ML18038B481

Brown Ferry Steam Electric Station Unit 2 Vessel Surveillance Matls Testing & Fracture Toughness Analysis.

ML18038B481
Person / Time
Site:	Browns Ferry
Issue date:	06/30/1995
From:	Branlund B, Carey R, Oza C GENERAL ELECTRIC CO.
To:
Shared Package
ML18038B480	List: ML18038B479 ML18038B481
References
GENE-B1100639, GENE-B1100639-0, GENE-B1100639-01, GENE-B1100639-1, NUDOCS 9510230407
Download: ML18038B481 (143)

Text

GE Nuclear Energy Technical Services Business GENE-B 110063 9-01 General Electric Company, June 1995 175 Curtner Avenue, San Jose, CA 95125 BROWNS FERRY STEAM ELECTRIC STATION UNIT 2 VESSEL SURVEILLANCEMATERIALSTESTING AND FRACTURE TOUGHNESS ANALYSIS Prepared by:

Chandra Oza, Prin rpal Engineer Engineering Services Verified by:

R.G.Carey, Engineer Engineering Services Approved by B. J. Branlund, Project Manager RPV Integrity 9510230407 951018 PDR ADOCK 05000260 P PDR

G~-31100639-01 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CARF FULLY This report was prepared by General Electric solely for the use of Tennessee Valley Authority {TVA). The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting information in this document are contained in the contract between the customer and General Electric Company, as identified in the purchase order for this report and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than the customer or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

GENE-B 110063 9-01 TABLE OF CONTENTS

~Pa e ACKNOWLEDGMENTS 1X

l. INTRODUCTION

2.

SUMMARY

AND CONCLUSIONS 2.1

SUMMARY

OF RESULTS

2.2 CONCLUSION

S

3. SURVEILLANCEPROGRAM BACKGROUND..

3.1 CAPSULE RECOVERY 3.2 RPV MATERIALSAND FABRICATIONBACKGROUND .

3.2.1 Fabrication Historv...

3.2.2 Material Pro tties of RPV at Fabrication 3.3 SPECIMEN DESCRIPTION.

J 3.3.2 Tensile S 'mens..

4. PEAK RPV FLUENCE EVALUATION. ....18 4.1 FLUX WIRE ANALYSIS ... .... 18 4.1.1 Procedure. ... 18 4.1.2 Results. ....... 19 4.2 DETERMINATIONOF LEAD FACTOR 19 4.2.1 Procedure ...... 19 4.2.2 Results. .......20 4.3 ESTIMATE OF 32 EFPY FLUENCE ..... .....21.

5. CHARPY V-NOTCH IMPACT TESTING. 28 5.1 IMPACT TEST PROCEDURE. 28 5.2 IMPACT 'I%ST RESULTS . 29 5.3 IRRADIATEDVERSUS UNIIUbG)IATEDCHARPY V-NOTCH PROPERTIES 30 5.4 COMPARISON TO PREDICTED IRRADIATIONEFFECTS 30

....30 5.4.2 Change in USE ....31

6. TENSILE TESTING. .50 6.1 PROCEDURE ....50 6.2 RESULTS ....51

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GENE-B 1100639-01 6.3 IRRADIATEDVERSUS Ul&VMDIATEDTENSILE PROPERTIES .....51

7. DEVELOPMENT OF OPERATING LIMTS CURVES ....58

7.1 BACKGROUND

. .......58 7.2 NON-BELTLINEREGIONS 58 7.3 CORE BELTLINEREGION. ...........58 7.4 EVALUATIONOF IRIUQ7IATIONEFFECTS .....59 7.4.1 ART Versus EFPY ....60 7.4.2 U r Shelf Ener at 32 EFPY. ....60 7.5 OPERATING LMITS CURVES VALIDTO 32 EFPY ....61

8. REFERENCES .......73 APPENDIX A - CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS .......75 APPENDIX B - EQUIVALENTMARGIN ANALYSIS .....81

GENE-31100639-01 LIST OF TABLES

~Pa e TABLE '1 CHEMICALCOMPOSITION OF RPV BELTLINEMA'TRIALS TABLE 3-2 MECHANICALPROPERTIES OF BELTLINEAND OTHER SELECIZD RPV MATERIALS........9 TABLE 4-1 SUMh1fARY OF DAILYPOWER HISTORY ....23 TABLE 4-2 SURVEILLANCECAPSULE FLUX AMP FLUENCE FOR IRRADIATIONFROM START-UP TO 10/1/94. 24 TABLE 5-1 VALLECITOSQUALIFICATIONTEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS 32 TABLE 5-2 IREE)IATED CHARPY V-NOTCH IMPACT TEST RESULTS 33 TABLE 5-3 UNIIUVJDIATEDCHARPY V-NOTCH IMPACT TEST RESULTS 34 TABLE 5-I SIGNIFICANTRESULTS OF IRRADIATEDAND UNIRRADIATEDCHARPY V-NOTCH DATA 35 TABLE 6-1: TENSILE TEST RESULTS FOR IRRADIATEDRPV MATERIALS. 52 TABLE 6-2: 'ENSILE TEST RESULTS FOR UNIRRADIATEDRPV MATERIALS 52 TABLE 6-3 COMPARISON OF UNIIGM3IATEDAND IRIM3IATEDTENSILE PROPERTIES AT ROOM 53 TABLE 6-4 COMPARISON OF UNIRIVJ3IATEDAND IRIMDIATEDTENSILE PROPERTIES AT 550'F ...53 TABLE 7-1 BROWNS FERRY 2 P - T CURVE VALUES 62 TABLE 7-2 BELTLINEART VALUES .66 TABLE 7-3 UPPER SHELF ENERGY ANALYSISFOR BELTLINEMATERIALS .67 TABLE B-1 EQUIVALENTMARGIN ANALYSISPLANT APPLICABIL1TYVEIUFICATIONFORM FOR, BROWN FERRY UNIT 2 - BWR 4/MK I.......................................................................... 82 TABLEB-2 EQUIVALENTMARGINANALYSISPLANT APPLICABILITYVERIFICATIONFORM FOR BROWNS FERRY UNIT 2 - BWR 4/MK I. 83'

GEiiE-B 1100639-01 LIST OF FIGURES P~ae FIGURE 3 1. SURVEILLANCECAPSULE HOLDER RECOVERED FROM BROWNS FERRY UNIT 2..........10 FIGURE 3-2. SCHEMATIC OF THE RPV SHOWING IDENTIFICATIONOF VESSEL BELTLINEPLATES AND WELDS .............. .11 FIGURE 3-3. FABRICATIONMETHOD FOR BASE METAL CHARPY SPECIMENS ................... 12 FIGURE 3-4. FABRICATIONMETHOD FOR WELD METAL CHARPY SPECIMENS .13 FIGURE 3 5. FABRICATIONMETHOD FOR HAZ METAL CHARPY SPECIMENS .14 FIGURE 3 6. FABRICATIONMETHOD FQR BASE METALTENSILE SPECMENS 15 FIGURE 3-7. FABRICATIONMETHOD FOR WELD METALTENSILE SPECIMENS .16 FIGURE 3-8. FABRICATIONMETHOD FOR HAZ METALTENSILE SPECIMENS ..... .17 FIGURE 4-1. SCHEMATIC OF MODEL FOR AZIMUTHALFLUX DISTRIBUTIONANALYSIS...................25 FIGURE 4-2. RELATIVE VESSEL FLUX VARIATIONWITH ANGULARPOSITION...................................26 FIGURE 4-3. RELATIVEVESSEL FLUX VARIATIONWITH ELEVATION. ....... .....27 FIGURE 5-1. BROWNS FERRY 2 UNHUVJ3IATED BASE METALIMPACT ENERGY..........................36 FIGURE 5-2. BROWNS FERRY 2 IRRADIATEDBASE METALIMPACTENERGY .37 FIGURE 5-3. BROWNS FERRY 2 Huber)IATED AND UNHNADIATEDBASE METALMPACT ENERGY38 FIGURE 5<. BROWNS FERRY 2 UNHNADIATEDBASE METALLATERALEXPANSION .-----------39 FIGURE 5-5. BROWNS FERRY 2 IRBADIATEDBASE METALLATERALEXPANSION.............................40 FIGURE 54 BROWNS FERRY 2 UNIRRADIATEDWELD METALIMPACT ENERGY.........""-"--".-""41 FIGURE 5-7. BROWNS FERRY 2 RADIATED WELD METALIMPACT ENERGY ..............42 FIGURE 5-8. BROWNS FERRY 2 IRRADIATEDAND UNHNADIATEDWELD METALIMPACT ENERGY43 FIGURE 5-9. BROWNS FERRY 2 UNHNADIATEDWELD METALLATERALEXPANSION ......................44 FIGURE 5-10. BROWNS FERRY 2 HulADIATED WELD METALLATERALEXPANSION..........................45 FIGURE 5-11. BROWNS FERRY 2 UNHGMDIATEDHAZ METALIMPACT ENERGY ................................46 5-12. BROWNS FERRY 2 RADIATED HAZ METALIMPACT ENERGY .47

'IGURE FIGURE 5-13. BROWNS FERRY 2 UNHNADIATEDHAZ METALLATERALEXPANSION........................48 FIGURE 5-14. BROWNS FERRY 2 IRRADIA'HZ)HAZ METALLATERALEXPANSION........-- --------49 FIGURE 6-1. TYPICALENGINEERING STRESS-STIVJN CURVE FOR HGVd3IATED RPV MATERIALS.54 FIGURE 6-2. FRACTURE LOCATION, NECKING BEHAVIORAND FRACTURE APPEARANCE FOR IRRADIA'IZDBASE METALTENSILE SPECIMENS 55 FIGURE 6-3. FRACTURE LOCATION, NECKING BEHAVIORAND FRACTURE APPEARANCE FOR IRRADIATEDWELD METALTENSILE SPECIMENS .56 FIGURE 6P. FRACTURE LOCATION, NECKING BEHAVIORAND FRACTIJRE APPEARANCE FOR IRRADIA'IZDHAZ METALTENSILE SPECIMENS 57

0 GEiiE-B 1100639-01 FIGURE 7-1. PRESSURE TEST P-T CURVES FOR UNIT 2 .68 FIGURE 7-2. HEAT-UP/COOLDOKVNP-T CURVES FOR UNIT ~ .69 FIGURE 7-~. CORE CRITICAL OPERATION P-T CURVES FOR UNIT 2 ....... 70 FIGURE 7A. COMBINED P-T CURVES FOR UNIT ~ .71 FIGURE 7-5. BROWNS FERRY 2 ART VERSUS EFPY FOR PLATE AND %VELD MATERIALS..... ....72

GENE-B 1100639-01 ABSTRACT The surveillance capsule at 30'zimuth location was removed from the Browns Ferry Unit 2 reactor in Fall 1994. The capsule contained flux wires for neutron fluence measurement and Charpy and tensile test specimens for material property evaluation. The flux wires were evaluated to determine the fluence experienced by the test specimens. Charpy V-Notch impact testing and uniaxial tensile testing were performed to establish the properties of the irradiated surveillance materials. Unirradiated Charpy and tensile specimens were tested as well to obtain the appropriate baseline data.

The irradiated Charpy data for the plate and weld specimens were compared to the unirradiated data to determine the shift in Charpy curves due to irradiation. The results are within the predictions of the Regulatory Guide 1.99 Revision 2.

I The irradiated tensile data for the plate and weld specimens were compared to the unirradiated data to determine the efFect of irradiation on the stress-strain relationship of the materials. The changes shown in the materials were consistent with the irradiation embrittlement efFects shown by the Charpy specimens.

