ML20138P355

Non-proprietary Analysis of 300 Deg Capsule from Carolina Power & Light Co Brunswick Unit 2 Reactor Vessel Radiation Surveillance Program

ML20138P355
Person / Time
Site:	Brunswick
Issue date:	11/30/1996
From:	Shaun Anderson, Burgos B, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20138P360	List: ML20138P355
References
WCAP-14774, NUDOCS 9703040246
Download: ML20138P355 (112)

Text

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POWER AND LIGHT LCOMPANY BRUNSWICK UNIT 2-REACTOR VESSEL l RADIATION SURVEILLANCE PROGRAM Westinghouse Energy Systems t'

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WESTINGHOUSE NON-PROPRIETARV CLASS 3 WCAP-14774 ANALYSIS OF THE 300 DEG CAPSULE FROM THE CAROLINA POWER AND LIGHT COMPANY

~

y BRUNSWICK UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM E. Terek B. N. Burgos S. L Anderson J. F. Williams J. D. Perock November 19C 7' Woric Performed Under Shop Order HCAP-106 Prepared by the Westinghouse Electric Corporation for the Carolina Power and Light Company r

Approved:

N M fort

o. E. s.

D. E. Boyle, Manager Reactor Equipment & Materials Engineering WESTINGHOUSE ELECTRIC CORPORATION Systems and Major Projects Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355

@ 1997 Westinghouse Electric Corporation All Rights Reserved

l TABLE OF CONTENTS SECTION TITLE PAGE f

1.0

SUMMARY

OF RESULTS 1

2.0 INTRODUCTION

3

3.0 BACKGROUND

4

4.0 DESCRIPTION

OF PROGRAM 6

10 TESTING OF SPECIMENS FROM THE 300 DEG CAPSULE 14 5.1 Overview 14 5.2 Charpy V-Notch Impact Test Results 16 5.3 Tensile Test Results 17 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 47 6.1 Introduction 47 6.2 Discrete Ordinates Analysis 48 6.3 Neutron Dosimetry 50 6.4 Projections of Pressure Vessel Exposure 54

7.0 REFERENCES

72 APPENDIX A - LOAD-TIME RECORDS FOR CHARPY IMPACT TESTS 75 r

APPENDIX B - TEST EQUIPMENT CALIBRATION RECORDS AND MODEL NUMBERS 89 APPENDIX C - EVALUATION OF CHARPY SPECIMEN 3EK 95 Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

ii LIST OF TABLES TABLE TITLE PAGE i

4-1 Unirradiated Chemical Content of the Brunswick Unit 2 Reactor 8

i Vessel Behline Materials 4-2 Chemical Composition of Four Irradiated Weld Meta! 2pecimens 9

Removed From the Brunswick Unit 2 300 Deg Capsule 4-3 Chemistry Results from the Low Alloy Steel NBS362 Certified Reference 10 Standard 5-1 Charpy V-notch impact Data for the Brunswick Unit 2 Lower Intermediate 19 i

Shell Plate 301 Irradiated to a Fluence of 4.06 x 10" n/cm (E > 1.0 MeV) 2 5-2 Charpy V-notch Impact Data for the Brunswick Unit 2 Surveillance Weld 19 Metal Irradiated to a Fluence of 4.% x 10" n/cm (E > 1.0 MeV) 2 5-3 Charpy V-notch Impact Data for the Brunswick Unit 2 HAZ Metal 20 Irradiated to a Fluence of 4.06 x 10" n/cm (E > 1.0 MeV) 2 5-4 Instrumented Charpy Impact Test Results for the Brunswick Unit 2 Lower 21 Intermediate Shell Plate 301 Irradiated to a Fluence of 4.06 x 10" n/cm2 (E > 1.0 MeV) l 5-5 Instrumented Charpy Impact Test Results for the Brunswick Unit 2 22 Surveillance Weld Metal Irradiated to a Fluence of 4.06 x 10" n/cm2 (E > 1.0 MeV) 5-6 Instrumented Charpy Impact Test Results for the Brunswick Unit 2 23 Reactor Vessel HAZ Metal Irradiated to a Fluence of'4.06 x 10" n/cm2 (E > 1.0 MeV)

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

lii LIST OF TABLES (Cont'd)

TABLE TITLE PAGE f

5-7 Measured Charpy Test Re sults for the Brunswick Unit 2 300 Deg.

24 Surveillance Materials 3

5-8 Tensile Properties of the Brunswick Unit 2 Reactor Vessel Surveillance 25 2

Materials Irradiated to a Fluence of 4.06 x 10" n/cm (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Rates at the Reactor Core Midplane 58 6-2 Relative Radial Distribution of Exposure Parameters within the 59 Pressure Vessel Wall 6-3 Relative Axial Distribution of Exposure Parameters 60 6-4 Fuel Cycle Specific Calculated Neutron Flux (E > 1.0 MeV) at the 61 Surveillance Capsule Center 6-5 Nuclear Parameters Used in the Evaluation of Neutron Sensors 62 6-6 Monthly Thermal Generation During the First Eleven Cycles of the 63 Bnmswick Unit 2 Reactor 6-7 Measured Sensor Activities and Derived Reaction Rates 66 6-8 Summary of Neutron Dosimetry Results for the Surveillance Capsule 67 3

6-9 Comparison of Measured and Ferret Calculated Reaction Rates at the 67 Surveillance Capsule Center 6-10 Adjusted Neutron Energy Spectmm at the Center of the Surveillance 68 Capsule AntJysis of Brunswick Unit 2 300 Deg. Capsule November 1996

iv i

LIST OF TABLES (Cont'd)

TABLE TITLE PAGE 6-11 Comparison of Calculated and Measured Neutron Exposure Levels for the 69 Surveillance Capsule l

t 6-12 Neutron Exposure Projections at the Inner Radius of the Reactor Pressure 70 Vessel Lower Intermediate Shell 6-13 Neutron Expose Projections at the Inner Radius of the Reactor Pressure 71 Vessel Circumferential Weld and Pressure Vessel Lower Shell I

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

v LIST OF FIGURF-S FIGURE TITLE PAGE e

4-1 Surveillance Capsule Holder from Brunswick Unit 2 11

'8 4-2 Surveillance Capsules Recovered from Brunswick Unit 2 12 4-3 Diagram Showing the Location of Specimens and Dosimeters within the 13 300 Deg. Capsule 5-1 Charpy V-notch Impact Energy vs. Temperature for the Brunswick Unit 2 26 Lower Intermediate Shell Plate 301 (Longitudinal Orientation) 5-2 Charpy V notch Lateral Expansion vs. Temperature for the Brunswick Unit 2 27 Lower Intermediate Shell Plate 301 (Longitudinal Orientation) 5-3 Charpy V-notch Percent Shear vs. Temperature for the Brunswick Unit 2 28 Lower Intermediate Shell Plate 301 (Longitudinal Orientation) 5-4 Charpy V-notch Impact Energy vs. Temperature for the Brunswick Unit 2 29 Surveillance Weld Metal 5-5 Charpy V-notch Lateral Expansion vs. Temperature for the Bmnswick Unit 2 30 Surveillance Weld Metal 5-6 Charpy V-notch Percent Shear vs. Temperature for the Brunswick Unit 2 31 Surveillance Weld Metal 5-7 Charpy V-notch Impact Energy vs. Temperature for the Brunswick Unit 2 32 Surveillance HAZ Metal 5-8 Charpy V-notch Lateral Expansion v.. Temperature for the Brunswick Unit 2 33 Surveillance HAZ Metal

- Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

vi LIST OF FIGURES (Cont'd)

FIGURE TITLE RGM 5-9 Charpy V-notch Percent Shear vs. Temperature for the Brunswick Unit 2 34 r

Surveillance HAZ Metal 5-10 Charpy Impact Specimen Fracture Surfaces of the Brunswick Unit 2 35 Reactor Vessel Lower Intermediate Shell Plate 301 (Longitudinal Orientation) 5-11 Charpy Impact Specimen Fracture Surfaces of the Brunswick Unit 2 36 Reactor Vessel Surveillance Weld Metal 5-12 Charpy Impact Specimen Fracture Surfaces of the Brunswick Unit 2 37 Reactor Vessel HAZ Metal 5-13 Tensile Properties for the Brunswick Unit 2 Reactor Vessel Lower 38 l

Intermediate Shell Plate 301 (Longitudinal Orientation) 5-14 Tensile Properties for the Brunswick Unit 2 Surveillance Weld Metal 39 5-15 Tensile Properties for the Brunswick Unit 2 Surveillance Program 40 HAZ Material 5-16 Fractured Tensile Specimens from Brunswick Unit 2 Reactor Vessel 41 Lower Intermediate Shell Plate 301 (longitudinal orientation) 5-17 Fractured Tensile Specimens from Brunswick Unit 2 Surveillance Weld Metal 42 5-18 Fractured Tensile Specimens from Brunswick Unit 2 Surveillance HAZ 43 Material Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

vii LIST OF FIGURES (Cont'd)

PAGE FIGURE TITLE 44 Engineering Stress-Strain Curves for Lower Intermediate Shell Plate 301 5 19 Tensile Specimens 5JE and 51D (longitudinal orientation) 45 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens 5JM 5-20 and 5K2 46 Engineering Stress-Strain Curves for HAZ Metal Tensile Specimens 5KD 5-21 and SKM 56 6-1 Brunswick Unit 2 R,0 Reactor Geometry 57 6-2 Surveillance Capsule Geometry 1

November 1996 Analysis of Brunswick Unit 2 300 Deg. Capsule

1 SECTION 1.0

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in the 300 Deg capsule, the first capsule to be removed from the Brunswick Unit 2 reactor pressure vessel, led to the following conclusions:

The capsule received an average fast neutron fluence (E > 1.0 MeV) of 4.06 x 10" n/cm after 2

10.9 Effective Full Power Years (EFPY) of plant operation.

Irradiation of the reactor vessel lower intermediate shell plate 301 (Ht. # C4489-1) Charpy specimens, oriented with the long axis of the specimen parallel to the major rolling direction (longitudinal orientation), to 4.06 x 10" n/cm (E > 1.0 MeV) resulted in a measured irradiated 2

30 ft-lb transition temperature of 29.4'F and a measured irradiated 50 ft-lb transition temperature of 73.2 F.

Irradiation of the surveillance weld metal Charpy specimens to 4.06 x 10" n/cm (E > 1.0 2

MeV) resulted in a measured irradiated 30 ft-lb transition temperature of 102.0 F and a measured irradiated 50 ft-lb transition temperature of 200.3 F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 4.06 x 10" n/cm2 (E > 1.0 MeV) resulted in a measured irradiated 30 ft-lb transition temperature of 26.3 F and a measured irradiated 50 ft-lb transition temperature of 108.5 F.

1 The measured upper shelf energy of lower intermediate shell plate 301 (Ht. # C4489-1)

(longitudinal orientation) resulted in a measured irradiated average upper shelf energy value of 113.7 ft-Ib after irradiation to 4.M x 10" n/cm (E > 1.0 MeV).

2 The measured upper shelf energy of the surveillance weld metal Charpy specimens resulted in a measured irradiated average upper shelf energy value of 67.0 ft-lb after irradiation to 4.06 x 2

0" n/cm (E > 1.0 MeV).

The measured upper shelf energy of the weld HAZ metal Charpy specimens resulted in a measured irradiated average upper shelf energy value of 66.52 ft-lb after irradiation to 4.06 x 10" n/cm (E > 1.0 MeV).

2 Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

2 The results of the tensile tests performed on the intermediate lower shell plate 301 (Ht. #

C4489-1) (longitudinal orientation) indicated that irradiation to 4.06 x 10" n/cm (E > 1.0 2

MeV) resulted in a measured 0.2 percent offset yield strength of 77.1 ksi at 100 F and 68.8 ksi at 550 F and a measured ultimate tensile strength of 97.8 ksi at 100 and 550 F.

The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 4.06 x 10" n/cm (E > 1.0 MeV) resulted in measured 0.2 percent offset yield 2

strengths of 82.5 ksi and 71.8 ksi at 125*F and 550 F, respectively, and measured ultimate tensile strengths of 95.7 ksi and 90.7 ksi at 125 F and 550 F, respectively.

The results of the tensile tests performed on the HAZ metal indicated that irradiation to 4.06 x i

10" n/cm (E > 1.0 MeV) resulted in measured 0.2 percent offset yield strengths of 87.1 ksi 2

and 71.3 ksi at 72 F and 550 F, respectively, and measured ultimate tensile strengths of 100.8 ksi and 90.7 ksi at 72 F and 550 F, respectively.

The calculated end-of-life (32 EFPY) maximum inner radius neutron fluence (E > 1.0 MeV) for the Brunswick Unit 2 reactor vessel lower intermediate shell plates 301 and 351 is:

Vessel inner radius

= 1.56 x 10'8 n/cm 2

28 2

Vessel 1/4 thickness = 1.117 x 10 n/cm l

The calculated end-of-life (32 EFPY) maximum inner radius neutron fluence (E > 1.0 MeV) for

=

the Brunswick Unit 2 reactor vessel lower shell plates 201 and 251 and circumferential weld is-Vessel inner radius

= 1.26 x 10'8 n/cm 2

Vessel 1/4 thickness = 9.024 x 10" n/cm2 l

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l 3

SECTION

2.0 INTRODUCTION

This repon presents the results of the examination of the 30u Deg capsule, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron e

irndiation on the Brunswick Unit 2 reactor pressure vessel materials under actual operating conditions.

3 A description of the surveillance program and the pre-irradiation mechanical propenies of the reactor vessel materials is presented in a report by General Electric entitled "Bmnswick Steam Electric Plant, Information on the Reactor Vessel Material Surveillance Program"I". The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" Westinghouse personnel were contracted to perform the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens at the Remote Metallographic Facility at the Westinghouse Science and Technology Center.

This repon summarizes the testing of and the post-irradiation data obtained from the 300 Deg surveillance capsule removed from the Bmnswick Unit 2 reactor vessel and discusses the analysis of the data.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

4 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 irriportant factor in ensuring safety in the nuclear industry. The beltline 5

region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class 1 (base material of the Brunswick Unit 2 reactor pressure vessel) are well documented in industry related literature.

Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code. The method uses fracture mechanics concepts and is m

based on the reference nil-ductility temperature (RTsm).

