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Transmittal of Technical Report 3002000078, BWRVIP-271NP: BWR Vessel & Internals Project Testing and Evaluation of the Browns Ferry Unit 2 120 Degrees Capsule
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' I 1 1I ELECTRIC POWERRESEARCH INSTITUTE 2013-056 BWR Vessel & Internals Project (BWRVIP)April 17, 2013Document Control DeskU. S. Nuclear Regulatory Commission 11555 Rockville PikeRockville, MD 20852Attention:

Joseph Holonich

Subject:

Project No. 704 -BWRVIP-271NP:

BWR Vessel and Internals

Project, Testing andEvaluation of the Browns Ferry Unit 2 120' Capsule"Enclosed are five (5) paper copies of the report "BWRVIP-271NP:

BWR Vessel and Internals

Project, Testing and Evaluation of the Browns Ferry Unit 2 120' Capsule,"

EPRI Technical Report 3002000078, April 2013. This report is being transmitted to the NRC for information only.This report describes testing and evaluation of the Browns Ferry Unit 2 1200 capsule.

Theseresults will be used to monitor embrittlement as part of the BWRVIP ISP.Please note that the enclosed report is non-proprietary and is available to the public by request toEPRI.If you have any questions on this subject please call Ron DiSabatino (Exelon, BWRVIPAssessment Focus Group Chairman) at 610-765-5753.

Sincerely, Dennis MadisonSouthern NuclearChairman, BWR Vessel and Internals Projectc: Gary Stevens, NRCTogether

... Shaping the Future of Electricity PALO ALTO OFFICE3420 Hillview Avenue, Palo Alto, CA 94304-1395 USA

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~CL e&-C I 4'~C p~ ýe&c ELECTRIC POWERa =2 I RESEARCH INSTITUTE 2013 TECHNICAL REPORTBWRVIP-271NP:

BWR Vessel and Internals ProjectTesting and Evaluation of the Browns Ferry Unit 2 1200 CapsulePREPAREDUNDER THENUCLEARPROGRAM BWRVIP-271 NP: BWR Vessel andInternals ProjectTesting and Evaluation of the Browns Ferry Unit 21200 Capsule3002000078 Final Report, April 2013EPRI Project ManagerR. CarterWork to develop this product was completed under the EPRI Nuclear Quality Assurance Programin compliance with 10 CFR 50, Appendix B and 10 CFR 21,~NOELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California 94304-1338

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& Testing, LLCTransWare Enterprises Inc.THE TECHNICAL CONTENTS OF THIS DOCUMENT WERE PREPARED IN ACCORDANCE WITHTHE EPRI NUCLEAR QUALITY ASSURANCE PROGRAM MANUAL THAT FULFILLS THEREQUIREMENTS OF 10 CFR 50 APPENDIX B, 10 CFR PART 21, ANSI N45.2-1977 AND/ OR THEINTENT OF ISO-9001 (1994). CERTIFICATION OF CONFORMANCE CAN BE OBTAINED FROMEPRI. BEFORE THE CONTENTS OF THIS DOCUMENT CAN BE APPLIED TO FULFILL QUALITYPROGRAM REQUIREMENTS, CONTRACTUAL ARRANGEMENTS BETWEEN THE USER ANDEPRI MUST BE ESTABLISHED.

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NOTEFor further information about EPRI, call the EPRI Customer Assistance Center at 800.313.3774 ore-mail askepri@epri.com.

Electric Power Research Institute, EPRI, and TOGETHER..

SHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc.Copyright

ACKNOWLEDGMENTS The following organizations prepared this report:Electric Power. Research Institute (EPRI)3420 Hillview AvePalo Alto, CA 94304Principal Investigators T. HardinM. O'ConnorMP Machinery

& Testing, LLC2161 Sandy DriveState College, PA 16803Principal Investigator Dr. M. P. Manahan, Sr.TransWare Enterprises Inc.1565 Mediterranean DriveSycamore, IL 60178Principal Investigator E. N. JonesThis report describes research sponsored by EPRI and its BWRVIP participating members.This publication is a corporate document that should be cited in the literature in the following manner:B WR VIP-2 71NP: B WR Vessel and Internals

Project, Testing and Evaluation of the BrownsFerry Unit 2 1200 Capsule.

EPRI, Palo Alto, CA: 2013. 3002000078.

iii

PRODUCT DESCRIPTION In the late 1990s, a Boiling Water Reactor Vessel and Internals Project (BWRVIP)

Integrated Surveillance Program (ISP) was developed to improve the surveillance of the U.S. BWR fleet.This report describes testing and evaluation of the Browns Ferry Unit 2 1200 capsule.

Theseresults will be used to monitor embrittlement as part of the BWRVIP ISP.Background The BWRVIP ISP represents a major enhancement to the process of monitoring embrittlement for the U.S. fleet of BWRs. The ISP optimizes surveillance capsule tests while at the same timemaximizing the quantity and quality of data, thus resulting in a more cost-effective program.

TheBWRVIP ISP provides more representative data that can be used to assess embrittlement inreactor pressure vessel beltline materials and improve trend curves in the BWR range ofirradiation conditions.

Challenges and Objectives Neutron irradiation exposure reduces the toughness of reactor vessel steel plates, welds, andforgings.

The objectives of this project were twofold:* To document the results of neutron dosimetry and Charpy V-notch ductility tests for thesurveillance materials (plate heat A098 1-1 and weld heat BF2 ESW) in the Browns FerryUnit 2 120' capsule" To compare the results with the embrittlement trend prediction of the U.S. NuclearRegulatory Commission (U.S. NRC) Regulatory Guide 1.99, Rev. 2ApproachThe Browns Ferry Unit 2 120' capsule had been irradiated in the reactor since plant startup.

Thesurveillance capsule contained flux wires for neutron flux monitoring, Charpy V-notch impacttest specimens, and tensile specimens.

The project team removed the capsule from the reactor in2010 and transported it to facilities for testing and evaluation.

The team used dosimetry to gatherinformation about the neutron fluence accrual of specimens from the capsule.

They thenperformed a neutron transport calculation in accordance with Regulatory Guide 1.190 andcompared it to the results from the dosimetry.

Testing of Charpy V-notch specimens wasperformed according to the American Society for Testing and Materials (ASTM) standards.

Results and FindingsThe report includes capsule neutron exposure and Charpy V-notch test results for Browns Ferry2 surveillance plate heat A0981-1 and surveillance weld BF2 ESW (an electroslag weld ofunknown heat number).

The project compared irradiated Charpy data to unirradiated data inorder to determine the shifts in Charpy index temperatures for the surveillance plate and weldmaterials due to irradiation.

For the surveillance plate, results indicate a shift about 32% higherv than the prediction of Regulatory Guide 1.99, Revision 2, but within (predicted shift + margin).For the weld, the shift was about 2.3 times higher than the predicted shift and about 15% higherthan (predicted shift + margin).

Researchers also measured flux wires and determined fluence forthe capsule.Applications, Value, and UseResults of this work will be used in the BWRVIP ISP that integrates individual BWRsurveillance programs into a single program.

The ISP provides data of high quality to monitorBWR vessel embrittlement.

The ISP results in significant cost savings to the BWR fleet andprovides more accurate monitoring of embrittlement in BWR vessels.KeywordsBWRCharpy testingMechanical properties Radiation embrittlement Reactor pressure vessel integrity Reactor vessel surveillance programvi CONTENTS1 INTRO DUCTIO N ..................................................................................................................

1-11.1 Im plementation Requirements

.......................................................................................

1-12 MATERIALS AND TEST SPECIM EN DESCRIPTIO N ..........................................................

2-12.1 Dosimeters

....................................................................................................................

2-12.2 Charpy V-Notch Specim ens .........................................

2-12.2.1 Capsule Loading Inventory

....................................................................................

2-12.2.2 Material Description and Properties

.......................................................................

2-42.2.3 Chem ical Com position

...........................................................................................

2-52.2.4 CVN Baseline Properties

.......................................................................................

2-52.3 Capsule Opening .........................................................................................................

2-123 NEUTRO N FLUENCE CALCULATIO N ................................................................................

3-13.1 Description of the Reactor System ................................................................................

3-23.1.1 Reactor System Mechanical Design .......................................................................

3-23.1.2 Reactor System Material

........................................................................................

3-43.1.3 Reactor Operating Data Inputs ...............................................................................

3-53.1.3.1 Core Loading ..................................................................................................

3-63.1.3.2 Power History Data ........................................................................................

3-63.1.3.3 Reactor State-Point Data ................................................................................

3-8Beginning of Operation through Cycle 16 State Points ...............................................

3-8Projected Reactor Operation

......................................................................................

3-8Lim itation of Fluence Projections

................................................................................

3-93.2 Calculation Methodology

.............................................................................................

3-103.2.1 Description of the RAMA Fluence Methodology

...................................................

3-103.2.2 The RAMA Geometry Model for the Browns Ferry 2 Reactor ...............................

3-113.2.2.1 The Geom etry Model ....................................................................................

3-143.2.2.2 The Reactor Core and Core Reflector Models ..............................................

3-15vii 3.2.2.3 The Core Shroud Model ...............................................................................

3-163.2.2.4 The Downcomer Region Model ....................................................................

3-163.2.2.5 The Jet Pump Model ....................................................................................

3-163.2.2.6 The Surveillance Capsule Model ..................................................................

3-163.2.2.7 The Reactor Pressure Vessel Model ............................

3-173.2.2.8 The Vessel Insulation Model .........................................................................

3-173.2.2.9 The Inner and Outer Cavity Models ..............................................................

3-173.2.2.10 The Biological Shield Model .......................................................................

3-173.2.2.11 The Above-Core Component Models .........................................................

3-17The Top Guide Model .............................................................................................

3-17The Core Spray Sparger Model ................................................................................

3-183.2.2.12 The Below-Core Component Models ..........................................................

3-183.2.2.13 Summary of the Geometry Modeling Approach

..........................................

3-183.2.3 RAMA Calculation Parameters

.............................................................................

3-183.2.4 RAMA Neutron Source Calculation

......................................................................

3-193.2.5 RAMA Fission Spectra .........................................................................................

3-193.3 Surveillance Capsule Activation and Fluence Results .............................................

3-193.3.1 Comparison of Predicted Activation to Plant-specific Measurements

...................

3-203.3.1.1 Cycle 1 30* Flux Wire Holder Activation Analysis

.........................................

3-203.3.1.2 Cycle 7 300 Surveillance Capsule Activation Analysis

..................................

3-213.3.1.3 Cycle 16 1200 Surveillance Capsule Activation Analysis

..............................

3-213.3.1.4 Surveillance Capsule Activation Analysis Summary .....................................

3-223.3.2 Capsule Peak Fluence Calculations and Lead Factor Determinations

.................

3-233.4 Capsule Fluence Uncertainty Analysis

.........................................................................

3-233.4.1 Comparison Uncertainty

.......................................................................................

3-243.4.1.1 Operating Reactor Comparison Uncertainty

.................................................

3-243.4.1.2 Benchmark Comparison Uncertainty

............................................................

3-243.4.2 Analytic Uncertainty

.............................................................................................

3-243.4.3 Combined Uncertainty

..........................................................................................

3-254 CHARPY TEST DATA ..........................................................................................................

4-14.1 Charpy Test Procedure

..................................................................................................

4-14.2 Charpy Test Data ..........................................................................................................

4-2Viii 5 CHA RPY TEST RESULTS ...............................................................................................