The flux wire results, combined with the lead factor determined from the last fuel cycle, were used to estimate the 32 EFPY fluence. The resulting estimate was about 43% lower than the previous estimate used to develop pressure-temperature curves. Therefore, new pressure-temperature curves were generated.

"vlu-

GENE-B 1100639-01 ACKNOWLEDGMENTS The author gratefully acknowledges the efforts of other people towards completion of the content.'f this report.

Charpy testing was completed by G. P. Wozadlo and G. E. Dunning. Tensile specimen testing was done by S. B. Wisner. Flux wire testing was performed by R. M. Kruger and R. D. Reager. Lead Factor calculations were performed by D. R. Rogers.

I G~-B 110063 9-01

1. INTRODUCTION Pwt of the eQ'ort to assure reactor vessel integrity involves evaluation of the fracture toughness of the vessel ferritic materials. The key values which characterize a material's fracture toughness are the reference temperature 0

of nil-ductility transition (RTt,~i) and the upper shelf energy (USE). These are de6ned in 10CFR50 Appendix G [1] and in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI [2]. These documents contain requirements used to establish the pressure- temperature operating limits which must be met to avoid brittle fracture.

Appendix H of 10CFR50 [3] and AS'185-66 [4] establish the methods to be used for surveillance of the Browns Ferry Unit 2 reactor vessel materials. Capsule removal and testing were don per the requirements of ASTM E185-82 [6] to the extent practical. The first vessel surveillance specimen capsule required by 10CFR50 Appendix H [3] was removed from Unit 2 in Fall 1994. The irradiated capsule was sent to the GE Vallecitos Nuclear Center (VNC) for testing. The surveillance capsule contained Qux wires for neutron Qux monitoring and Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials &om or representative of the vessel materials nearest the core (beltline). The impact and tensile specimens were tested to establish properties for the irradiated materials. Unirradiated Charpy and tensile specimens were sent from site to GE Vallecitos Nuclear Center (VNC) and tested using the same testing methods.

The results of the surveillance specimen testing are presented in this report, as required per 10CFR50 Appendices G and H [1 2 3]. The irradiated material properties are compared to the unirradiated properties to determine the efFect of irradiation on the tensile properties, through tensile testing, and on material toughness, through Charpy testing. Flux wire results and updated lead factor analyses are used to determine the need for changes to the pressure-temperature (P-T) curves.

GENE-B 1100639-01

2. SUii'IMARYAND CONCLUSIONS 2.1 SUiziMARY OF RESULTS The 30'zimuth surveillance capsule was removed and shipped to VNC. The fiux wires, Charpy V-Notch and tensile test specimens removed from the capsule were tested according to ASTM E185-82 [6]. The methods and results of the testing are presented in this report as follows:

a. Section 3: Surveillance Program Background

b. Section 4: Peak RPV Fluence Evaluation

c. Section 5: Charpy V-Notch Impact Testing

d. Section 6: Tensile Testing Section 7: Development of Operating Limits Curves The signi6cant results of the evaluation are below:

The 30'zimuth position capsule was removed from the reactor. The capsule contained 9 flux wires: 3 copper (Cu), 3 iron (Fe), and 3 nickel (Ni). There were 36 Charpy V-Notch specimens in the capsule: 12 each of plate material, weld material and heat afFected zone (HAZ) material. The 8 tensile specimens removed consisted of 3 plate, 2 weld, and 3 HAZ metal specimens.

The chemical compositions of the beltline materials were determined from data obtained from GE QA records. The copper (Cu) and nickel (Ni) contents were determined for all beltline heats of plate material. The values for the limiting beltline plate are 0.16% Cu and 0.52% Ni. The limiting beltline weld values are 0.28% Cu and 0.35% Ni.

The purpose of the flux wire testing was to determine the neutron flux at the surveillance capsule location. The Qux wire results show that the Quence (fiom E >1 MeV Qux) received by the surveillance specimens was 1.52x101 n/cm at removal.

GENE-B 110063 9-01

d. A neutron transport computation was performed, based on the performance of the last fuel cycle. Relative flux distributions in the azimuthal and axial directions were developed. The lead factor, relating the surveillance capsule flux to the peak inside surface flux, was 0.98.

The surveillance Charpy V-Notch specimens were impact tested at temperatures selected to define the transition of the fracture toughness curves of the plate, weld, and HAZ materials. Measurements were taken of absorbed energy, lateral expansion and percentage shear. From absorbed energy and lateral expansion curve-fit results (for plate and weld metal only), the values of USE and of index temperature for 30 ft-lb, 50 ft-lb and 35 mils lateral expansion (MLE) were obtained (see Table 5-4). Fracture surface photographs of each specimen are presented in Appendix A.

f. The curves of irradiated Charpy specimens and unirradiated Charpy specimens established the 30 ft-lb index temperature irradiation shift and the decrease in USE.

~ The surveillance plate material showed a measured 38'F shift and a 6 ft-lb decrease (4% decrease) in USE. The weld material showed a 1'F shift and essentially no decrease in USE.

g. The measured shifts of 38 F for plate and 1'F for weld, for a fluence of 1.52x1017 n/cm2, were within their respective Reg. Guide 1.99 [7] range predictions (dRT~+2a) of -20'F to 48'F, and -39'F to 73'F.

The irradiated tensile specimens were tested at room temperature (70'F), reactor operating temperature (550'F). The results in comparison to unirradiated data were tabulated (see Tables 6-3 and 6-4) for each specimen including yield and ultimate tensile strength, uniform and total elongation, and reduction of area. The results generally showed increasing strength and decreasing ductility, consistent with expectations for irradiation embrittlement.

The 32 EFPY fluence prediction of 6.05x1017 n/cm2, based on the flux wire test and lead factor results presented here, was about 43% lower than that previously established (1.1x1018 n/cm ) for development of P-T curves.

I 0

r GENE-B1100639-01 As a part of the development of the pressure-temperature (P-T) operating limits curves, the adjusted reference temperature (ART = initial RT<~Y+ dZTq~T+

Margin) was predicted for each beltline material, based on the methods of Reg.

Guide 1.99. The ARTs for the limiting material, weld ESW, at 32 EFPY is 92.1'F.

The beltline material USE values at 32 EFPY were predicted using the methods of Reg. Guide 1.99, with initial beltline USE values based generic USE values (see Table 7-3). It is expected that the actual 32 EFP Y USE will be in excess of 50 ft-lbs for all beltline plated and welds. In addition, the results of the USE for the surveillance materials show that the BWROG equivalent margin 'esting analysis is applicable.

P-T curves were developed for three reactor conditions: pressure test (Curve A),

non-nuclear heatup and cooldown (Curve B), and core critical operation (Curve C). The curves are valid for 32 EFPY of operation. The beltline curve is more limiting for curve A. For curve B and curve C, the non-beltline curves are limiting for pressures less than approximately 1100 psig. The P-T curves are shown in Figures 7-1 through 7-3. Figure 7-4 shows the combined Curves A, B, and C P-T curves.

2.2 CONCLUSION

S The requirements of 10CFR50 Appendix G [1] deal basically with vessel design life conditions and with limits of operation designed to prevent brittle Gacture. However, based on the evaluation of surveillance testing results, and the associated analyses, the following conclusions are made:

a. The 30 ft-Ib shifts and decreases in USE measured were within Regulatory Guide 1.99 Revision 2 predictions.

b. The values of ART and USE for the reactor vessel beltline materials are expected to remain within limits in 10CFR50 Appendix G [1] for at least 32 EFPY of operation.

4

GENE-B 1100639-01 I

3. SURVEILLANCEPROGRAM BACKGROUND 3.1 CAPSULE RECOVERY The reactor pressure vessel (RPV) originally contained three surveillance capsules at 30; 120', and 300'zimuths at the core midplane. The specimen capsules are held against the RPV inside surface by a spring loaded specimen holder. Each capsule receives equal irradiation f

because c core symmetrv. During the Fail 1994 outage, the 30'ositioned capsule was removed. The capsule was cut from its holder assembly and shipped by cask to the GE Vallecitos Nuclear Center (VNC), where testing was performed.

Upon arrival at VNC, the capsules were examined for identification. The drawing number 117C406)G001 Part P6 isstamped on the Browns Ferry Unit 2 30'urveillance capsule basket.

The general condition of the basket as received is shown in Figure 3-1. The capsule contained three impact (Charpy) specimen capsules and four tensile specimen capsules. Each tensile specimen capsule contained two tensile specimens. Each Charpy specimen capsule contained 12 plate, weld or HAZ Charpy specimens and 3 flux wires (one iron, one copper, and one nickel) in a sealed helium environment.

3.2 RPV MATERIALSAND FABRICATIONBACKGROUND 3.2.1 Fabrication Histo The Browns Ferry 2 RPV is a 251 inch diameter BWR/4 design. Construction was performed by Ishikawajima-Harima Heavy Industries Co. (IHI) to the Summer 1965 Addenda of the 1965 edition of the ASME Code. The shell and head plate materials are ASME SA 302, Grade B, MOD. 1339 Class 1 low alloy steel (LAS). The nozzles and closure flanges are ASME SA 508 Class 2. The vessel plates were heat treated as follows:

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GENE-B 1100639-01 The base metal specimens were cut from Heat A0981-1. The test plate received the same heat treatment beltline plates, see Section 3.2.1. The Charpy specimens were removed from the test plate and machined as shown in Figure 3-3. Specimens were machined from the 1/4 T and 3/4 T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction). The base metal Charpy specimens from the surveillance capsule were stamped as shown in Figure 3-3; the stamp code is taken from GE Drawing Number 921D277.

a The weld metal and HAZ Charpy specimens were fabricated by welding together two piece of the surveillance test plate Heat C-2884 and C-2868. The two plates were electroslag-welded (BOW Weld Procedure WR-12-4) and heat treated the same as the core region plates.

The weld specimens and HAZ specimens were fabricated as shown in Figures 3-4 and 3-5, respectively. The base metal orientation in the weld and HAZ specimens'was longitudinal. The specimens were stamped on one end as shown in Figure 3-3; the stamp code is taken from GE Drawing Number 921D277.

3.3.2 Tensile S ecimens Fabrication of the surveillance tensile specimens is also described in the GE purchase specification f8]. The materials, and thus the compositions and heat treatments for the base, weld and HAZ tensiles are the same as those for the corresponding Charpy specimens. The specimens were stamped on one end as shown in Figure 3-6; the stamp code is taken from GE Drawing Number 921D276.

The base metal specimens were machined &om material at the 1/4 T and 3/4 T depth. The specimens, oriented along the plate rolling direction, were machined to the dimensions shown in Figure 3-6. The gage section was tapered to a minimum diameter of 0.250 inch at the center.

The weld metal tensile specimen materials were cut &om the welded test plates, as shown in Figure 3-7. The specimens were machined entirely from weld metal, scrapping material that might include base metal. The fabrication method for the HAZ tensile specimens is illustrated in Figure 3-8. The specimen blanks were cut &om the welded test plates such that the gage section minimum diameters were machined at the weld fusion line. The finished HAZ specimens are approximately half weld metal and half base metal oriented along the plate rolling direction.

