) fg '

i '

)

N~

r h #1%W .

l a .

• ; . j

.e

...4 u% . , . . .

, 1

.v . ' A ,

. .r

c-4 s

,a

. u: .. i, ., ,. ,,{-

l _ _ _

Specimen YL2 300'F m ...

' ','  ?:

, Y7yu,g' f:V x. *.:;m gi;,k:g  ;;2k l.L yg -};- , g ' > ,9 9_

i; \\ %

"a f vi c

, 6-

,.r, i L ! Y:- ,

• t p v ,  ;'

%wGep...,-:, r,

< . , ,; ;d

• l -

m,

, q:; ,,,;

l 1

l l

V41 x ,r, e

g- gy .e r b ~,

e #1pir ,

• Mw  ; e Specimen YL3 550'F Figure 5-12. Fractured Tensile Specimens from Byron Unit 2 Reactor Vessel Shell forging MK24-3 (Tangential Orientation) l l

l l

l neo.-oisee so 5-24 RM-21319 l

l . _ _-____ -_. _- -__ _ _ _ _ _ _ _ - _ _ _

I 1

mwyn m+r , w..,s,x 7 ,, a:. ,wn, y m . n", n;!!

.y o ~m it/,[(i q s 1 :][gb dS.WM ' .s9#{fy , rihw %h3M.i '/'@l_i.f QJ.yd_f - ,2 - (;(

milTspj

+n .. u ) .s. w wa,a,..sn, , a- g ;wye x un a,

. , -n, + :m, n.

w e. ,.1.,,c4y4 w ey: uw m r_ms, w py.

.ew ,

.s.

p 4 e q uie p. hW%jf e2:41R) a N,n hdhk

,:~z 9- kn h[dh[NdQi%p%qp.a 0 y , 4 . W. &- w k $ 9d

-,n -

Specimen YT1 78'F

,, p f; m w e a. - y- +,;. y; .-w jw m,

'i ,< " * *. j . qg ~ yG..,hjGly 99 -

QM sg

~. '; ..

s.

g ,,uso h.4ly y%

aM 4@v%v c - + wm n .n;r. m%;fy:?pr el?p %.9. . . . .

. + fA-am- .e: m%, m, a; m .e :. ., - ;.

~,

_,v ,

r

?. O gywd j,[ )g *y gA,3p w, +:jll j  ? *'

f m +

y_. ._ . y n 1;Mkwk z , -($

, ., o . . t~

,.g

~ ~

f

. . . . . Mr .

. .. . x

. si vag t y .kq A':,

p>

• 3wm w4ye%p aw;. y$%w.%.

- ws .

.y,w Specimen YT2 300*F

. , +r.

%-)# - . ,,, ', ~.,} .1 -

i' <s,

2. .

y ,

1

p  %

1 @-

qn

.? #

(. .

se..

.a . <-

a ., \ .'

Specimen YT3 550*F Figure 5-13. Fractured Tensile Specimens from Byron Unit 2 Reactor Vessel i

Shell Forging MK24-3 (Axial Orientation)  ;

i l

l seo. scisse so t;.2 5 RM-21318 e-___--_-___-.-__.._

l 1

- umyh,ggJ x p p:n, , ;s ,;,.eage@.

n y

,w e,w

,we< .:, w

.aye yJ,-lli;:. w~

e, v, u

.ma , ; v. sg.w n s, .;e ,

sq ,w'y; n:wnMwy , ;ic w\ ,3. Q, t,g v  ; .\ .i. ,

.-o - mum p#,,,p ct p~ ' ma _. . r /.c, ,2,q;::a" t.r %,o ..?r,w . ;w lg:ps?:M, v,pu, p,98

+

w' Q' M y, i ,

• u

..un j #

,u A

p .

+- n

<-o. 44 . . .

fv $ Q.

e ei

@o+,.,y,4 k

<t

-- s,:

%Lg,,

wv tit $rMtM

><1w qy.n p% uy g $g $e 4w$

94 c,

x . +aL . ;.

,~

j j,M@j', . Aq; , NWW_p.s M.;, eh @ 0 M A.y diosthdi ' M S MaikKddL &l? W.1ax:ky d r lS Nr 3 l Specimen YW1 78'F 1

1

[A Sh5INM %M M vn- ~M ggy 't %j a'Nbbh2Uhkb.Nd.p:dt%ih'MW'IM'

. .y A g y ,1w,id ,h ;

,q y."M w.

;w owhJM*Y AMY I 5 >< c

'0' ~

n  :

,m;-w.n.

i pp k .h vMm; - a e r mM&%;-n g

+ m. y c::,,

nd l):

• n?q!)n%@'F-t 15g'-@'!V, f '-w k Wpfj;h; M fM:/ k;xN Y N.,Q W h $0 k B h ,

J

' q e

p. J rny;'4MI

,wgewa wgcu 'd r$q fM y A qg@W:gy;af.

l y-m M - .

us. nws&.a.A.M,w:+p%y"

. w 44 - ,

Specimen YW2 300*F 1

1 as. 3san. y  ; u. se m v--  % ,.

3m

.%en.,,l4j f pp.tx %y1H..N.yyy asi g ity^- . ' t i; p , W "

I PAftJ,vR.