5-15.1 Analysis of Im pact Test Results .....................................................................................

5-15.2 Irradiated Versus Unirradiated CVN Properties

.............................................................

5-16 REFERENCES

.....................................................................................................................

6-1A A PPENDIX A -DO SIM ETER A NA LYSIS ......................................................................

A-1A.1 Dosim eter M aterial Description

................................................................................

A-1A.2 Dosim eter Cleaning and M ass M easurem ent ...........................................................

A-1A.3 Radiom etric Analysis

....................................................................................................

A-4ix

LIST OF FIGURESFigure 2-1 Photograph of 1200 capsule for Browns Ferry Unit 2 showing positiveidentification m arkings ......................................................................................................

2-3Figure 2-2 Schematic drawing showing the locations of test specimens in the BrownsFerry Unit 2 1200 surveillance capsule .............................................................................

2-4Figure 2-3 Browns Ferry 2 plate A0981-1 (LT) unirradiated Charpy energy plot .......................

2-8Figure 2-4 Browns Ferry 2 weld BF2 ESW unirradiated Charpy energy plot ..........................

2-10Figure 3-1 Planar view of Browns Ferry 2 at the core mid-plane elevation

...............................

3-3Figure 3-2 Planar view of the Browns Ferry 2 RAMA quadrant model at the core mid-p la ne e levatio n ...............................................................................................................

3-12Figure 3-3 Axial view of the Browns Ferry 2 RAMA model .....................................................

3-13Figure 4-1 Illustration of digital optical comparator measurement of shear fracture area ..........

4-2Figure 5-1 Irradiated plate A0981-1 (Browns Ferry 2 120* capsule)

Charpy energy plot ..........

5-2Figure 5-2 Irradiated plate A0981-1 (Browns Ferry 2 1200 capsule) lateral expansion plot ....... 5-3Figure 5-3 Irradiated weld BF2 ESW (Browns Ferry 2 1200 capsule)

Charpy energy plot ........

5-4Figure 5-4 Irradiated weld BF2 ESW (Browns Ferry 2 1200 capsule) lateral expansion p lo t ...................................................................................................................................

5 -5Figure A-1 Packet G9 Cu dosimeter wire: prior to cleaning (left); and afterclea ning/co iling (right) ......................................................................................................

A -2Figure A-2 Packet G9 Fe dosimeter wire: prior to cleaning (left); and aftercleaning/coiling (right) .....................................................................................................

A -2Figure A-3 Packet G9 Ni dosimeter wire: prior to cleaning (left); and after cleaning/coiling (rig h t) ...............................................................................................................................

A -2Figure A-4 Packet G10 Cu dosimeter wire: prior to cleaning (left); and afterclea ning/co iling (rig ht) ......................................................................................................

A -3Figure A-5 Packet G10 Fe dosimeter wire: prior to cleaning (left); and aftercleaning/coiling (right) ......................................................................................................

A -3Figure A-6 Packet G10 Ni dosimeter Wire: prior to cleaning (left); and aftercleaning/coiling (right) ......................................................................................................

A -3xi

LIST OF TABLESTable 2-1 Browns Ferry 2 1200 surveillance capsule specimen inventory

................................

2-2Table 2-2 Best estimate chemistry of Browns Ferry 2 surveillance plate A0981-1 ....................

2-5Table 2-3 Best estimate chemistry of Browns Ferry 2 surveillance weld BF2 ESW ..................

2-5Table 2-4 Unirradiated Charpy impact test data for surveillance plate A0981 -1 (LT) ................

2-6Table 2-5 Unirradiated Charpy impact test data for surveillance weld BF2 ESW ......................

2-7Table 2-6 Baseline CVN properties

..........................................................................................

2-7Table 3-1 Summary of material compositions by region for Browns Ferry 2 .............................

3-5Table 3-2 Summary of the Browns Ferry 2 core loading inventory

...........................................

3-7Table 3-3 State-point data for each cycle in Browns Ferry 2 ....................................................

3-9Table 3-4 Comparison of specific activities for Browns Ferry 2 Cycle 1 30' flux wireholder w ires (C /M ) ..........................................................................................................

3-20Table 3-5 Comparison of specific activities for Browns Ferry 2 cycle 7 30' surveillance capsule flux w ires (C /M ) .................................................................................................

3-2 1Table 3-6 Comparison of specific activities for Browns Ferry 2 cycle 16 1200 surveillance capsule flux w ires (C /M ) .................................................................................................

3-22Table 3-7 Comparison of activities for Browns Ferry 2 surveillance capsule flux wires ...........

3-22Table 3-8 Lead factors for Browns Ferry 2 300, 1200, and 3000 surveillance capsules

...........

3-23Table 3-9 Combined capsule fluence uncertainty for energy >1.0 MeV ..................................

3-25Table 4-1 Irradiated Charpy V-Notch impact test results for base metal (heat A0981-1)specimens from the Browns Ferry 2 1200 surveillance capsule ........................................

4-3Table 4-2 Irradiated Charpy V-Notch impact test results for weld metal (heat BF2 ESW)specimens from the Browns Ferry 2 1200 surveillance capsule ........................................

4-4Table 4-3 Irradiated Charpy V-Notch impact test results for HAZ specimens from theBrowns Ferry 2 1200 surveillance capsule ........................................................................

4-4Table 5-1 Effect of irradiation (E>1.0 MeV) on the notch toughness properties

........................

5-6Table 5-2 Comparison of actual versus predicted embrittlement

..............................................

5-6Table 5-3 Percent decrease in upper shelf energy ...................................................................

5-7Table A-1 Wire dosimeter masses ..........................................................................................

A-4Table A-2 GRSS specifications

...........................................................................................

A-5Table A-3 Counting schedule for the dosimeter materials

.......................................................

A-6Table A-4 Neutron-induced reactions of interest

.....................................................................

A-6Table A-5 Results of the radiometric analysis

.........

...............................................................

A-6xiii

IINTRODUCTION Test coupons of reactor vessel ferritic beltline materials are irradiated in reactor surveillance capsules to facilitate evaluation of vessel fracture toughness in vessel integrity evaluations.

Thekey values that characterize fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the upper shelf energy (USE). These are defined in lOCFR50 AppendixG [1] and in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI [2].Appendix H of IOCFR50 [1] and ASTM E185-82 [3] establish the methods to be used for testingof surveillance capsule materials.

In the late 1990s the BWR Vessel and Internals Project (BWRVIP) initiated the BWRVIPIntegrated Surveillance Program (ISP) [4], and the BWRVIP assumed responsibility for testingand evaluation of ISP capsules.

The surveillance plate and weld from the Browns Ferry NuclearPlant Unit 2 (hereinafter, Browns Ferry 2 or BF2) were designated as "ISP representative surveillance materials" to be tested by the ISP according to an approved capsule withdrawal andtest schedule.

This report addresses the withdrawal and test of the Browns Ferry 2 1200 capsule.

The capsulewas irradiated for 16 cycles of operation.

The capsule was removed in February 2010 and testingand evaluation was completed in September 2012. The surveillance capsule contained flux wiresfor neutron flux monitoring, Charpy V-notch impact test specimens, and tensile specimens.

Thecapsule was shipped to MP Machinery

& Testing, LLC for opening and testing of the CharpyV-notch surveillance specimens.

Evaluation of the fluence environment was conducted byTransWare Enterprises, Inc. Final evaluation of the Charpy test data and irradiated materialproperties and compilation of this report were performed by EPRI. The Charpy V-notchsurveillance materials were tested per ASTM E 185-82, and the information and the associated evaluations provided in this report have been performed in accordance with the requirements of1OCFR50 Appendix B [5].This report compares the irradiated material properties of surveillance plate heat A098 1-1 andsurveillance weld BF2 ESW to their baseline (e.g., unirradiated) properties.

The observedembrittlement (as characterized by AT30) is compared to that predicted by U.S. NuclearRegulatory Commission (U.S. NRC) Regulatory Guide 1.99, Rev. 2 [6]. Other BWRVIP ISPreports will integrate these shift results with the previous Browns Ferry 2 surveillance capsuleresults for a broader characterization of embrittlement behavior.

1.1 Implementation Requirements The results documented in this report will be utilized by the BWRVIP ISP and by individual utilities to demonstrate compliance with 10CFR50, Appendix H, Reactor Vessel MaterialSurveillance Program Requirements.

Therefore, the implementation requirements of I OCFR50,Appendix H govern and the implementation requirements of Nuclear Energy Institute (NEI)03-08, Guideline for the Management of Materials Issues [7], are not applicable.

1-1 2MATERIALS AND TEST SPECIMEN DESCRIPTION The General Electric (GE) designed Browns Ferry Unit 2 (BF2) 1200 surveillance capsule wasremoved from the plant for analysis and testing during the February 2010 refueling outage. Thecapsule was a GE standard single basket design and contained a total of two Charpy packets andfour tensile tubes. Within each Charpy packet were a total of 12 Charpy V-notch specimens andthree high purity dosimetry wires. Each of the tensile tubes contained two tensile specimens.

The1200 capsule is an original plant capsule, and has been irradiated in the plant since initial startup.This is the second surveillance capsule to be removed from BF2 and tested. The 300 capsule wasremoved during the Fall 1994 outage and was tested by GE [8].2.1 Dosimeters The dosimetry wires were located along the ends of the Charpy specimens during irradiation.

Each of the two Charpy packets contained one high purity iron, copper, and nickel wire forfluence determination.

Further details on the exact wire locations during the irradiation areprovided in the capsule opening discussion given in Section 2.3. A detailed discussion of theradiometric analysis of the capsule dosimetry wires is provided in Appendix A.2.2 Charpy V-Notch Specimens The BF2 1200 capsule Charpy V-notch and tensile specimen inventory, material descriptions, unirradiated (baseline)

Charpy impact data, and previously measured capsule data aresummarized in this section of the report.2.2.1 Capsule Loading Inventory The BF2 1200 surveillance capsule inventory is provided in Table 2-1. All of the capsulespecimens, which include tensile specimens, Charpy specimens, and dosimeters, were recovered from the capsule basket. Testing was only performed on the 24 Charpy specimens, and thedosimetry wires were counted and weighed to determine specific activities.

All eight of thetensile specimens (three base metal, three weld, and two HAZ) remain untested and are beingheld in reserve for future testing since there is no near-term use for tensile data. The technical advantage of storing the tensile specimens untested is that there will be options in the future forhow these specimens will be used to obtain useful data. For example, the tensile specimengeometry is conducive to fabrication of subsize Charpy as well as miniaturized Charpy V-notchspecimens.

Further, research is currently underway to develop testing methods which will enablethe determination of plane-strain fracture toughness data from Charpy-sized specimens.

Withthese new technologies in view, there may be a need in the future for static and!or dynamictensile data for use in the calculation of fracture toughness from experimental data obtained fromCharpy specimens.

Therefore, all of the tensile specimens have been placed into the archivestorage so that they can be tested when necessary in the future. Similarly, the broken Charpy2-1 Materials and Test Specimen Description specimen halves have been added to long-term archive storage for future use in miniature mechanical behavior specimen

testing, chemistry

analysis, and microstructural studies.As indicated in Table 2-1, there were a total of two Charpy packets in the capsule, and eachcontained three dosimetry wires (one Fe wire, one Cu wire, and one Ni wire) and 12 Charpyspecimens.

Charpy packet G9 was found to contain all of the base metal test specimens as wellas 4 weld specimens.