GENE-31100639-01 TABLE 3-1 CHEMICALCOMPOSITION OF RPV BELTLINEMATERIALS E

Com osition bv Wei ht Percent'eat/Lot Identilication C Mn P S Si Ni Mo Cu No Lower Shell Plates:

6-127-14 C2467-2 0.20 1.36 0.008 0.013 0.20 0.52 0.47 0.16 6-127-15 C2463-1 0.21 1.33 0.008 0.015 0.16 0.48 0.47 0.17 6-127-17 C2460-2 0.21 1.29 0.012 0.014 0.17 0.51 0.45 0.13 Lower-Intermediate Shell Plates:

6-127-6 A0981-1 0.20 1.35 0.007 0.011 0.19 0.55 C.49 0.14 6-127-16 C2467-1 0.20 1.36 0.008 0.013 0.20 0.52 0.47 0.16 6-127-20 C2849-1 0.21 1.30 0.010 0.015 0.23 0.50 0.46 0.11 Surveillance Plate: A0981-1 see above for the plate with the same heat number Welds:

ES Weld 0.016 0.35 0.28 Axial'ircumferential D55733 0.08 1.70 0.014 0.005 0.40 0.65 0.45 0.09 Surveillance Weld 0.15 1.49 0.010 0.011 0.09 0.33 0.49 0.20 Data Rom the 92-01 response [9] except where noted.

'etter from J.Valente to T.R.Mcintyre [11]

G~-B 1100639-01 TABLE 3-2 MECHANICALPROPERTIES OF BELTLINEAND OTHER SELECTED RPV liATERIALS Initial ID. Heat RTmv Locati n 'hfo Number ~F Beltlinea & b.

Lower Shell Plates 6-127-14 C2467-2 -20'F 6-127-15'-127-17 C2463-1 -20'F C2460-2 O' Lower Intermediate 6-127-6 A0981-1 -10'F Shell Plates 6-127-16 C2467-1 -10'F 6-127-20 C2849-1 -10'F Welds:

Longitudinal ESW 10'F Circumferential D55 733 -40'F Non'-Beltlinea + b:

Head Dome B5524-2 +10 Top Head Flange AKU75 +10 Closure Head Segment C2426-2 +10 C2426-3 +10 C 1717-3 +10 C 1722-3 +10 Bottom Head Dome C-2669-2 +42 Bottom Head Upper Torus B-6747-1 +40 B-6776-2 +40 C-2369-1 +40 Jet Pump Nozzle 214484 +54 a Test data information from GE-NE-523-A65-0594

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GENE-B 110063 9-01

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Ci FIGURE 3-1. SURVEILLANCE CAPSULE HOLDER RECOVERED FROM BROWNS FERRY UNIT 2 GENE-B 1100639-01 Vessel Range Shell Course 5 MK-60 Upper Shell

~ Longitudinal Welde-Girth Welds ~

Shell Course 4 MK-16 Upper Intermediate Shell Intermediate Shell Shell Course 3 MK-59 Shell Course 2 MK-58 Plate Heats: 40981-1 Lower Intermediate Shell C2467-1 Core C2849-1 Beltline Region I

Shell Course 1 MK-57 Plate Heats: C2467-2 Lower Shell C2463-1 C2460-2 Bottom Head Enclosure FIGURE 3-2. SCHEMATIC OF THE RPV SHOWING IDENTIFICATIONOF VESSEL BELTLINE PLATES AND WELDS GENE-B I 100639-01 PLATE, HEAT A0981-1 STAMP 1.0S2 0.010 CODE NUMBER 2.'l65 0.015 0.01~.001R 45'+1'394a0.001 0.394%.001 FIGURE 3>>3. FABRICATIONMETHOD FOR BASE METALCHARPY SPECIMENS

GENE-B 110063 9-01 ROLUNG DIRECTION PLATE

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SCRAP THE WELD VESSEL WALL ROOT THICKNESS MATERIAL MACHINEDAS SHOWN ON BASE METAL CHARPY SPECIMEN FIGURE 4

FIGURE 3-4. FABRICATIONMETHOD FOR WELD METALCHARPY SPECIMENS GENE-B 1100639-01

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THICKNESS VESSEL WALL THICKNESS MACHINEAS SHOWN ON BASE METAL CHARPY SPECIMEN FIGURE VESSEL WALL THICKNESS p>> .1 8.

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,p(QQ 449 FIGURE 3-5. FABRICATIONMETHOD FOR HAZ METALCHARPY SPECIMENS 0

GENE-B 1100639-01 PLATE VESSEL WALL THICKNESS 1/4T r r/r OA375-14 UNC-2A BOTH ENDS 1.$ XH).005 GAGE LENGTH STAMP CODE 0.375R NUMBER iTYP) 1/2

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GAGE MARKS 30 deg+ Dl TV~P 0.~.02 1-1/4 REDUCED SECTION 3 1/16 D'OTES:

1. D ~ 02508).001 0 AT CENTER OF REDUCED SECTION 2 D'~ACTUAL "D 0+0.002TO0.005ATENDSOF REDUCED SECTION, TAPERING TO "D" AT CENTER FIGURE 3-6. FABRICATIONMETHOD FOR BASE METALTENSILE SPECIMENS L

GENE-31100639-01 WELD CUT WELD OUT OF TEST PLATE

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py+p SCRAP THE WELD ROOT MATERIAL IRM FIGURE 3-7. FABRICATIONMETHOD FOR WELD METALTENSILE SPECIMENS GENE-B 1100639-01 PLATE 6

+pa PLATE VESSEL WALL THICKNESS WELD DIMENSIONS AS SHOWN IN BASE METALTENSILE FIGURE SCRAP THE VESSEL WALL WELD ROOT THICKNESS MATERIAL FIGURE 3-8. FABRICATIONMETHOD FOR HAZ METALTENSILE SPECIMENS

GENE-B 110063 9-01

4. PEAK RPV FLUENCE EVALUATION Flux wires removed from the 30'apsule were analyzed. as described in Section 4.1, to determine flux and fluence received by the surveillance capsule. The lead factor. determined as described in Section 4.2. was used to establish the peak vessel fluence from the flux wire results.

Section 4.3 includes 32 EFPY peak fluence estimates.

4.1 FLUX WIRE ANALYSIS 4.l.l ~Pr cpu g The surveillance capsule contained 9 flux wires: 3 iron, 3 copper, and 3 nickel. Each wire was removed from the capsule, cleaned with dilute acid. weighed, mounted on a counting card, and analyzed for its radioactivity content by gamma spectrometry. Each iron wire was analyzed for Mn-54 content, each nickel wire for Co-58 and each copper wire for Co-60 at a calibrated 4-cm or 10-cm source-to-detector distance with 100-cc Ge(Li) and 170-cc Ge detector systems.

To properly predict the flux and fluence at the surveillance capsule from the activity of the fiux wires, the periods of full and partial power irradiation and the zero power decay periods were considered. Operating days for each fuel cycle and the reactor average power fraction are shown in Table 4-1. Zero power days between fuel cycles are listed as well.

From the flux wire activity measurements and power history, reaction rates for Fe-54 (n,p) Mn-54, Cu-63 (n,u) Co-60 and Ni-58 (n,p) Co-58 were calculated. The E I MeV )

fast flux reaction cross sections were determined Rom past testing at Browns Ferry 3 [10], also a 251 inch, 764 bundle plant, using multiple dosimeter and spectrum unfolding techniques. The cross sections for the iron, copper and nickel wires are 0.213 barn, 0.00374 barn and 0.274 barn, respectively. These values are consistent with other measured cross section functions determined at GE's Vallecitos Nuclear Center from more than 65 spectral determinations for BWRs and for the General Electric Test Reactor using activation monitors and spectral unfolding techniques.

These data functions are applied to BWR pressure vessel locations based on water gap (fuel to vessel wall) distances. The cross sections for E )0.1 MeV flux were determined &om the measured l-to-0.1 MeV cross section ratio of 1.6.

GENE-31100639-01 4.1.2 Results The measured activity, reaction rate and full-power flux results for the 30'urveillance capsule are given in Table 4-2. The E >1 MeV Qux values were calculated by dividing the wire reaction rate measurements by the corresponding cross sections, factoring in the local power I

history for each fuel cycle. The fluence result, 1.52xl017 n/cm (E >1 MeV) was obtained by multiplying the full-power flux value for copper, iron, and nickel by the operating time and full power fraction, shown in Table 4-1.

The accuracies of the values in Tables 4-2 for a 2cr deviation are estimated to be:

+ 5% for dps/g (disintegrations per second per gram)

+ 10% for dps/nucleus (saturated)

+ 20% for flux and fluence E >1 MeV

+ 20% for Qux and fluence E >0.1 MeV 4.2 DETERh'GNATION OF LEAD FACTOR The flux wires detect Qux the location of the surveillance capsule. The wires will reQect the power Quctuations associated with the operation of the plant. However, the Qux wires are not at the location of peak vessel Qux. Alead factor is required to relate the flux at the wires'location to the peak Qux. The lead factor is the ratio of the Qux at the surveillance capsule to the Qux at the peak vessel inside surface location. The lead factor is a function of the core and vessel geometry and of the distribution of power density and voids in the core. The lead factor was generated for the Browns Ferry geometry, using a typical fuel cycle to determine power shape and void distribution. The methods used to calculate the lead factor are discussed below.

4.2.1 Procedure Determination of the lead factor for the RPV inside wall was made using a combination of two separate two-dimensional neutron transport computer analyses. The Qrst of these established the azimuthal and radial variation of Qux in the vessel at the fuel midplane elevational.

-"19-

GENE-B 1100639-01 established the azimuthal and radial variation of Qux in the vessel at the fuel midplane elevational.

The second analysis determined the relative variation of fiux with elevation. The azimuthal and axial distribution results were combined to provide the ratio of flux, or the lead factor, between the surveillance capsule location and the peak flux locations.

The DORT computer program, which utilizes the discrete ordinates method to solve the Boltzmann transport equation in two dimensions, was used to calculate the spatial Qux distribution produced by a fixed source of neutrons in the core region. The azimuthal distribution was obtained with a model specified in (R,B) geometry, assuming eighth-core symmetry with reflective boundary conditions at 0'nd 45'. Calculations were performed using neut-,on cross-sections from a 26 energy group set, with angular dependence of the scattering cross-sections approximated by a third-order Legendre polynomial expansion.

A schematic of the (R,B) vessel model is shown in Figure 4-1. A total of 132 radial intervals and 90 azimuthal intervals were used. The model consists of an inner and outer core rey'on, the shroud, water regions inside and outside the shroud, and the vessel wall. The core region material compositions and neutron source densities were representative of conditions at an elevation 75 inches above the bottom of active fuel, which is near the elevation of the wires. Flux as a function of azimuth and radius was calculated in order to establish the azimuth of the peak Qux and its magnitude relative to the Qux at the wires'ocation of 30'.

The calculation of the axial flux distribution was performed in (R,Z) geometry, using a simplified cylindrical representation of the core configuration and realistic simulations ofthe axial variations of power density and coolant mass density. The core description was based on conditions near the azimuth angle of 25'here the edge of the core is closest to the vessel wall.

The elevation of the peak Qux was determined, as weH as its magnitude relative to the Qux at the surveillance capsule elevation.

4.2.2 Results The two-dimensional computations indicate the Qux to be a maximum 25.75'ast the RPV quadrant references (0', 90', etc.), at an elevation about 77 inches above the bottom of active fuel. The peak closest to the 30'ocation of the surveillance capsule removed is at 25.75',

as shown in Figure 4-2. The relative Qux distribution versus elevation is shown in Figure 4-3.

The calculated Qux at the capsule (R,B) position along the midplane was modified by an GENE-B 110063 9-01 position. Theresultingsurveillancecapsulefluxis 8.8x10 n/cm2-s. Thepeakfluxatvessel surface from the transport calculation, incorporating the axial adjustment factor obtained from the (R.Z) calculation is 9.0x108 n/cm2-s. Therefore the lead factor is 8.8/9.0=0.98.

The transport calculation of surveillance capsule flux. 8.8x108 n/cm"-s, is about 49%

higher than the dosimetry result of 5.9xl08 n/cm=-s. This is attributed to conservatism incorporated in the transport calculation model and may, in part. result from the use of nominal rather than as-built radius. A difference in vessel radius has little, ifany, effect on the calculated lead factor. since the difference would affect both capsule radius and vessel radius and would not significant;y alter the ratio of fluxes at the two locations.