)[; *er.

i ,- , -

p.ns,, ,

, . y+:n g, , , ; s 3,9, ,

,s

+<

J l

l s .: M c$f , , - (3 ,eh 4 j

1. s

q. . -I'a

=

Qf(6 M. ' ,

\ Q #4 1 e -

Specimen YW3 550*F Figure 5-14. Fractured Tensile Specimens from Byron Unit 2 Reactor Vessel Weld Metal me.*ue o 5-2A* RM-21320

r i

l t

I r

,e

.? f' i .* . -

100 l

{- I 80 -

• g 60 A

$ i t'

5 40 t t.. .

' 20 -

SPEC.TT3 550F .

t 0 I I i e 1 e O 0.04 0.00 0.12 0.16 0.2 0.24 0.28 Strain, in/in i

f(

v l

Figure 5-15. Typical Stress-Strain Curve for Commonwealth Edison Company Byron Station Unit 2 Shell Forging MK24-3 Tension Specimens.

noe.-ioiun io 5-27 l

l l

t r

i SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY l u

6.1 Introduction

, Knowledge of the neutron environment with,in 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 i 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 tie present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and  !

that experienced by the test specimens. The former requirement is normally '

met by employing a combination of rigorous analytical techniques and

, measurements obtained with passive ' neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from

. analysis.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for 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 survelliance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.

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

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

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

l me,mm. i.

6-1 a

n ,

[ .

(E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to

• the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the

~

Regulatory Guide 1.99, " Radiatio'n Damag'e to Reactor Vessel Materials."

This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule U. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the ,

projected exposure of the pressure vessel are provided. .

6.2 Discrete Ordinates Analysis

• 1 A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillance

! program. The 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.

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

! 1-inch and approximately 56 inches in height. The containers are positioned L axially such that the specimens are centered on the core midplane, thus .

spanning the central 5 feet of the 12-foot high reactor core.

1 ano notus io g.g

I from a neutron transport standpoint, the surveillance capsule structures are l

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.

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 parameters (v(E > 1.0 Mev), e(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/t(E > 1.0 MeV), within the pressure vessel geometry.

. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations.

The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule ,

positions, and several azimuthal locations on the pressure vessel. inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the >

l locations of interest for the cycle 1 irradiation; and established the means to perform similar predictions and docimetry evaluations for all subsequent l 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 me.nomi so 6-3

f M1 qy 4 G hp ' T  ;

L L s E i L -t .

i b of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the. bur up of individual fuel assemblies increased. ,

h i The absolute cycle specific data from the udjoint ("p uations together with l relative neutron energy spectra and radial distribution information from the -

l

' forward calculation provided the means to:  :

q , .

1. Evaluateneutrondodimetryobtainedfromsurveillancecapsule; ,

I locations. ,

1

2. Extrapolate dosimetry results to M y Iceations at the inner radius -

andthrough'thethicknessofthepressdravesselwall. .l

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

4. Establish a mecharnsm for projection of pressure vessel exposure as #

the design of each new fuel cycle evolves. .

The forward transport calculation -for the rehetor model summarized in ,

figures 4-1 and 6-1 was carried out in R, e geometry using the DOT ,

two-dimensional discrete ordinates code (41 and tho SAILO? cross-section library'[5). ' The SAILOR library is a 47 group E@FB-IV based data set produced specifically for light water reactor applications. In these analyses anisotopic scattering was treated with a P3 expansion of.the cross sections and the angular discretization was modeled with an S8 rder of angular .

quadrature.

i The reference core po%r distribution utilized in the forward analysis was i derived from statistical studies of long-term operation of Westinghouse 4-loop .

plants. Inherent in the development of this reference core power distribution .

is the use of an out-ir, fuel management strategy; i.e., fresh fuel on the core l

periphery, furthermore, for the peripheral fue' assemblies, a 2e i 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 +2e ,

4 l

l

. s J

. n.o.noin

6-4 l

M level for a large number of fuel cycles, the use of this reference

' distribution is expected to yield somewhat conservative results.

1*f All adjoint analyses were also carried out using an S8 rder of angular-quadrature and the P3 cross-section approximation from the SAll.OR library.

.f

, Adjoi,nt source locations were chosen at several aziw thal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, e geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, e (E > 1.0 MeV). Having the impor-tance functions and appropriate core source distributions, the response of interest could be calculated as:

R(r,0)=fr#0 E # I(r, 0, E) S (r, 0, E) r dr de dE where: R(r,0) = , (E > 1.0 MeV) at radius r end azimuthal angle e i

I(r,0,E) = Adjoint importance function at radius, r, azimuthal angle e, and neutron source energy.E.

S (r, 0, E) = Neutron source strength at core location r, e and t energy E.

Although the adjoint importance functions used in the Byron 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 ratio of I dpa/$ (E > 1.0 MeV) is insensitive to changing core source distributions.

In the application of these adjoint important functions to the Byron Unit 2 l 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/4 (E > 1.0 MeV) and 9 (E > 0.1 MeV)/, (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific & (E > 1.0 MeV) solutions from the individual adjoint evaluations.

sm.noins in 6-5 L1 ' ,

gggy '

e y V s y , g

, N -:, S i

I 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. Byron Unit 2 [6). The relative power levels in fuel i asserrolies 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 forn rd calculation is also illustrited in Figure 6-2.