RTum is defined as the greater of either the drop weight nil-ductility transition temperature (NDTF per ASTM E208'*) 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 plate. The RT m of a given material is used to index that material to a reference s

stress intensity factor curve (K ai curve) which appears in Appendix G to the ASME Code. The K a i

curve is a lower bound of dynamic, crack anest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K ai curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

RTum and, in tum, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor vessel surveillance program in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTsm) due to irradiation is added to the initial RTym to adjust the RTsm for radiation embrittlement. This adjusted reference temperature (ART = IRTum + ARTsm + Margin) is l

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

5 used to index the material to the Km curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

3

- Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

6 SECTION

4.0 DESCRIPTION

OF PROGRAM Three surveillance capsules for monitoring the effects of neutron exposure on the Brunswick Unit 2 reactor pressure vessel core region materials were insened in the reactor vessel prior to initial plant start-up. The three capsules were positioned in the reactor vessel between the jet pumps and the vessel wall at the 30,120, and 300* azimuthal locations. The venical center of the capsules is opposite the venical center of the core.

The 300 Deg capsule was removed after 10.9 Effective Full Power Years (EFPY) of plant operation.

The capsule contained specimens made from lower intermediate shell plate 301 (Ht. # C4489-1). A search of CB&I files revealed that the weld was fabricated in a manner similar to that of the longitudinal welds using plate 301 (Heat Number C4489-1) and the post weld heat treatment of the welded piece was at 1150 F (+25/-50*F) for a total of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. No additional information was found discussing the fabricatien of the weld.

All plate test specimens were machined from the 1/4 thickness location of the plate after performing a simulated postweld stress-relieving treatment. Specimens were machined from weld metal and the heat-affected-zone (HAZ) metal of a stress-relieved weldment from plate C4489-1.

Charpy V-notch impact specimens from lower intermediate shell plete 301 (Ht. # C4489-1) were macl0 net in the longitudinal orientation (longitudinal axis of specimen parallel to major working direction). The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of the Charpy was perpendicular to the weld direction and parallel to the plate s ' ce; the notch was machined to be perpendicular to the plate surface. The HAZ material impact Charpy specimens were machined with the long dimension perpendicular to the length of the weld and parallel to the top surface of the plate; the axis of the notch was perpendicular to the original plate surface.

The chemical composition of the beltline materials are presented in Table 4-1. The chemistry reponed in Table 4-1 was obtained from unirradiated material for the plate and weld.m In addition, chemical analyses were performd on four irradiated Charpy specimens to identify the surveillance weld chemistry and are presented in Table 4-2. The chemistry results from the NBS cenified reference standards are reported in Table 4-3.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

7 All plate and test specimens were heated to 1650*F 25 F, held I hr/in. minimum and water quenched for approximately 23 minutes. They were then tempered at 1220 F z 10*F, held 30 min /in.

minimum and air cooled. Test specimens were stress relieved by heating with a rate of 72 F/hr to 1150 F +25/-50 F, held for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, furnace cooled with a rate of 90*F/hr to 600 F and air cooled.

The post-weld heat treatment of the weld specimens was at 1150 F +25/-50 F for a total of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

S The 300 Deg capsule contained dosimeter wires of pure copper, nickel, and iron. These were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters [$ (E > 1.0 MeV), $ (E > 0.1 MeV), and dpa/sec].

The locations of the test specimens and dosimeters within the capsule is shown in Figure 4-3.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

a

~

8 Table 4-1 Materials Unirradiated Chemical Content of the Brunswick Unit 2 Re l

I

=

Cu (%)I'l Ni (%)t'l n-Material Description l

0.15 0.54 Lower Shell Plate (Heat # C4500-2) l 0.11 0.60 Lower Shell Plate (Heat # C4550-2) 0.12 0.60 Lower Intermediate Shell Plate (Heat # C4489-1) 0.12 0.57 Lower Intermediate Shell Plate (Heat # C4521-2) 0.16 0.82 Nozzle N16A (Heat # Q2Q1VW) 0.16 0.82 Nozzle N16B (Heat # Q2Q1VW) t 0.05 0.96 Axial Welds G1 & G2 *!

(Heat # S3986) t 0.05 0.96 Axial Welds F1 & F2 *!

(Heat # S3986) 0.02 0.90 Circumferential Weld (Heat # 3P4000)

"i Cu and Ni chemistry values as reported in Generic Letter 92-01 Subm f b i ted A search of the GE and CB&I files indicated that (a]

heat

[5]

il l sis performed on number could be conclusively identified. In addition, the chem ca ana yh the d surveillance weld metal specimens 3BK,3C5 and 3D2 does not matc chemistries of the beltline longitudinal welds.

November 1996 Analysis of Brunswick Unit 2 300 Deg. Capsule

9 Chemical Composition of Four Irradiated Weld Metal Specimens R Table 4-2 300 Deg Capsule emoved from the Brunswick Unit 2 7

Chemical Analyses (wt%) for Weld Metal Specimens Element

-y 1

3BK*

• t 3EK*

3C5**

Al 3D2**.

i

<0.02 0.023

<0.02 As

<0 2

<0.02

<0.02 g

<0.02 B

<0.02

<0.004 Co

<0.004

<0 or4

<0.004 0.010 0.011 0.011 Cr 0.011 0.048 0.062 0.056 Cu 0.050 0.18 0.16 0.19 Mn 0.18 1.32 1.30 1.54 Mo 1.39 I

0.46 0.54 0.53 Ni 0.48 0.84 0.65 0.90 P

0.88 0.014 0.012 0.015 Si 0.014 0.391 0.342 0.317 Sn 0.339

<0.01

<0.01

<0.01 Ti

<0.01

<0.002

<0.002

<0.002 V

<0.002

<0.004

<0.004

<0.004 Za

<0.004

<0.01

<0.01

<0.01 C

<0.01 0.071 0.243 0.096 S

0.070 0.015 0.012 0.016 0.016 placed in a location within the capsule that was desi specimen. However, it was evaluation (Appendix C of this repon) was performed to determi metal specimen. An weld m:tal specimen. This evaluation lead to the conclusion th t hine if this sp material make-up in the notch region that can not be explained and th s specimen has an odd

/,

at appear questionable.

e chemical analysis results

• • A search of the GE and CB&I files indicated that the surv ill weld wire similar to that of the longitudinal welds in the vee ance weld was fabricated be conclusively identified. In addition, the chemical analysisssel, however, no h longitudinal welds. metal specimens 3BK,3C5 and 3D2 does not match the d e c emistries of the bekline Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996-

10 Table 4-3 Chemistry Results from the Low Alloy Steel NBS362 Certified Reference Standard 3

ELEMENT Certified Measured (wt %)

(wt %)

l Al 0.09 0.07 As 0.09 0.09 B

0.003

<0.004 Co 0.30 0.30 Cr 0.30 0.29 Cu 0.50 0.48 Mn 1.04 1.01 Mo 0.068 0.%2 Ni 0.59 0.56 P

0.04 0.03 Si 0.39 0.403 Sn 0.016 0.017 Ti 0.08 0.02 V

0.040 0.037 Zr 0.19 0.21 Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

11 i

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jf Figure 4-1. Surveillance Capsule Holder from Brunswick Unit 2 Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

i 3

12 I,

1

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Figure 4-2. Surveillance Capsules Recovered from Brunswick Unit 2 1

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 i

4 13 i

k i

i a

CHARPY V-NOTCH SPECIMENS EC l

I I

2 S

2 l

117c4654G10 117C465409

n eo ln l

eo l l

$o &

0*

Y*

E' eo 3

(HAZ) pyELD) 3EK 3J3 p(AZ) l M

3c3 3DL M

(WELD) 3AU

~

3DT ggAZ)

M 3AM 337 pgAq 37 m

333 ggq 375 g

33D ptA4 36Y (BASE) 3h otA4 36K (SABE)

I (WELD) 3c5 3sc (BASE) h g

(WELD) 3E h8 58 I8

~

3An g

@ WELD) sce 37A (SASE)

M 3D2 3sg g

U U

m m

I""

r""

C C

X X

E 2

3 m

m

' V)

V5 Figure 4-3. Diagram Showing the Location of Specimens and Dosimeters within the 300 Deg. Capsule Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l l

14

)

l I

SECTION 5.0

?

l TESTING OF SPECIMENS FROM THE 300 DEG CAPSULE l

i 5.1 Overview f

The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at i

the Remote Metallographic Facility at the Westinghouse Science and Technology Center (STC).

W l

Testing was performed in accordance with 10 CFR Part 50, Appendix H, ASTM Specification E185-82*, and Westinghouse Procedure MHL 8402, Revision 2, as modified by Westinghouse RMF Procedures 8102, Revision 1 and 8103, Revision 1.

Upon receipt of the capsale at the hot cell laboratory, the specimens and spacer blocks were crirefelly removed, inspected for identification number, and checked against the master list as supplied 'ay the Carolina Power and Light Company. No discrepancies were found. Westinghouse STC used parts of RMF Procedure 8804 in opening the capsule. This procedure pertains to Westinghouse surveillance capsules, however, the pertinent parts of the procedure are to maintain permanent record and specimen and capsule identification. These aspects of the procedure were applied to the GE designed capsule.

m and RMF Procedure The Charpy impact tests were performed per ASTM Specification E23-93a 8103, Revision 1, on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy machine is instrumented with a GRC 830-I instrumentation system, feeding into an IBM compatible 486 computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (Eo). From the load-time curve (Appendix A), the load of general yielding (Pay), time to general yielding (tcy), maximum load (Pu), and time to maximum load (tu) 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 (P,),

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

4 The energy at maximum load (Eu) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E,) is the difference between the total energy to fracture (Eo) and the energy at maximum load (Eu).

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

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

15 oy = P * { L / [ B * ( W - a )2

• C ] }

(1) ay where L is 2e distance between the specimen supports in the impact testing machine; B is the width of the specimen measured parallel to the notch; W is the height of the specimen, measured perpendicularly to the notch; and a is the notch depth. The constant C is dependent on the notch flank angle ($), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending).

In three-point bending a Charpy specimen in which & = 45 and p = 0.010 inches, Equation 1 is valid with C = 1.21. Therefore (for L = 4W),

cy = Pay * { L / [ B * ( W - a )2

• 1.21 ] } = [ 3.31 Pay W ] / [ B ( W - a )2 ]

(2)

For the Charpy specimens, B is 0.394 inP1, W is 0.394 in.'", and a is 0.079 in.l" Equation 2 then reduces to:

cy = 33.3

• Pay (3) where cy is in units of psi and Pay is in units of Ib. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in S

compliance with ASTM Specification A370-92 l. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests were performed on a 20,000-pound Instron Model 1115, split-console test machine, per ASTM Specification E8-93t'l and E21-92" 1, and RMF Procedure 8102, Revision 1. The upper pull rod of the test machine was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Extension measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through j

specimen failure. The extensometer gage length is 1.00 inch. De extensometer is rated as Class B-2 per ASTM E83-93"U Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

t 16 Elevated test temperatures were obtained with a three-zone electric resistance split-tube furn nine-inch hot zone. All tests were conducted in air.

Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the fol procedure was used to monitor specimen temperature. Chromel-alumel thermoco ff shallow holes in the center of each end of the gage section of a dummy specimen and in e the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upp l l and lower grip and controller temperatures was developed over the range of room temperature to 550 F (288'C). The upper grip was used to control the fumace temperature. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experime 2 F.

indicate that this method is accurate to The yield load, ultimate load, fracture load, total elongation, and uniform elongation we directly from the load-extension curve. The yield strength, ultimate strength, and fracture s were calculated using the original cross-sectional area. _he final diameter and final gage length w determined from post-fracture photographs. The fracture area used to calculate the fracture st stress at fracture) and percent reduction in area was computed using the final diameter measurem 5.2 Charoy V-Notch Impact Test Results The results of Charpy V-notch impact tests performed on the various materials contained in t Deg Capsule are presented in Tables 5-1 through 5-7. Charpy curve-fits were develope CVGRAPHn21 program, and the fitted results are shown in Figures 5-1 through 5-9. The fracture appearance of each irradiated Charpy specimen from the various materials is shown through 5-12. The fractures show an increasingly ductile or tougher appearance with increa temperature. Load-time records for the individual instrumented Charpy specimens are c Appendix A.

Irradiation of the reactor vessel lower intermediate shell plate 301 (Ht. # C4489-1) Charpy specim oriented with the long axis of the specimen parallel to the major rolling direction (longitudinal

~

orientation), to 4.06 x 10" n/cm (E > 1.0 MeV) resulted in a measured irradiated 30 ft-Ib transitio 2

temperature of 29.4*F and a measured irradiated 50 ft-lb transition temperature of 73.2 F.

November 1996 Analysis of Brunswick Unit 2 300 Deg. Capsule

---

~

17 Irradiation of the surveillance weld metal Charpy specimens to 4.06 x 10" n l

2 1

I resulted in a measured irradiated 30 ft-lb transition temperature of 102.0 F an 50 ft-lb transition temperature of 200.3 F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens 2

3 1.0 MeV) resulted in a measured irradiated 30 ft-lb transition temperature of 26 3 irradiated 50 ft-lb transition temperature of 108.5 F.

sured

~

The measured upper shelf energy of lower intermediate shell plate 301 (H orientation) resulted in a measured irradiated average upper shelf energy value nal irradiation to 4.% x 10" n/cm (E > 1.0 MeV).

2 er The measured upper shelf energy of the surveillance weld metal Charpy specim measured irradiated average upper shelf energy value of 67.0 ft-lb after irradiation t 2

(E > 1.0 MeV).

cm The measured upper shelf energy of the weld HAZ metal Charpy specimens r irradiated average upper shelf energy value of 66.5 ft-Ib after irradiation to 4 06 x 10 re 2

MeV).

n cm (E > 1.0 Charpy specimen 3EK test results are suspect :ad were not used in this ev this report).

n x C of 5.3 Tensile Test Results The results of tensile tests performed on the Intermediate Lower Shell Plate 301 metal and the weld HAZ material are shown in Table 5-8. Fractured tensile spe nce weld materials tested are shown in Figures 5-16 through 5-18. Engineering stress-strain e

~~

tensile specimens are shown in Figures 5-19 through 5-21. All work was pe f e

r ormed according to the approved procedures and requirements of Purchase Specification FDSD-SRPLO-25 October 22,1991 with exception of the capsule opening procedure as discussed previousl i i

y n this report.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

I 18 The results of the tensile tests performed on the intermediate lower shell plate 301 (Ht. # C4489-1) 2 i

(longitudinal orientation) indicated that irradiation to 4.06 x 10" n/cm (E > 1.0 MeV) resulted in a measured 0.2 percent offset yield strength of 77.1 ksi at 100*F and 68.8 ksi at 550 F and a measured t

i ultimate tensile strength of 97.8 ksi at 100 and 550 F.