The fracture toughness analysis is based on a 1/4 T depth flaw in the beltline region, so the attenuation of the fiux to that depth is considered. This attenuation is calculated according to Reg. Guide 1.99 requirements, as shown in the next section.

4.3 ESTIMATE OF 32 EFPY FLUENCE The inside surface fluence (fs~) at 32 EFPY is determined from the flux wire fluence for 8.2 EFPY of 1.52x1017 n/cm-", using the lead factor of 0.98. The time period 32 EFPY is based on 40-year operation at an 80% capacity factor. The resulting 32 EFPY fluence value at the peak vessel inside surface is:

fsurf = 1.52x10 7*(32/8.2)/0.98 fsurf = 6.05 x1017 n/cm2 The peak inside surface fluence of 6.05 x1017 n/cm2 is about 43% lower than that used in previous analyses (1.1x1018 n/cm2 ) [11]. Therefore, the previous numbers were quite conservative.

The 1/4 T fluence (f) is calculated according to the following equation &om Reg. Guide 1.99 [7]:

f= fs~e-0.24x) (4-1) where x = distance, in inches, to the 1/4 T depth.

GENE-B 1100639-01 For a vessel beltline lower-intermediate shell and lower shell of 6.13 inches thick, the corresponding depth x is 1.53 inches. Equation 4-1 evaluated for these values of x gives:

f = 0.6923 fs~, or f= 4.19x1017 nicm2 The impact of these revised fluences on the P-T curves is discussed in Section 7.

GENE-B 1100639-01 TABLE 4-1

SUMMARY

OF DAILYPO%'ER HISTORY Cvcle C cle Dates

'perating D~ss Full Power Fraction Days Between

~Ccles 7/20/74 - 3/18/78 1338 0.355 41 4/28/78 - 4/27/79 365 0.723

+4 6/1/79 - 9/30/80 488 0.759 31 11/1/80 - 7/3 1/82 638 0.784 229 3/18/83 - 9/15/84 548 0.759 2478 7/1/91 - 1/31/93 581 0.849 121 5/31/93 - 10/1/94 489 0.972 4447 (total) 0.743 (average)

TABLE 4-2 SURVEILLANCECAPSULE FLUX AND FLUENCE FOR IRRADIATIONFROM START-UP TO 10/I/94 Full Power Flux Fluence Fluence dps/g Bement Reaction Rate (n/cm2 s) (ll/cm2) (n/cm2)

Wire Element at end of Irrad>at>on Ld s/nucleus saturated E >1 MeV L >1 MeV E >0.1 MeV Iron 6.05E+04 1.23E-16 5.80E+08 1 49E.].17 2.39EH. I 7 Nickel 1.07E+06 1.67E-16 6 1 1E+08 1.57E+17 2.52 r<<17 Copper 5.62 EH 03 2.15 E-18 5.75 E+08 1.48E+17 2.37 E+17 Average 1.52E+17 2 43E+17 a Full power flux, based on thermal power of 3293 Mwt

• Average values of the tests reported.

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5. CHARPY V-NOTCH IMPACT TESTING The 36 Charpy specimens recovered from the surveillance capsule were impact tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials. In addition, unirradiated base, weld, and HAZ metal specimens recovered from the Browns Ferry site were tested for baseline data. Testing was conducted in accordance with ASTM E23-88 [12].

5.1 lMPACT TEST PROCEDURE The Vallecitos testing machine used for irradiated and unirradiated specimens was a Riehle 5'Iodel PL-2 impact machine, serial number R-89916. The pendulum has a maximum velocity of 15.44 ft/sec and a maximum available hammer energy of 240 ft-lb.

The test apparatus and operator were qualified using NIST standard reference material specimens. The standards consist of sets of high and low energy specimens, each designed to fail at a specified energy at the standard test temperature of -40'F. According to ASTM E23-88 [12],

the test apparatus averaged results must reproduce the NIST standard values within an accuracy of+5% or+1.0 ft-lb, whichever is greater. The qualification of the Riehle machine and operator is summarized in Table 5-1. The calibration tests are valid for one year.

Charpy V-Notch tests were conducted at temperatures between -80'F and 300'F. The cooling Quid used for both irradiated and unirradiated specimens tested at temperatures below 70'F was ethyl alcohol. At temperatures between 70'F and 200'F, water was used as the temperature conditioning Quid. The specimens were heated in silicon oil above 200'F. Cooling of the conditioning Quids was done by heat exchange with liquid nitrogen; heating was done'by an immersion heater. The bath of Quid was mechanically stirred to maintain uniform temperatures.

The fluid temperature was measured with a calibrated thermocouple. Once at test temperature, the specimens were manually transferred with centering tongs to the Charpy test machine and impacted within 5 seconds.

For each Charpy V-Notch specimen the test temperature, energy absorbed, lateral expansion, and percent shear were evaluated. In addition, for the irradiated specimens, photographs were taken of fracture surfaces. Lateral expansion and percent shear were measured

GENE-B 110063 9-01 according to specified methods [12]. Percent shear was determined using method number 1 of Subsection 11.2.4.3 of ASTM E23-88 [12], which involves measuring the length and width of the fracture surface and determining the percent shear value from Table 2 of ASTM E23-88 [12].

5.2 IMPACT TEST RESULTS Twelve Charpy V-Notch specimens each of irradiated base, weld, and HAZ material were tested at temperatures (-80'F to 300'F) selected to define the toughness transition and upper shelf portions of the &acture toughness curves. The absorbed energy, lateral expansion, and percent shear data are listed for each material in Table 5-2. Plots of absorbed energy data for base and weld materials are presented in Figures 5-2 and 5-7, respectively. Plots of absorbed energy and lateral expansion data for HAZ material, Figures 5-12 and 5-14, did not fit a hyperbolic curve because of the scatter in the data. Lateral expansion plots for base and weld materials are presented in Figures 5-5 and 5-10, respectively. The irradiated curves are plotted along with their corresponding unirradiated curves in Figures 5-3 and 5-8. The fracture surface photographs and a summary of the test results for each specimen are contained in Appendix A.

Twelve Charpy V-Notch specimens each of unirradiated base, weld and HAZ material were tested at temperatures (-80'F to 300'F) selected to define the toughness transition and upper shelf portion of the &acture toughness curves. The absorbed energy, lateral expansion,'nd percent shear data are listed in Table 5-3. Plots of absorbed energy data for base and weld metals are presented in Figures 5-1 and 5-6, respectively. Lateral expansion plots for base and weld metals are presented in Figures 5-4 and 5-9, respectively. Plots of absorbed energy and lateral expansion data for HAZ material, Figures 5-11 and 5-13, did not fit a hyperbolic curve because of the scatter in the data.

The plate and weld data sets are fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook [13]:

Y=A+B *TANH[(T- T0)/C],

where Y = impact energy or lateral expansion T = test temperature, and A, B, T0 and C are determined by non-linear regression.

GERS-B 1100639-01 The TANH function is one of the few continuous functions with a shape characteristic of low alloy steel fracture toughness transition curves. Typically the curve Gts were generated by setting both shelves free with a default lower shelf energy of 5 ft-lbs or lateral expansion of 4 mils.

5.3 IRBADIATEDVERSUS UNIKRADIATEDCRAPPY V-NOTCH PROPERTIES As a part of the RPV surveillance test program, extra Charpy V-Notch specimens were fabricated,and delivered to the site. Specimens were recovered from storage at the site and forwarded to GE for impact testing. This was done because GE had no records ofunirradiated baseline test results for this surveillance program.

The irradiated and unirradiated Charpy V-Notch data curves were used to estimate the values given in Table 5-4: 30 ft-lb, 50 ft-lb and 35 MLE index temperatures, and the USE for the sets of base and weld metal irradiated material data and for the base and weld metal unirradiated material data. Transition temperature shift values are determined as the change in the temperature at which 30 ft-lb impact energy is achieved, as required in ASTM E185-82 [6]. The resulting shifts in Charpy curves are discussed in the next section..

5.4 COMPARISON TO PREDICTED IRRADIATIONEFFECTS 5.4.1 Irradiation Shift The measured transition temperature shifts for the plate and weld materials were compared to the predictions calculated according to Regulatory Guide 1.99, Revision 2 [7]. The inputs and calculated values for irradiated shift are as follows:

Plate: Copper = 0.14%

Nickel = 0.55%

CF= 98 fluence = 1.52x1017 n/cm2 Reg. Guide 1.99 dRT~ = 14'F Reg. Guide 1.99 bRT~+ 2czg(34'F) = 48'F max., -20'F min.

Measured Shift = 37.9 'F

GENE-B 1100639-01 Weld: Copper = 0.20%

Nickel = 0.33%

CF = 120 fluence = 1.52x1017 n/cm2 Reg. Guide 1.99 MT~q = 17'F Reg. Guide 1.99 MT~z + 2'(56'F) = 73'F max., -39'F min.

Measured Shift = 1.3'F The weight percents of Cu and Ni are based on Table 3-1. CF shown above is the chemistry factors from Tables or 2 of Reg. Guide 1.99. The fluence factor is 0.141. The 1

measured shift of 37.9'F for the plate is above the predicted shifts of 14'F and mea"ured shift of 1.3'F for the weld is below the predicted shift of 17'F. The measured shiQs for the plate and weld are within the bounds (-20'F to 48'F for the plate material and -39'F to 73'F for the weld material; respectively) of the Reg. Guide 1.99 uncertainty of 2a.

5.4.2 Change in USE Using the copper and fluence data above with Figure 2 of Reg. Guide 1.99, decreases in USE of 9% are predicted for the plate and decreases in USE of 13% are expected for the weld.

i The measured decrease in the USE value of 4% for the plate is below the predicted value. The weld material shows essentially no change in the USE value, which is less than the 13% decrease in USE predicted by the Reg. Guide 1.99.

GENE-B 1100639-01 TABLE 5-1 VALLECITOS QUALIFICATIONTEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS Test Energy Acceptable Specimen Temperature Absorbed Range Vallecitns HH-40 229 Alcohol -40 75.0 Riehle hfachine HH-40 384 Alcohol Q0 74.5 (tested 6/28/94) HH-40 980 Alcohol -40 70.5 HH-40 1152 Alcohol -40 72.5 HH-40 1172 Alcohol -40 ~7 Average 73.5 74.9+ 3.7 pass LL-39 080 Alcohol -40 13.5 LL-39 095 Alcohol -40 13.0 LL-39 631 Alcohol -40 13.5 LL-39 775 Alcohol -40 13.5 LL-39 930 Alcohol -40 7

Average 13.3 13.2+ 1.0 pass V

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GENE-B 1100639-01 TABLE 5-2 IRRADIATEDCHA3G'Y V-NOTCH IMPACT TEST RESULTS Test Fracture Lateral Percent Shear Temperature Energy Expansion (Method 1)

Identification ~oF ~ft-1b ~mils Base: ESC -80 10.5 10.0 3 Heat A0981-1, ESY -40 17.0 13.5 ll Longitudinal, E7Y -20 33.0 30.5 13 f=1.52xl017 n/cm~ E7K 0 38.5 33.0 19 E71 40 60 50 40 n/cm'pecimen E7D 60 82.5 61.0 59 E64 80 94.5 70.0 68 ESU 100 121.0 91.0 85 E72 120 120.5 88.0 100 Esl 160 130.0 91.0 100 QE57 200 136.0 94.0 100 ESS 300 131.5 88.0 100 Weld: EB7 -80 2.0 5.0 2 Heats D55733 EBS -40 13.0 12.5 4 f=1.52x1017 EBK -20 37.5 31.0 9 EAP 0 50.0 42.0 15 EBD 20 59.5 52.0 22 EBB 40 59.5 50.0 30 EB1 80 59.0 52.0 42 EAM 100 76.5 64.5 50 EBE 120 87.0 65.0 68 EB4 160 107.0 87.0 100 EB2 200 107.5 84.5 100 EBA 300 113.0 88.5 100 HAZ: ED 6 -80 3.5 6.0 1 f=1.52x10 n/cm'J3 -60 37.0 30.0 12 EEY -40 54.0 44.0 24