Charpy packet G 10 contained the remaining four weld specimens alongwith all of the HAZ specimens.

The capsule and a schematic showing the position of thespecimens within the capsule are shown in Figures 2-1 and 2-2, respectively.

Table 2-1Browns Ferry 2 1200 surveillance capsule specimen inventory Charpy Number of Number of RelativePacket Charpy Specimens Flux Wires Vertical PositionNo.2Base Weld HAZ Fe Cu NiG9 8 4 0 1 1 1 Lowest in basketG10 0 4 8 1 1 1 Highest in basket1 The capsule included tensile specimens, but the tensile specimens were not tested. Four tensilespecimens for this capsule were located at the lowest axial position below Charpy packet G9 and fourtensile specimens were located between Charpy packets G9 and G10.2 The packet numbers in this table are organized by axial position in the capsule with packet G9 at thelowest elevation in the reactor and packet G10 at the highest.2-2 Materials and Test Specimen Description Figure 2-1Photograph of 1200 capsule for Browns Ferry Unit 2 showing positive identification markingsUpper photograph shows the binary reactor code 26 in the upper left corner of the basket. The lowerphotograph shows the GE drawing number engraved on the basket outer surface.

The Side whichFaced the Pressure Vessel in the Plant is Facing Down in this Photograph.

2-3 Materials and Test Specimen Description G10 Weld/HAZ Qijpy ViGm Dmsiol117C4058Tensile Tube G9Tensile Tube G8__(G9 G9 Base!Weld Charpy wir Ication117C4058Tensile Tube G7Tensile Tube G6Figure 2-2Schematic drawing showing the locations of test specimens in the Browns Ferry Unit 21200 surveillance capsule2.2.2 Material Description and Properties The material descriptions, chemical compositions, and unirradiated (baseline)

Charpymechanical properties of the materials irradiated in the BF2 120°capsule are summarized below.The BF2 surveillance base metal specimens were machined from plate heat number A0981-1 inthe longitudinal orientation (LT). Unirradiated baseline data are available for the plate, weld andHAZ materials.

All of the specimens were stamped on one end with the BF2 three-digit FABcodes assigned by GE.The weld Charpy surveillance specimens were made by the electroslag-weld (ESW) processaccording to Babcock and Wilcox Weld Procedure WR-12-4 and subsequently heat treated tosimulate the core region plate treatments

[8]. The records regarding the identity of the weld wireheat number/flux lot combination used for the surveillance weld are not available

[8,9];therefore, the surveillance weld is referred to as BF2 ESW.2-4 Materials and Test Specimen Description 2.2.3 Chemical Composition The best estimate chemical compositions of the surveillance plate A098 l-1 and weld BF2 ESWare shown in Tables 2-2 and 2-3, respectively.

Chemical compositions are presented in weightpercent.

If multiple measurements exist for a single specimen, those measurements are firstaveraged to yield a single value for that specimen, and then the different specimen averages areaveraged to determine the best estimate for the surveillance heat.Table 2-2Best estimate chemistry of Browns Ferry 2 surveillance plate A0981-1Cu Nt P S 1 Seie DSuc(wt%) (Wiom% (W*o/) (wt%/) (WtV) $c.nEouo014 0.55 0.007 0.011 0.19 N/A Refs. 8 and 90.14 0.55 0.007 0.011 0.19 *-Best Estimate AverageTable 2-3Best estimate chemistry of Browns Ferry 2 surveillance weld BF2 ESW0.20 0.33 0010 0.011 0.09 N/A Refs. 8 and 90.20 0.3 0.010]i 0.011 0.09 +[ Best Estimate Average2.2.4 CVN Baseline Properties As noted above, the Browns Ferry 2 surveillance plate Charpy specimens are longitudinal (LT)specimens.

Table 2-4 provides the unirradiated (baseline)

Charpy test data for the A098 1-1surveillance plate material and Table 2-5 provides the unirradiated data for the weld [8].The baseline test data were fit to a hyperbolic tangent curve using the computer programCVGRAPH [10]. Figures 2-3 and 2-4 show the fitted Charpy energy data curves for theunirradiated plate and weld, respectively.

Table 2-6 summarizes the baseline (unirradiated)

Charpy V-notch properties (index temperatures) of plate heat A098 1-1 and weld heat BF2 ESW.In this table and throughout this report, T30 is the 30 ft-lb (41 J) transition temperature; T50 is the50 ft-lb (68 J) transition temperature; T35..il is the 35 mil (0.89 mm) lateral expansion temperature; and USE is the average energy absorption at full shear fracture appearance.

2-5 Materials and Test Specimen Description Table 2-4Unirradiated Charpy impact test data for surveillance plate A0981-1 (LT)E7BE7JE7CE63E75E74E62E5Ka All specimen-105.3 4.33 5.0(-76.3) (5.87) (0.13)-39.5 21.92 15.0(-39.7) (29.72) (0.38)6.3 23.08 20.5(-14.3) (31.29) (0.52)40.1 73.68 54.0(4.5) (99.90) (1.37)64.0 99.02 72.5(17.8) (134.25)

(1.84)97.2 107.05 70.5(36.2) (145.14)(1.79)174.7 140.88 92.0(79.3) (191.01)

(2.34)327.9 144.72 93.5(164.4) (196.21)

(2.37)IlDs have a dot over the middle alphanumeric character.

3.711.720.838.258.860.3100.0100.02-6 Materials and Test Specimen Description Table 2-5Unirradiated Charpy impact test data for surveillance weld BF2 ESWEBJ -49.7 5.47(-45.4) (7.42)EALEAEEC3EBUEB6EAKEAD14.4(-9.8)50.0(10.0)81.3(27.4)119.1(48.4)204.3(95.7)301.6(149.8)373.8(189.9)28.06(38.04)47.89(64.93)22.40(30.37)83.90(113.75)102.15(138.49)113.53(153.92)110.62(149.98)6.5(0.17)21.5(0.55)40.0(1.02)25.0(0.64)63.5(1.61)79.0(2.01)88.0(2.24)84.5(2.15)4.612.722.528.350.379.3100.0100.0a All specimen IDs have a dot over the middle alphanumeric character.

Table 2-6Baseline CVN properties A0981-1 Browns Ferry 2 -46.7(LT) Surveillance Plate (-43.7)-14.8 -25.4(-26.0) (-31.9)137,8(186.8)BF2 ESW Browns Ferry 2Surveillance Weld-52.5 -19.5 -29.5(-46.9) (-28.6) (-34.2)116.0(157.3)2-7 Materials and Test Specimen Description PLATE HEAT A0981-1 (BF2)CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 09/09/2002 08:29 AMPage 1Coefficients of Curve 1A = 70.15 B = 67.65 C = 84.96 TO = 11.25 D = O.OOE+00Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=137.8(Fixed)

Lower Shelf Energy=2.5(Fixed)

Temp@30 ft-lbs=-46.7 Deg F Temp@50 ft-lbs=-14.8 Deg FPlant: Browns Ferry 2 Material:

SA302BM Heat: A098 1-1Orientation:

LT Capsule:

UNIRRAFluence:

0.0 nrcm^2.0IT.0u300 .250 i- .I--200"15010050-o T4- ....-* iT-t010 20 I100 200 300-I-S -----I400 500 6000 "-300-200 -100Temperature in Deg FCharpy V-Notch DataTemperature

-80. 00-60. 00-40. 00-20. 00.0020. 0040. 0080. 00100. 00120. 00200. 00Input CVN8.5017.5035.5040. 0097. 0068. 0073. 00104.50137.00134.50146.50Computed CVN16.6423 .8033. 6646. 3361. 2477. 0992. 21115. 42122. 90128. 09136 .23Differential

-8.14-6.301.84-6.3335. 76-9.09-19.21-10.9214. 106.4110.27Figure 2-3Browns Ferry 2 plate A0981-1 (LT) unirradiated Charpy energy plot2-8 Materials and Test Specimen Description PLATE HEAT A0981-1 (BF2)Page 2Plant: Browns Ferry 2 Material:

SA302BM Heat: A0981-1Orientation:

LT Capsule:

UNIRRA Fluence:

0.0 n/cm^2Charpy V-Notch DataTemperature 300. 00Input CVN133. 00Correlation Coefficient

= .956Computed CVN137. 65Differential

-4. 65Figure 2-3, continued Browns Ferry 2 plate A0981-1 (LT) unirradiated Charpy energy plot2-9 Materials and Test Specimen Description WELD HEAT ESW (BF2)CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 09/09/2002 08:23 AMPage 1Coefficients of Curve 1A = 59.25 B = 56.75 C = 81.22 TO = -6.21 D = 0.OOE+00Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1 16.0(Fixed)

Lower Shelf Energy=2.5(Fixed)

Temp@30 ft-lbsý52.5 Deg F Temp@50 ft-lbs=-

19.5 Deg F' Plant: BROWNS FERRY 2 Material:

ESW Heat: ESWOrientation:

NA Capsule:

UNIRRA Fluence:

0.0 n/cm^2300250.200 +/- -0L.>150 -1uJ500-300I 4 --l00i0 I0 100 200 300 400 500 60Temperature in Deg F-200 -1000Charpy V-Notch DataTemperature

-80. 00-60.00-40. 00-20. 00.0020. 0040. 0060. 0080. 00120.00200. 00Input CVN3.5037.0054. 0030. 0043. 50106.0093. 50107.5082. 0097. 50107.50Computed CVN18.3726.3436.9149. 7063. 5876. 9588.4597. 41103. 87115.14115.30Differential

-14.8710.6617.09-19.70-20. 0829. 055.0510.09-21.87-13.64-7. 80Figure 2-4Browns Ferry 2 weld BF2 ESW unirradiated Charpy energy plot2-10 Materials and Test Specimen Description WELD HEAT ESW (BF2)Page 2Plant: BROWNS FERRY 2 Material:

ESW Heat: ESWOrientation:

NA Capsule:

UNIRRA Fluence:

0.0 n/cm^2Charpy V-Notch DataTemperature 300. 00Input CVN143. 00Correlation Coefficient

= .891Computed CVN115.94Differential

27. 06Figure 2-4 (continued)

Browns Ferry 2 weld BF2 ESW unirradiated Charpy energy plot2-11 Materials and Test Specimen Description 2.3 Capsule OpeningThe surveillance capsule was opened on April 11, 2012. As shown in Figure 2-1, the 1200capsule consisted of a single basket attached to the lead tube. The outside of the capsule hadidentification markings which could be clearly read. The capsule was marked with the correctbinary reactor code number 26, which agrees with the markings reported on GE Drawing129B3578

entitled, "Surveillance Program Number Identification of GE Reactors."

The capsulebasket was also engraved with drawing number 1 17C4060B007, Part Number 7, and thismarking is consistent with the markings observed on the 300 capsule basket described in [8]. Asindicated in Figure 2-2, the capsule basket contained two Charpy packets and four tensile tubes.Referring to Figure 2-1, the lead tube is positioned on the top surface of the basket in thephotograph.

Therefore, the surface that is facing down in the photograph was facing the vesselduring irradiation.

The hook at the top of the photograph is the vessel lower attachment hook,and it was on the bottom of the capsule when the capsule was installed in the plant. The Charpypacket end tabs are on the right side in Figure 2-1. Moving up from the bottom of the capsule,the first item in the capsule was tensile tube G6, then tensile tube G7, followed by Charpy packetG9. Above Charpy packet G9, the remaining two tensile tubes G8 and G9 were loaded withCharpy packet G 10 at the highest elevation in the plant.Attention was paid to the location of the Charpy specimens and the dosimetry wire locations during disassembly of the Charpy packets.