The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 4.06 x 10" n/cm (E > 1.0 MeV) resulted in measured 0.2 percent offset yield strengths of 82.5 ksi 2

and 71.8 ksi at 125 F and 550 F, respectively, and measured ultimate tensile strengths of 95.7 ksi and 90.7 ksi at 125 F and 550*F, respectively.

The results of the tensile tests performed on the HAZ metal indicated that irradiation to 4.06 x 10" n/cm (E > 1.0 MeV) resulted in measured 0.2 percent offset yield strengths of 87.1 ksi and 71.3 ksi at 2

72 F and 550 F, respectively, and measured ultimate tensile strengths of 100.8 ksi and 90.7 ksi at j

72*F and 550 F, respectively.

l t

.=

I Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

19 TABLE 5-1 Charpy V-notch Impact Data for the Brunswick Unit 2 Lower Intermediate Shell Phte 301 Irradiated to a Fluence of 4.06 x 10" n/cm (E > 1.0 MeV) 2 (Longitudinal Orientation) s Sample s

Temperature Impact Energy Lateral Expansion Shear Number F

l C

ft-lbs l Joules mils j mm

1. /c 37A

-30

-34 13 18 10 0.25 10 36K 25

-4 27 36 19 0.48 15 375 50 10 40 54 28 0.71 25 3AB 72 22 54 73 34 0.86 30 36C 150 66 72 97 54 1.37 50 37B 200 93 113 153 74 1.88 100 36Y 250 121 115 155 81 2.06 100 36J 300 149 113 153 85 2.16 100 TABLE 5-2 Charpy V-notch Impact Data for the Brunswick Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 4.06 x 10" n/cm (E > 1.0 MeV) 2 Sample Temperature Impact Energy Lateral Expansion Shear Number F

l C

ft lbs l Joules mils l mm

• /o 3BK

-30

-34 6

8 2

0.05 0

3EK*

25

-4 23 31 16 0.41 10 3C5 72 22 32 44 24 0.61 30 3D2 150 66 37 51 34 0.86 35 3AM 200 93 32 43 28 0.71 20 3AU 208 98 56 75 47 1.19 75 3CE 250 121 71 96 64 1.63 100 3CP 300 149 63 86 60 1.52 100

~~

• Not used in evaluation (See Appendix C of this report)

{

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

20 TABLE 5-3 Charpy V-notch Impact Data for the Brunswick Unit 2 HAZ Metal t

2 Inadiated to a Fluence of 4.% x 10" n/cm (E > 1.0 MeV)

Sample Temperature

!_mnact Energy Lateral Expansion Shear j

Number F

l C

ft-Ibs Joules mils mm

)

3J7

-30

-34 11 15 8

0.20 20

)

3J2 0

-18 18 25 15 0.38 20 3DT 25

-4 43 59 34 0.86 70 l

3DK 50 10 34 46 24 0.61 40 3JD 72 22 44 60 32 0.81 40 f

3E3 150 66 51 69 60 1.52 70 3JA 200 93 58 78 52 1.32 95 3DL 275 135 71 97 64 1.63 100 i

F l '.'-

l i

l kWysis of Brunswick Unit 2 300 Deg. Capsule November 1996 I

21 TABLE 5-4 Instrumented Charpy Impact Test Results for the Brunswick Unit 2 Lower Intennediate Shell Plate 301 Irradiated to a Fluence of 4.06 x 10" n/cm (E > 1.0 MeV) 2 (Longitudinal Olientation)

Sample Test Temp Charpy L.

laed Energies (It-Ibhn')

Yseid Load Time to Max. Imad Time to Fracture Arrest Imad Yield Stress Flow Stress Number Energy Charpy Max.

Prop.

Yield Max.

Imad

(*F)

(ft-Ib)

QA E,,/A FgA (Ib)

(msec)

(ib)

(msec)

(!b)

(Ib)

(ksi)

(ksi) 37A

-30 13.44 105 72 36 3604 0.14 3526 0.22 3743 205 120 123 36K 15 26.57 216 167 49 3433 0.16 4233 0.41 4174 515 l14 127 375 50 39.54 3I5 22t>

92 3410 0.14 4477 0.51 4449 535 113 131 3AB 72 53.95 434 307 127 3341 0.14 4485 0.67 4450 1741 lit 130 36C 150 71.65 577 290 257 3142 0.14 4311 0 67 4220 2230 104 124 37B 200 112.73 905 255 620 3029 0.14 4295 0.67 101 122 36Y 250 114.54 922 279 643 2596 0.14 4091 0.67 96 lib 36J 300 112.7 907 266 642 2715 0.13 3975 0.66 90 til I

• Fully ductile fracture, no brittle fracture load and no arrest load.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

[e 1

.g.'

a..

N TABLE 5-5 Instrumented Charpy Impact Test Results for the Brunswick Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 4.06 x 10" n/cm (E > 1.0 MeV) 2 Sampic Test Temp Oterpy Normalized Emergscs (R-lhhn')

Yseld Load Time to Man. Imad Time to Fractwe Artest Imed Yseld Stress Flow Stress Number Energy Charpy Max.

Prop.

Yield Max.

Imad

(*F)

'(R-b)

E/A E,/A F/A (b)

(tmec)

(b)

(msec)

(b)

(b)

(ksi)

(ksi)

JBK

-30 5.51 47 25 22 3335 0.13 JJJa 0.13 JJJ5 0

111 Ill JEK" 25 22.53 151 135 44 3545 0.15 4004 0.35 3521 243 115 125 3C5 72 32.11 259 159 69 3476.

0.16 4073 0 46 4069 1101 115 125 3DZ 150 37.32 301 153 115 3202 0.14 3535 0.47 3755 1465 IOb Il7 3AM 200 31.53 254 374 50 Juus

0. I 4 3655 0.47

.3664 550 100 Ill -

JAU 205

. 55.56 447 203 245 3106 0.15 J913 0.51 354l 2743 103 117 3CE 250 70.53 570 202 365 3014 0.54 3533 0.52 100 144 3CP 300 63.31 510 194 3t6 2920 0.14 3706 0.51 97 110

• Fully ductile fractum, no brittle fracture load and no armst load.

• • Not used in evaluation (See Appendix C of this report)

[

I Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 t

i

_ _,,. _. _ t

23 TABLE 5-6 Instrumented Charpy Impact Test Results for the Brunswick Unit 2 Reactor Vessel HAZ Metal Irradiated to a Fluenca of 4.06 x 10" n/cm (E > 1.0 MeV) 2 Sampic Test Temp Charpy Normalized Energies (ft-lb/in')

Yield Load Time to Max. Lead Time to Fracture Arrest Iml Ykid Stress How Stress Number I M gy Charpy Max.

Prop.

Yield Max.

Imad

(*F)

(R-lb)

E/A EJA PgA (Ib)

(msec)

(Ib)

(msec)

(Ib)

(ib)

(ksi)

(ksi) 3J7

-30 11.13 90 44 46 3834 0.14 3984 0.16 3894 544 127 130 3J2 0

18.2 147 94 52 3685 0.15 3949 0.26 3949 665 122 127 3DT 25 43.47 350 121 229 3596 0.14 4027 0.33 4027 3209 119 127 3DK 50 34.14 275 166 109 3525 0.14 4096 0.41 4095 1255 117 127 JJD 72 44.49 358 286 142 3433 0.14 4206 0.51 4206 2635 114 127 3E3 150 55.1 411 205 207 3'27 0.16 4017 0.51 3802 1770 104 119 3JA 200 57.85 466 201 265 3152 0.14 3863 0.55 2053 1384 105 117 3DL 275 71.49 576 198 378 2931

0. I4 3830 0.51 97 112

• Fully ductile fracture, no brittle fracture load and no arrest load.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

,s

24 i

TABLE 5-7 Measured Charpy Test Results for the Brunswick Unit 2 300 Deg. Surveillance Materials i

Matenal Measured 30 ft-lb Measured 50 ft-lb Measured 35 mil Measured Temperature")

Temperature")

Lateral Expansion Upper Shelf Temperature")

Energy.23 o

Lower Inter Shell Plate 301 29.4 F 73.2*F 75.4 F 113.7 ft-Ib (Longitudinal)

(73.9 ft-lb)*

Surveillance Program Weld 102.0 F 200.3"F 165.8 F 67.0 ft-lb Metal Weld HAZ Metal 26.3*F 108.5 F 68.7 F 66.5 ft-lb NOTES:

1. The information provided in this table is not available for the unitradiated surveillance program materials.

2. The measured irradiated upper shelf energy values presented in this table are the average of the measured ft-lb values from specimens that exhibited at least 95% shear.

3. The number in parentheses is 65% of the measured longitudinal USE value.

l.

I e

l f

(

l Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

25 i

l 1

+

TABLE 5-8 I

l Tensile Propetties of the Brunswick Unit 2 Reactor Vessel Surveillance Materials

(

Irradiated to a Fluence of 4.06 x 10'? n/cm (E > l.0 MeV)

[

2 i

i L

Sampic Test Temp 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reducten i

Material Number (T)

Strength Strength Load Stress Streagth Elongation Elongation in Area

[

(ksi)

(ksi)

(kip)

(ksi)

(ksi)

(%)

(%)

(%)

l Base i

L Intermediate SJE 100 77.1 97.8 3.15 204.6 64.2 9.0 233 69 l

Lower Shell 5JD 550 68.8 97.8 3.05 178.5 62.1 10.5 24 3 65 Plate 301 r

{

Weld 5JM 125 82.5 95.7 3.40 155.0 69.3 11 3 22.6 55 Metal SK2 550 71.8 90.7 3.60 133.9 733 8.7 17.0 45 i

HAZ 5KU 72 87.1 100.8 3.36 156.9 68.3 10.1 23.4 56 Metal 5KM 550 71 3 90.7 3.60 130.4 73 3 7.0 12.2 44 f

a k

i i

i P

t i

Analysis of Brunsuck Unit 2 300 Deg. Capsule gg g

.s

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

.- m.

-.~

. ~.

26 Brunswick Unit 2 Surv. Plate O' GRAPH 43 Hyperbolic Tangent Curve Printed at 113909 on 11-04-1996 Page!

Coefficients of Curve 1 A = 57J0 B = 55.73 C = 10735 10 = 81L59 l

~

Equation is 03 = A + B ' [ tanh((T - 10)/C) ]

Upper Shelf Energy: t!3f4 Fixed Temp. at 30 ft-lbs 29.4 Temp at 50 ft-lha 732 lower Shelf Energy 2.19 Fixed i

Material: PLATE SA533B1 Heat Numben C445-1 Orientation: LT Capsule 300DEG Total Fluence 4.06e+17 300 m asu

,Q I

a i

x 200 h

I tto a

250 c) c Gr.1 "A '

100 Z

o O

ej

\\

/

J u

l 1

l

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant BW2 Cap: 300DEG Lienah P1JLTE MM Ori: LT But f. C445-1 Charpy V-Notch Data Temperature input CVN Energy Computed 03 Energy Differential

-30 13

~ 1322

-22 25 27 283

-13 50 40 3a71 i28 e

72 54 49JB 4fl i

15 0 72 86.73

-14.73 200 113 10123 IL76 1

250 115 10fL4 659 300 113 11133 L46 SUM of PJSIDUAIS = 9.46 Figure 5-1. Charpy V-notch Impact Energy vs. Temperature for the Brunswick Unit 2 Lower Intennediate Shell Plate 301 (Longitudinal Orientation)

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

27 Brunswick Unit 2 Surv. Plate CVGRAPH 43 Hyperbohe Tangent Curve Printed at It4919 on 11-04-1996 Page1 Coefficients of Curve !

& = 405 B = 395 C = 107:15 1D = 90.46 Equation is I.F. = A + B ' ( tanb((T - 1D)/C) l Upper Shelf 1.E:110 Fixed Temperature at 1.F. 3fx 7E4 lower Shelf 1.E: 1 Fixed Material: PLATE SA533B1 Heat Number: C4489-1 Orientation: LT Capsule 300DID Total Fluence: 4D6e+17 20u rn

=

150 c.

x 100 o

h or a

su e

2 U

-300

-200

-100 0

100 200 300 400 600 600 Temperature in Degrees F Data Set (s) Plottai Plant: BW2 Cap: 300DBG Matenah PIATE SA533Bt Ori: LT Heat f. C4489-1 Charpy V-Notch Data Temperature input Iteral Erpannon Computal 1.F.

Differential

-30 10 857 L42 25 19 19D1

-D1 50 25 26 21 L72 72 34 3177 22 150 54 e0.4

-64 2[D 74 70.91 328 250 81 1 15 434 300

[6 M43 656 SL1 of RESIDUAIS = IL45 Figure 5-2. Charpy V-notch Lateral Expansion vs. Temperature for the Brunswick Unit 2 Lower Intermediate Shell Plate 301 (Longitudinal Orientation)

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

3 Brunswick Unit 2 Surv. Plate j

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 115&33 on 11-04-1996 Page1 Coefficients of Cune 1 A = 50 B=50 C = 99.15

'm = 116.01 Equation is Shear /. = A + B ' [ tanh((T - 1D)/C) l Temperature at 50x Shear:

116 Materiah PLATE SA533B1 Heat Numbec C4489-1 Orientation: LT Capsule 300DE Total Fluence 4D6e+17 100 2

b-

)

u c

e C*

eu a

C o

eO D

4 o

O J

U

-300

-200

-100 0

100 200 300 400 500 000 Temperature in Degrees F Data Set (s) Ntted Plant: BW2 Cap 300DE Matenat PLATE SA533B1 Ori: LT Heat f. C4489-1 Charpy V-Notch Data Temperature.