-20 30.0 22.5 7 0 43.5 36.5 19 EJS 20 106.0 81.5 65 EJC 40 93.5 67.0 48 EJ1 60 107.5 86.0 75 80 82.0 73.0 60 120 97.5 78.0 100 200 107.5 82.0 100 300 143.0 92.0 100

GENE-B 1100639-01 TABLE 5-3 UMFQhG)IA.TED CXKARPY V-NOTCH EVlPACT TEST RESULTS Test Fracture Lateral Percent Shear Specimen Temperature Energy Expansion (Method 1)

Identification ~f% ~fi-Ib ~mils Base: ESJ -80 8.5 5.5 2 Heat A0981-1 E7A -60 17.5 14 9 Longitudinal E61 -40 ""35.5 29 17 E66 -20 40 37 19 E7M 0 97 69 47 E56 20 68 56 37 E6U 40 73 56 47 E76 80 104.5 77 86 E77 100 137 89 100 E7L 120 134.5 93.5 100 ESE 200 146.5 90 100 E6T 300 133 84 100 Weld: ED6 -80 3.5 6 1 Heats D55733 EJ3 -60 37 30 12 EEV'DB

-40 54 44 24

-20 30 22.5 7 EJJ 0 43.5 36.5 19 EJS 20 106 81.5 67 EJC 40 93.5 67 48 EJl 60 107.5 86 76 EDC 80 82 73 60 EJB 120 97.5 78 100 EJD 200 107.5 82 100 EEC 300 143 92 100 ED4 -80 13 11.5 3 EDD -60 44 34.5 12 EE1 -40 53 42.5 25 ED7 -20 25.5 24.5 30 EE7 0 104.5 79 55 EDE 40 120.5 84 74 EJ4 60 121.5 74.5 84 EEB 80 139.5 88.5 100 EES 100 130 88 100 ED2 120 121 92 100 EDL 200 126.5 88 100 EDM 300 110.5 89 100 GENE-B 1100639-Ol TABLE 5-4 SIGNIFICANT RESULTS OF IRRADIATEDAND UNIRRADIATED CHARPY V-NOTCH DATA Index Index Temperature Temperature Index Upper ShelP

('F) ('F) Temperature Energy

~Q~e F&lfiJh ~F= >~1 DfLU)

PLATE: Heat A0981-1, Longitudinal f=1.52x1017

-48.4 -14.'1 -25.2 n/cm'nirradiated 141.8/92.1 Irradiated 82 '>lEU.

Difference 37.9 35.9 33.4 6.3/4 0 (4%)

Reg. Guide 1.99, Rev 2 dRTgprb: 14 1.99, Rev 2% Decrease in USE: (9%)

Reg. Guide 1.99, Rev 2 (d+2a)": -20 to 48 WELD: Heat D55733 f=1.52x1017

-26.9 -7.7 n/cm'nirradiated 10.9 112.0 Irradiated 2RE Difference 1.3 15.9 10.5 3 3 (-3%)

Reg. Guide 1.99, Rev 2 dRT>prb: 17 1.99, Rev 2% Decrease in USEc: (13%)

Reg. Guide 1.99, Rev 2 (b+2cz)h: -39 to 73 a USE values &om Longitudinal/Transverse oriented Charpies; values are equal for weld metal.

Longitudinal USE from data shown in Figure 5-2.

Transverse plate USE is taken as 65% of the longitudinal USE, per USNRC MTEB 5-2 [16].

" Determined in section 5.4.1 c See section 5.4.2

UN IRRADIATED CHARP Y Base Energy 160 140 120 100 So 8 60 40 20 I

-200 -100 0 100 200 300 4oo Test Temperature, 'F C)

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6. TENSILE TESTING Eight round bar tensile specimens were recovered &om the surveillance capsule and six were tested. Uniaxial tensile tests were conducted in air at room temperature (70'F)'and RPV operating temperature (550'F). Six unirradiated specimens, sent from the Browns Ferry site to GE-~K San Jose, were tested at the same temperatures. The tests were conducted in accordance with ASTM ES-89 [14].

6.1 PROCEDURE All tests were conducted using a screw-driven Instron test frame equipped with a 20-kip load cell and special pull bars and grips. Heating was done with a Satec resistance clamshell furnace centered around the specimen load train. The test temperature was monitored and controlled by a chromel-alumel thermocouple spot-welded to an Inconel clip that was Giction-clipped to the surface of the specimen at its midline. Before the elevated temperature tests, a profile of the furnace was conducted at the test temperature of interest using an unirradiated steel specimen of the same geometry. Thermocouples were spot-welded to the top, middle, and bottom of a central 1 inch gage of this specimen. En addition, the clip-on thermocouple was attached to the midline of the specimen. When the target temperatures of the three thermocouples were within+5'F of each other, the temperature of the clip-on thermocouple was noted and subsequently used as the target temperature for the irradiated specimens.

Alltests were conducted at a calibrated crosshead speed of 0.005 in/min until well past yield, at which time the speed was increased to 0.05 inch/min until fi'acture. Crosshead displacement was used to monitor specimen extension during the test.

The test specimens were machined with a minimum nominal diameter of 0.250 inch at the center of the gage length. The yield strength (YS) and ultimate tensile strength (UTS) were calculated by dividing the measured area (0.0491 in>) into the 0.2% offset load and into the maximum test load, respectively. The values listed for the uniform and total elongations were obtained Rom plots that recorded load versus specimen extension and are based on a 1.5 inch gage length. Reduction of area (RA) values were determined Rom post-test measurements of the necked specimen diameters using a calibrated blade micrometer and employing the following formula:

RA = 100% * (Ao - Ag/A GENE-B 1100639-01 After testing, each broken specimen was photographed end-on. showing the fracture surface, and lengthwise, showing the fracture location and local necking behavior.

6.2 RESULTS Irradiated tensile test properties of Yield Strength (YS), Ultimate Tensile Strength (UTS), Reduction of Area (RA), Uniform Elongation (UE), and Total Elongation (TE) are presented in Table 6-1; all but UE are presented in Table 6-2 for unirradiated specimen's. A stress-strain curve for a 550'F base metal irradiated specimen is shown in Figure 6-1. his curve is typical of the stress-strain characteristics of all the tested specimens. The surveillance materials generally follow the trend of decreasing properties with increasing temperature.

Photographs of the fracture surfaces and necking behavior are given in Figures 6-2 through 6-4.

6.3 IRRADIATEDVERSUS UNIRIVQ)IATEDTENSILE PROPERTIES Unirradiated tensile test data was tested to provide direct comparison with the irradiated data at room temperature, shown in Table 6-3. The unirradiated and irradiated plate and weld data at 550' was compared to determine the irradiation effect, shown in Table 6-4. The trends of increasing YS and UTS and of decreasing TE and for the weld decreasing RA, characteristic of irradiation embrittlement, are seen in the data.

GEiiE-81100639-01 TABLE 6-1: TENSILE TEST RESULTS FOR IRRADIATEDRPV MATERIALS Test Yielda Ultimate Uniform Total Reduction Specimen Temp. Strength Strength Elongation Elongation of Area N~1~e ~O ~k~~ MiC ~0/ ~0/ ~0/

Base: EKA 70 71.2 92.5 9.3 19.5 71.4 EKJ 550 68.9 90.1 7.6 16.8 72.2 Weld: EL1 70 72.4, 92.2 9.0 18.7 68.7 ELC 550 67.5 87.0 7.3 15.0 '1.2 HAZ: EMB 70 70.9 97.6 8.3 17.5 64.5 EM3 550 65.9 86.8 7.0 14A 63.9 a Yield Strength is determined by 0.2% offset.

TABLE 6-2: TENSILE TEST RESULTS FOR UNIRRADIATEDRPV MATERIALS Test Yielda Ultimate Uniform Total Reduction Specimen 19umh:c.

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~

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70 dml Strength 66.9 Strength lksD 88.9 Elongation Elongation

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19.7 of Area 70.3 EKK 550 60.6 83.3 17.0 67.9 Weld: ELB 70 64.2 84.4 20.7 70.5 ELA 550 62.3 81.9 15.1 62.5 HAZ: EM2 70 64.6'50 84.9 16.3 68.3 EM7 63.1 83.9 13.9 64.6 a Yield Strength is determined by 0.2% offset.

GEiiE-B 110063 9-01 TABLE 6-3 COMPARISON OF UNIRRADIATEDAND IRRADIATEDTENSILE PROPERTIES AT ROOM TEMPERATURE Yield Ultimate Strength Total Elongation Reduction of Strength ~o/ Area

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Base: Unirradiated 66.9 88.9 19.7 70.3 Irradiated 71.2 92.5 19.5 71.4 Difference a 6.4% 4.0 -1.0% 1.6%

Weld: Unirradiated 64.2 84.4 20.7 70.5 Irradiated 72.4 92 2 18.7 68.7 Difference a 12.8% 93% -9.7%

a Difference = [(Irrad. - Unirrad.)/Unirrad.]

• 100%

TABLE 6A COMPARISON OF< UNIRRADIATE<DAND ItuMDIATEDTENSILE PROPERTIES AT 550 F Yield Strength Ultimate Strength Total Elongation Reduction of Area

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Base: Unirradiated 60.6 83.3 17.0 67.9 Irradiated 68.9 90.'1 16.8 72.2 Difference a 13.7% 8.2% 1.2% 6.33%

Weld: Unirradiated 62.3 81.9 15.1 62.5 Irradiated '7.5 87.0 15.0 61.2 Difference a 8.3% 62% -0.7% -2.1%

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~k!'.

P ELC 550oF I

FIGURE 6-3. FRACTURE LOCATION, NECKING BEHAVIOR AND FRACTURE APPEARANCE FOR IRRADIATEDWELD METALTENSILE SPECIMENS GENE-B 110063 9-01 Sg.

I ~

)

70'F EM3 550'F FIGURE 6A. FRACTURE LOCATION, NECKING BEHAVIOR AND FRACTURE APPEARANCE FOR IRRADIATEDHAZ METALTENSILE SPECIMENS

GENE-B 110063 9-01

7. DEVELOPMENT OF OPERATING LMITS CURVES P-T cur ves for Unit 2 were previously developed in GE report 523-A65-0594 [15].

Therefore. only the aspects of the curves which have changed, as a result of the testing presented here and as a result of ASME Code changes are discussed below.

7.1 BACKGROUND

The revised fluence value in Section 4 (6.05x10'/cm ), which is about 43% lower the fluence used in the previous report (1.1x10" n/cm ), is used in this section to revise the adjusted reference temperatures (ARTs), which are subsequently used to revise the beltline P-T curves.

The P-T curve revision includes consideration of the change to the aHowable &acture toughness equation in ASME Code Section XI, Appendix G, which occurred in 1992. The coefBcient 1.233 in the KIR/Kla equation in Figure G-2210-1, became 1.223. The result of the revision is an increase of about 1/2'F to the calculated temperature for a given pressure on the P-T cur ves (i.e., all curved portions of the P-T curves shift 1/2'F to the right).