Each packet was found to contain one Fe, one Cu, andone Ni dosimetry wire along the ends of the Charpy specimens at the locations shown in Figure2-2. The wire locations along the ends of the Charpy specimens were on the top side of theCharpy packets when the capsule was in the plant. Therefore, the wires were irradiated in ahorizontal position in the reactor.

The identifications assigned to the dosimetry wires indicate theCharpy packet from which they were recovered.

The wires and Charpy specimens were placed inindividually marked containers for positive identification throughout the work.2-12 3NEUTRON FLUENCE CALCULATION The Browns Ferry 2 1200 capsule was irradiated for 16 cycles of ofieration.

The surveillance capsule was placed in the reactor's 120' capsule holder prior to cycle 1 and was removedfollowing cycle 16 for a total irradiation period of 22.9 effective full power years (EFPY). Thesurveillance capsule included copper, iron, and nickel flux wire dosimetry specimens.

Evaluation of the surveillance capsule specimens requires knowledge of the neutron irradiation environment.

The neutron flux density, neutron energy spectrum, and neutron fluence arerequired at the surveillance capsule location.

The NRC has established guidelines in Regulatory Guide 1.190 [11] for determining best estimate values of flux, energy spectrum, and fluence forRPV damage assessments using particle transport methods.

These guidelines are not specifically intended for use in surveillance capsule evaluations;

however, they do provide a suitableframework to support the development of accurate neutron transport analysis models forsurveillance capsule evaluations.

This report documents the application of the modeling and analysis guidelines provided in [ 11]to determine the surveillance capsule accumulated irradiation and capsule specimen neutronfluence of the Browns Ferry 2 120' ISP capsule flux wires. Additionally, the accumulated irradiation for the 30' capsule, removed at the end of cycle 7, and the 300 flux wires, removed atthe end of cycle 1, were determined.

The neutron fluence was also calculated for the 3000capsule at the projected time of removal, and for the 120' capsule at the end of cycle 16 and atthe end of the reactor's extended design life of 54 EFPY. The fluence and activation valuespresented in this report were calculated using the RAMA Fluence Methodology

[12] (hereinafter referred to as "RAMA").

The specific activities predicted by RAMA are compared to the activitymeasurements reported in Appendix A.RAMA was developed for the Electric Power Research Institute, Inc. (EPRI) and the BoilingWater Reactor Vessel and Internals Project (BWRVIP) for the purpose of calculating neutronfluence in Boiling Water Reactor (BWR) components.

As prescribed in Regulatory Guide 1.190,RAMA has been benchmarked against industry standard benchmarks for both pressurized waterreactor (PWR) and BWR designs.

In addition, RAMA has been compared with several plant-specific dosimetry measurements and reported fluence from several commercial operating reactors.

The results of the benchmarks and comparisons to measurements show that RAMAaccurately predicts specimen activities, RPV fluence, and vessel internal component fluence inall light water reactor types. Under funding from EPRI and the BWRVIP, the RAMAmethodology has been reviewed by the U.S. NRC and subsequently given generic approval fordetermining fast neutron fluence in BWR pressure vessels [13] and vessel internal components that include the core shroud and top guide [14].3-1 Neutron Fluence Calculation 3.1 Description of the Reactor SystemThis section provides an overview of the reactor design and operating data inputs that were usedto develop the Browns Ferry 2 reactor fluence model. All reactor design and operating datainputs used to develop the model were plant-specific and were provided by Tennessee ValleyAuthority (TVA). The inputs for the fluence geometry model were developed from design andas-built drawings for the reactor pressure vessel, vessel internals, fuel assemblies, andcontainment regions.

The reactor operating data inputs were developed from core simulator datathat provided a historical accounting of how the reactor operated for cycles 9 through 16. Coresimulator data was not available for cycles 1 through 8. Data for these cycles was approximated using information from the following sources:

1) cycle summary reports,

2) spreadsheet data,and 3) cycle 9 data for axial power shapes, water densities, and fuel assembly orientation.

Projections for cycle 17 to the end of the reactor's extended operating life were based on thereactor's operating history for cycle 16.3.1.1 Reactor System Mechanical DesignBrowns Ferry 2 is a General Electric BWR/4 class reactor with a core loading of 764 fuelassemblies.

Browns Ferry 2 began commercial operation in 1975 with a design rated power of3293 MWt. At the beginning of cycle 11 the rated power was increased to 3458 MWt. At thetime of this fluence analysis, Browns Ferry 2 had completed 16 cycles of operation.

Figure 3-1 illustrates the basic planar configuration of the Browns Ferry 2 reactor at an axialelevation near the reactor core mid-plane.

All of the radial regions of the reactor that are requiredfor fluence projections are shown. Beginning at the center of the reactor and projecting outward,the regions include:

the core region, including control rod locations and fuel assembly locations (fuel locations are shown only for the 0 to 90 degree quadrant);

core reflector region (bypasswater); central shroud wall; downcomer water region including the jet pumps; reactor pressurevessel (RPV) wall; cavity between RPV and insulation; insulation; cavity region between theinsulation and biological shield; and biological shield (concrete wall).The mechanical design inputs that were used to construct the Browns Ferry 2 fluence geometrymodel included as-built and nominal design dimensional data. As-built data for the reactorcomponents and regions of the reactor system is always preferred when constructing plant-specific models; however, as-built data is not always available.

In these situations, nominaldesign information is used.3-2 Neutron Fluence Calculation Reacto NodhS. West0*33W 3W Surveilance CaosUle" Jt Pumnp A~ssemnbly

++ + +,+ + .,+ +

27------------(D+::.:~::i Cot 9Rellctorta

+ + + '..+

g;+++ + + -,+,#,,g 4; WA ; ,9 ;9X+ + + +

-Sg oI0 + + + +- 't+++r~lrP~

I 1L*Reator PEacu270 4- 0 .. ." -----+ ,."- + + ," + t .." .Reactor lado+ + ,+ ' + +4e' + ++ + essel and Clod0 + + -+ + + 0,+++++ + + +4 .... ..Vssd Irsulaioud ClodN,'÷+ + + + + ,24 F 12(rThm~~avt dnRgksnn t meF= Fud unde im

um u 1W(Locvfns slm =*I += ("m~nI md bnlýFigure 3-1Planar view of Browns Ferry 2 at the core mid-plane elevation For the Browns Ferry 2 fluence model, the predominant dimensional information used toconstruct the fluence model was nominal design data. As-built data was available for thefollowing dimensions:

" Jet pump mixer pipe radial location* Reactor pressure vessel cladding inner radiusAnother important component of the fluence analysis is the accurate description of thesurveillance capsules in the reactor.

It is shown in Figure 3-1 that three surveillance capsuleswere initially installed in the Browns Ferry 2 reactor.

The capsules were attached radially to theinside surface of the RPV (looking outward from the core region) at the 300, 1200, and 300'azimuths.

Surveillance capsules are used to monitor the radiation accumulated in the reactor overa period of time. The importance of surveillance capsules in fluence analyses is that they containflux wires that are irradiated during reactor operation.

When a capsule is removed from thereactor, the irradiated flux wires are evaluated to obtain activity measurements.

Thesemeasurements are used to validate the fluence model. Three sets of flux wires have beenremoved from the Browns Ferry 2 reactor and analyzed.

(See Section 3.4, which presents acomparison of the calculated-to-measured capsule results.)

3-3 Neutron Fluence Calculation 3.1.2 Reactor System MaterialEach region of the reactor is comprised of materials that include reactor fuel, steel, water,insulation,

concrete, and air. Accurate material information is essential for the fluence evaluation as the material compositions determine the scattering and absorption of neutrons throughout thereactor system and, thus, affect the determination of neutron fluence in the reactor components.

Table 3-1 provides a summary of the materials for the various components and regions of theBrowns Ferry 2 reactor.

The material attributes for the steel, insulation,

concrete, and aircompositions (i.e. material densities and isotopic concentrations) are assumed to remain constantfor the operating life of the reactor.

The attributes of the fuel compositions in the reactor coreregion change continuously during an operating cycle due to changes in power level, fuelbumup, control rod movements, and changing moderator density levels (voids).

Because of thedynamics of the fuel attributes with reactor operation, several state-point data sets are used todescribe the operating states of the reactor for each operating cycle. The number of data sets usedin this analysis is presented in Section 3.1.3.3.3-4 Neutron Fluence Calculation Table 3-1Summary of material compositions by region for Browns Ferry 2Control Rods and Guide TubesCore Support PlateFuel Support PieceFuel Assembly Lower Tie PlateReactor CoreReactor Coolant / Moderator Core Reflector Fuel Assembly Upper Tie PlateTop GuideCore Spray Sparger PipesCore Spray Sparger Flow AreasShroudDowncomer RegionJet Pump Riser and Mixer Flow AreasJet Pump Riser and Mixer MetalJet Pump Riser Brace and PadsSurveillance Capsule Specimens Reactor Pressure Vessel CladReactor Pressure Vessel WallCavity RegionsInsulation CladInsulation Biological Shield CladBiological Shield WallStainless Steel, B4CStainless SteelStainless SteelStainless Steel, Zircaloy, Inconel235U, 238U, 239Pu, 240PU, 241Pu, 242pu, Ofue 0 ZircaloyWaterWaterStainless Steel, Zircaloy, InconelStainless SteelStainless SteelWaterStainless SteelWaterWaterStainless SteelStainless SteelCarbon SteelStainless SteelCarbon SteelAir (Nitrogen)

Stainless SteelAluminum FoilCarbon SteelReinforced Concrete3.1.3 Reactor Operating Data InputsAn accurate evaluation of reactor vessel and component fluence requires an accurate accounting of the reactor's operating history.

The primary reactor operating parameters that affect thedetermination of fast neutron fluence in light water reactors include reactor power levels, corepower distributions, coolant water density distributions, and fuel material (isotopic) distributions.

3-5 Neutron Fluence Calculation 3.1.3.1 Core LoadingIt is common in BWRs that more than one fuel assembly design may be loaded in the reactorcore in any given operating cycle. For fluence evaluations, it is important to account for the fuelassembly designs that are loaded in the core in order to accurately represent the neutron sourcedistribution at the core boundaries (i.e. peripheral fuel locations and the top and bottom fuelelevations).

Eleven different fuel assembly designs were loaded in the Browns Ferry 2 reactor during theperiod included in this evaluation.

Table 3-2 provides a summary of the fuel designs loaded inthe reactor core for each evaluated operating cycle. The cycle core loading provided by TVA wasused to identify the fuel assembly designs in each cycle and their location in the core loadinginventory.

(Note that fuel loadings for cycle 12 were divided into three individual periods,identified as 12A, 12B and 12C.) For each cycle, appropriate fuel assembly models were used tobuild the reactor core region of the Browns Ferry 2 RAMA fluence model.3.1.3.2 Power History DataReactor power history is the measure of reactor power levels and core exposure on a continual orperiodic basis. For this fluence evaluation, the power history for the Browns Ferry 2 reactor wasdeveloped from power history inputs provided by TVA. The power history data showed thatBrowns Ferry 2 started commercial operation with a design rated thermal power of 3293 MWtfor cycles 1 through 10 and implemented a measurement uncertainty recovery (MUR) poweruprate of 3458 MWt at the beginning of cycle 11. It was assumed in this analysis that all futurecycles would operate at the 3458 MWt power level.The power history data for Browns Ferry 2 included daily power levels for most cycles. Whendaily power histories were not available, average power levels were constructed based onexposure accumulation using the core simulator codes. This data was used to calculate thecapsule and vessel fluences.