Input Pertent Shear Computed Percent Shear Ddferential

-30 10 499 5

25 15 1175 124 50 25 20 5 4 11 72 30 2915 34 150 50 81L49

-16.4 9 i

200 10 0 8447 1552 250 IJO 9171 E2B 3 10 100 IrlEl 22

~.

i SUM of REDUAIS = 189 1

Figure 5-3. Charpy V-notch Percent Shear vs. Temperature for the Brunswick Unit 2 Lower Intermediate Shell Plate 301 (Longitudinal Orientation)

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

29 Brunswick Unit 2 Surv. Weld CVCPJLPH 4.1 Hyperbolic Tangent Curve Printed at 07:49:30 on 12-20-1996 Page!

Coefficients of Curve 1 A 3459 B = 314 C = 148f18 1V = 123.33

~

Equation is CVN : A + B

• l tanh((T - 1D)/C) }

lipper Shelf Energ 67 Fixed Temp. at 30 ft-lbs 102 Temp. at 50 ft-Ibs 2003 Lower Shelf Energy: 219 Fixed Material: WELD Heat Number:

Orientation:

Capsule 300DE Total Fluence 41EE+17 300

<n asu

.c T

aw 20u N

t:W L.

150 cua r.c 100 a

U f

o o

i

/.

U i

t i

.i i

-300

-200

-100 0

10 0 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Bf2 Cap: 300DE Matenal: TELD Ori Heat f.

Charpy V-Notch Data Temperature Input O'N Energy Computed O'N Energy Differential

-30 6

,951

-351 72 32 2334 ILl5 150 37 4033

-333 200 32 4934

-1734 208 56 5126 4.73 250 71 57 1199 300 63 61.47 12 SUM of EESIDt'AIS = 3S1 Figure 5-4. Charpy V-notch Impact Energy vs. Temperature for the Brunswick Unit 2 Surveillance Weld Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l l

30 i

i l

Brunswick Unit 2 Surv. Weld i

I

(

CYCRAPH 4j Hyperbolic Tangent Curve Printed at 076155 on 12-20-1996 I

Page1 Coefficients of Curve 1

~

A = 47.79 B = (L79 C = 185.99 70 = 200.91 l

Equation is II : A + B ' I tanh((T - TD)/C) )

Upper Shelf 1.E: 842 Temperature at II 35: 165,8 lower Shelf II: 1 Fixed Matenal: TELD Heat Number Orientation:

Capsule 300D10 Total Fluence 4217 200

<n

=

15 0 a

X W

100

.c b

,_a so y

l o

a

/

\\

U j

i j

l 1

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Bf2 Cap: 300DEC Materiat YELD Ori:

Heatf Charpy V-Notch Data Temperature input lateral Expansion Computed 11 Diffenntial

-30 2

7.44

-5.44 72 24 17.71 63

^.

150 34 3132 237 200 28 4258

-14 2 208 47 44 2 2S1 l50 64 5357 1042 300 60 6 3.16

-3.16 SUM of PISIDUAIS :-L49 l

r I

i i

i l

Figure 5-5. Charpy V-notch Lateral Expansion vs. Temperature for the Brunswick Unit 2 Surveillance Weld Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l l

l 31

\\

Burnswick Unit 2 Surv. Weld i

CVGRAPfl (J Hyperbolic Tangent Curve Printed at 11 fib 23 on 11-04-1996 Page1 Coefficients of Curve 2 A = 50 B = 50 C = 10258 3 = 18P.22 l

~- !

Equation is Shearx = A + B ' l tanh((T - M)/C) 1

[

l Temperature at 50z Shear: 1822 j

Materiah WED lient Number:

Orientation:

Capsule 300 DIE Total Fluence 4a6E+17 1N

[

/

~

\\

u ce e

.C rn ou r

>co i

Oy 40 O

f

,0 w

U

-300

-200

-100 0

100 200 300 400 500 000 Temperature in Degrees F Data Set (s) Plotted Plant: Bf2 Cap 300 DIE Matanah FEla SA533B1 Ori:

llent f.

Charpy V-Notch Data Temperature input Percent Shear Computed Perant Shear Differential

-30 0

157

-157 25 10 4.45 554 72 30 la44 1955 150 35 34.75 21 200 20 5857

-3&57 l

2DB 75 823 1289 250 10 0 7&94 2LD5

. l i

300 100 9085 934.

i Sl'M of IESIDUAIS = 2ED6 i

I Figure 5-6. Charpy V-notch Percent Shear vs. Temperature for the Brunswick Unit 2 SurveiHance Weld Metal l

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

{

32 BrunswickUnit 2 HAZ Surv. Mat'l 0' GRAPH 4.1 Hyperbolic Tangent Curve Printed at it3909 on 1141-1996 Page1 Coefficients of Curve 3 A = 3436 B = 3P.16 C = l'222

11) = 431?

l Equation is CVN = A + B ' [ tanh((T - TO)/C) l Upper Shelf Energy 6652 Temp. at 30 ft-lbs 263 Temp at 50 ft-lbs 10&5 lower Shelf Energy: 22 Fixed Materiah HEAT AFFD ZONE SA53381 Heat Number.

Orientation:

Capsule 300 DEL Total Fluenz (D6E+17 30u 5

W l

g 20u him

$3 c

cc 100 Z

l e

o e,

Su e

A u

s

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Data Setis) Plotted Plant: BY2 Cap; 300DID Material HEAT AFFD ZONE SA533B1 Ori:

Heatf.

Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential

-30 11 1724

-624 0

18 2354

-554 25 43 29fe 1333 50 34 3G.15

-2.15 72 44 4L76 223 150 51 56E7

-557 200 58 6134

-334 275 71 65D6 5.93' SUM of PJSIDUAIS = -2J5 Figure 5-7. Charpy V-notch Impact Energy vs. Temperature for the Brunswick Unit 2 Surveillance HAZ Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

33 i

Brunswick Unit 2 HAZ Sury. Mat'l CVCPJLPli U ifyperbolic Tangent Curve Printed at it4219 on 11-04-1996 l

Page1 Coefficients of Curve 3 A = 3135 B = 30 2 C = 1072 W = $812 l

Equation is 2 = A + B 'I tanh((T - E)/C))

}-

Upper Shelf LE: 62.9 Temperature at 11. 35c 6&7 tower Shelf I.E: 1 Fixed Matenat HEAT AWD ZDNE SA533B1 Ileat Number:

Orientation:

Capsule 30001E Total Fluence: 4116E+17 200 rn

=

1s0 6

c.

M M

im es b0 a

W a

so r

+

i o

i i

-300

-200

- 10 0 0

100 200 300 400 500 600 Temperature in Degrees F Data Setis) Plotted Plant: BW2 Cap: 300 DIE Matenat HEAT AFD 20NE SA51381 Ori:

liest f.

Charpy V-Notch Data j

Temperature input Lateral Expansion Computed 2 Differential

-30 8

ItD3

-3D3 l

0 15 16 5

-15 25 34 2239 113 50 24 29El

-5S1 72 32 2Ul3

-333 15 0 EO 5144 655 200 52 5&79

-6.79.

l 35 64 6133 2J6 SUM of RISIDUAIS = -l i

Figure 5-8. Charpy V-notch Lateral Expansion vs. Temperature for the Brunswick Unit 2 Surveillance HAZ Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

34 Brunswick Unit 2 HAZ Surv. Mat'l CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1158:Il on 114:-1E Page!

Coefficients of Curve 3 A = 50 B = 50 C = 150.59 TO = 6L4 l

~

Equation is Shearz = A + B ' [ tanh((T - 70)/C) 1 Temperature at 50x Shear. 6L4 Wateriah HEAT AFD ZONE SA53::B1 Heat Number Orientatics:

Capsule 300DE Total Fluenm 4.06E+17 100 g

o

/

so h

4

+

.c m

acop g

40 g-CL.

/

/

/~

U i j

i

-300

-200

- 100 0

100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted i

Plant: BW2 Cap; 300DE Matenah HEAT AFFD ZONE SA533B1 Ori Beat f.

Charpy V-Notch Data Temperature Input Pen:ent Shear Computed Peent Shear Differential

-30 20 229

-29 0

20 30E7

-1047 25 70 3ttl4 3125 50 40 4622

-622 72 40 5351

-13 51 15 0 70 7tL43

-6 43 200 95 8tL3 BE9-275 10 0 94.46 553 SL'M of RESIDUAIS = 63 Figure 5-9. Charpy V-notch Percent Shear vs. Ternperature for the Brunswick Unit 2 Surveillance HAZ Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l 35 i

i 4

f 37A 36K 375 i

r f

i 3AB 6C 37B l

)

36Y 36J I

Figure 5-10.

Charpy Impact Specimen Fracture Surfaces of the Brunswick Unit 2 Reactor Vessel Lower Intermediate Shell Plate 301 (Longitudinal Orientation)

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

36

@j i 3BK 3EK 3C5 BBB 3D2 3AM 3AU 3CE 3CP Figure 5-11.

Charpy Impact Specimen Fracture Surfaces of the Brunswick Unit 2 Reactor Vessel Surveillance Weld Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

i i

37 i

l l

l L

l 337 3J2 3DT i

B

'j ' i h$I i

I l

I l

3DK 3JD 3E3 1

I i

4 L

BB l

3JA 3DL i

j 4

d

}

i Figure 5-12.

Charpy Impact Specinien Fracture Surfaces of the Brunswick Unit 2 Reactor Vessel HAZ Metal j

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

38 120 Ultimate Strength 100 < -

0 1

l l

80 <

.~-

M

\\

n i

c.

x l

0.2% Yield Stre.'gth g*

60<

,c 3

40 20 w

w w

w y

0 50 100 150 200 250 300 350 400 450 500 550 600 650 Temperature (Deg F) 80 Reduction in Area 70-60

_ 50 l

~j 40 2

n. 30<

Total Elongation 20<

10=

C

~

Uniform Elongation o

0 50 100 150 200 250 300 350 400 450 500 550 600 650 Temperature (Deg F)

Figure 5-13.

Tensile Properties for the Brunswick Unit 2 Reactor Vessel Lower Intermediate Shell Plate 301 (Longitudinal Orientation)

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

39 120-I 100<

Ultimate Strength 4

-h 80<

+

m a

g 0.2% Yield Strength 60< -

g c

g 40< -

20 O

0 50 100 150 200 250 3h0 3ho 400 450 Sbo Sh0 600 650 Temperature (Deg F) 80 i

70 i

i 60 Reduction in Area 50<

$ 40 <

E i

e Q.

30<

Total Elongation 20<

O 10<

~

~

Uniform Elongation 0

0 50 100 150 200 250 300 350 400 450 500 550 600 650 Temperature (Deg F)

Figure 5-14.

Tensile Propenies for the Brunswick Unit 2 Surveillarice Weld Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

- - - - - - - ~ _ _ _

40 120 Ultimate Strength

+

80 <

C A

O.2% Yield Strength 6

60 en c.

g u) 40 20<

0 0

50 100 150 200 250 300 350 400 450 500 550 600 650 Temperature (Deg F) 80 70<

60< -

50< -

j 40 <

1f c.,

30-W Total Elongation 20 <

--c 10-Uniform Elongation o

0 50 100 150 200 250 300 350 400 450 500 550 600 650 Temperature (Deg F)

Figure 5-15.

Tensile Properties for the Brunswick Unit 2 Surveillance Program HAZ Material 1

November 1996 Analysis of Brunswick Unit 2 300 Deg. Capsule

41 l

i d

1 i

1 1

]..

1 I..-

p ; ',, \\ ' \\'.

J i,a C395

.t 3

\\\\

1 3 4 1

4

.,1,

}

,., f f i <

v --

l i

i i

Specimen 5JE tested at 100oF I

i l

t i

1 l

i l

i

~

ff..

.~

o s

, r j

m' i

l w-4 f

,.,c, :, v -.

l leasi i

ia Specimen 5JD tested at 550oF i

l i.-

I i

a Figure 5-16. Fractured Tensile Specimens from Brunswick Unit 2 Reactor Vessel i

)

Lower Intermediate Shell Plate 301 (longitudinal orientation)

{

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 i

42 1,. -

kij a m i e w L. a n & W Specimen 5JM tested at 125aF gggy,g,ngstAvd6E%-

I Specimen 5K2 tested at 550"F Figure 5-17. Fractured Tensile Specimens from Bmnswick Unit 2 Surveillance Weld Metal Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l 43 4

4 I

i 1

1 i

+

pr.

l y

A q

z-x IJ i

.1 s,

_ y..

, qj i,,.

.i

$?

6 h

Specimen 5KD tested at 70oF I

y

, p 4

D (I e

I.. j e e p.'

1 Specimen 5KM tested at 550oF j

i i

Figure 5-18. Fractured Tensile Specimens from Brunswick Unit 2 Surveillance HAZ Material Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

44 STRESS-STRAIN CURVE BRUNSWICK UNIT 2 120.00 100.00-v>

80.00-VI 60.00-Ew 40.00-5JE 20.00-100 F 0.00 0.00 0.10 0.20 STRAIN, IN/IN STRESS-STRAIN CURVE BRUNSWICK UNIT 2 100.00 90.00-80.00-70.00-60.00-VI 50.00-mg 40.00-5JD 20.00-550 F 10.00-0.00 0.00 0.10 0.20 STRAIN, IN/IN Figure 5-19. Engineering Stress-Strain Curves for Lower Intermediate Shell Plate 301 Tensile Specimens 5JE and 5JD (Longitudinal Orientation)

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l 45 I

STRESS-STRAIN CURVE BRUNSWICK UNIT 2 100.00 90.00-80.00-

~

70.00-ct) 80.00-d 50.00-mg 40.00-30.00-i 5JM 20.00-l 10.00-0.00 0.00 0.10 0.20 STRAIN, IN/IN STRESS-STRAIN CURVE BRUNSWICK UNIT 2 100.00 90.00-80.00-70.00-(t) 90.00-6 50.00-eg 40.00-30.00-

~

SK2 20.00-550 F 10.00-0.00 0.00 0.04 0.06 0.12 0.16 STRAIN, IN/IN l

l Figure 5-20. Engineering Stress-Strain Curves for Weld Metal Tensile Specimens 5JM and SK2 1

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

46 STRESS-STRAIN CURVE BRUNSWICK UNIT 2 120.00 100.00-25 80.00-lu:

d 00.00-40.00-SKD 20.00-72 F 0.00 0.00 0.10 0.20 STRAIN Ik/IN i

STRESS-STRAIN CURVE BRUNSWICK UNIT 2 100.00 90.00-i 80.00-70.00-00.00-i 50.00-i eg 40.00-j 30.00-SKM 20.00-550 F 10.00 O.00 0.00 0.04 0.08 0.12 STRAIN, IN/IN Figure 5-21. Engineering Stress-Strain Curves for HAZ Metal Tensile Specimens SKD and SKM Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

47 SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction

-l Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral pan of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectmm, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the pressure 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 generally derived solely frorn analysis.