7.2 NON-BELTLINEREGIONS The non-beltline Curve B curves are developed for two regions: the upper vessel region, governed by the jet pump nozzle limits, and the bottom head region, governed by the bottom head dome limits. Table 3-2 has the limiting initial RT~r values which are: 54'F for the jet pump nozzle and 42'F for the bottom head dome. The 1/2'F adjustment was made to the curved portions of the non-beltline curves, but not to the straight line and step portions, which are based on 10CFR50 Appendix G.

Although bottom head Curve B is not limiting, it is included in Figure 7-2, as there may be transients where the bottom head is cooler than the upper vessel regions.

GENE-B 1100639-01 7.3 CORE BELTLZNE REGION C

..he decreased fluence has an impact on the beltline P-T curves, by decreasing the ARTs of the beltline plates and welds. Figures 7-1 through 7-4 show the beltline curves at 32 EFPY.

Table 7-1 shows the beltline curve data points. As with the non-beltline curves, the 1/2'F adjustment was made to the curved portions of the beltline curves.

7.4 EVALUATIONOF IRRADIATIONEFFECTS The impact on adjusted reference temperature (ART) due to irradiation in the beltline materials is determined according to the methods in Reg. Guide 1.99 [7], as a function of neutron fluence and the element contents of copper (Cu) and nickel (Ni). The speciflc relationship &om Reg. Guide 1.99 [7] is:

ART = Initial RT~r+ ZEST~~+ Margin (7-1) where:

~T~ [CF]*$0.28 - 0.10 log f) (7-2)

Margin=2 (crP+ag ) (7-3)

CF = chemistry factor 6'om Tables 1 or 2 of Reg. Guide 1.99 [7],

f= 1/4 T fluence (n/cm2) divided by 1019, cq = standard deviation on initial RT~q, ag = standard deviation on MT~z, is 28'F for welds and 17'F for base material, except that vg need not exceed 0.50 times the MT~q value.

Once two sets of surveillance capsule data are available, the CF values in Reg.

Guide 1.99 [7] can be modifled to reflect the results. However, this is only the &st set of surveillance data &om Unit 2, so only the results of the flux wire tests are factored into beltline ART calculations.

Each beltline plate and weld BATED~ value is determined by multiplying the CF 6'om Reg. Guide 1.99, determined for the Cu-Ni content of the material, by the fluence factor for the EFP Y being evaluated. The Margin term and initial RT~q are added to get the ART ofthe GENE-B 1100639-01 material. The 32 EFPY ART values are shown in Table 7-2. Results for all of the beltline plates and the electroslag weld are shown.

7.4.1 .ART Versus EFPY The results in Table 7-2 show that the most limiting beltline plate is C2467-1 at 32 EFPY. The resulting ARTs at 32 EFPY are 49.7'F for the plate and 92.1'F for the weld.

Figure 7-5 shows the ART as a function of EFPY.

7.4.2 U er Shelf Energy at 32 EFPY Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy (USE) of the beltline materials. The USE must be above 50 ft-Ib at all times during plant operation, assumed here to be up to 32 EFPY. According to the BAW-1845 report the initial USE ofthe plates was not tested during fabrication, as there was no requirement to do so at that time.

Therefore, USE was determined for surveillance material plate and the same USE was applied to corresponding vessel plate material. For the other plates a generic USE value was estimated based on four surveillance plate material USEs. Calculations of 32 EFPY USE, using Reg. Guide 1.99 methods, are summarized in Table 7-3.

The equivalent transverse USE of the plate material is taken as 65% of the longitudinal USE, according to USNRC MTEB 5-2 [16]. Although the plate surveillance data show the decrease in USE to be considerably less than the prediction for the corresponding copper content (see Table 5-4), the USE decrease prediction values &om Reg. Guide 1.99 were used for the beltline plates in Table 7-3.

'ccording to the BAW-1845 report the weld metal initial USE values were determined Rom a generic USE value based on three surveiHance weld values. Unlike the plate, the weld metal USE has no transverse/longitudinal correction, because weld metal has no orientation e6ect. The weld surveillance data also show the decrease in USE to be considerably less than the prediction for the corresponding copper content, however, the USE 'decrease prediction values lrom Reg. Guide 1.99 were still used in Table 7-3.

Based on the results in Table 7-3, it is expected that the beltline materials will have USE values above 50 ft-lb at 32 EFPY, as required in 10CFR50 Appendix G [1]. Since USE and ART

GENE-31100639-01 requirements are met, irradiation e6ects are not severe enough to necessitate additional analyses or preparations for RPV annealing before 32 EFPY. Moreover, TVA is a participant in a BWR Owners'roup program to perform analyses to demonstrate equivalent margin [17j in cases as low as 35 ft-lb. Tables B-1 and B-2 in Appendix B show a decrease in surveillance plate and weld USE less than what is predicted in RG 1.99 and that the conclusions of the equivalent margin analysis are fully applicable.

7.5 OPERATING LIMITS CURVES VALIDTO 32 EFPY Figures 7-1 through 7-3 show P-T curves valid to 32 EFPY. The P-T curves are developed by considering the requirements applicable to the non-beltline, beltline and closure flange regions. The beltline curve is more limiting for curve A. For curve B and curve C, the non-beltline curves are limiting for pressures less than approximately 1100 psig. Curve B for the bottom head has been included to provide the appropriate limits for any transients where some bottom head stratiflcation might occur.

tt I'I

GENE-B 110063 9-01 TABLE 7-1 BROWNS FERRY 2 P -T CURVE VALUES

+ + + + + + > + + '+ + ~ + + + > + + '+ + + '+ + 4 + + + + + + + + + + ++++++++++++++++++++4++4+++++

REQUIRED T EM PE RATU RES + + +

32 EFPY NON- BOTTOM 32 EFPY UPPER 32 EFPY NON-PRESSURE BELTLINE BELTLINE HEAD BELTLINE VESSEL BELTLINE BELTLINE CURVE A CURVE A CURVE B CURVE B CURVE B CURVE C CURVE C 0 82.0 82.0 82.0 10 82.0 82.0 82.0 2Q 82.0 82.0 82.0 30 82.0 82.0 82.0 4Q 82.0 82.0 94.6 50 82.0 82.0 107.6 60 82.0 82.0 118.6 70 82.0 88.1 128.1 80 82.0 96.3 136.3 90 82.0 103.3 143.3 100 82.0 109.4 149.4 110 82.0 115.0 155.0 120 82.0 119.9 159.9 130 82.0 124.7 164.7 140 82.0 129.3 169.3 150 82.0 133.6 173.6 160 82.0 137.5 177.5 170 82.0 140.9 180.9 180 82.0 143.9 183.9 190 82.0 146.7 186.7 200 82.0 149.4 81.2 189.4 210 82.0 152.1 91.4 192.1 220 82.0 154.6 100.4 194.6 230 82.0 157.0 108.3 197.0 240 82.0 159.3 115.3 199.3 250 82.0 161.5 121.8 201.5 260 82.0 163.6 127.6 203.6 270 82.0 165.6 133.1 205.6 280 82.0 167.6 138.1 207.6 290 82.0 169.5 142.8 209.5 300 82.0 107.2 171.3 147.2 211.3 310 82.0 114.3 173.1 154.3 213.1 312.5 82.0 115.3 173.5 155.3 213.5 312.5 112.0 115.3 173.5 155.3 213.5 320 112.0 118.2 174.8 158.2 214.8 330 112.0 121.8 176.4 161.8 216.4 340 112.0 125.2 178.0 165.2 218.0 350 112.0 128.5 179.6 168.5 219.6 360 112.0 131.6 181.1 171.6 221.1 370 112.0 134.6 182.6 174.6 222.6 GENE-B 1100639-01 Table 7-1 Browns Ferry 2 P - T Curve Values (Continued) 1 1 0 >i* 1 1 0 8 II 1 0 4 ~ 0 0 I 0 0 to 0 0 ~ 1 0 4I 4t 0 1 1 4 REQUIRED TEMPERATURES 0 0 0 0 0 0 lit 0 '4i 0 0 0 1 0 0 t 32 EFPY NON- BOTTOM 32 EFPY UPPER 32 EFPY NON-PRESSURE BELTLINE BELTLINE HEAD BELTLINE VESSEL BELTLINE BELTLINE CURVE A CURVE A CURVE B CURVE B CURVE B CURVE C CURVE C 380 112.0 137.4 184.1 177.4 224.1 390 112.0 140.2 185.6 180.2 225.6 4CO 112.0 72.6 142.8 187.1 182.8 227.1 410 112.0 81.6 145.4 188.6 185A 228.6 420 112.0 88.6 147.8 190.0 1873 230.0 430 112.0 94.6 150.2 191.4 190.2 231.4 440 112.0 99.6 152.5 192.8 192.5 232.8 450 112.0 103.6 154.7 194.1 194.7 234.1 460 112.0 107.1 156.8 195.4 196.8 235.4 470 '112.0 110.2 158.9 196.7 198.9 236.7 480 112.0 113.1 160.9 197.9 200.9 237.9 490 112.0 115.9 162.9 199.1 202.9 239.1 500 87.0 112.0 118.6 164.8 200.3 204.8 240.3 510 91.2 112.0 121.3 166.7 201.4 206.7 241.4 520 95.2 112.0 123.9 168.5 202.5 208.5 242.5 530 99.0 112.0 126.5 170.3 203.6 210.3 243.6 540 102.5 112.0 129.0 172.0 204.6 212.0 244.6 550 105.9 112.0 131.4 173.7 205.6 213.7 245.6 560 109.2 112.0 133.7 175.3 206.6 215.3 246.6 570 112.3 112.0 135.9 176.9 207.5 216.9 247.5 580 115.2 112.0 138.0 178.5 208.4 218.5 248.4 590 118.1 112.0 139.8 180.0 209.3 220.0 249.3 600 120.8 112.0 141.6 181.5 210.1 221.5 250.1 610 123.4 112.0 143.3 182.9 210.9 222.9 250.9 620 126.0 114.0 145.1 184.4 211.7 . 224.4 251.7 630 128.4 116.2 146.7 185.8 212.4 225.8 252.4 640 130.8 118.3 148.4 187.2 213.1 227.2 253.1 650 133.0 120.3 149.9 188.5 213.7 228.5 253.7 660 135.2 122.3 151.4 189.8 214.4 229.8 254.4 670 137.4 124.2 152.9 191.1 215.0 231.1 255.0 680 139.4 126.0 154.4 192.4 215.5 232.4 255.5 690 141.5 127.8 155.8 193.7 216.1 233.7 256.1 700 143.4 129.6 157.2 194.9 216.6 234.9 256.6 710 145.3 131.3 158.6 196.1 217.1 236.1 257.1 720 147.2 132.9 159.9 197.3 217.5 237.3 257.5 730 149.0 134.6 161.2 198.4 218.0 238.4 258.0 740 150.7 136.1 162.4 199.6 218.4 239.6 258.4 750 152.4 137.7 163.6 200.7 218.9 240.7 258.9 760 154.1 139.2 164.7 201.8 219.3 241.8 259.3 770 155.7 140.7 165.8 202.9 219.7 242.9 259.7 780 157.3 142.1 166.9 204.0 220.1 244.0 260.1 GENE-B 1100639-01 Table 7-1 Browns Ferry 2 P - T Curve Values (Continued) 1 11 ~ 1 ~ 1111 1 1 1 1 1 1 1111 1 1*111 1111 1 1 1 1REQUIRED TEQPERATURES1 1 1 1 1 1 111 1 11 11111 11 11 1 1 11 1111 1 11 32 EFPY NON- BOTTOM 32 EFPY UPPER 32 EFPY NON-PRESSURE BELTLINE BELTLINE HEAD BELTLINE VESSEL BELTLINE BELTLINE CURVE A CURVE A CURVE B CURVE B CURVE B CURVE C CURVE C 790 158.9 143.6 168.0 205.1 220.5 245.1 260.5 800 160.4 144.9 169.1 206.1 220.9 246.1 260.9 810 161.9 146.3 170.2 207.1 221.3 247.1 261.3 820 163.4 147.6 171.3 208.1 221.7 248.1 261.7 S30 164.8 148.9 172.3 209.1 222.1 249.1 262.1 840 166.2 150.2 173.4 210.1 222.4 250.1 262.4 850 167.6 151.5 174.4 211.1 222 8 251.1 262.8 860 168.9 152.7 175.5 212.0 223 I 252.0 263.1 870 170.2 153.9 176.5 213.0 223.5 253.0 263.5 889 171.5 155.1 177.6 213.9 2~&.8 253.9 263.8 890 172.8 156.3 178.6 214.8 224.2 254.8 264.2.