Periods of reactor shutdown due to refueling outages and otherevents were also accounted for in the model. The power history data was verified by comparing the calculated energy production in effective full power years with power production recordsprovided by TVA. Table 3-3 lists the accumulated EFPY at the end of each cycle for this fluenceevaluation.

3-6 Neutron Fluence Calculation Table 3-2Summary of the Browns Ferry 2 core loading inventory 1 764 GE32 632 132 GE33 364 168 232 GE34 124 168 232 240 GE35 123 153 488 GE46 70 354 336 4 GE77 212 388 4 160 GE78 4 580 4 176 GE79 348 216 200 GE710 148 200 200 216 GE911 48 200 516 GEl1 I12A 764 GEl 312B 764 GE1312C 49 715 GEl313 111 281 372 GEI114 149 335 280 GE1315 111 653 GE1416+2 764 ATRIUM-10 Due to lack of detailed information on QUAD+ design, these lead test assemblies (LTAs) were modeled as GE6 bundles.2 Cycles 17 and beyond use cycle 16 data for projecting fluence to the end of the extended plant license period.3-7 Neutron Fluence Calculation 3.1.3.3 Reactor State-Point DataCycles 1 through 8 of Browns Ferry 2 were derived from Cycle Summary Reports and plant dataspreadsheets, which contained bundle average exposures.

Radial power shapes were calculated based on this bundle average exposure data. Data from cycle 9 was used to provide axial powershapes, water densities, and fuel assembly orientation for cycles 1 through 8. This represented the best available information for these cycles.Core simulator data was provided by TVA to characterize the historical operating conditions ofBrowns Ferry 2 for cycles 9 through 16 and cycle projections.

The data calculated with coresimulator codes represents the best-available information about the reactor core's operating history over the reactor's operating life. In this analysis, the core simulator data provided by TVAwas processed by TransWare to generate state-point data files for input to the RAMA fluencemodel. The state-point files included three-dimensional data arrays that described core powerdistributions, fuel exposure distributions, fuel materials (isotopics),

and coolant water densities.

A separate neutron transport calculation was performed for each of the state points tallied inTable 3-3. The calculated neutron flux for each state point was combined with the appropriate power history data described above in order to provide an accurate accounting of the fast neutronfluence for the reactor pressure vessel. Fluence projections to the end of the reactor's design lifeand extended design life were performed using projected equilibrium cycles. Equilibrium cyclesare discussed below.Beginning of Operation through Cycle 16 State PointsA total of 170 state points were used to represent the operating history for the first 16 operating cycles of Browns Ferry 2. These state points were selected from hundreds of exposure points thatwere calculated with the core simulator code. The hundreds of exposure points were evaluated and grouped into a fewer number of exposure ranges in order to reduce the number of transport calculations required to perform the fluence evaluation.

Several criteria were used in thedetermination of the exposure ranges, including evaluations of core thermal powers, core flows,core power profiles, and control rod patterns.

In determining exposure ranges, it is assumed thatthere will be at least one exposure step in that range that would accurately represent the averageoperating conditions of the reactor over that range. This single exposure step is then referred toas the "state point". Table 3-3 shows the number of state points used for each cycle in thisfluence evaluation.

Projected Reactor Operation Projections of plant operations beyond cycle 16 are represented with an "equilibrium" cycle thatincorporates the best-available information on expected cycle length, fuel bundle loading, andoperating strategies for future cycles. Cycle 16, at a thermal power level of 3458 MWt, is used asthe equilibrium cycle for this analysis to project fluence to the end of the extended plant designlife of 54 EFPY.3-8 Neutron Fluence Calculation Table 3-3State-point data for each cycle in Browns Ferry 21 3 3293 1.32 3 3293 2.03 3 3293 3.04 3 3293 4.4567891011121314151617+ 2333320171917211917161632933293329332933293329334583458345834583458345834585.56.88.19.410.712.214.015.717.619.421.322.954.0'The rated thermal power is listed for each cycle. The actual power levels were usedfor the individual state-point calculations for cycles 1-16.2 Cycles 17 and beyond use cycle 16 data for projecting fluence to the end of theextended plant license.Limitation of Fluence Projections Some of the fluence values presented in this report are based on projections of the operation ofBrowns Ferry 2 beyond the current operating cycle. Projections are performed using an assumedequilibrium cycle. The significance of the equilibrium cycle is that it defines the flux profiles thatare used to project fluence into the future. Providing that the design basis for the equilibrium cycle does not change appreciably, projections based on the equilibrium cycle should remainbounding through 54 EFPY to support licensing, in-service inspection, and flaw evaluation activities.

3-9 Neutron Fluence Calculation If the design basis for the equilibrium cycle changes at any point in time that would result in asignificant change to the flux profiles for the equilibrium cycle, then a new evaluation is needed.Operating conditions, if changed, that could impact the validity of the equilibrium cycle includepower uprates, introduction of new fuel designs, changes in projected cycle lengths, changes incore loading strategies, changes in reactor flow, or other changes that could alter the flux profilesused in the fluence projections.

3.2 Calculation Methodology The Browns Ferry 2 capsule evaluation was performed using the RAMA Fluence Methodology software package [12]. RAMA and the application of RAMA to the Browns Ferry 2 reactor aredescribed in this section.3.2.1 Description of the RAMA Fluence Methodology The RAMA Fluence Methodology (RAMA) is a system of computer codes, a data library, and anuncertainty methodology that determines best-estimate fluence in light water reactor pressurevessels and vessel components.

The primary codes that comprise the RAMA methodology include model builder codes, a particle transport code, and a fluence calculator code. The datalibrary contains nuclear cross sections and response functions that are needed for each of thecodes. The uncertainty methodology is used to determine the uncertainty and bias in the best-estimate fluence calculated by the software.

The primary inputs for RAMA are mechanical design parameters and reactor operating historydata. The mechanical design inputs are obtained from plant-specific design drawings, whichinclude as-built measurements when available.

The reactor operating history data is obtainedfrom reactor core simulator codes, system heat balance calculations, daily operating logs, andcycle summary reports that describe the operating conditions of the reactor over its operating lifetime.

The primary outputs from RAMA calculations are neutron flux, neutron fluence,dosimetry activation, and an uncertainty evaluation.

The model builder codes consist of geometry and material processor codes that generate input forthe particle transport code. The geometry model builder code uses mechanical design inputs andmeshing specifications to generate three-dimensional geometry models of the reactor.

Thematerial processor code uses reactor operating data inputs to process fuel materials, structural materials, and water densities that are consistent with the geometry meshing generated by thegeometry model builder code.The particle transport code performs three-dimensional neutron flux calculations using adeterministic, multigroup, particle transport theory method with anisotropic scattering.

Theprimary inputs prepared by the user for the transport code include the geometry and material datagenerated by the model builder codes and numerical integration and convergence parameters forthe iterative transport calculation.

The transport solver is coupled with a general geometrymodeling capability based on combinatorial geometry techniques.

The coupling of generalgeometry with a deterministic transport solver provides a flexible,

accurate, and efficient tool forcalculating neutron flux in light water reactor pressure vessels and vessel components.

Theprimary output from the transport code is the neutron flux in multigroup form.3-10 Neutron Fluence Calculation The fluence calculator code determines fluence and activation in the reactor pressure vessel andvessel components over specified periods of reactor operation.

The primary inputs to the fluencecalculator include the multigroup neutron flux from the transport code, response functions for thevarious materials in the reactor, reactor power levels for the operating periods of interest, thespecification of which components to evaluate, and the energy ranges of interest for evaluating neutron fluence.

The fluence calculator includes treatments for isotopic production and decaythat are required to calculate specific activities for irradiated materials.

The reactor operating history is generally represented with several reactor state points that represent the various powerlevels and core power shapes generated by the reactor over the life of the plant. These detailedstate points are combined with the daily reactor power levels to produce accurate estimates of thefluence and activations accumulated in the plant.The uncertainty methodology provides an assessment of the overall accuracy of the RAMAFluence Methodology.

Variances in the dimensional data, reactor operating data, dosimetry measurement data, and nuclear data are evaluated to determine if there is a statistically significant bias in the calculated results that might affect the determination of the best-estimate fluence for the reactor.

The plant-specific results are also weighted with comparative results fromexperimental benchmarks and other plant analyses and analytical uncertainties pertaining to themethodology to determine if the plant-specific model under evaluation is statistically acceptable as defined in Regulatory Guide 1.190.The RAMA nuclear data library contains atomic mass data, nuclear cross-section data, andresponse functions that are needed in the material processing, transport,

fluence, and reaction ratecalculations.

The cross-section data and response functions are based on the BUGLE-96 nucleardata library [15] and the VITAMIN-B6 data library [16].The RAMA Fluence Methodology is described in the Theory Manual [ 17]. The generalprocedures for using the methodology are presented in the Procedures Manual [ 18].3.2.2 The RAMA Geometry Model for the Browns Ferry 2 ReactorSection 3.1 described the design inputs that were provided by TVA for the Browns Ferry 2.reactor fluence evaluation.

These design inputs were used to develop a plant-specific, three-dimensional computer model of the Browns Ferry 2 reactor with the RAMA FluenceMethodology.

Figures 3-2 and 3-3 provide general illustrations of the primary components, structures andregions developed for the Browns Ferry 2 fluence model. Figure 3-2 shows the planarconfiguration of the reactor model at an elevation corresponding to the reactor core mid-plane elevation.

Figure 3-3 shows an axial configuration of the reactor model. Note that the figures arenot drawn to scale. They are intended only to provide a perspective for the layout of the model,and specifically how the various components, structures, and regions lie relative to the reactorcore region (i.e., the neutron source).Because the figures are intended only to provide a general overview of the model, they do notinclude illustrations of the geometry meshing developed for the model. To provide such detail isbeyond the scope of this report.3-11 Neutron Fluence Calculation 0A,435. 81 (17.n431.17 (teGIln3K 113640 Oute r c" v Bkd o gd k349 .2 ( N -. 5 ") --O u_ S hC lNCO (01345 -33L0(3.3)-Inner

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--22N.O1 0 .0)23 .I ...o22 1 '29 I lI I I 9"her Fud Peiberii Fud Co ShiM RPV Clad Bd SlkwAssemblies Assemblies Reactor InnerClad Nobs. Th dmwmg a wtta s=do-Dmmsins m gvein I ceimelmg (uclus)_h noundt swmebv messcamsties vmnsmt Im W. 120 Mnid 3W? capsuesFigure 3-2Planar view of the Browns Ferry 2 RAMA quadrant model at the core mid-plane elevation 3-12 Neutron Fluence Calculation SPEOid HeW FfSWOW HOWd FI NL- wPSW PfWUUbP Gdd PloUfmF~d IW hd -z.93043C(3S31-)

Em--EIL0Fromca00caftd;Rod Ddm;: RIOnIEa,P--V~dC-2-3-9941(3OI~nPumplr BraeF,'EMC-1-25&A592a0000,02,ECL_E0a.U0aU549.43M(26.31-)

Lm- -F.Ph WEWVa .ccm St RobLamner dR~ " o RAMA& ModdF III- 1dnMbgig ntdI sCieknialsios me mmie a inge s (kinceFigure 3-3Axial view of the Browns Ferry 2 RAMA modelThe following subsections provide an overview of the computer models that were developed forthe various components, structures, and coolant flow regions of the Browns Ferry 2 reactor.3-13 Neutron Fluence Calculation 3.2.2.1 The Geometry ModelRAMA uses a generalized three-dimensional geometry modeling system that is based on acombinatorial geometry technique, which is mapped to a Cartesian coordinate system. In thisanalysis, an axial plane of the reactor model is defined by the (x,y) coordinates of the modelingsystem, and the axial elevation at which a plane exists is defined along a perpendicular z-axis ofthe modeling system. Thus, any point in the reactor model can be addressed by specifying the(x,y,z) coordinates for that point.Figure 3-1 illustrates a planar cross-section view of the Browns Ferry 2 reactor design at an axialelevation corresponding to the reactor core mid-plane elevation.