'Ihe use of fast neutron fluence (E > 1.0 MeV) to correlate measured material propeny changes to the neutron exposure of the material has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853D'1, " Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693"'1,

" 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 Regulatory Guide 1.99nsi,

" Radiation Embrittlement of Reactor Vessel Materials."

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

1 l

{

This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in the surveillance capsule withdrawn at the end of the lith fuel cycle (B212R1 outage). The dosimetry evaluation is based on current state-of-the-art methodology and nuclear data including neutron transport and dosimetry cross-section libraries derived from the ENDF/B-VI data base.

In the capsule dosimetry evaluation, fast neutron exposure parameters in terms of neutron fluence (E >

1.0 MeV), 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 wall.

Also, uncertainties associated with the derived exposure parameters at the surveillance capsules and with the projected exposure of the pressure vessel are provided.

6.2 Discrete Ordinates Analysis A plan view of the Brunswick Unit 2 reactor geometry at the core midplane is shown in Figure 6-1.

Since the reactor exhibits 1/8th core symmetry, only a 0-45 degree sector is depicted. In addition to the core, reactor internals, pressure vessel, and sacrificial shield, the geometry also included representations of the jet pumps located intemal to the pressure vessel. Development of the geometric model shown in Figure 6-1 made use of nominal design dimensions throughout. The jet pumps were modelled as homogeneous zones characteristic of the pump geometry opposite the axial midplane of the reactor core.

An elevation view of the surveillance capsule attached to the pressure vessel wall is shown in Figure 6-2. From a neutronic standpoint, the surveillance capsule itself is a significant structure. The presence of this material has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the shroud and the pressure vessel. In order to accurately determine the neutron environment at the dosimetry location, the capsule itself was also included in the analytical model.

In performing the fast neutron evaluations for the surveillance capsules and reactor vessel geometry a series of three transport calculations were carried out. The first, a two-dimensional forward R,0 calculation was used to provide distributions of neutron flux throughout the geometry depicted in Figures 6-1 and 6-2. The second computation, a two-dimensional analysis in R,Z geometry, was used to synthesize a three-dimensional distribution of the neutron environment throughout the pressure l

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

49 vessel geometry. The use of this R,O/R,Z synthesis approach allows the axial variation in void fraction characteristic of BWR cores to be accounted for in the overall analysis.

l The third computation, an R,0 evaluation in the adjoint mode, was employed to establish a set of importance functions relating the neutron flux at the center of the surveillance capsule to the neutron source distribution in the reactor core. These importance functions were then used with the individual fuel cycle source distributions to compute the neutron flux at the center of the surveillance capsule for each of the 1I operating fuel cycles. The cycle dependent values of neutron flux were, in turn, input to the dosimeter reaction rate calculations described in Section 6.3 All of the transport calculations for Brunswick Unit 2 were carried out using the DORT discrete ordinates code"0 and the BUGLE-93 cross-section library"R The BUGLE-93 library is a 47 group ENDF/B-VI based data set produced specifically for light water reactor applications. In the analyses, anisotropic scattering was treated with a P expansion of the cross-sections and the angular 3

discretization was modelled with an S, order of angular quadrature.

Core power distributions including fuel assembly burnups and relative axial distributions of both fuel burnup and core void fractions were provided by Carolina Power and Light Company"U for each of the eleven fuel cycles utilized at Brunswick Unit 2. For the current evaluations, these data sets were processed to provide a bumup weighted average of power distribution and core void fraction for use in the transport calculations.

4 4

Selected results from the neutron transport analyses are provided in Tables 6-1 and 6-2. In Table 6-1, the calculated exposure parameters [$(E > 1.0 MeV), $(E > 0.1 MeV), and dpal are given at the j

geometric center of the surveillance capsule as well as at several azimuthal locations along the inner radius of the pressure vessel. The tabulated data for both the capsule and vessel are representative of an axial elevation corresponding to the reactor core midplane.

Radial gradient information applicable to 4(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec is given in Table 6-2. The gradient data are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Table 6-2.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

50 Axial gradient information applicable to all three exposure parameters is provided in Table 6-3.

Again, the gradient data are presented on a relative basis indexed to the midplane of the reactor core; and may be used in a multiplicative fashion to translate the midplane calculations to any desired axial elevation.

The fuel cycle specific results of the adjoint transport analysis at the surveillance capsule location are provided in Table 6-4.

(-

6.3 Neutron Dositnetry The passive neutron sene included in the Brunswick Unit 2 surveillance program are listed in Table 6-5. Also given in Table 6-5 ve the primary nuclear reactions and associated nuclear constants that i

were used in the evaluation of tae neutron energy spectrum within the surveillance capsules and in the subsequent determination of tho various exposure parameters of interest [$(E > 1.0 MeV),

$(E > 0.1 MeV), dpa/sec].

l The use of passive monitors such as those listed in Table 6-5 does not yield a direct measure of the energy dependent neutron flux 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 fmm the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

-The measured specific activity of each monitor.

-The physical characteristics of each monitor.

-The operating history of the reactor.

-The energy response of each monitor.

~

-The neutron energy spectrum at the monitor location.

The specific activity of each of the neutron monitors was determined using established ASTM procedures"' *"mso21 Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the Brunswick Unit 2 reactor was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report," and the Nucleonics Week Data Sheets for the cycles 1 through 11 Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

51 i

\\

operating period. The irradiation history applicable to the exposure of the capsule is given in Table 6-6.

Having the measured specific activities, the physical characteristics of the sensors, and the operating

-f history of the reactor, reaction rates referenced to full power operation were determined from the following equation:

A R=

l C [1-e */] [e *']

N FY{

o j

rel

\\

where:

1.t= Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,,,(rps/ nucleus).

A= Measured specific activity (dps/gm).

No= Number of target element atoms per gram of sensor.

F= Weight fraction of the target isotope in the sensor material.

Y= Number of product atoms produced per reaction.

l P = Average core power level during irradiation period j (MW).

j P,,, = Maximum or reference power level of the reactor (MW).

C;= Calculated ratio of $(E > 1.0 MeV) during irradiation period j to the time weighted average 4(E

> 1.0 MeV) over the entire irradiation period.

A= Decay constant of the product isotope (1/sec).

t,= Length of irradiation period j (sec).

tf ecay time following irradiation period j (sec).

D J

and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [P,]/[P,,,) accounts for month by month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio C. which can be calculated for each fuel cycle using the adjoint transport technology 3

discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

52 For the irradiation history of the surveillance capsule, the flux level term in the reaction rate calculations was developed from the adjoint transport calculation results provided in Table 6-4.

j Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Table 6-7.

1 i

Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment codet331. The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as ir:put and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured l

reaction rate data. The " measured" exposure parameters along with the associated uncertainties were then obtained from the adjusted spectrum.

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 linearly related to the flux $ by some response matrix A:

ff*'*) = [ A * &}*)

q 8

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

)

l R=[og &,

g 5

relates a set of measured reaction rates R, to a single spectrum $, by the multigroup reaction cross-section o,,. The log-normal approach automatically accounts for the physical constraint of positive

~

I fluxes, even with large assigned uncertainties.

l In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were appmximated in a multi-group format consisting of 53 energy groups. The trial input spectmm was converted to the FERRET 53 group structure using the SAND-II code'2'3 This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

l i

l 53 l

l SPLINE interpolation procedure in regions where group boundaries do not coincide. ne 620 point spectrum was then re-collapsed into the group structure used in FERRET.

l t

j De sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry file *3, were also l.'

collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded to 620 groups, was emr'syed as a weighting function in the cross-section j

collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/B-VI data l

l files. These matrices included energy group to energy group uncertainty correlations for each of the l.

individual reactions. However, correlations between cross-sections for different sensor reactions were i

not included. The omission of this additional uncertainty information does not significantly impact the l

results of the adjustment.

)

i i

l Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectmm at the sensor set locations, the neutron spectrum input to the FERRET evaluation 1

l was taken from the center of the surveillance capsule modeled in the forward R,0 transport i

l calculation. While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-l sections were developed from the ENDF/B-VI data files, the covariance matrix for the input ' trial j

spectrum was constructed from the following relation:

]

M,1 = R,*, + R, R,1 P,1 where R, specifies an overall fractional normalization uncenainty (i.e., complete correlation) for the set of values. He fractional uncertainties R, specify additional random uncenainties for group g that are correlated with a correlation matrix given by:

P,i = (1-0] 6,i + 0 e *"

i where:

)

t i

i H=N i.

2 2y L

i l

l l

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 i

i 54 I

i i

i The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (0 specifies the strength of the i

latter term). The value of 5 is I when g = g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice implies that I

neighboring groups are strongly correlated when 0 is close to 1. Strong long range correlations (or

/

anti-correlations) were justified based on information presented by R. E. Maerker". Maerker's results am closely duplicated when y = 6.

't i

The uncertainties associated with the measured reaction. rates included both statistical (counting) and j

r systematic components. The systematic component of the overall uncertainty accounts for counter --

t efficiency, counter calibrations, irradiation history corrections, and corrections for competing reactions in the individual sensors.

5 I

Results of the FERRET evaluations of the surveillance capsule dosimetry are given in Table 6-8. He-l t

data summarized in this table include fast neutron exposure evaluations in terms of $(E > 1.0 McV),

$(E > 0.1 MeV), and dpa. The measured and FERRET adjusted reaction rates for each reaction are -

l given in Table 6-9. An examination of Table 6-9 shows that, in all cases, reaction rates calculated.

with the adjusted spectra match the measured reaction rates to better than 5% The adjusted spectra from the least squares evaluation are given in Table 6-10 in the FERRET 53 energy group structure.

i In Table 6-11, absolute comparisons of the measured and calculated fluence at the center of the-capsule are presented. The results for the evaluation indicate M/C ratios of 0.870,0.866, and 0.908

- for $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa, respectively.

l i

6.4 Proiections of Pressure Vessel Exoosure Neutron exposure projections at several azimuthal locations on the pressure vessel inner radius are given in Tables 6-12 and 6-13. Along with the current (end of Cycle 11 exposure), projections are also provided for exposure periods of 16 and 32 EFPY. The data in Table 6-12 are applicable to the j

maximum exposure locations on the vessel upper shell above core midplane. The data in Table 6-13

' *)

apply to the beltline circumferential weld located 2.2 feet below the midplane of the active reactor core. In computing these vessel exposures, the calculated values from Table 6-1 were scaled by the average M/C ratios observed from the evaluations of the capsule dosimetry. Projection for future operation were based on the assumption that the exposure rates characteristic of the average of fuel Cycles 1-11 would continue to be applicable throughout plant life.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

55 All of the capsules remaining in the Brunswick Unit 2 reactor are located azimuthally at 30 relative to the core cardinal axes. Thus, each of the capsules will experience the same lead factor relative to the maximum neutron exposure of the pressure vessel. Based on the evaluations provided in this report,.

I the lead factor associated with the remaining capsules is 0.76. That is, the fluence accrued by the surveillance capsules lags the maximum fluence experienced by the inner wall of the pressure vessel.

~.

i t

i I

i 1

i l

1 1

1

)

i

]

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

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

56 i

i i

1 i

l I

t l

l l

l

.i l

g

., *~

3 sa j

s Insulation y

Pressure Vessel Jet Pumps e

[

t Shroud

• s.

/

g l

O t.

e#.

• 1 Q,,

4 '*, 3 S

'\\

l

)

Core

./ s t

l

/

a e.

/, g a

/>,

3. e I,. f.

• s 0

50 100 150 200 250 350 400 Radius (cm) r Figure 6-1 Brunswick Unit 2 R,0 Reactor Geometry Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

57 h

C 108.75 in. to VESSEL CENTER com 0.88in.

r/

BASKET

/

3.5 in.

/

\\

_/> >.

+ 0.75in.

L Capsule e core

/

% TENSILE CAPSULE NPLUXWIRES VESSEL WALL j/

l b CHARPY CAPSULE

!.M-!

72 in.

////

I N

/[/

LOWER BRACKET s

L/

a Bottom of Vessel Figure 6-2. Surveillance Capsule Geometry Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

58 TALLE t;-l CALCULATED FAST NEUTRON EXPOSURE RATES AT THE REACTOR CORE MIDPLANE

$(E > 1.0 MeV)

$(E > 0.1 MeV) 2 2

LOCATION (n/ctn -sec)

(n/cm -sec) daa/sec CAPSULE CENTER 1.35e+09 2.32e+09 2.05e-12 VESSEL IR 0 Degrees 8.08e+08 1.48e+09 i.27e-12 15 Degrees 1.10e+09 2.0le+09 1.72e-12 30 Degrees 1.03e+09 2.04e+09 1.62e-12 45 Degrees 1.69e+09 3.10e+09 2.62e-12 VESSEL 1/4T 0 Degrees 5.99e+08 1.42e+09 9.71e-13 15 Degrees 8.17e+08 1.91e+09 1.31e-12 30 Degrees 7.51e+08 1.91e+09 1.23e-12 45 Degrees 1.25e+09 2.92e+09 1.99e-12 VESSEL 3/4T 0 Degrees 2.53e+08 9.34e+08 4.76e-13 15 Degrees 3.32e+08 1.19e+09 6.13e-13 30 Degrees 3.16e+08 1.23e+09 6.03e-13 i

45 Degrees 5.lle+08 1.79e+09 9.26e-13 Note: Calculated exposure rates are based on an average over 11 fuel cycles.

knalysis of Brunswick Unit 2 300 Deg. Capsule November 1996

59 i

}

TABLE 6-2 i

RELATIVE RADIAL DISTRIBUTION OF EXPOSURE PARAMETERS j

WITHIN THE PRESSURE VESSEL WALL

$(E > 1.0 MeV)

$(E > 0.1 MeV)

LOCATION (n/cm -sec)

(n/cm'-sec) doa/see i

2 l

5 VESSEL IR O Degrees 1.00 1.00 1.00 15 Degrees 1.00 1.00 1.00 i

30 Degrees 1.00 1.00 1.00 45 Degrees 1.00 1.00 1.00 l

l VESSEL 1/4T 0 Degrees 0.74 0.%

0.76 l

1 15 Degrees 0.74 0.95 0.76 30 Degrees 0.73 0.94 0.76 j

45 Degrees 0.74 0.94 0.76 VESSEL 3/4T

.?