900 174.0 157.4 179.7 215.7 224 5 255.7 264.5 910 175.3 158.6 180.7 216.6 224.8 256.6 264.8 920 176.5 159.7 181.7 217.5 225.2 257.5 265.2 930 177.6 160.8 182.7 218.4 225.5 258.4 265.5 940 178.8 161.8 183.7 219.3 225.9 259.3 265.9

'50 180.0 162.9 184.7 220.1 226.2 260.1 266.2 960 181.1 163.9 185.7 220.9 226.5 260.9 266.5 970 182.2 165.0 186.7 221.8 226.9 261.8 266.9 980 183.3 166.0 187.7 222.6 227.2 262.6 267.2 990 184.3 167.0 188.6 2H.4 227.6 263.4 267.6 1000 185.4 167.9 189.6 224.2 227.9 264.2 267.9 1010 186.4 168.9 190.5 225.0 228.2 265.0 268.2 1020 187.5 169.9 191.4 225.8 228.6 265.8 268.6 1030 188.5 170.8 192.2 226.6 228.9 266.6 268.9 1040 189.5 171.7 193.0 227.3 229.2 267.3 269.2 1050 190.5 172.6 193.8 228.1 229.6 268.1 269.6 1060 191.4 173.5 194.6 228.8 229.9 268.8 269.9 1070 192.4 174.4 195.4 229.6 230.2 269.6 2702 1080 193.3 175.3 196.2 230.3 230.5 270.3 270.5 1090 194.2 176.2 196.9 231.0 230.9 271.0 270.9 1100 195.2 177.0 197.7 231.7 231.2 271.7 271.2 1110 196.1 177.9 198.4 232.5 231.5 272.5 271.5 1120 197.0 178.7 199.1 233.2 231.9 273.2 271.9 1130 197.8 179.5 199.8 233.9 232.2 273.9 2722 1140 198.7 180.3 200.5 234.5 232.5 274.5 272.5.

1150 199.6 181.1 201.2 235.2 232.9 275.2 272.9 1160 200.4 181.9 201.9 235.9 233.2 275.9 273.2 1170 201.3 182.7 202.6 236.6 233.5 276.6 273.5 1180 202.1 183.5 203.2 237.2 233.8 277.2 273.8 1190 202.9 184.2 203.9 237.9 234.1 277.9 274.1

GENE-B 110063 9-01 Table 7-1 Browns Fenv 2 P - T Curve Values (Continued)

~ + + '~ 'i + + '~ + + + + + + + + ~' i 'i ie e + i

+ ++ + i' + < '~

REQ UIRED TEMPERATURES +

'" + '>> + + + 'i + 'i + + + + i ++++e e e e e 4>> 4' + 'i 32 EFPY NON- BOTTOM 32 EFPY UPPER 32 EFPY NON-PRESSURE BELTLINE BELTLINE HEAD BELTLINE VESSEL BELTLINE BELTLINE CURVE A CURVE A CURVE B CURVE B CURVE B CURVE C CURVE C 1200 203.7 185.0 204.6'05 238.5 234 4 278.5 274.4 1210 204.5 185.7 2 2392 234.8 '779 2 274.8 1220 205.3 186.5 205.9 239.8 235.1 279.8 275.1 1230 206.1 187.2 206.5 240.5 235.4 280.5 275.4 1240 206.9 187.9 ')07 2 241.1 235.7 281.1 275.7 1250 207.6 188.7 207.8 241.7 236.0 281.7 276.0 1260 208.4 189.4 208.5 242.3 236.3 282.3 276.3 1270 209.1 190.1 209.1 242.9 236.6 282.9 276.6 1280 209.9 190.8 209.8 243.5 237.0 283.5 277.0 1290 210.6 191.4 210A 244.1 237.3 284.1 277.3 1300 211.3 ]92.1 211.1 244.7 237.6 284.7 277.6 1310 212.0 192.8 211.7 245.3 237.9 285.3 277.9 1320 212.7 193.5 9]'7 4 245.9 238.2 285.9 278.2 1330 213.4 194.1 213.0 246.5 238.5 286.5 278.5 1340 214.1 194.8 213.7 247.0 238,8 287.0 278.8 1350 214.8 195.4 214.3 247.6 239.1 287.6 279.1 1360 215.5 196.1 215.0 248.2 239.4 288.2 279.4 1370 216.2 196.7 215.6 248.7 239.7 288.7 279.7 1380 216.8 197.3 216.3 249.3 240.0 289.3 280.0 1390 217.5 197.9 216.9 249.8 240.3 289.8 280.3 1400 218.2 198.6 217.6 250.4 240.6 290.4 280.6

Table 7-2 BELTLINE ART VALUES FOR BROWNS FERRY 2 Low-Int Shell Low-Int Shell:

Thickness = =

6.13 inches 32 EFPY Peak I.D. fluence = 6.05E+17 32 EFPY Peak 1/4 T fluence = 4.19E+17 Lower Shell Lower Shell:

Thickness = 6.13 inches 32 EFPY Peak I.D. fluence = 6.05E+17 32 EFPY Peak 1/4 T fluence = 4.19E+17 Initial 32 EFPY 32 EFPY 32 EFPY COMPONENT I.D. HEAT %Cu %Ni CF RTndt Del ta RTndt Margin Shift ART PLATES:

Lower Shell 6-127-14 C2467-2 0.16 0.52 112.4 -20 29.9 29.9 59.7 39.7 Lower Shell 6-127-15 C2463-1 0.17 0.48 116.8 -20 31.0 31.0 62.1 42.1 Lower Shell 6-127-17 C2460-2 0.13 0.51 88.3 0 23,5 23.5 46.9 46.9 Low-Int Shell 6-127-6 A0981-1 0.14 0.55 97.8 -10 26.0 26.0 52.0 42.0 Low-Int Shell 6-127-16 C2467-1 0.16 0.52 112.4 -10 29.9 29.9 59.7 49.7 Low-Int Shell 6-127-20 C2849-1 0.11 0.5 73 -10 19.4 19.4 38.8 28.8 WELDS:

Long. 0.28 0.35 154.5 10 41.0 41.0 82.1 92.1 D55733 ESW'ircumferential 0.09 0.65 116.7 -40 31.0 31.0 62.0 22.0

" ESW chemistry based on (average+ 1 sigma) of several qualification weld chemistries.

tb C)

CO Ch

0 GENE-B 1100639-01 TABLE 7-3 UPPER SHELF ENERGY ANALYSIS FOR BELTLINEMATERIALS Initial Initial 32 EFPY 32 EFPY 32 EFPY Test Longit. Trans. 1/4T Fluence %DECR Trans.

Location Heat Temp. USE USE %CU (x 10~17) USE USE Lower C2467-2 USE 120 78 0.16 4.2 12 68.6 Shell C2463-1 USE 120 78 0.17 4.2 13 '7.9 C2460-2 USE 120 78 0.13 42 10 70.2 Int Shell A0981-1 USE 142 92.3 0.14 4.2 ll 82.1 C2467-1 USE 120 78 0.16 4.2 12 68.6 C2849-1 USE 120 78 0.11 4.2 9.5 70.6 Welds:

Axial AllESW USE 95 0.28 19 76.0 Circumfer- AS A ential weld USE 145 0.09 4.2 129.1 89 g prrg ~op so~ng g-g gsoL o~nssug pg o~n5rg 4.) aVnXmZdnm ave@ aaSSaA ZOXOma WnWINIIN 0'009 0'00t008 0'OOZ 0'001 0'0 NOLLVHHdOd0 hdd2 SI H03 GI IVAHA%13 HNI'LL'IHH-NON NOLLVH2dO 30 bc'E 'H03 GI IVA SI HAHl13 HNI'LL'HH daZ8 dllL IOH OOZ MIHS So I 'Z8 'BNI11128 SIlhl MH

'D ddV 098900I CINV SJllhll t 3NI1J.138 NON OISd ZIC 000 0

XI ill

/ 009

/

/ z O

I I 008 0

'IGSSHA HEL NI 13M KLM Il I

I (

fll LINIILSD.OHCKH NKLShS - V CO OOOL I

I xm I O I I GVHH NOLJ.OH '803 doZt'NV,

'HS SPA H2ddll H03 dokS OOZED

'HNI'LLTIH303 doOI HHV SHll'IVA>PULH IVLLINI V HA%13 00>l Z >tul1 Xrrog mmozH I

0091

G~-B 110063 9-01 1600 l Browns Ferry Unit 2 I I 1400 CURVE B INITIALRTndt VALUES ARE 10'F FOR BELTLINE, 1200 54'F FOR UPPER VESSEL.

(3 AND 42F FOR BOTTOM HEAD CI z

n- 1000 0

B - NON-NUCLEARHEAT-UP/

COOLDOWN LIMIT

)

Ill 800 8O tu Q: UPPER VESSEL LIMITS z I AND 10CFR50 APP G I- I REQ'MTS I BELTLINE,82;1'F SHI 600, I UJ I K I - - BOTTOM HEAD LIMITS D

400

/

/

/

CURVES ARE VALIDFOR 32 EFPY 200 OF OPERATION EXCEPT WHERE THE BOLTUP NON-BELTLINECURVE IS NOT LIMITING.

82 F IN THIS CASE THE NON-BELTLINECURVE IS VALIDFOR 20 EFPY OF OPERATION 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUMREACTOR VESSEL METALTEMPERATURE ('F)

Figure 7-2. Heat-up/Cooldown P-T Curves for Unit 2

\1 fl

GENE-B 110063 9-01 1600 Browns Feny Unit 2 1400 I

CURVE C I

I INITIALRTndt VALUES ARE 1200 10'F FOR BEL'lIINE, 54OF FOR UPPER VESSEL.

D AND 42'F FOR BOTTOM z

0I- 1000 K

)

0 800 C - NUCLEAR O (CO K CRITICAL)LIMIT Ul K

Z I-ltl K

600 I

I I

NON-SELTLINE LIMITS AND 10CFR50 APP G D REQ'MTS BELTLINE, 82.1'F SHIFT (0

tll K XIINIMUM CRITICALITY 1 400 WITH /

NORMALWATER LEVEL //

820F CURVES ARE VALIDFOR 32 EFPY 200 OF OPERATION EXCEPT WHERE THE IN THIS CASE ~

NON-BELTLINECURVE IS NOT LIMITING.