It is shown for this one elevation that the reactor design is a complex geometry composed of various combinations of rectangular, cylindrical, and wedge-shaped bodies. When the reactor is viewed in three dimensions, thevarying heights of the different components, structures, and regions create additional geometrymodeling complexities.

An accurate representation of these geometrical complexities in apredictive computer model is essential for calculating

accurate, best-estimate fluence in thereactor pressure vessel, the vessel internals, and the surrounding structures.

Figures 3-2 and 3-3 provide general illustrations of the planar and axial geometry complexities that are represented in the Browns Ferry 2 fluence model. For comparison

purposes, the planarview illustrated in Figure 3-2 corresponds to the same core mid-plane elevation illustrated inFigure 3-1. The computer model for Browns Ferry 2 assumes azimuthal quadrant symmetry inthe planar dimension.

Figure 3-2 illustrates the quadrant geometry that was modeled in this analysis.

In terms of themodeling coordinate system, the "northeast" quadrant of the geometry is represented in themodel. The 00 azimuth, which has a "north" designation, corresponds to the 00 azimuthreferenced in the plan drawings for the reactor pressure vessel. Degrees are incremented clockwise.

Thus, the 900 azimuth is designated as the "east" direction.

All other components, structures, and regions have been appropriately mirror reflected or rotated to this quadrant basedupon their relationship to the pressure vessel orientation to ensure that the fluence isappropriately calculated relative to the neutron source (i.e., the core region).

Although symmetryis a modeling consideration, the results presented in this report for the different components andstructures are given at their correct azimuths in the plant.Figure 3-3 illustrates the axial configuration of the primary components, structures, and regionsin the fluence model. For discussion

purposes, the same components, structures, and regionsshown in the planar view of Figure 3-2 are also illustrated in Figure 3-3. Figure 3-3 shows thatthe axial height of the fluence model spans from a lower elevation just below the jet pump riserinlet to above the core shroud head flange. This axial height covers all areas of the reactorpressure vessel that are expected to exceed a fluence threshold of l.OE+17 n/cm2 at 54 EFPY.As previously noted, Figures 3-2 and 3-3 are not drawn precisely to scale. They are intendedonly to provide a perspective of how the various components, structures, and regions of thereactor are positioned relative to the reactor core region (i.e., the neutron source) and each other.The following subsections provide details on the modeling of individual components, structures, and regions.

Please refer to the figures for visual orientation of the components and regionsdescribed in the following subsections.

3-14 Neutron Fluence Calculation 3.2.2.2 The Reactor Core and Core Reflector ModelsThe reactor core contains the nuclear fuel that is the source of the neutrons that irradiate allcomponents and structures of the reactor.

The core is surrounded by a shroud structure thatserves to channel the reactor coolant through the core region during reactor operation.

The regionbetween the core and the core shroud is the core reflector, and it contains coolant.

The reactorcore geometry is rectangular in design and is modeled with rectangular elements to preserve itsshape in the analysis.

The core reflector region interfaces with the rectangular shape of the coreregion and the curved shape of the core shroud. It is, therefore, modeled using a combination ofrectangular and cylindrical elements.

The core region is centered in the reactor pressure vessel and is characterized in the analysis withtwo fundamental fuel zones: interior fuel assemblies and peripheral fuel assemblies.

Theperipheral fuel assemblies are the primary contributors to the neutron source in the fluencecalculation.

Because these assemblies are loaded at the core edge where neutron leakage fromthe core is greatest, there is a sharp power gradient across these assemblies that requiresconsideration.

To account for the power gradient, the peripheral fuel assemblies are sub-meshed with additional rectangular elements that preserve the pin-wise details of the fuel assemblygeometry and power distribution.

The interior fuel assemblies make a lesser contribution to thereactor fluence and are, therefore, modeled in various homogenized forms in accordance withtheir contributions to the reactor fluence.

For computational efficiency, homogenization treatments are used in the interior core region primarily to reduce the number of mesh regionsthat must be solved in the transport calculation.

The meshing configuration for each fuelassembly location in the core region is determined by parametric studies to ensure an accurateestimate of fluence throughout all regions of the reactor system.Each fuel assembly design, whether loaded in the interior or peripheral locations in the core, isrepresented with four axial material zones: the lower tie plate/end plug zone, the fuel zone, thefuel upper plenum zone, and the upper tie plate/end plug zone. The structural materials in the topand bottom nozzles for each unique assembly design are represented in the model to address theshielding effects that these materials have on the components.

above and below the core region.The fuel zone contains the nuclear fuel and structural materials for the fuel assemblies.

Thematerials for each fuel assembly are unique during reactor operation and are incorporated intothe model using reactor operating data from core simulator codes. The upper plenum regioncaptures fission gases during reactor operation.

The Browns Ferry 2 reactor core region has a nominal elevation for the bottom of active fuel at549.43 cm (216.31 in.) and an active fuel height of 381.00 cm (150 in.). Browns Ferry 2 loadedfuel designs with active fuel heights ranging from 146" to 150". The core simulator codes usedby Browns Ferry 2 modeled the core as 150" in all situations, so this value was also used in theRAMA model. Since the predominant peripheral fuel designs throughout Browns Ferry 2'shistory were 150" in height, the effect of this approximation on the component fluence will benegligible.

From an isotopic standpoint, the core is modeled using quadrant symmetry.

For the 300 and 1200capsule evaluations, as well as the peak RPV fluence calculations, the NE fuel quadrant wasused.3-15 Neutron Fluence Calculation 3.2.2.3 The Core Shroud ModelThe core shroud is a canister-like structure that contains the reactor core and channels the reactorcoolant and steam produced by the core into the steam separators.

Axially the shroud extendsfrom the lower shroud wall to the top of the shroud head rim in the model. The core shroud iscylindrical in design and is modeled with pipe elements.

3.2.2.4 The Downcomer Region ModelThe downcomer region lies between the core shroud and the reactor pressure vessel. It isbasically cylindrical in design, but with some geometrical complexities created by the presenceof jet pumps and surveillance capsules in the region. The majority of the downcomer region ismodeled with pipe segments.

The areas of the downcomer containing the jet pumps andspecimen capsules are modeled with the appropriate geometry elements to represent their designfeatures and to preserve their radial, azimuthal, and axial placement in the downcomer region.These structures are described further in the following subsections.

3.2.2.5 The Jet Pump ModelThere are ten jet pump assemblies in the downcomer region of Browns Ferry 2, which providethe main recirculation flow for the core. The jet pumps are modeled at azimuths 30, 60, and 900in the downcomer region. When symmetry is applied to the model, the 30' location represents the jet pump assemblies that are positioned azimuthally at 30, 150, 210, and 3300; the 600location represents those at 60, 120, 240, and 300'; and the 90' location represents the jet pumpassemblies at 90 and 2700. Note that there are no jet pumps present at the 0- and 1800 azimuthsof the reactor.The jet pump model includes representations for the riser, mixer, and diffuser pipes; nozzles;rams head; riser inlet pipe; and riser brace yoke, leafs, and pads. The jet pump assembly design ismodeled using cylindrical pipe elements for the jet pump riser and mixer pipes. The riser pipe iscorrectly situated between the centers of the mixer pipes. The riser brace assembly modelincludes two leaf structures that attach to the yoke and pad elements.

3.2.2.6 The Surveillance Capsule ModelSection 3.1 describes the three surveillance capsules installed in the Browns Ferry 2 reactor.

Thesurveillance capsules are installed near the inner surface of the pressure vessel wall. Thesurveillance capsules are rectangular in design. Because of this shape, the capsules are not easilyimplemented in the otherwise cylindrical elements of the downcomer region model. Withreference to Figure 3-1, it is observed that the capsules are of small dimensions in the planargeometry and they reside a long distance (view factor) from the core region. Based on thesefactors, the otherwise rectangular shape of the surveillance capsules can be reasonably approximated in the model with arc elements.

The surveillance capsule model also includes arepresentation for the downcomer water that surrounds the capsule on all sides.The surveillance capsules are correctly modeled behind the jet pump riser pipes at the 30- and600 azimuths.

When symmetry is applied to the model, the 300 location represents the capsuleinstalled at 300, while the 600 location represents the capsules at 120 and 3000.3-16 Neutron Fluence Calculation The surveillance capsules are modeled at their correct axial position and height relative to thecore region. The surveillance capsules cover about nine percent of the total core height.3.2.2.7 The Reactor Pressure Vessel ModelThe reactor pressure vessel and vessel cladding lie outside the downcomer region and each iscylindrical in design. Both are modeled with pipe elements.

The cladding-pressure vesselinterface is a key location for RPV fluence calculations and is preserved in the model. Thisinterface defines the inside surface (OT) for the pressure vessel base metal where the RPVfluence is calculated.

Browns Ferry 2 has cladding only on the inside surface of the pressurevessel wall.3.2.2.8 The Vessel Insulation ModelThe vessel insulation lies in the cavity region outside the pressure vessel wall. The insulation iscylindrical in design and follows the contour of the pressure vessel wall. It is modeled with pipeelements.

3.2.2.9 The Inner and Outer Cavity ModelsThe cavity region lies between the pressure vessel and biological structures.

As previously described, the vessel insulation lies in the cavity region; thus creating two cavity regions.

Theinner cavity region lies between the vessel and the insulation.

The outer cavity region liesbetween the vessel insulation and biological shield cladding.

The boundaries of the cavityregions follow the contours of the pressure vessel, vessel insulation, and biological shield. Thecavity regions are essentially cylindrical in design and are modeled with pipe segments.

3.2.2.10 The Biological Shield ModelThe biological shield (concrete) defines the outer most region of the fluence model. Thebiological shield is basically cylindrical in design and is modeled with pipe segments.

There iscladding on the inside and outside surface of the biological shield.3.2.2.11 The Above-Core Component ModelsFigure 3-3 includes illustrations of other components and regions that lie above the reactor coreregion. The predominant above-core components represented in the model include the top guideand core spray spargers.

The Top Guide ModelThe top guide component lies above the core region. The top guide is appropriately modeled byincluding representations for the vertical fuel assembly parts and top guide plates. The upper fuelassembly parts that extend into the top guide region are modeled in three axial segments:

the fuelrod plenum, fuel rod upper end plugs, and fuel assembly upper tie plate. The fuel assembly partsand top guide plates are modeled with rectangular elements.