0 Degrees 0.31 0.63 0.37 i

15 Degrees 0.30 0.59 0.36 l

30 Degrees 0.31 0.60 0.37 45 Degrees -

0.30 0.58 0.35 l

t t

I i

f I

i i

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

60 TABLE 6-3 RELATIVE AXIAL DISTRIBUTION OF EXPOSURE PARAMETERS HEIGHT (ft)

F(z)

+5.38 0.52

+4.38 0.80

+3.38 0.98

+2.38 1.04

+1.38 1.05

+0.38 1.02 0.0 1.00

-0.38 0.98

-1.38 0.92

-2.38 0.84

-3.38 0.73

-4.38 0.58

-5.38 0.35 Note: Z = 0.0 is referenced to the midplane of the reactor core.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

61 i

I TABLE 6-4 i

i FUEL CYCLE SPECIFIC CALCULATED NEUTRON FLUX (E > 1.0 MeV)

AT THE SURVEILLANCE CAPSULE CENTER I

[

CYCLE FLUX in/cm -sec1 I

2 r

1 1.62E+09 2

1.60E+09 i

3 1.15E+09 l

l 4

1.32E+09 l

t 5

1.28E+09 6

1.42E+09 l

7 1.44E+09

.j 8

1.33E+09 9

1.28E+09 l

10 1.20E+09 11 1.24E+09 Average 1.35E+09 l

l i

i l.

F Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 l

62 TABLE 6-5 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS 1

REACTION TARGET OF-WEIGIri' PRODUCT INTEREST FRACTION HALF-LIFE Cu-63 (n,a) Co-60 0.6917 5.271 yrs Fe-54 (n,p) Mn-54 0.0580 312.5 dys Ni-58 (n p) Co-58 0.6827 70.78 dys Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

~_

63 TABLE 6-6 l

MONTHLY THERMAL GENERATION DURING THE FIRST ELEVEN CYCLES OF THE BRUNSWICK UNIT 2 REACTOR THERMAL THERMAL THERMAL MONTH (MW-hr)

MONTH LMW-hr)

MONTH (MW-hr) 0405 0012500 11n7 0000000

%/80 0000000 05n5 0148844 12n7 0001563 07/80 0000000 06n5 0125202 Oln8 0883123 08/80 0000000 07n5 0212731 02n8 1403666 09/80 0150571 08n5 0569225 03n8 1106317 10/80 1251271 09US 0300888 04n8 1292585 11/80 1262123 10n5 1197319 05n8 1586110 12/80 0597993 11n5 0979833 06n8 0531921 01/81 1098990 1205 1172157 0708 1455046 02/81 0667171 i

{

Oln6 0946428 08n8 1508960 03/81 0178109 02n6 0817172 09n8 M73488 04/81 0545735 03n6 0591218 1008 1736819 05/81 0147279 l

0406 0000000 11n8 0939249 06/81 0620517 4

0506 0016003 12n8 1588841 07/81 0775438 j

06n6 1078926 01n9 1564392 08/81 1281900 i

0766.

0788028 02n9 12N143 09/81 1471176 08n6 1194128 03n9 0084172 10/81 0891180 09n6 0834401 04n9 0000000 11/81 1333517 19n6 0789789 05n9 0098855 12/81 1375144 1L'i6 0149363 06n9 0637191 01/82 0866323 12Sa 0602508 07n9 1163758 02/82 1186107 0167 1341472 08n9 1293212 03/82 1347896 02n7 0585761 09n9 0911926 N/82 1067640 03n7 1147344 10n9 1561066 05/82 0000000 0467 0539119 11n9 1416365 06/82 0000000 05n7 0941144 12n9 1351546 07/82 0000000 06U7 -

1245437 01/80 1229990 08/82 0000000 0707 0963128 02/80 1370499 09/82 0000000

~

08n7 0917672 03/80 00M471 10/82 0557346 i

09n7 0280N5 04/80 0000000 11/82 0000000 1067 0000000 05/80 0000000 12/82 1274129 4

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

64 l

TABLE 6-6 (continued)

MONTHLY THERMAL GENERATION DURING THE FIRST ELEVEN CYCLES OF THE BRUNSWICK UNIT 2 REACTOR THERMAL THERMAL THERMAL MONTH (MW-hr)

MONTH (MW-br)

MONTH (MW-hr) 01/83 1352780 08/85 1690778 03/88 0000000 02/83 0783031 09/85 1305150 04/88 0029407 03/83 1773155 10/85 0828663 05/88 0924667 04/83' 0428342 11/85 1374167 06/88 1688154 05/83 1007177 12/85 0000000 07/88 1215057 06/83 1000553 01/86 0000000

- 08/88 1752219 07/83 1573880 02/86 0000000 09/88 1687861 08/83 1062818 03/86 0000000 10/88 1776949 09/83 1477730 04/86 0000000 11/88 1502852 10/83 1764999 05/86 0000000 12/88-1805567 11/83 0055272 06/86 0255060 01/89 1799843 12/83 -

0000000 07/86 1512107 02/89 1632092 01/84 1275456 08/86 1653338 03/89 1760771 02/84 1468832 09/86 1747085 04/89 1746448 03/84 0610832 10/86 0804552 05/89 1790313 04/84 0000000 11/86 1657489 06/89 10708 %

05/84 0000000 12/86 1802263 07/89 1633033 06/84 0000000 01/87 1106782 08/89 1566162 07/84 0000000 02/87 1306273 09/89 0379168 08/84 0000000 03/87 1616751 10/89 0000000 09/84 0000000 04/87 1482956 11/89 0000000 10/84 12388 05/87 1682944 12/89 0000000 11/84 0453055 06/87 1736392 01/90 0000000 12/84 0713582 07/87 1781111 02/90 0000000 01/85 1708488 08/87 1779556 03/90 0701814 02/85 1580591 09/87 1625574 04/90 1427305 03/85 1048984 10/87 1534183 05/90 1140720

,- l 04/85 1228603 11/87 1376827 06/90 1064343

• I 05/85 1780229 12/87 1297778 07/90 1731393

. 06/85 1606491 01/88 0037715 08/90 0850289 07/85 1565455 02/88 0000000 09/90 1329384 Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

i 65 TABLE 6-6 (continued)

MONTHLY THERMAL GENERATION DURING THE FIRST ELEVEN CYCLES OF THE BRUNSWICK UNIT 2 REACTOR THERMAL THERMAL THERMAL MONTH (MW-hr)

MONTH (MW-hr)

MONTH (MW-hr) 10/90 1280210 08/92 0000000 06/94 0007869 11/90 1720660 09/92 0000000 07/94 1674473 12/90 1789238 10/92 0000000 08/94 1796669 01/91 1382159 11/92 0000000 09/94 1733499 02/91 1607538 12/92 0000000 10/94 1806409 03/91 1636922 01/93 0000000 11/94 1736615 04/91 0000000 02/93 0000000 12/94 1802773 05/91 1319661 03/93 0000000 01/95 1795891 06/91 1703179 G4/93 0000000 02/95 1628602 07/91 1778176 05/93 0572959 03/95 1802790 08/91 1758328 06/93 1731856 04/95 1743087 09/91 0612987 07/93 1791963 05/95 1811378 10/91 0000000 08/93 1772780 06/95 1701665 11/91 0000000 09/93 1733054 07/95 1759158 12/91 0000000 10/93 1644008 08/95 1632473 01/92 1136312 11/93 1735434 09/95 1655589 02/92 0780272 12/93 1788958 10/95 1597090 03/92 1348484 01/94 1772716 11/95 1501116 04/92 0931086 02/94 1584865 12/95 1305829 l

05/92 0000000 03/94 1437659 01/96 1110027 06/92 0000000 04/94 0000000 02/96 0064340 07/92 0000000 05/94 0000000 TOTAL = 2.334E+08 i

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

66 TABLE 6-7 L

MEASURED SENSOR ACTIVITIES AND DERIVED REACTION RATES MEASURED SATURATED REACTION ACTIVITY ACTIVITY RATE SENSOR (dos /em)

(dos /em)

(ros/ atom).

Cu-63 (n.a) Co-60 96-0519 1.64e+04 3.18e+04 4.86e-18 j

96-0522 1.83e+04 3.55e+04 5.42e-18 AVERAGE 1.74e+04 3.37e+04 5.14e-18 Fe-54 (n.o) Mn-54 96-0521 8.87e+04 1.50e+05 2.40e 96-0524 9.18e+04 1.56e+05 2.49e-16 AVERAGE 9.03e+04 1.53e+05 2.45e-16 Ni-58 (n.o) Co-58 96-0520 5.75e+05 1.89e406 2.70e-16 96-0523 5.66e+05 1.86e+06 2.66e-16 AVERAGE 5.71e+05 1.88e+06 2.68e-16 l

Note: Measured data are referenced to a counting date of 05/01/%.

i

+

I i

). -

1 1

4 4

i Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 r

67 3

i l

TABLE 6-8 l

l

SUMMARY

OF NEUTRON DOSIMETRY RESULTS FOR THE SURVEILLANCE CAPSULE I

Measured Flux and Fluence i

o,.

I Flux Fluence 1

2 2

Quantity (n/cm -sec)

(n/cm )

Uncertainty l

. 6 (E > 1.0 MeV) 1.18E+09 4.06E+17 10 %

$ (E > 0.1 MeV) 2.01E+09 6.92E+17 17 %

dpa/sec 1.86E-12 6.41E-04 9%

I i

i i

i Note: Fluence values were based on 10.9 effective full power years l

of operation through the completion of fuel cycle 11.

j l

i i

i TABLE 6-9 COMPARISON OF MEASURED AND FERRET CALCULATED

{

REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER REACTION RATE RATIO l

(RPS/ NUCLEUS)

MEAS / CALC l

Reaction Meas Adi Calc

' Adi Calc

?

Cu63(n,(x)Co60 5.14E-18 4.90E-18 1.05 Fe54(n,p)Mn54 2.45E-16 2.42E-16 1.01 i

[-

NiS8(n,p)CoS8 2.68E-16 2.85E-16 0.94 l

1 l

\\

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

68 i

TABLE 6-10 i

ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF THE SURVEILLANCE CAPSULE l

" d 2

2 Group #

Energy (MeV)

Flux (n/cm -sec)

Group # Energy (MeV) Flux (n/cm sec) 1 1.73E401 1.34E+06 28 9.12E-03 7.33E+07 2

1.49E+01 2.73E+06 29 5.53E-03 7.48E407 3

1.35E+01 8.13E+06 30 3.36E-03 2.35E+07 4

1.16E401 1.78E407 31 2.84E-03 2.31E+07 5

1.00E+01 3.45E+07 32 2.40E-03 2.31E+07 i

6 8.61E+00 4.78E+07 33 2.04E-03 6.95E+07 7

7.41E+00 1.07E+08 34 1.23E-03 6.95E+07 8

6.07E+00 1.19E+08 35 7.49E-04 6.84E+07

)

9 4.97E+00 1.55E+08 36 4.54E-04 6.48E+07 10 3.68E+00 1.19E+08 37 2.75E-04 6.76E+07 11 2.87E+00 1.54E+08 38 1.67E-04 7.06E+07 12 2.23E+00 1.36E+08 39 1.01E-04 6.89E+07 13 1.74E+00 1.31E+08 40 6.14E-05 6.80E+07 14 1.35E+00 9.80E+07 41 3.73E-05 6.82E+07 15 1.11E+00 1.39E+08 42 2.26E-05 6.79E+07 16 8.21E-01 1.28E+08 43 1.37E-05 6.62E+07 17 6.39E-01 1.17E+08 44 8.32E-06 6.52E+07

]

18 4.98E-01 8.44E+07 45 5.04E-06 6.70E+07

)

19 3.88E-01 1.03E+08 46 3.06E-06 6.88E+07 20 3.02E-01 1.51E408 47 1.86E-06 6.95E+07 21 1.83E-01 1.30E+08 48 1.13E-06 6.74E+07 J

22 1.11E-01 1.04E+08 49 6.83E-07 6.47E+07 23 6.74E-02 9.05E+07 50 4.14E-07 8.92E407 24 4.09E-02 6.46E+07 51 2.51E-07 3.16E+08 l

25 2.55E-02 5.19E+07 52 1.52E-07 5.37E+08 i

I 26 1.99E-02 4.20E+07 53 9.24E-08 1.17E+09 27-1.50E-02 7.07E+07 Note: Tabulated energy levels represent the upper energy in each group.

~.!

j Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

.. ~.

69 TABLE 6-11 l

COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FOR THE SURVEILLANCE CAPSULE i

i Calculated Measured M/C Fluence (E > 1.0 Mev) [n/cm2]

4.67E+17 4.06E+17 0.870 Fluence (E > 0.1 Mev) [n/cm2]

7.99E+17 6.92E+17 0.866 dpa 7.M E-04 6.41E-04 0.908 t

f I

t f

t 9

s Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

70 TABLE 6-12 NEUTRON EXPOSURE PROJECTIONS AT THE INNER RADIUS OF THE REACTOR PRESSURE VESSEL LOWER INTERMEDIATE SHELL 4(E > 1.0 MeV)

@(E > 0.1 MeV)

LOCATION (n/cm )

(n/cm )

2 2

da 2

EOC 11 NEUTRON EXPOSURE (10.9 EFPY)

VESSEL IR 0 Degrees 2.55e+17 4.65e+17 4.19e-04 1

15 Degrees 3.48e+17 6.30e+17 5.67e-04 30 Degrees 3.25e+17 6.38e+17 5.32e-04 45 Degrees 5.33e+17 9.74e+17 8.63e-04 3

16 EFPY NEUTRON EXPOSURE i

VESSEL IR 0 Degrees 3.73e+17 6.80e+17 6.13e-04 15 Degrees 5.09e+17 9.27e+17 8.29e-04 1

30 Degrees-4.76e+17 9.34e+17 7.79e-04 45 Degrees -

7.80e+17 1.43e+18 1.26e-03 32 EFPY NEUTRON EXPOSURE VESSEL IR 0 Degrees 7.46e+17 1.36e+18 1.23e-03 15 Degrees 1.02e+18 1.85e+18 1.66e-03 30 Degrees 9.52e+17 1.87e+18 1.56e-03 45 Degrees 1.56e+18 2.85e+18 2.53e-03 Note: These exposures' represent the maximum levels experienced by the vessel Lower Intermediate shell.