Selected results from the neutron transport analyses performed for the Byron i Unit 2 rLactor are provided in Tables 6-1 through 6 :i. The data listed in

. these tables, establish the means for absolute con .ari:ons of analysis and measurement-for the capsule irradiation period and provide the means to correlate dosintry results wit'b the corresponding neutron exposure of the pressure vessel wall.

In Table 6-1, the calculated exposure parametdrs le (E > 1.0 MeV),

, (E > 0.1.MeV), and dpa) are given at the geonetric center of the two ,

surveillance capsule positions for both the design basis and the plant .

specific core power distributions. The plant specific data, based on the ,

adjoint transport analysis, are meant to establish the absolute c'omparison-of l measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar-data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycle 1 plant specific power distr.ibutions. It is important to note that the data for the vessel L

inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself.

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

L -(E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport *

' calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall J 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.

3990s/101989 10 Q

l l

m ,

n' ,

1 L

y:

For example, the neutron flux-(E > 1.0 HeV) at the 1/4T position on the 45'  ;

azimuth is given by:-  !

l

' = v(220.27, 45') F (225.75, 45')

1/4T(45')

Twhere- # = Projected neutron flux at the 1/4T position on 1/4T(45') ,

the 45' azimuth 1 a

e(220.27,45')

= Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth. I

'F (225.75, 45') =. Relative radial distribution function from Table 6-3.

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

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

!However,-for the neutron pad arrangement in Byron 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* to 45'  !

-(12.5') which produces the maximum vessel flux. Other octants have neutron pads extending 22.5* or 20' which provide more shielding. For the octant with the-12.5* pad, the maximum flux to the vessel occurs near 25' and the values in the tables for the 25' angle are vessel maximum values. Exposure values for 0*, 15*, and 45' can be used for all octants; values in the tables for 25' and 35' are maximum values and only apply to octants with a 12.5' neutron pad 1 extent.

6.3 Neutron Dosimetry l The passive neutron sensors included in the Byron Unit 2 surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation ano. nome io 6-7 a a

of.the neutron energy. spectrum within the capsule and.the subsequent determination of the'various exposure parameters of interest

[e (E >-1.0 Mev), , (E > 0.1 MeV), dpa).

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,wereplacedinholesdrilledinspacersat'severaiaxialleveiswithin the ca'psules. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.

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

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

known. In particular, the following variables are of interest:

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

o The energy response of the monitor, o The neutron energy spectrum at the monitor location.

-o The physical characteristics of the monitor.

The specific activity of each of the neutron monitors was determined using established ASTM procedures (7 through 20j. Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drif ted germanium, Ge(Li), gamma spectrometer. The irradiation -

history of the Byron Unit 2 reactor during cycle 1 was obtained from NUREG-0020. " Licensed Operating Reactors Status Summary Report" for the ,

applicable period.

l me. noms io 6-8

i The; irradiation history applicable to capsule U is given .in Table 6-7.

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

values were derived using the pertinent data

• from Tables 6-6 and 6-7.

> Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code [21). 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 along with associated uncertainties where then obtained from the adjusted spectra.-  ;

1 In the FERRET evaluations, a log normal least-squares algorithm weights both  !

the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are r ,

. linearly related to the flux 9 by some response matrix A: j

\

. f (s,a) = I A (s) ,g(a) i g ig i

where i indexes the measured values belonging to a single data set s, g 1 designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, R I

$=9 o$g #g l

[  :

E. relates a set of measured reaction rates R $

to a single spectrum 9g by the multigroup cross section o$g. (In this case, FERRET also adjusts the

. cross-sections.) The lognormal approach automatically accounts for the  !

physical constraint of positive fluxes, even with the large assigned uncertainties.

m e.no m eio 6-9

T In1the FERRET analysis of the dosimetry. data, the continuous quantities (i.e.,

fluxes-and cross-sections) were ' approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group; structure using the SAND-II code (22). This precedure was carried out

• by.first expanding the a priori spectrum into the SAND-II 620 group structure ',

using'a SPLINE interpolation procedure for interpolation in regions where  ;

group boundaries do not coincide. The 620 point spectrum was then easily- I collapsed to the group' scheme used in FERRET.

The cross-sections were also collapsed into the 53 energy group structure usingSAND11withcalculatedspectra(asexpandedto620 groups)asweighting 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 wer'e neglectea due-to data and code limitations, but are expected to be unimportant, For each set of data or a priori values, the inverse of the corresponding ,'

relative covariance matrix M is used as a statistical weight. In some cases,

• as for the cross sections, a multigroup covariance matrix is used. More ,

~

often, a simple parameterized form is used:  ;

N gg,=Rh_+R R,Pg gg, where RN specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The j fractional uncertainties Rgspecify additional random uncertainties for group g that are correlated with a correlation matrix:

Pgg. = (1 - 0) 6gg, + 0 exp (~ ] '.

i

- The first term specifies purely random uncertainties while the second term l describes short-range correlations over a range r (0 specifies the

! strength of the latter term.)

l l

me. noms ,o g.19 l l

t.

Y:

For the a priori calculated fluxes, a short-range correlation of r =-6 groups was used. This choice implies that neighboring groups are strongly correlated when 8 is close to 1. Stronglong-rangecorrelations(or

,. _anticorrelations)werejustifiedbasedoninformationpresentedby

, R. E. Maerker (23). Maerker's results are closely duplicated when T = 6.