3-17 Neutron Fluence Calculation The Core Spray Sparger ModelThe core spray spargers include upper and lower sparger pipes and a vertical inlet pipe. The corespray spargers are appropriately represented as torus structures in the model. The sparger pipesreside inside the upper shroud wall above the top guide. The spargers are modeled as pipe-like structures and include a representation of reactor coolant inside the pipes.3.2.2.12 The Below-Core Component ModelsFigure 3-3 includes illustrations of other components and regions that lie below the reactor coreregion. The fuel support piece, core support plate, and core inlet regions appropriately include arepresentation of the cruciform control rod below the core region. The lower fuel assembly partsinclude representations for the fuel rod lower end plugs, lower tie plate, and nose piece. Thebelow-core components are modeled with rectangular elements with the exception of the coresupport plate. The core support plate is modeled using both rectangular and cylindrical elementsto provide an appropriate representation of that component.

3.2.2.13 Summary of the Geometry Modeling ApproachTo summarize the reactor modeling

process, there are several key features of the RAMA codesystem that allow the reactor design to be accurately represented for RPV and capsule fluenceevaluations.

Following is a summary of some of the key features of the model." Rectangular, cylindrical, and wedge bodies are mixed in the model in order to provide anaccurate geometrical representation of the components and regions in the reactor.* The reactor core geometry is modeled with rectangular bodies to represent its actual shape inthe reactor.

The fuel assemblies in the core region are also sub-meshed with additional rectangular bodies to represent the pin cell regions in the assemblies.

A combination of rectangular and cylindrical bodies is used to describe the transition partsbetween the rectangular core region and the cylindrical outer core regions." Cylindrical and wedge bodies are used to model the components and regions that extendoutward from the core region (core shroud, downcomer, RPV, etc.).* The surveillance capsules are modeled at their correct radial, azimuthal, and elevational positions behind the jet pumps in the downcomer region.* The above-core region includes accurate representations of the top guide and core sprayspargers.

The below-core region includes appropriate representations for the fuel support piece, coresupport plate, core inlet regions, cruciform control rods, and control rod drives.* The biological shield is appropriately represented as a cylindrical body.3.2.3 RAMA Calculation Parameters The RAMA transport code uses a three-dimensional deterministic transport method to calculate the neutron flux. The accuracy of the transport method is based on a numerical integration technique that uses ray-tracing to characterize the geometry, anisotropy treatments to determine 3-18 Neutron Fluence Calculation the directional flow of particles, and convergence parameters to determine the overall accuracyof the flux solution between iterates.

The code allows the user to specify values for each of theseparameters.

The primary input parameters that control the ray-tracing calculation are the distance betweenparallel rays in the planar and axial dimensions, the depth that a particle is tracked when areflective boundary is encountered, and the number of equally spaced angles in polar coordinates for tracking the particles.

Plant-specific values are determined for each of the parameters.

TheRAMA transport calculation employs a treatment for anisotropy that is based on a Legendreexpansion of the scattering cross sections.

By default, the RAMA transport calculation uses themaximum order of expansion that is available for each nuclide in the RAMA nuclear datalibrary.

For the actinide and zirconium

nuclides, a P5 expansion of the scattering cross sections isused. For all other nuclides, a P7 expansion of the scattering cross sections is used.The overall accuracy of the neutron flux calculation is determined using an iterative technique toconverge the flux iterations.

The convergence criterion used in the evaluation was determined byparametric study to provide an asymptotic solution for this model.3.2.4 RAMA Neutron Source Calculation RAMA calculates a unique neutron source distribution for each transport calculation using theinput relative power density factors for the fuel region and data from the RAMA nuclear datalibrary.

The source distribution changes with fuel burnup; thus, the source is determined usingcore-specific three-dimensional burnup distributions at frequent intervals throughout a cycle. Forthe fluence model, the peripheral fuel assemblies are modeled to preserve the power gradient atthe core edge that is formed from the pin-wise source distributions in these fuel assemblies.

3.2.5 RAMA Fission SpectraRAMA calculates a weighted fission spectrum for each transport calculation that is based on therelative contributions of 235U, U, 23 Pu, 2 Pu, Pu, and 242Pu isotopes.

The fission spectra forthese isotopes are derived from the BUGLE-96 nuclear data library.3.3 Surveillance Capsule Activation and Fluence ResultsThis section documents the fluence and activation results for the Browns Ferry 2 reactor.

Theactivation results also form the basis for the validation and qualification of the application of theRAMA Fluence Methodology to the Browns Ferry 2 reactor in accordance with the requirements of U. S. NRC Regulatory Guide 1.190 (Reg. Guide 1.190). Reg. Guide 1.190 requires fluencecalculational methods to be validated by comparison with measurements from operating reactordosimetry for the specific plant being analyzed or for reactors of similar design.Three flux wire activation analyses were performed for the Browns Ferry 2 reactor.

Flux wireswere removed from the 300 capsule flux wire holder and analyzed at the end of cycle 1(irradiated for 1.3 EFPY); surveillance capsule flux wires were removed at the end of cycle 7from the 30' capsule (irradiated for 8.1 EFPY); and surveillance capsule flux wires wereremoved at the end of cycle 16 from the 1200 capsule (irradiated for 22.9 EFPY). Details of thedosimetry specimens and analysis are presented below.3-19 Neutron Fluence Calculation Peak fluence was calculated for each of the two removed capsules and the 300 capsule flux wireholder. Additionally, peak fluence was calculated for the 3000 capsule still in the reactor insupport of lead factor calculations.

Lead factors are determined and reported for all capsules.

3.3.1 Comparison of Predicted Activation to Plant-specific Measurements The comparison of predicted activation for the Browns Ferry 2 cycles 1, 7, and 16 flux wires tomeasurements is presented in this subsection.

Fluence values are also calculated and reported inSection 3.4.2 for each of the capsule flux wires.3.3.1.1 Cycle 1 300 Flux Wire Holder Activation AnalysisCopper and iron flux wires were irradiated in the Browns Ferry 2 surveillance capsule flux wireholder at the 30' azimuth during the first cycle of operation.

The wires were removed after beingirradiated for a total of 1.3 EFPY. Activation measurements were performed following irradiation for the following reactions

[19]: 63Cu (n,at) 60Co and 54Fe (n,p) 54Mn. The preciselocation of the individual wires within the surveillance capsule flux wire holder is not known,therefore, the activation calculations were performed at the center of the holder.Table 3-4 provides a comparison of the RAMA calculated specific activities and the measuredspecific activities for the flux wire specimens.

The cycle 1 total flux wire average calculated-to-measured (C/M) value is 0.91 with a standard deviation of +0.02.Table 3-4Comparison of specific activities for Browns Ferry 2 Cycle 1 30* flux wire holder wires(C/M)IronFe-1 58.11 54.05 0.93 --Fe-2 58.80 54.05 0.92 ---Fe-3 57.44 54.05 0.94 --Average 58.12 54.05 0.93 0.01CopperCu-1 3.198 2.808 0.88 ---Cu-2 3.130 2.808 0.90Cu-3 3.053 2.808 0.92 ---Average 3.127 2.808 0.90 0.02Total FluxWire --- ... 0.91 0.02Average3-20 Neutron Fluence Calculation 3.3.1.2 Cycle 7 30' Surveillance Capsule Activation AnalysisCopper, iron, and nickel flux wires were irradiated in the Browns Ferry 2 surveillance capsule atthe 300 azimuth during the first 7 cycles of operation.

The wires were removed after beingirradiated for a total of 8.1 EFPY. Activation measurements were performed following irradiation for the following reactions

[8]: 63Cu (n,a) 60Co, 54Fe (n,p) 54Mn, and 58Ni (n,p) 58Co.Table 3-5 provides a comparison of the RAMA calculated specific activities and the measuredspecific activities for the surveillance capsule flux wire specimens.

The cycle 7 capsule total fluxwire average C/M value is 1.20 with a standard deviation of +/-0.06. It is noted that the EOC 7capsule comparison shows significantly more conservatism than the two other dosimetry evaluations at EOC 1 and EOC 16. It was noted in the original capsule evaluation

[8] that theevaluators also showed a similar overestimate that could imply that some other circumstance exists outside both predictive models that caused the capsule to experience less activation thanpredicted, such as a dislocation of the capsule container.

A 20% change in activity can be causedby as little as a 1/4" variation in radial positioning of the capsule.Table 3-5Comparison of specific activities for Browns Ferry 2 cycle 7 30* surveillance capsule fluxwires (C/M)IronAverage' 6.05E+04 7.74E+04 1.28CopperAverage' 5.62E+03 6.49E+03 1.15NickelAverage' 1.07E+06 1.26E+06 1.18 -Total Flux Wire....1.20 0.06AverageThe source document for the flux wire measurements only provided an average activity thatrepresents the average of three wires for each wire type.3.3.1.3 Cycle 16 1200 Surveillance Capsule Activation AnalysisCopper, iron, and nickel flux wires were irradiated in the Browns Ferry 2 surveillance capsule atthe 1200 azimuth during the first 16 cycles of operation.

The wires were removed after beingirradiated for a total of 22.9 EFPY. Activation measurements were performed following 63 60 54 54588irradiation for the following reactions:

6Cu (na1) Co, Fe (n,p) Mn, and 58Ni (n,p) "Co.Table 3-6 provides a comparison of the RAMA calculated specific activities and the measuredspecific activities for the surveillance capsule flux wire specimens.

The cycle 16 capsule totalflux wire average C/M value is 1.08 with a standard deviation of +/-0.07.3-21 Neutron Fluence Calculation Table 3-6Comparison of specific activities for Browns Ferry 2 cycle 16 1200 surveillance capsuleflux wires (CIM)IronFe-G9 80.04 91.95 1.15 --Fe-G10 85.15 91.95 1.08 ---Average 82.60 91.95 1.11 0.05CopperCu-G9 13.71 13.70 1.00 --Cu-G10 13.74 13.70 1.00 ---Average 13.73 13.70 1.00 0.00NickelNi-G9 1076.30 1226.63 1.14 --Ni-GlO 1120.09 1226.63 1.10 ---1098.20 1226.63 1.12 0.03Total Flux...... 1.08 0.07Wire Average3.3.1.4 Surveillance Capsule Activation Analysis SummaryTable 3-7 presents a summary of the total average calculated-to-measured result of specificactivities for all Browns Ferry 2 flux wires. Combining all flux wires (copper, iron, and nickel),the total average C/M is 1.08 with a standard deviation of +/-0.13.Table 3-7Comparison of activities for Browns Ferry 2 surveillance capsule flux wires30' Flux Wire (EOC 1) 6 0.91 0.02300 Capsule (EOC 7) 9 1.20 0.061200 Capsule (EOC 16) 6 1.08 0.07Total 21 1.08 0.133-22 Neutron Fluence Calculation 3.3.2 Capsule Peak Fluence Calculations and Lead Factor Determinations Peak fast neutron (E > 1.0 MeV) fluences were calculated for each of the capsules originally installed in the Browns Ferry 2 reactor.

Of the three original

capsules, two have been removed,those being the 30- and 120-degree capsules.

The third capsule, located at 300', remains in thereactor.

The peak fluences for the 30-degree capsule are reported at the time of their respective

removal, while the 120-degree capsule has fluence reported at the end of cycle 16, and at the endof the reactor's extended operating life of 54 EFPY. Additionally, the lead factor for each capsuleis calculated by dividing the peak capsule fluence by the respective peak RPV fluence at a givenreporting time. The results of these calculations are presented in Table 3-8. Note that since the3000 capsule has not yet been removed, the lead factor and fluence are estimated.

It is observed in Table 3-8 that the lead factors vary between cycles and capsules.