1 l 1 Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

71 i

TABLE 6-13 l

i NEUTRON EXPOSURE PROJECTIONS AT THE IhWER RADIUS

[

OF THE REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELD

[

AND PRESSURE VESSEL LOWER SHELL.

i t

(

@(E > 1.0 MeV) 4(E > 0.1 MeV) i 2

2

. LOCATION (n/cm )

(n/cm )

doa f

EOC 11 NEUTRON EXPOSURE (10.9 EFPY) t VESSELIR i

0 Degrees 2.06e+17 3.76e+17 3.39e-04 l

15 Degrees 2.81e+17 5.10e+17 4.59e-04

{

30 Degrees 2.63e+17 5.17e+17 4.31e-04 45 Degrees 4.31e+17 7.88e+17 6.98e-04 16 EFPY NEUTRON EXPOSURE f

VESSEL IR i

0 Degrees 3.02e+17 5.50e+17 4.96e44 1

15 Degmes 4.12e+17 7.47e+17 6.71e-04 30 Degrees.

3.85e+17 7.56e+17 6.31e-04 45 Degrees 6.31e+17 1.15e+18 1.02e-03 32 EFPY NEUTRON EXPOSURE

[

VESSEL IR O Degrees 6.04e+17 1.10e+18 9.93e-04

[

15 Degrees 8.23e+17 1.49e+18 1.34e-03

('

30 Degrees 7.71e+17 1.51e+18 1.26e-03

'45 Degrees 1.26e+18 2.31e+18 2.04e-03 Note: 7hese exposures represent the maximum levels experienced by the vessel circumferential weld and lower shell.

.l i

l t

.I i

l i

l

{

i l

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 i

~.

t

I 72 SECTION

7.0 REFERENCES

1.

NEDO-24157, " Brunswick Steam Electric Plar,t 2: Information on Reactor Vessel Material Surveillance Program," General Electric, June 1994, Revision 2.

2.

ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels", E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.

3.

Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, " Fracture Toughness Criteria for Protection Against Failure".

4.

ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Deternune Nil-Ductility Transition Temperature of Ferritic Steels", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

5.

BSEP 95-0572, " Brunswick Steam Electric Plant, Unit Nos. I and 2 Docket Nos. 50-325 and 50-324/ License Nos. DPR71 and DPR-62 Response to NRC Generic Letter 92-01, Revision 1, Supplement 1 Reactor Pressure Vessel Integrity."

6.

Code of Federal Regulations,10 CFR Part 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements", U.S. Nuclea-Regulatory Commission, Washington, D.C.

7.

ASTM E23-93a, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials",

in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.

8.

ASTM A370-92, " Standard Test Methods and Definitions for Mechanical Testing of Steel Products",in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.

9.

ASTM E8-93, " Standard Test Methods for Tension Testing of Metallic Materials", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.

10. ASTM E21-92, " Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.

11. ASTM E83-93, " Standard Practice for Verification and Classification of Extensometers", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.

12. CVGRAPH, Hyperbolic Tangent Curve-Fitting Routine, Version 4.1, developed by ATI Consulting, March 1996.

13. ASTM E853-87, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

73

14. ASTM E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

15. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S.

Nuclear Regulatory Commission, May,1988.

16. RSIC Computer Code Collection CCC-543, " TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport, Version 2.7.3", May 1993.

17. RSIC Data Library Collection DLC-175, " BUGLE-93, Production and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron / Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data", April 1994.

18. CP&L Letter 96A-0283, " Brunswick Unit 2 Fluence Data", J. M. Given to J. W. Voss Jr., May 17,1996.

19. ASTM Designation E482-89, Standard Guidefor Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

20. ASTM Designation E560-84, Standard Recommended Practicefor Extrapolating Reactor Vessel Surveillance Dosimetry Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

21. ASTM Designation E693-79, Standard Practicefor Characterizing Neutron Exposures in Ferritic Steels in Terms ofDisplacements per Atom (dpa), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

22. ASTM Designation E706-87, Standard Master Matrixfor Light-Water Reactor Pressure Vessel Surveillance Standard, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

23. ASTM Designation E853-87, Standard Practicefor Analysis and Interpretation of Light-Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

24. ASTM Designation E261-90, Standard Practicefor Determining Neutron Fluence Rate, Fluence, and Spectra by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

25. ASTM Designation E262-86, Standard Methodfor Determining Thermal Neutron Reaction and Fluence _ Rates by Radioactivation Techniques,in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

26. ASTM Designation E263-88, Standard Methodfor Measuring Fast-Neutron Reaction Rates by Radioactivation ofIron, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

74

27. ASTM Designation E264-92, Standard Methodfor Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

28. ASTM Designation E481-92, Standard Methodfor Measuring Neutron-Fluence Rate by Radioactivation of Cobalt and Silver, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

29. ASTM Designation E523-87, Standard Test Methodfor Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

30. ASTM Designation E704-90, Standard Test Methodfor Measuring Reaction Rates by Radioactivation of Uranium-238, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

31. ASTM Designation E705-90, Standard Test Methodfor Measuring Reaction Rates by Radioactivation of Neptunium-237, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.

32. ASTM Designation E1005-84, Standard Test Methodfor Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society.or Testing and Materials, Philadelphia, PA,1993.

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

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

35. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section l

Compendium", July 1994.

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

Analysis of Brunswick Unlt 2 300 Deg. Capsule November 1996

75 APPENDIX A LOAD-TIME RECORDS FOR CHARPY IMPACT TESTS Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 l~

s Cwve 784472-AF13 Wg Wp y

-PM = Maximum Load P = Fast Fracture Load ir p

P General I

GY = Yield Load u

I t

i I

.3 l

P = Fast Fracture A

u i

Arrest Load l

l 1

l l

I I

I i

1 I

I 1

i I

i t

I I

l I

I I

I I

I I

I I

4-- tGY e

tM tg Time W, = Fracture initiation Region i

t Wp = Fracture Propagation Region GY = Time to GeneralYielding

[

t

= Time to Maximum Load M

i g

t

= Time to Fast (Brittle) Fracture Start g

Fig. A-1-Idealizedload-time record l

t e

v

77 anunsuscx unit et un ansc i

i i

0

?

A i

x

.f 5

\\

l.

~

$d s

g M Auna_m a,_ ______,

e i

s s

.3 a.e e.4 3.6 4.s e,o TIE

( MIEC >

BRUNSWICK UNIT #2 SPECIMEN NUMBER

37A MATERIAL

BASE CAPSULE

BRUNSWICK #2 85R#tsWICK 1.pt!T 82 3sg east

)

8 4

4 a

e-mIr a

3 a

l e

b^Aa^_

-^-

4..~__m e

4 a

e 4

a%

3.2 2.4 3.6 4.s

6. 0 rinc c nace >

ERUNSWICK UNIT #2 SPECIMEN NUMBER

36K MATERIAL

BASE CAPSULE

BRUNSWICK #2 Figure A-2. Load-time records for Specimens 37A and 36K.

78 entmsu!cK urtIT et 3P5 test

}

i i

a i

4 7

I.

{

a

.N 4

w al u ~,-

4 4

i

.n 1.a r.4 s.s 4.s s.o T!E

( MSEC 3 BRUNSWICK UNIT #2 SPECIMEN NUMBER

375 MATERIAL

BASE CAPSULE

BRUNSWICK #2 anunsu!CK urt!T et ans angg

,I 8

i a

a a

1s a

e-I w

m

=-

-^^ ^

.9 3.2

2. 4

3. 6 4.8 s.O TIE

( Peste )

BRUNSWICK UNIT #2 SPECIMEN NUMBER

3AB MATERIAL

BASE CAPSULE

BRUNSWICK #2 Figure A-3. Load-time records for Specimens 375 and 3AB.

79 anUMWICK UMIT et sac enet i

i a

J 4

7 S"

gr q.

l 4

4 3

.9 1.2 2.4 3.6

4. 0

6. 0 TIE C MIEC D BRUNSWICK UNIT #2 SPECIMEN NUMBER

36C MATERIAL

BASE CAPSULE

BRUNSWICK #2 SAWERWICK UMIT et 375 anK

~

=

a i

e i

4 M

1 7

~-

S

.i q.

i a

i i

.3 s.a a.4 s.s 4.s 6.0 T!PE

( MEC 3 BRUNSWICK UNIT #2 SPECIMEN NUMBER

37B MATERIAL BASE CAPSULE

BRUNSWICK #2 Figure A-4. Load-time records for Specifnens 360 and 37B.

81 samniscK mir.e

=

• ~

m Iz d

,C

~

1r w=

es

• A AA

_Aa, aa

..A.

,a$_,,_,

.D 1.2 2.4 3.6 4.8 6.4 TIE

( reEC )

BRUNSWICK UNIT #2 SPECIMEN NUMBER

3BK MATERIAL

WELD CAPSULE

BRUNSWICK #2 anuMailcK UNIT.4 3EK EELD I

i 6

i w

M 1x 3

e, -

a w=

\\

b m.. _....

.9 1.2 2.4 3.6

4. 0

6. 0 TIE

( Matt 3 BRUNSWICK UNIT #2 SPECIMEN NUMBER

3EK MATERIAL

WELD CAPSULE

BRUNSWICK #2 Figure A-6. Load-time records for Specimens 3BK and 3EK.

i j

8

BALMSu!CK utlt at 3CS e

8 g

3 e

a R'

e M

1z

^e-3 O

e" g.

Id 9-b^^^^a

-a a.

i i

.9 8.2 2.4 3.6 4.8

6. 0

?!*

( pistC 3 BRUNSWICK UNIT #2 SPECIMEN NUMBER

3C5 MATERIAL

WELD CAPSULE

• BRUNSWICK #2 anUMsWICK witT et 302 651.0 6

6 i

I e

e e

1 x

3 e

l

[

^

g l

4 4

5

.3 3.2 2.4 3.6 8.s

6. 0 TIE C PtKC D BRUNSWICK UNIT #2 SPECIMEN NUMBER

3D2 MATERIAL iWELD CAPSULE

BRUNSWICK #2 Figure A-7. Load-time records for Specimens SC5 and 3D2.

83 anuMSWICK LMI? et Set acts

}

4 i

a a

M i

z N

as w

.C 4

• t~

~

/

s 4

4 a

.9 8.2 2.4 3.6

4. 8 6.0 f!E

( MBEC )

BRUNSWICK UNIT #2 SPECIMEN NUMBER.

23AM MATERIAL

WELD CAPSULE 2 BRUNSWICK #2 EALMEWICK LMIT et 3au agLa w.

4 a

a a

e m

I r

A ea w

4 q

i a

a i

.9 1.2 E.4 3.6 4.0 6.0 ras

< note >

BRUNSWICK UNIT #2 SPECIMEN NUMBER

3AU MATERIAL

WELD CAPSULE

BRUNSWICK #2 Figure A-8. Load-time records for Specimens 3AM and 3AU.

84 amschilcx tsett et scg asLD i

a i

8 1

1 M

1 s

N aa 4

w *"

l4

~

r

?.

q.

1 s

6 4

a

.3 1.2 2.4 3.6 4.s 6.0 Tlft i NEC 3 BRUNSWICK UNIT #2 SPECIMEN NUMBER

3CE MATERIAL WELD CAPSULE

BRUNSWICK #2 AAtssWICx 48617 et Sir agLa 4

5 a

6 f

e M

k A

g.m S"

{

\\

w I,

4 i

an =

E i

t e

a s

a i

.9 let 2.4 3.6 4.0

6. 0 t st

< Marc >

BRUNSWICK UNIT #2 SPECIMEN NUMBER

3CP MATERIAL

WELD CAPSULE 2 BRUNSWICK #2 Figure A-9. Load-time records for Specifnens 3CE and 3CP.

3 i

85 m acx mIt.e w

}

4 4

i i

9" M

i e-3 w"

=

s AA-7^_____

4 _ _ _

.D 1.2 2.4 3.6 4.0

6. 0 TIN

( nug )

BRUNSWICK UNIT #2 SPECIMEN NUMBER 33J7 MATERIAL

8-16-96 CAPSULE BRUNSWICK #2 MMSWICK Imtf et 3Je w

J 6

a e-i A

g eu ei 4

w 95" 9m M*

=

AAAA-a

__x_

e 4

4 6

6

.8 1.2 2.4 3.6 4.8 6.0 71E

( MIEC )

BRUNSWICK UNIT #2 SPECIMEN NUMBER 23J2 MATERIAL

HAZ CAPSULE

BRUNSWICK #2 Figure A-10. Load-time records for Specimens 3J7 and 3J2.

86 namourer wrr.e

=

w 3

i i

MI z

N e

e n

).

w g.-

a ei l

e

.D 1.2 2.4 3.6 4.8 6.0 3

7:r.

c Mice >

BRUNSWICK UNIT #2 I

SPECIMEN NUMBER 33DT MATERIAL

HAZ CAPSULE

BRUNSWICK #2 anuMsWICK WIT et Eut sent i

i i

4 l

1 m

1 7

a e

e en W

I-e6 q-

.3 a.:

a.4 s.6 4.s 6.o TIPE

( Marc )

c BRUNSWICK UNIT #2 SPECIMEN NUMBER

3DK MATERIAL

HAZ CAPSULE

BRUNSWICK #2 Figure A-11. Load-time records for Specimens 3DT and 3DK.

)

~

~... _..

.a m u aurer mit.e em 87 y

e R

1x a

$ ci W

.(

i l

g,-

aw 6

f P

I u-f t

1.8 8.4 3.6 6.0 TIE

( MIEC D BRUNSWI 3 UNIT #2 I

SPECIMEN NUMBER

3JD MATERIAL HAz CAPSULE BRUNSWI G #2 masedicK un!T se 23 sent i

4 a

a a

j I

i I"

a

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e-3 w

i e.

l

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~

g i

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4

.3 8.2 E.4 3.6 4.s

6. 0 TIE

( WC D BRUNSWI G UNIT #2 SPECIMEN NUMBER 23E3 MATERIAL inAz CAPSULE

BRUNSWIG #2 Figure /s-12. Load-time records for Specimens 3JD and SES.