-For the integral reaction rate covariances, simple normalization and random

+ uncertainties were combined as deduced from experimental uncertainties. 4 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 3.96 x 10 18 n/cm2 (E > 1.0 MeV) with-an associated uncertainty of 1 8%. Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the adjusted spectrum are provided in Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.

A summary of the measured and calculated neutron exposure of capsule U is presented in Table 6-12. The agreement between calculation and measurement falls within 1 12% for all exposure parameters listed. The calculated fast neutron exposure (9 (E > 1.0 MeV), 9 (E > 0.1 MeV), dpa) values agreed with the measurements to within 1-3% whereas, the thermal neutron exposure calculated for cycle 1 exceeded the measured value by 12 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 Table 6-2 were used to perform projections beyond the end of cycle 1.

^

l l

mo, nom to 6-11

7 , , -

w, .

.- py _

4 L _

l

.]

,l In the calculation 'of exposure- gradients for: use in the development _of heatup -  ;

[ and cooldown curves _for the Byron Unit-2 reactor coolant system, exposure i <

projections to 16 EFPY an'd 32 EFPY were employed. Data based'on both al  :

fluence (E>1.0MeV):slopeandaplantspecific~dpaslopethrough.the .

vessel wall are provided in Table'6-14. In order to access RT NDT ~ V 8 ' "

' fluence . trend curves, dpa equivalent fast neutron fluence levels for the 1/4Tl and 3/4T' positions.were defined by the relations v' (1/4T) = 4 (Surface) { dp ( e) }

d 4 d'('3/4T)=$(Surface)-(dpla

,) )

Using th'is approach results.in the dpa equivalent fluence values listed in

-Table 6-14.

In Table 6-15 updated lead factors are listed for each cf the Byron Unit'2 ,1 surveillance capsules. These data may be used as a guide in establishing +

future withdrawal schedules for the remaining capsules. ,

e 6

me.noins io 6-12

q.

+

p d,,

- (TYPICAM h - 88.50 - 61.00

-i

- 81.625 IN. '

J' #h 7///p/> M i }

{ k N

A nlA 'EUTRONPAD\  %% , b I J l

Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule mo. nam. io 6-13 e  !

7

~

q 4'

1 0.74 9.79- 9.78 9.59 Cycle 1 1,91 1.94 0.96 9.77 Design Basis 9.99 1.92 9.97 9.95 9.94 9.57 1.92 1.19 1.00 1.95 1.19 9.71 1.13 1.99 1.97 1.95 9.98 1.91 1.95 9.97 9.87 1.97 1.99 1.95

, 1.14 1.13 1.13 1.14 1.98 -

1.99 1.96 9.88 1.10 1.94 ,

1.18 1.14 1.14 1.29 9.99- 1.94 1.12 9.92 Figure 6-2. Core Power Distributions Used in Transport Calculations for Byron Unit 2 l

neo.noiu ,io 6-14

g j', .

[ *

, j.-

7 TABLE 6-l' I!

CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT-THE SURVEILLANCE CAPSULE-CENTER i t .:.

', DESIGN BASIS CYCLE 1 29.0* 31.5' 29.0' 31.5' ,

r(E > 1-0 MeV)

. 1.13 x~10 ll 1.21 x 10 11 8.84 x 10 10 9.51 x 10 10--  :

2 '

(n/cm -sec)

. '4 (E > 0.1 MeV) 5.07-x 10 11 5.44 x 10 11 3.97 x 10 ll 4.28 x 10 11

^

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

~l i

8 is b

he e

.-g a

me.noisas io 6-15 o .c

7.- _ ..

~

h TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT

THE' PRESSURE VESSEL CLAD / BASE' METAL INTERFACE ,

0ESIGN BASIS 0* - 15' 25' 35' 45' o

1.78 x 10 10 10 3.01 x 10 10 2.45 x 10 10 2.81 x 10 10

-v(E>1.0Mev) 2.66 x 10 2

(n/cm-sec) 10

1. (E>0.1Mev). 3.70'x 10 10 5.60 x 10 10 8.22 x 10 6.96 x 1010 7.04 x 1010 2

(n/cm-sec) dpa/sec 2.77 x 10 -11 4.12 x 10'11 5.04 x 10

-11 4.15 x 10 -11 4.48 x 10 -11 ,

CYCLE 1 SPECIFIC ,

0* 15' 25' 35' 45'

1. (E>1.0Mev) 1.32.x 1010' 2.06 x 1010 2.38 x 10 10 1.98 x 10 10 2.31 x 10 10 2

(n/cm -sec)

((E>0.1Mev) 2.74 x 10 10 4.34 x 10 10 6.50 x 10 10 5.62 x 10 10 5.79 x 10 10 2

-(n/cm sec) 2.05 x 10'11 3.19 x 10 -11 3.99 x 10

-11

- dpa/sec 3.35 x 10-11 3.68 x 10 -11 .

aeo.neine io 6-16  :

p, p

nj =

e .

TABLE 6-3

.u RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV)- s lV WITHIN THE PRESSURE VESSEL WALL Eo

... u, .

Radius v (cm)- 0* 15' 1 35' 45' 220.27(1) 1.00 1.00 1.00 1.00 1.00 .