NON-BELTLINECURVE IS VALIDFOR 20 EFPY OF OPERATION 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUMREACTOR VESSEL METALTEMPERATURE ('F)

Figure 7-3. Core Critical Operation P-T Curves for Unit 2

GE1lE-B 110063 9-01 1600 I I I

BROWNS FERRY UNIT 2 I I

I f AA' B'I 1400 CC'200 I I I T,

I k

I I II A', B', C' CORE BELTLINE I

I I I( AFTER ASSUMED 82.1oF SHIFT FROM AN INITIAL WELD RTndt OF IO'F 0 I I

x I I

1000 I-Lll A. B, C - NON-BELTLINELIMITS tO WHZ JET PUMP NOZZLE

)0' Lll I

I RTndt OF 54'F FOR B&C 800 I BOTI'OM HEAD DOME I RTndt OF 42'F O I Lll K I I I I z / I I- / I 600 / I I I LLI tL I I D

(0

/ /

to A - SYSTEM HYDROTEST LMIT LLI LL WITH FUEL IN VESSEL 400 B - NON-NUCLEARHEATUP/

/ COOLDOWN LIMIT 312 PS IG /

/ C - NUCLEAR (CORE CRITICAL)LIMIT 200 BOLTLIP 2oF I

S A',B',C'RE VALIDFOR 32 EFPY OF OPERATION 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUMREACTOR VESSEL METAlTEMPERATURE ('F)

Figure 7<. Combined P-T Curves for Unit 2 100.0 AIITol'late IVcld C246 I~ I FSIV 80.0 0 -10.0 10.0 I 48 17.1 2 -0.8 22.6 3 2.6 273 4 5.6 31.4 5 83 35.2 6 10.9 38.7 60.0 7 13,2 41.9

& 15.4 45.0 9 17.5 47.8 10 19.5 50.6 II 21.4 53.2 12 23.2 55.6 I 40.0 13 24.9 58.0 14 26.6 603 IS 28.2 62.6 16 29.8 64.7 17 31.3 66.8 18 32.8 68,8 19 34.2 70.7 20.0 Plate Material 20 35.6 72.6 Wetd Matertat 21 36.9 74.5 22 38.2 76.2 23 39.5 78.0 24 40.7 '79.7 25 41.9 81.4 0.0 26 43.1 83.0 27 44.3 84.6 28 45.4 86.2 29 46.5 87.7 30 47.6 89.2 31 48.7 90.6

-20.0 32 49.7 92.I EFPY (Years)

Figure 7-5. Browns Ferry 2 ART Versus EFPY for Plate and Nfeld Materials

II I GENE-B 1100639-01

8. REFERENCES

[1] "Fracture Toughness Requirements." Appendix G to Part 50 of Title 10 of the Code of Federal Regulations. July 1983.

[2] "Protection Against Non-Ductile Failure." Appendix G to Section XI of the 1992 ASME Boiler dc Pressure Vessel Code.

P] "Reactor Vessel Material Surveillance Program Requirements," Appendix H to Part 50 of Title 10 of the Code of Federal J

Regulations, July 1983.

[4] "Surveillance Test for Nuclear Reactor Vessels," American Society for Testing and iVfaterials. Philadelphia, PA, (ASTM E185-66).

[5] "Browns Ferry Nuclear Plant Updated Final Safety Analysis Report, Section 4.2,"

Tennessee Valley Authority.

[6] "Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards, American Society for Testing and Materials, Philadelphia, PA, July 1,1982, (ASTM E185-82).

[7] "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

[8] "Reactor Pressure Vessel Purchase Specification, Rev. 9," GE-NED, San Jose, CA, (21A111 1).

[9] Letter &om Ralph H. Shell to U.S. Nuclear Regulatory Commission, "Browns Ferry Nuclear Plant (BFN), Sequoyah Nuclear Plant (SQN), and Watts Bar Nuclear Plant (WBN) - Response to Generic Letter 92-01 (Reactor Vessel Structural Integrity),

Tennessee Valley Authority, Chattanooga, Tennessee, July 7, 1992.

[10] G. C. Martin, "Browns Ferry Unit 3 In-Vessel Neutron Spectral Analysis," GENE, Pleasanton, CA, August 1980, (NEDO-24793).

GENE-B 110063 9-01

[11] Letter from J. Valente (Restart Engineering Manager - ATH 3A-BFN) to T. R. McIntyre, "Browns Ferry Nuclear Plant - Units 1, 2, and 3 Pressure Temperature Limits Calculation

- Attachment 1 and Attachment 2." (Attachment 1 is Branch / Project IdentiGer ND-Q0999-900054 RO, RIMs Accession Number B22 '90 1227 102, Rev. 0 page 4 of 9).

[12] "Standard Methods for Notched Bar Impact Testing of Metallic Materials," Annual Book of ASTM Standards, American Society for Testing and Materials, Philadelphia, PA, (ASTM E23-88).

[13] "Nuclear Plant Irradiated Steel Handbook," Electric Power Research Institute, Palo Alto, CA, September 1986, (EPRI Report NP-4797).

[14] "Standard Methods of Tension Testing of Metallic Materials," Annual Book of ASTM Standards, American Society for Testing and Materials, Philadelphia, PA, (ASI M E8-89).

[15] G. W. Contreras, "Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Pressure Temperature Operating Limits," GENE, San Jose, CA, June 1994, (GENE-523-A65-0594).

[16] "Fracture Toughness Requirements," USNRC Branch Technical Position MI'EB 5-2, Revision 1, July 1981.

[17] H. S. Mehta, T. A. Caine, and S. E. Plaxton, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels," GENE, San Jose, CA, February 1994, (NEDO-32205-A, Revision 1).

GENE-B 1100639-01 APPENDIX A - CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS Photographs of each Charpy specimen fracture surface were taken per the requirements of ASTM E185-82. 'l'he pages following show the fracture surface photographs along with a summary of the Charpy test results for each irradiated specimen. The pictures are arranged in the order of base, weld, and HAZ materials.

GENE-B 1100639-01 BASE: E71 BASE: E7D Temp: 40 'F Temp: 60 'F Energy: 60.0 ft-lb \ Energy: 82.5 ft-lb MLE: 50.0 mils MLE: 61.0 mils I

Shear: 40 % Shear: 59%

BASE: E64 BASE: ESU Temp: 80 'F Temp: 100 F Energy~: 94.5 ft-lb Energy: 121.0 ft-lb MLE: 0.0 mils MLE: 91.0 mils Shear: 68 % Shear: 85%

BASE: E72 BASE: ES1 Temp: 120 'F Temp: 160 'F Energy: 120.5 ft-lb v ij Energy:. 130.0 ft-lb ter MLE: 91.0 mils MLE: 88.0 mils Shear: 100 % Shear: 100%

BASE: E57 BASE: ESS Temp: 200 'F Temp: 300 'F Energy: 136.0 ft-lb Energy: 131.5 ft-lb MLE: 94.0 mils MLE: 88.0 mils Shear: 100 % Shear: 100%

GENE-B 110063 9-01 BASE: ESC BASE: ESY Temp: 80 oF I Temp: -40 'F 10.5 ft-Ib

-' i Energy: 17.0 ft-lb Energy:

MLE: 10.0 mils MLE: 13.5 mils Shear: Shear: 11%

BASE: E7Y BASE: E7K Temp: -20 'F Temp: 0 oF Energy: 33.0 ft-lb Energy: 38.5 ft-Ib MLE: 30.5 mils MLE: 33.0 mils Shear: 13  % Shear: 19%

~

• WELD: EB7 WELD: EBS h

Temp: -80 'F Temp: -40 'F

'A ~

Energy: 13.0 ft-lb Energy: 2.0 ft-lb M~

MLE: 5.0 mils MLE: 12.5 mils Shear: 2% Shear 4 o/o WELD: EBK WELD: EAP Temp: 20 oF Temp: 0 'F Energy: 37.5 ft-Ib Energy: 50.0 ft-Ib MLE: 31.0 mils MLE: 42.0 mils Shear: 9% Shear. 15 %

GENE-B 1100639-01 WELD: EBD WELD: EBB Q3 I Temp: 20 'F Temp: 40 'F Energy: 59.5 ft-lb Energy: 59.5 ft-lb MLE: 52.0 mils MLE: 50.0 mils Shear: 22% Shear: 30%

WELD: EB1 WELD: EAM Temp: 80 'F Temp: 100 'F Energy: 59.0 ft-Ib Energy: 76.5 ft-Ib MLE: 52.0 mils MLE: 64.5 mils Shear: 42 % Shear: 50%

WELD: EBE WELD: EB4 Temp: 120 'F Temp: 160 'F Energy: 87.0 ft-lb Energy: 107.0 ft-lb MLE: 65.0 mils MLE: 87.0 mils Shear: 68 % Shear: 100 %

WELD: EB2 WELD: EBA Temp: 200 'F Temp: 300 'F Energy: 107,5 ft-Ib Energy: 113.0 ft-Ib MLE: 84.5 mils MLE: 88.5 mils Shear: 100 % Shear. 100%

GENE-B 1100639-01 HAZ: ED6 HAZ: EJ3 Temp: 80 oF Temp: -60 'F Energy: 3.5 ft-lb Energy: 37.0 ft-lb MLE: 6.0 mils MLE: 30.0 mils Shear: 1% Shear: 12%

HAZ: EEY HAZ: EDB Temp: -40 'F Temp: -20 'F Energy: 54.0 ft-lb Energy: 30.0 ft-lb MLE: ~4.0 mils MLE: 22.5 mils Shear: 24% Shear: 7%

HAZ: EJJ HAZ: EJS Temp: 0 'F Temp: 20 'F Energy: 43.5 ft-lb Energy: 106.0 ft-lb MLE: 36.5 mils MLE: 81.5 mils Shear: 19 % Shear: 65 %

HAZ: EJC HAZ: EJ1 Temp: 40 'F Temp: 60'F Energy: 93.5 ft-lb Energy: 107.5 ft-lb MLE: 67.0 mils MLE: 86.0 mils Shear: 48% Shear: 75%

v ~

~I

GENE-B 1100639-01 HAZ: EDC HAZ: EJB Temp: 80 'F Temp: 120 'F Energy: 82.0 ft-Ib Energy: 97.5 ft-lb MLE: 73.0 mils %LE: 78.0 mils Shear: 60% Shear: 100 %

HAZ: E3D HAZ: EEC Temp: 200 'F Temp: 300 'F Energr: 107.5 ft-lb 4 b1 Energy: 143.0 ft-lb

', ~

i~E: 82.0 mils MLE: 92.0 mils Shear: 100 % Shear: 100 %

1' GENE-B 110063 9-01 APPENDIX B EQUIVALENTMARGIN ANALYSIS

GE1'K-B 1100639-01 TABLE B-I EQUIVALENTMARGINANALYSIS PLANT APPLICABILITY VERIFICATIONFORM FOR BROWNS FERRY UNIT 2 - 8%'R 4/MK I BWR/3-6 PLATE Surveillance Plate USE:

%Cu = 0,14 Capsule Fluence =1.52 x 10'~ n/cm-Measured % Decrease = 4 (Chatpy Curves)

RG. 1.99 Predicted% Decrease = 9 (R.G. 1.99, Figure 2)

Limitin Beltline Plate USE:

%Cu = 0.1/

32 EFPY 1/4T Fluence M.2x 10'/cm-RG. 1.99 Predicted% Decrease ~ 13 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (RG. 1.99, Position 2.2) 13 % < 21'/+ so vessel plates are bounded bv eauivalent marmn analvsis Q~

4

GENE-B 1100639-01 TABLE B-2 EQUIVALENTMARGINANALYSISPLANT APPLICABILITY VERIFICATIONFORM FOR BROWNS FERRY UNIT 2 - BWR 4/MK I BWR/2-6 WELD Surveillance Weld USE:

%Cu = 0.20 Capsule Fluence = 1.52 x 10'/cm Measured % Decrease = -3 (Charpy Curves)

R.G. 1.99 Predicted% Decrease = 13 (R.G. 1.99, Figure 2)

Limitin Beltline Weld USE:

%CU = 0.28 32 EFPY 1/4T Fluence = 4.2 x 10'/cm R.G. 1.99 Predicted % Decrease = 21 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 21 % < 34%, so vessel welds are bounded by equivalent margin analysis i

Q