88 e-rcx im:T w am e

e a

1 e

A ns -

s

^

s a

s

.o 8.8 2.4 3.s 4.0

s. 0 TIE

( Metc 3 BRUNSWIG UNIT #2 SPECIMEN NUMBER 3JA MATERIAL

• HAZ CAPSULE 2BRUNSWI G #2 MushitCK LMIT et ML seg 9

i i

a e

e M

M 1

7 S"

a u~

~

4 i

s c

.o s.2 2.4 s.s e.e

s. s TIE

( sec )

i BRUNSWIM UNIT #2 I

SPECIMEN NUMBER 23DL MATERIAL

HAZ CAPSULE 2BRUNSWI G #2 Figure A.13. Lcad-time records for Specimens 3JA and 3DL.

i i

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

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

89 APPENDIX B TEST EQUIPMENT CALIBRATION RECORDS AND MODEL NUMBERS

• 1 I

1 i

i i

t l

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4 i

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996 Y

V

90 WESTINGHOUSE SCIENCE & TECHNOLOGY CENTER CALIBRATION RECORD DA7E 8/15/96 INSTRUMENT BEING CALIBRATED: Tinius-Olsen Mod. 74.Ser.# 123159 IMPAcr MACHINE LOCATION: Blde. 302A Low 12 vel Cell OPERATOR: F.X.GRADICH CALIBRATION METHOD: NIST standards ner ASTM E23-91 RESULTS:

DIAL DYNATUP NOT SERIES SRM No. AVE.VALUE AVE.VALUE EXPECIED RANGE ACCEPTABLE ACCEPTABLE ft-lbs ft-lbs ft-lbs LL52 2092 13.2 12.95 9-15 X

HH53 2096 72.4 72.77 65-85 X

CALIBRATION :

X ACCE. lABLE NOTACCEPTABLE COMMENTS: THE CALIBRATION RESULTS ARE ACCEPTABLE FOR BOTH THE DIAL AND DYNATUP OUTPUT.

CALIBRATION DUE DATE: 8/15/97 APPROVAL:

/3 b

~

C. DEFLITCH DA~IE SUPERVISOR REMOTE METALLOGRAPHIC FACILITY SCIENCE & TECHNOLOGY CENTER e

e

=

SCIENTIFIC INSTRUMENTS Page Of 91

//hfCond.7/

an-asoli CAllBRATION/ SERVICE DATA Last Cat.

E- //.

/94 Date I C 79 T

Serial No.

Unit Under Test (UUT)

~

l Make n,--

Inv. No.

J Ob7 f Model dd Job.No.

Customer Name/Div.

T T' C Temp.

93 C NO Humid.

O Notes:

Specs.

Procedure No.

Equip. Used (e.g., Descrp., S/N, NIST #)

M 375

/

^0

0. d.T

,F

- 3AS 4

-3 *Nf

-se.

- 30 a 200

- 20s

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Remarks:

J

SCIENTIFIC INSTRUMENTS page

/ ot

/

92 RM-tS011 CAllBRATION/ SERVICE DATA Last cal. 9-64fcond. b Date 9 - F-9 4-Unit Under Test (UUT) viGsTAL i ne w oueTf C Serial No.

I7 (r/9 I

Make OM6G4 inv. No.

Model 1%%

Job.No.

NO Customer Name/Div.

KemTT 7

Temp.

DC Humid.

YOb Notes:

Specs.

z.T " F

~d up 0 I, 12 #c 4 Procedure No.

O#

Equ;p. Used (e.g., Descrp., S/N, NIST #)

E,e /gg, Yaf z.7,,3 ) g"

.6TDs T Rd -5

- I 50,000 - I 70

-50,000

~ SO 0.0000 000 50.000 0 '/9 100 000 to e 4,5 eevn E ll,00 IlI UMITED CAUBRATION i l

.790.00 3'l$

SgANDARDS-l. ABOR _ATORY 500,00 499

= e=

750.00 iTI mr,,o. / 7uf 9 -ir-yr mo, cesy Ma 9" I ~

cA DUE CAUBRATED PERFORMANCE

-/.5 0 /* '6 70o'F

~

t % 5'f j

4 i

i Service Technician f

• y Hours V Approvai /

Remarks:

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/

9q SCIENTIFIC INSTRUMENTS page CALIBRATION / SERVICE DATA Last Call'%~ f6 C:nd. A Date 8~ I~ 0 C M

MtcRoMcTER Serial No.

7602 Unit Under Test (UUT) 3ROlAnj (SHARPC Inv. No.

Make Job.No.

C43B6 Model PETER BARSoTr;

/ 5 TG Temp.

72 Humid. -% /0 Customer Name/Div.

C 'DAT A Specs.

Notes:

E-/b Procedure No.

Equip. Used (e.g., Descrp., S/N, NIST #)

G AttG c Bloc K

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( Avt. cF 3l2 oderecheasi.)

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2 Approval

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four Asset Number: 985537

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Certificate of Calibration i

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Equipment Owner Calibration Facility

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Ii WESTINGHOUS E ;, ELEC..CORPpyg

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Calibration traceable to the National institute of Standards and Technology in L:

accordance with MIL-STD-45662A has been accomplished on the instrument listed below by comparison with standards maintained b E.I.L. Instruments. The i:

accuracy and stability of all standards maintained b E.I.L. Instruments are traceable to national standards maintained by the NationalInstitute of Standards and Technology in Washington, D.C. and Boulder, Colorado. Complete record of t:

all work performed le maintained by E.I.L. Instruments and is available for inspection upon request.

_j :

NS m N 20.000 LBS.

MANUFACTURER:

MODEL NUMDER:

ENSRE TESTER SERIAL NUMBER:

NOMENCLATURE:

MB-44323-H CALIBRATiggl4291 Item No.

1 P.O. NUMBER:

REPORT:

c:

C 09Q3 6

4 6*.

72F Date of Calibration RH Temp.

t Certified By:

C In Tolerance When Received Date Due:

Out Of Tolerance When Received i:

Traceability GAGE BLOCK SET, NIST #821/256567-96, MAY 20, 1996 i:

LIGHT SOURCE. NIST #534/221461-1, 1/11/96 WEIGHT SET, NIST #5316/665733, 11/21/94

[

TEMPERATURE, NIST #253699, 7/20/95 5

PRESSURE, NIST #490746, 2/19/93

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FLOW. NIST #731/241140-88 b

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ISO 9002 CERTIFIED

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APPENDIX C.

j EVALUATION OF CHARPY SPECIMEN 3EK 4

?

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,1 i

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Analysis of Brunswick Unit 2 300 Dog. Capsule November 1996 1

4*

.. ~ -

96 Westinghouse received the Brunswick Unit 2 300* surveillance capsule at the Science and Technology Center on April 15, 1996. When the capsule was disassembled it was discovered that Charpy specimen 3EK was in a location that was designated for weld metal Charpy specimens. However, 3EK was an identification number for a HAZ Charpy specimen. This resulted in Westinghouse etching the specimen to determine if it was a HAZ specimen or weld metal specimen. This evaluation resulted in the preliminary determination that this specimen was a weld metal specimen. Hence, Charpy specimen 3EK was tested as if it were a weld metal specimen and the results were included in the weld metal test data.

A chemical analysis was performed on four weld metal specimens as pan of this effort (see Table 4-2). The chemical analysis was performed on weld metal Charpy specimens 3BK,3C5,3D5 and 3EK.

The results of this chemical analysis indicated that Charpy specimen 3EK had a slight difference in the results than Charpy specimens 3BK,3C5 and 3D5 (particularly in the copper, nickel and carbon).

A review of the chemical analysis results of Charpy specimen 3EK by CP&L indicate that the chemical analysis results are very similar to the chemistry data for the Brunswick Unit 2 plate material.

The chemistry results of Charpy specimen 3EK lead to a further investigation as to why the results were different than the other three weld metal specimens. Hence, the two halves of Charpy specimen 3EK were etched and reexamind. The results of the reevaluation are shown in Figure C-1. Figure c.

I shows that the entire specimen was made of weld metal except in the region of the notch. The notch region appears to be made of part base metal and part weld metal.

The odd material make-up in the notch region can not be explained and the chemical analysis test results appear to be questionable. Since it can not be determined if the fracture represents a weld fracture, a plate fracture or a HAZ fracture; the data was eliminated from the evaluation of the Charpy test data presented in this report.

The test results from Charpy specimen 3EK are presented in Tables 4-2,5-2 and 5-5 of this repon for completeness.

Contained in Figures 5-4 through 5-6 are plots of the measured Charpy data for the weld metal without the results of specimen 3EK included and contained in Figures C-2 through C-4 of this appendix are the plots of the measured Charpy data for the weld metal with the results of specimen 3EK included. Contained in Table C-1 of this appendix is a comparison of the weld metal Charpy test results when specimen 3EK is and is not included in the evaluation of the data.

1 l

l l

l l

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

97 Base Metal f

W

/

V Weld Weld Base Metal o

Weld J

Weld

^

Base Metal

[

[

As Received

/

Chemistry Analysis Cut Weld 7

~.

Figure C 1.

Sketch of Charpy Specimen 3EK After Etching Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

98 Brunswick Unit 2 Surv. Weld CVGPAPH 4.1 Hyperbolic Tangent Curve Printed at 112D9 on 11-04-1996 Page1 Coefficients of. Curve 2 A = 3459 B = 24 C = 163fi6 10 = 114.97 Equation is CVN : A + B 'I tanhKT - TO)/C) ]

Upper Shelf Energy: 67 Fixed Temp. at 30 ft-lbs 915 Temp at 50 ft-lbs 1995 Lower Shelf Energy: 219 Tued Materiah YELD Heat Number:

Orientation:

Capsule 300DEG Total Fluence 426E+17 300 4

m 250

,Q Iax 2m X

tio 1.

150 C)c r.c 100 0

C of

)

Y o

i i

i i

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: BY2 Cap: 300DE Matenah WELD SA5EB1 Ori; Beat f.

Charpy V-Notch Data Temperature input CVN Energy Computed CVN Ex gy Differential j

-30 6

1141

-551 25 23 18.38 4f1 72 32 3128 E71 i

15 0 lrl 4L42

-4.42 l

200 32 50.06

-18.06 206 56 5125 4.74 250 71 NiS5 14.44 a

300 63 60 2 2.11 SL'M of REIDUAIS = 351 Figure C-2.

Charpy V-notch Impact En:,cy vs. Temperature for the Brunswick Unit 2 Surveillance Weld Metal Including Spaimen 3EK Results Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

99 J

Brunswick Unit 2 Surv. Weld CVCRAPH (j Hyper %lic Tangent Curve Printed at It4tJ9 on 11-04-1996 Page!

Coefficients of Curve 2 A = 315 B = 305 C = 146.41 TO=1345 Equation is E : A + B ' [ tanh((T - 11))/C) l P

~

U per Shelf II: 62 Fixed Temperature at 2 3fx 1513 lower Shelf LE: 1 Fixed Material: YELD Heat Number.

Orientation:

)

Capsule 300D'S Total Fluence (DGE+17 200 I

)

20 b

n M

10 0

-e k0 a

o e

y a

su y

u i

-:10 0

-200

-100 0

100 200 300 400 soo soo j

Temperature in Degrees F

]

Data Set (s) Plotted i

Plant: BT2 Cap: 300DED Maten'ah TElh SA533B1 ori:

liest f.

Charpy V-Notch Data Temperature input Lateral Expansion Computed E Differential

-30 2

6J13

-4El 25 16 12J6 331 72 24 1921 4.78 150 34 34.71

.71 200 28 443

-16 3 208 47 45S4 13 250 64 515 6 1143 300 60 5623 176 SUM of RISIDUAIS = 432 Figure C-3.

Charpy V-notch Lateral Expansion vs. Temperature for the Brunswick Unit 2 Surveillance Weld Metal including Specimen 3EK Results Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

100 Brunswick Unit 2 Surv. Weld CVGRAPH 4J Hyperbolic Tangent Curve Printed at 075342 on 12-20-1996 Page1 Coefficients of Curve 1 A = 50 B = 50 C = 92l38 E = 185El Equation is Shearx = A + B ' I tanh((T - E)/C) j Temperature at 5&/. Shear: 1852 Material TELD Heat Numben Orientation-e Capsule 300DEG Total Fluenx 4D6E+17 100

=

u o

ce 60

(

acoOb 40 O

a o

/

U l

l a

g

-300

-200

- 100 0

100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant BW2 Cap 300DEG Material TELD Ori:

Heat l:

Charpy V-Notch Data Temperature input Perwnt Shear Computed Percent Shear Differential

-30 0

.92

.92 72 30 734

?2.15 150 35 3L53 3.46 200 20 57f1

-37El 208 75 EL78 1321 250 100 80D5 19.H 300 100 9221 7.78 SGI of PSDUAIS : 2&D2 Figure C-4.

Charpy V-notch Percent Shear vs. Temperature for the Brunswick Unit 2 Surveillance Weld Metal Including Specimen 3EK Results Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

I 101 i

TABLE C-1 Comparison of the Weld Metal Charpy Test Results when Specimen 3EK is and is not Included in the Evaluation of the Data Parameter With 3EK Without 3EK Measured 30 ft-lb Temperature 91.5 F 102.0 F Measured 50 ft-lb Temperature 199.5 F 200.3'F Measured 35 mil Lateral Expansion 151.3 F 165.8'F Temperature Measured Upper Shelf Energy 67.0 ft-lb 67 ft-lb e

Analysis of Brunswick Unit 2 300 Deg. Capsule November 1996

ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 NRC DOCKET NO. 50-324 i

OPERATING LICENSE NO. DPR-62 REACTOR VESSEL MATERIAL SURVEILLANCE SPECIMEN TEST RESULTS i

+

i LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Carolina Power & Light Company in this document. Any other actiom discussed in the submittal represent intended or planned actions by Carolina Power & Light Company. They are described to the NRC for the NRC's information and are not regulatory ::ommitments. Please notify the Manager-Regulatory Affairs at the Brunswick Nuclear Plant of any questions regarding this document or any associated regulatory commitments.

Commitment Committed

)

date or outage 1.

Submit proposed lic ense amendments for each unit to incorporate 8/1/97 a combined surveilhnce program into the Technical Specifications.

l