220.64 0.976 0.979- 0.980. 0.977 -0.979 221.66 0.888 0.891 0.893 0.891 0.889 i 1

-222.99 0.768 0.770 0.772 0.770 0.766 t

'224.31- 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 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 I.- 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 0.211 233.57 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 0.0828- 0.0817 0.0846 0.0835 0.0762-

,: 240.19 0.0671 0.0660 0.0689 0.0679 0.0608 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 l . NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius me.noine io 6-17 gv. ..

o

' TABLE 6-4

~

RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > O.1 MeV)-

'WITHIN THE PRESSURE VESSEL WALL e

Radius *l (cm) 0' 15' 25' 35' 45' 220.27( ) 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' O.927 0.907 0.901 225.63- 0.874 0.865 0.874 0.850 0.842 226.95 0.818 0.808 0.818 0.792 0.782 228.28- 0.761 0.750 0.716 0.734- 0.721 229.60 0.705 0.693 0.704 0.677 0.662 ,

230.92~ 0.649 0.637 0.649' O.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 0.465 0.443 L236.22 LO.436_ 0.428 0.440l 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) 0.233 0.226 0.237 0.223 0.186 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius -

m o.m m a io 6-18

b , ~ .

i

((' p 1 b  :;c

\>

., . . . TABLE 6 i L

RELATIVE: RADIAL OlSTRIBUTIONS OF IRON DISPLACEMENT RATE-(dpa) j f* WITHIN THE-PRESSURE VESSEL WALL ,

q.

.e, Radius-t (cm)' 0'- 15*- 25' 35' 45' 220.27-(1) 1.00 1.00- 1.00 1.00 1.00 .

220.64 0.984 0.981 0.984 0.983 0.984 0.921 0.915 221.66 0.912: 0.909- 0.917 222.99 0.815 0.812. 0.826 0.833 0.821 224.31 0.722 0.719 0.737 0.747 0.730 225.63 0.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572

. ~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 ;7 233.57 0.300 0.297 0.322. 0.334 0.305 234.89 0.263 0.261 0.285 0.295 0.266 236.22 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 0.183 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 L 242.17(2) 0.116 0.113 0.128 0.134 0.106 l-l . . .-

T NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius i

I 2..o.noi... io 6-19 a .

I?.I W'

TABLE 6 i NUCLEAR PARAMETERS FOR NEUTRON-FLUX MONITORS Reaction Target Fission- l Monitor of Weight Response Product -Yield Material Interest Fraction Range Half-Life- (%)

Copper Cu63(n a)Co60 0.6917 E'> 4.7 MeV 5.272 yrs Iron' Fe54(n,p)Mn54 0.0582 E > 1.0 MeV 312.2 days Nickel NiS8(n,p)CoS8- 0.6830 E > 1.0 MeV 70.90 days Uranium-238* ' U238(n,f)Cs137 1.0 .E > 0.4 MeV 30.12 yrs 5.99-Neptunium-237* Np237(n,f)Cs137 1.0 E > 0.08 MeV 30.12 yrs 6.50 '

Cobalt-Aluminum

• CoS9(n,y)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobalt-Aluminum CoS9(n,r)Co60 0.0015- E-> 0.015 MeV' 5.272 yrs

.L

• Denotes that monitor is cadmium shielded.

e i

nowioia, io -

6-20 l l

l

.6___-----___________-__ i

a e

TABLE 6-7 ,

1 IRRADIATION HISTORY OF NEUTRON SENSORS-CONTAINED IN CAPSULE-U  ;

("[

I Irradiation P P g

Irradiation Decay-3 ,

Period' (MWt ) Pp ,f, Time (days) Time'(days)- ,

2/87 842 .247 18 809 3/87 1902 .558 31 778 4/87 1731 .508 30 748 5/87- 2553 .749 31 717 6/87 1299 .381 30 687 7/87 2161 .634 31- 656 8/87 854 .250, 31 625

~*

. 9/87 2440 .715 30 ~595 10/87 2684 .787 31 '564 C- 11/87 2411 .707 30 534

~

12/87 468 .137 31 503 1/88- 2424 .711 31 472 2/88- -

2537 .744 29 -443 3/88 2521 .739 31 412 4/88 3077 .902 30 382 'i

. 5/88 2897 .849 31 351 6/88 2701 .792 30 321 7/88. 2745 .805 31 290 8/88 2280- .668 31 259 9/88 1911 .560 30 229

.10/88 1633 .479 31 198 y- 11/88 1805 .529 30 168 12/88 1296 .380 31 137 1/89 931 .273 7 130 NOTE: Reference Power = 3411 MW t

l 2990s/101989 to g.g}

L. 3

E v ,

TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES

~

,J Measured Saturated Reaction -

Monitor and Activity Activity- Rate ,

Axial Location (dis /see gm) *(dis /see gm). (RPS/ NUCLEUS) $

Cu-63-(n,a)Co-60 4

Top 5.36 x 10 4.18 x 10 5 Middle 5.02 x 10 4

3.92 x 10 5 Average 5.19 x 10 4 4.05 x 10 5 6.18 x 10'17 l Fe-54(n.p)Mn-54 Top 1.43 x 10 6 3.99 x 10 6 6

3.71 x 10 6 Middle 1.33 x 10 '

Bottom 1.29 x 10 6 3.60 x 10 6 Average- 1.35 x 10 6 3.77 x 10 6 6.01 x 10 -15 Ni-58(n,p)Co-58 Top- 8.64 x 10 6 5.79 x 10 7 l Middle- 7.73 x 10 6 5.18 x 10 7 Bottom 7.80 x 10 6 5,23 x 10 7 Average 8.06 x 10 6 5.40 x 10 7 7.71 x 10 -15

-U-238(n,f)Cs-137(Cd)

Middle 1.45 x 10 5 3.65 x 10 6 3.70 x 10 -14 l

l 2e90s/101e83 to - g_g; 1

L

TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd

p 19.

'? ,

Measured- Saturated Reaction Monitor and Activity Activity Rate

. Axial location (dis /sec om) (dis /sec-gm)' (RPS/ NUCLEUS).

Np-237(n,f) Cs-137-(Cd).

Middle l'.34 x 10 6 5.21 x 10 7 3.15 x 10 -

Co-59 (n,r) Co-60 ,

7 7 Top 1.11 x 10 8.66 x 10 1.13 x 10 7 8.81 x 10 7

, Middle Bottom 1.12 x 10 7 8.74 x 10 7 1.12 x 10 7 5.70 x 10 -12

- . Average 8.74 x 10 7

~Co-59-(n,r) Co-60 (Cd)

L Top 5.63 x 10 6 4.39 x 10 7 Middle 5.88 x 10 6 4.59 x 10 7 Bottom 5.82 x 10 6 4.54 x 10 7 Average 5.78 x 10 6 4.51 x 10 7 2.94 x 10 -12 f.-. .

o a ouioi ., io 6-23 I

'w ,

o.

.gih l '

- TABLE'6-9 .

SUMMARY

OF NEUTRON DOSIMETRY RESULTS~

TIME AVERAGED EXPOSURE RATES .

2 9 (E> 1.0_MeV) (n/cm -sec)- l'09 x 10 11 -- 8%

2

1. (E> 0.1 MeV) (n/cm -sec) 4.64 x 10 11 1 15%

2.05 x 10 -10 11g dpa/see .

10

,(E< 0.414 eV) (n/cm2 -sec). 4.44 x 10 30%

INTEGRATED CAPSULE EXPOSURE-2 3.96 x 10 18

+-(E>1.0MeV)-(n/cm) 8%- ,.

2

+ (E> 0.1 MeV) (n/cm ) 1.69 x 10 19 1 15% /

dpa 7.45 x-10 -3 11g -

+ (E< 0.414 eV) (n/cm )

2 1.61 x 10 18 30%

NOTE: Total Irradiation Time = 1.15 EFPY 9

7 e

m e.no m eio 6-24

g ,

4 tt>

r p .

TABLE 6-10 L

s COMPARISON OF MEASURED AND FERRET CALCULATED'

?i REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER-  ;

0 -!

.. l

'*- ' ~

Adjusfed -

fi R6 action- Measured Calculation C/M

~17

~

1Cu-63(n,a)Co-60 6.18x10'II 6.25x10 1.011 -

" -15 -15 Fe-54(n.p)Mn-54 6.01x10 5.93x10 0.99 N1-58(n,p)Co-58

~

-15 -15 7.71x10 7.84x10 1.02 .,

-14 ~14 U-238(n,f)Cs-137(Cd) 3.70x10 3.39x10 0.92

-13 -13

.Np-237'(n.f) Cs-137 (Cd) 3.15x10 -

3.29x10 1.04

-12 -12

.00-59 (n,r) Co-60 (Cd) 2.94x10 2.94x10 1.00

-12 -12

-Co-59(n,r)Co-60 5.70x10 5.70x10 1.00 .

t D

9 I.

u.a.noi . '"

6-25

• ......_....._._x_,

0 p .

Y TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER -

9 4

Energy AdjusgedFlux Energy AdjusgedFlux .

Group (Nev) (n/cm -sec) Group -(Mev) -(n/cm -sec)  !

' 1 6 ~3 10 1 1.73x10 8.75x10 28 9.12x10 2.12x10 1 7 ~3 10 i 2 1.49x10 1.98x10 29 5.53x10 2.76x10 1 7 -3 9 3 1.35x10 - 7.67x10 30 3.36x10 8.64x10 8 -3 4 1.16x10 1

1.71x10 31 2.84x10 8.28x10 9 1 8 -3 9 5 1.00x10 3.77x10 32 2.40x10 8.01x10 0 8 ~3 10 6 8.61x10 6.42x10 33 2.04x10 2.27x10 0 9 -3 10 7 7.41x10 1.47x10 34 1.23x10 2.10x10 0 9 -4 10 8 6.07x10 2.08x10 35 7.49x10 1.96x10 0 9 -4 10

.9 4.97x10 4.37x10 36 4.54x10 1.88x10 0 9 -4 10 10 3.68x10 5.77x10 37 2.75x10 2.03x10 0 10 ~4 10 11 2.87x10 1.21x10 38 1.67x10 2.24x10 ,

0 10 -4 10 12 2.23x10 1.67x10 39 1.01x10 2.18x10 0

2.33x10 19 6.14x10

-5 2.15x10 10 f.

p 13 1.74x10 40 0 10 -5 10 14 1.35x10 2.57x10 41 3.73x10 2.08x10 10 -5 10 -

15 ' 1.'11x100 4.67x10 42- 2.26x10 1.99x10

-1 10 1.37x10

-5 10 16 8.21x10 5.29x10 43 1.91x10

~1 10 8.32x10

-6 1.80x10 10 17 6.39x10 5.46x10 44

~1 10

.5.04x10

-6 10.

18 4.98x10 3.94x10 45 1.63x10

~1 10 -6 10 19 3.88x10 5.52x10 46 3.06x10 1.51x10 3.02x10

~1 5.67x10 10 47 1.86x10

-6 1.37x10 10 20 1.13x10 -6

-1 10 10 21 1.83x10 5.61x10 48 1.00x10

~1 10 -7 10 22 1.11x10 4.48x10 49 6.83x10 1.01x10

-2 10 ~7 10 23 6.74x10 3.12x10 50 4.14x10 1.09x10 .

-2 10 -7 9 24 4.09x10 1.77x10 51 2.51x10 9.21x10

-2 10 1.52x10

~7 9 25 -- 2.55x10 2.33x10 52 7.90x10 -

~

-2 1.15x10 10 9.24x10

-8 10 26 1.99x10 53 1.63x10

-2 10 27 1.50x10 1.47x10 NOTE: Tabulated energy levels represent the upper energy of each group.

me.noiseno 6-26

w J, ,

4 9 N t P

TABLE 6-12 COMPARISDN OF CALCULATED AND MEASURED .

'" << EXPOSURE LEVELS FOR CAPSULE V.

r

i 1

. Calculated Measured- C/M 2 18 18 f(E>~1.0MeV)(n/cm) 3.32 x 10 3.96 x 10 0.97 .-

6(E>0.1'MeV)-(n/cm) 2 1.49 x 10 19 1.69 x 10 19 1.01-dpa 6.51'x 10 ~3 7.45 x 10'3 1.01 2 1,58 x 10 18 1.61.x 10 18 f(E< 0.414 lev)-(n/cm ) 1.12 n'

14 9 >

.'p.

-S l .

m o.noin .io 6-27

F y '

({

i ' - <

qu

. TABLE 6-13' 3 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS - .

j ON THE PRESSURE ~ VESSEL CLAD / BASE METAL INTERFACE FOR BYRON UNIT 0 2 AZIMUTHAL ANGLE i

(

O'- 15' 25.(a) 35' 45' '

1~15 EFPY

-6(E> 1.0-MeV) ' 5.49'x 10 17 8.57 x 10 17 9.90 x 10 17 8.24 x 10 17 9.61 x 10 17 2

(n/cm )

4(E>0.1MeV)-- 1.08 x 10 18 1.71 x 10 18 2.56 x 10 18 2.21 x 10 18 2.29 x 10 18 2

-(n/cm ) j l

dpa. 8.24 x 10'4 1.28 x 10'3 1.60 x 10-3 1.34 x 10-3 1.47 x 10 -3 -

-16.0 EFPY  :

+(i> 1.0 MeV) 8.89 x 10 18 1.33 x 10 19 1.'51 x 10 19 1.23 x 10 19 1,41 x 10 19 2

(n/cm )- ,,

~

19 19 19 19 19 6(E> 0.1 MeV) 1.84 x 10 2.80 x 10 4.11 x 10 3.48 x 10 3.53 x 10 2

(n/cm )

i 2.25 x 10 -2'

-2 -2 -2 -2 dpa: 1.38 x 10 2.06 x.10 2.52 x 10 2.08 x 10 32.0 EFPY 6(E> 1.0 MeV) 1.79 x 10 19 2.68 x 1019- 3.03 x_1019 2.47 x 10 19 2.83 x 10 18 2 .

(n/cm )

4(E> 0.1 MeV) 3.71 x 10 19 5.62 x 10 19 8.26 x 10 19 7.00 x 10 19 7.08 x 10 19 I 2

(n/cm ) ,

dpa 2.78 x 10 -2 4.14 x 10 -2 5.07 x 10 -2 4.17 x 10 -2 4.51 x 10 -2 (a) Maximum point on the pressure vessel 6-28

m TABLE S-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES 16 EFPY NEUTRON FLUENCE (E> 1.0 MeV) SLOPE dpa SLOPE 2

.(n/cm ) .(equivalent n/cm2 ) ,

Surface 1/4 T 3/4 T Surface _ 1/4 T 3/4 T 0* 8.89 x 10 18 4.83 x 10 I9 1.03 x 10 18 8.89 x 10 18 5.61 x 10 18 1.95 = 10 I8 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 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 25*(a) 18 1.23 x 10 19 8.15 x 10 18 35* 1.23 x 10 19 6.69 x 10 18 1.43 x 10 18 3.05 x 10 45* 1.41 x 10 19 7.54 x 10 18 1.52 x 10 18 1.41 x 10 19 9.02 x 10 18 3.09 x 10 18 32 EFPY NEUTRON FLUENCE (E> 1.0 MeV) SLOPE dpa SLOPE (n/cm ) (equivalent n/cm )

Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 0* 1.79 x 10 19 9.72 x 10 18 2.07 x-10 18 1.79 x 10 19 1.13 x 10 19 3.92 x 10 10 15* 2.68 x 10 19 1.45 x 10 19 3.05 x 10 18 2.68 x 10 19 1.69 x 10 I9 5.81 x 10 18 25*(a) 3.03 x 10 19 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 19 1.35 x 10 19 2.86 x 10 18 2.47 x 10 19 1.64 x 10 19 6.12 x 10 18 45* 2.83 x 10 19 1.52 x 10 19 3.06 x 10 18 2.83 x 10 19 1.81 x'10 19 6.20 x 10 18 (a) Maximum point on the pressure vessel 3990s/10t989 IO

- . - - . _. -....-____m ,. mm._ ._- -

{l '

i.

p 1

TABLE 6-15 t

I UPDATED LEAD FACTORS FOR BYRON UNIT 2 SURVEILLANCE CAPSULES

! be 't r,

Capsule Lead Factor V 4.00(8) it X 4.02 W 4.02 2 4.02

/!

V 3.75 .

Y 3.75 ,

(a) Plant specific evaluation ,

l

)

e l.

l l

l-noo.noi n io 6-30 1

l J

,7 ,

4 m

SECTION 7.0  ;

. SURVEILLANCE-CAPSULE REMOVAL SCHEDULE-I. The fo11'owing removal schedule meets ASTM E185-82 and is. recommended for 4 - future capsules to be removed from the Byron Unit 2 reactor vessel:

Capsule Estimated ,

Location Lead Fluence 2

Capsule. (deg.) Factor Removal Time (a) (n/cm )

V' ~58.5 4.00 1.15(Removed) 3.96 x 10 18 X 238.5 4.02 4.5 1.71 x 1019(b)

V 61 3.75 8.5 3.02 x 1019(c) j Y- 241 3.75 15 5.33 x 10 19 1 W 121.5 4.02 Standby -

f Z- 301.5~ 4.02 -

Standby -

1

. i 4

(a) Effective full power years from plant startup.

(b) Approximate fluence at 1/4 thickness reactor vessel wall at end of life.

(c) . Approximate fluence at reactor vessel inner wall at end'of life. '

i

.i b

- t 4

-w i

3..o.noi nio 71 o

1 L. ._ _

L

p. +

J7rg m 3 - ~-

, ps.

y ,

a, .

.Y +

SECTION8.0; REFEREFCE5 n

2.i 1. L. R'. Singer, "Comacnwealth Edison Company Byron Station Unit No. 2,, -

ReactorVesselRadiationSurveillanceProgram,"WCAP-10398, December 1983.=.

7r e i '

Codo of federal Regulations,10CFR50, Appendix G, " Fracture' Toughness -

2.-

i Requirements", and Appendix H, " Reactor Vessel Material Sunaillance q- $

Program Requirements,." U.S. Nuclear kegulatory Comission, Washington,

'D. C. >

Regulatory Guide 1.99, Proposed Revioken 2, " Radiation Damage to Reactor

{ -3.

<4 Vessel Materials", U.S. Nuclear Repulatory Commission, February, 1986.

t Nh j

4. R. G; Soltesz, R.; K. Disney, J. Modhrh, h d S. L. Ziegler, " Nuclear Rocket Shielding Met @ ads,9edificatich, IIpdeting and Input Data 2

Preparation. 'Vol. 5 -Two-Dimensional Disc' rete Ordinates Transport j I Technique",;WANL-PR(LL)-034, Vol . 5, Aurp st 1970. '

.t

-5. "0RNL RSCI Data Library Collection DLC-76 Ski' LOR Coupled Self-Shie'lded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section. Library for Light Water Reactors".

64 J.V. Alexander,et.al.,'"CorePhysicsParametersandPlantOperations-Data for the' Byron Generating Station Unit 2 Cycle 1", WCAP-11136, June 1986. (Proprietary) i

7. ASTM Designation E482-82, " Standard Guide for Application of Neutron [

Transport Methods for Reactor Vessel Surveillance", in ASTM Standards,

,- Section 12, American Society for Testing and Materials, PhilcJeiphia, PA, 1984.

.. -l a,o,noisse io g1 l

i

_-._______________._._______m. _-____._r

7 4

i

8. ASTM Designation EE60-77, " Standard Recommended Practice for h Extrapolating Reactor vessel Surveillance Desimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, j Philadelphia, PA, 1984. ,;

9. ASTM Designation E603-79, " Standard Practice for Characterizing Neutron ,

Exposures in Ferritic Steels in Terms of Displacements per Atom (d p )", '(

i in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984,

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

11. ASTM Designation E853-84, " Standard Practice for Analysis and ,

Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, ,;'

Philadelphia, PA, 1984.

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

13. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standarcs, Section 12, American Society for Testing and M6terials, Philadelphia, PA, 1984.

14. ASTM Designation E263-82, "Staridard Method for DetermTning Fast-Neutron .

Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 304. .

15. ASTM Designation E264-82, " Standard Mathod for Determining Fast-Neutron Flux' Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984, a.wicius,o 8-2 ,

h,,

i p

j 16. 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.

17. ASTM Designation E523-82, " Standard Method for Determining fast-Neutron l Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materic1s, Philadelphia, PA, 1984. *

18. 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. ,

i

19. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutror ,

Flux Density by Radioactivation of Neptunium-237", in ASIM Standards, i Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

6

20. ASTM Designation E1005-84, " Standard Method for Application and Analysis I

of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, Aaerican Society for Testing and Materials,  !

Philadelphia, PA, 1984.

21. F. A. Schmittroth, FERRET Data Analysis Core _, HEDL-THE 79-40, Hanford Engineering Developent Laboratory, Richland, WA, September 1979.

22. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spettra Determined by Foil Activation, AFWL-TR-67-41, Vol. 1-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967. ,

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

mw,mse no g.3 e