In theory, aplant running with a consistent fuel loading pattern and a symmetric power shape will havesimilar lead factors for all capsules, since the capsules usually reside in symmetric locations.

Inthe case of Browns Ferry 2, the decreased lead factor between the EOC 7 30' capsule and theEOC 16 1200 capsule can be attributed to changing fuel designs, as seen in Table 3-2. Like otherfluence predictions, any future changes in any of the items listed in "Limitations of FluenceProjections" will impact the 3000 capsule lead factor predictions.

Table 3-8Lead factors for Browns Ferry 2 300, 1200, and 300° surveillance capsules300 EOC 7 8.1 EFPY 2.40E+17 2.29E+17 1.051200 EOC 16 22.9 EFPY 6.44E+17 6.30E+17 1.023000 EOXL1 (est.) 54 EFPY 1.60E+18 1.58E+18 1.01'EOXL represents the end of the extended design life, which is assumed to represent 54 EFPY.3.4 Capsule Fluence Uncertainty AnalysisThis section presents the combined uncertainty analysis and bias determination for the BrownsFerry 2 capsule fluence evaluation.

The combined uncertainty is comprised of the comparison uncertainty factors and an analytic uncertainty factor developed in this section.

When combined, these components provide a basis for determining the overall uncertainty (1 a) and bias in thecapsule fluence for this analysis.

The requirements for determining the combined uncertainty and bias for light water reactorfluence evaluations are provided in Regulatory Guide 1.190. The method implemented fordetermining the combined uncertainty and bias for reactor component fluence is described in theRAMA Theory Manual [ 17]. Regarding the determination of a bias in the fluence, Regulatory Guide 1.190 provides that an adjustment to the calculated fluence for bias effects is needed if astatistically significant bias exists in the fluence computation.

3-23 Neutron Fluence Calculation The results presented in this section show that the combined uncertainty for the Browns Ferry 2capsule fluence evaluation is 12.2% and that no adjustment for bias effects is required to thecalculated capsule fluence reported in Section 3.3 of this report.The following subsections describe the comparison uncertainties determined in Section 3.3, thedetermination of the analytic uncertainty, and the determination of the overall combineduncertainty and bias for the Browns Ferry 2 capsule fluence evaluation.

3.4.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activitymeasurements.

For capsule fluence evaluations, two comparison uncertainty factors areconsidered:

an operating reactor comparison uncertainty factor and a benchmark comparison uncertainty factor. The determination of a comparison uncertainty factor based on measurements involves the combination of two measurement components.

One component is the variation inthe comparison of the calculated-to-measured (C/M) activity ratio and the other accounts for theuncertainty introduced by the measurement process.3.4.1.1 Operating Reactor Comparison Uncertainty The operating

reactor, or plant-specific, comparison uncertainty for the Browns Ferry 2 reactor isdetermined by combining the standard deviation for the activity comparisons with themeasurement uncertainty for the plant-specific activity measurements.

3.4.1.2 Benchmark Comparison Uncertainty The benchmark comparison uncertainty used in the Browns Ferry 2 uncertainty analysis is basedon a set of industry standard simulation benchmark comparisons.

3.4.2 Analytic Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters needs to be tested for its contribution to the overall fluence uncertainty.

The uncertainty values for the geometry parameters are based upon uncertainties in thedimensional data used to construct the plant geometry model. The uncertainty values for thematerial parameters are based upon uncertainties in the material densities for the water andnuclear fuel materials and the compositional makeup of typical steel materials.

The uncertainty values for the fission source parameters are based upon uncertainties in the fuelexposure and power factors for the fuel assemblies loaded on the core periphery.

The transport method used in the fluence analysis employs a fission source calculation that accounts for therelative contributions of the uranium and plutonium fissile isotopes in the fuel and the relativepower density of the fuel in the reactor.

Both fission source parameters are derived directly frominformation calculated by three-dimensional core simulator codes. The uncertainty values for thenuclear cross-section parameters are based upon uncertainties in the number densities for thepredominant nuclides that make up the reactor materials.

3-24 Neutron Fluence Calculation The uncertainty parameters for the fluence model inputs are based upon geometry meshing andnumerical integration parameters used in the neutron flux transport calculation.

The process fordetermining the geometry meshing and numerical integration parameters involves an exhaustive sensitivity study that is described in the RAMA Procedures Manual [ 18].3.4.3 Combined Uncertainty The combined uncertainty for the capsule fluence evaluation is determined with a weighting function that combines the analytic, plant-specific comparison, and benchmark comparison uncertainty factors developed in Sections 3.4.1 and 3.4.2. Table 3-9 shows that the combineduncertainty (la) determined for the Browns Ferry capsule fluence is 12.2% for energy >1.0MeV.Table 3-9 also shows that, in accordance with Regulatory Guide 1.190, no bias term exists and itis not necessary to adjust the RAMA predicted capsule fluence in this analysis for bias effects.

Itis also demonstrated in Table 3-9 that the combined uncertainty is within the limits prescribed inU.S. NRC Regulatory Guide 1.190 (i.e. < 20%).Table 3-9Combined capsule fluence uncertainty for energy >1.0 MeVCombined Uncertainty (lo) 12.2%Bias None11 The bias terms are less than their constituent uncertainty values,concluding that no statistically significant bias exists.3-25 4CHARPY TEST DATA4.1 Charpy Test Procedure Charpy impact tests were conducted in accordance with American Society for Testing andMaterials (ASTM) Standards E 185-82 and E23-02. The 1982 version of E 185 has been reviewedand approved by NRC for surveillance capsule testing applications.

This standard references ASTM E23. The tests were conducted using a Tinius Olsen Testing Machine Company, Inc.Model 84 impact test machine with a 300 ft-lb (406.75 J) energy capacity.

The Model 84 isequipped with a dial gage as well as the MPM optical encoder system for accurate absorbedenergy measurement.

The machine is also equipped with an instrumented

striker, so a total ofthree independent measurements of the absorbed energy were made for every test. In all cases,the optical encoder measured energy was reported as the impact energy. The optical encoderenergy is much more accurate than the analog dial. The optical encoder can resolve the energy towithin 0.04 ft-lbs (0.054 J), whereas, for the dial, the resolution is approximately 0.25 ft-lbs (0.34J). The impact energy was corrected for windage and friction for each test performed.

Thevelocity of the striker at impact was nominally 18 ft/s (5.49 m/s). The MPM encoder systemmeasures the exact impact velocity for every test. Calibration of the machine was verified asspecified in ASTM E23, and verification specimens were obtained from the National Institute forStandards and Technology (NIST) and tested in accordance with the standard.

The ASTM E23 procedure for specimen temperature control using an in-situ heating and coolingsystem was followed.

The advantage of using the MPM in-situ heating/cooling technology is thateach specimen is thermally conditioned right up to the instant of impact. Thermal lossesassociated with liquid bath systems, such as those resulting from transfer from a liquid bath tothe test machine, are completely eliminated.

Each specimen was held at the desired testtemperature for at least 5 minutes prior to testing, and the fracture process zone temperature washeld to within + 1.80 F (+/- 10 C) up to the instant of strike. Precision calibrated tongs were usedfor specimen centering on the test machine.Lateral expansion (LE) was determined from measurements made with a lateral expansion gage.The lateral expansion gage was calibrated using precision gage blocks which are traceable toNIST. The percentage of shear fracture area was determined by integrating the ductile and brittlefracture areas using the MPM Digital Optical Comparator (DOC) image analysis system. Asshown in Figure 4-1, each fracture surface image is captured, outlined to delineate the brittlearea, and outlined to define the outer ductile fracture region. The DOC software then performs apixel area integration and automatically calculates the shear fracture area. This method for sheararea determination is the most accurate method given in ASTM E23 and is far superior to thecommonly used photograph comparison method.The number of Charpy specimens for measurement of the transition region and upper shelf waslimited.

Therefore, the choice of test temperatures was very important.

Prior to testing, the4-1 Charpy Test DataCharpy energy-temperature curve was predicted using embrittlement models and previous data.The first test was then conducted near the middle of the transition region, and test temperature decisions were then made based on the test results.

Overall, the goal was to perform two tests onthe upper shelf, and to use the remaining specimens to characterize the 30 ft-lb (41 J) index. Thisapproach was successful and the transition region and upper shelf energy are well defined.Figure 4-1Illustration of digital optical comparator measurement of shear fracture areaFirst, the Brittle Fracture Area is outlined (within green line). Next, the Outer Ductile Fracture Area isoutlined (within red line). Finally, the software integrates the areas and calculates the Percent ShearFracture Area.4.2 Charpy Test DataA total of eight irradiated base, weld, and HAZ metal specimens, respectively, were tested overthe transition region temperature range and on the upper shelf. The data are summarized inTables 4-1 through 4-3. In addition to the energy absorbed by the specimen during impact, themeasured lateral expansion values and the percentage shear fracture area for each test specimenare provided in the tables. The Charpy energy was acquired from the optical encoder and hasbeen corrected for windage and friction in accordance with ASTM E23. The impact energy is theenergy required to initiate and propagate a crack in the Charpy specimen.

The optical encoderand the dial cannot correct for tossing energy or losses in the test machine, and therefore thissmall amount of additional energy, if present, may be included in the data for some tests. Theinstrumented striker energy does not include tossing energy or machine vibration energy since4-2 Charpy Test Datathe energy, in this case, is measured only during a few milliseconds of contact between thestriker and specimen.

Based on comparison between the instrumented striker energy and theoptical encoder energy, it has been shown that the tossing energy, and other losses, are small formost tests.The lateral expansion is a measure of the transverse plastic deformation produced by the contactedge of the striker during the impact event. Lateral expansion is determined by measuring themaximum change of specimen thickness along the sides of the specimen.

Lateral expansion is ameasure of the ductility of the specimen.

The nuclear industry tracks the embrittlement shiftusing the 35 mil (0.89 mm) lateral expansion index. In accordance with ASTM E23, the lateralexpansion for some specimens, which could be broken after the impact test, should not bereported as broken since the lateral expansion of the unbroken specimen is less than that for thebroken specimen.

Therefore, when these conditions exist, the value listed is the unbrokenmeasurement and a footnote is included to identify these specimens.

All of the 1200 capsulespecimens that did not separate during the test could be broken by hand under the ASTM E23requirements.

The percentage of shear fracture area is a direct quantification of the transition in the fracturemodes as the temperature increases.

All metals with a body centered cubic lattice structure, suchas ferritic pressure vessel materials, undergo a transition in fracture modes. At low testtemperatures, a crack propagates in a brittle manner and cleaves across the grains. As thetemperature increases, the percentage of shear (or ductile) fracture increases.

This temperature range is referred to as the transition region and the fracture process is mixed mode. As thetemperature increases

further, the fracture process is eventually completely ductile (i.e., no brittlecomponent) and this temperature range is referred to as the upper shelf region.Table 4-1Irradiated Charpy V-Notch impact test results for base metal (heat A0981-1) specimens from the Browns Ferry 2 1200 surveillance capsule--F (OC) ft-lb (J) mils (mm) % ShearE7B -105.3 (-76.3) 4.33 (5.87) 5.0 (0.13) 3.7E7J -39.5 (-39.7) 21.92 (29.72) 15.0 (0.38) 11.7E7C 6.3 (-14.3) 23.08 (31.29) 20.5 (0.52) 20.8E63 40.1 (4.5) 73.68 (99.90) 54.0 (1.37) 38.2E75 64.0 (17.8) 99.02 (134.25) 72.5 (1.84) 58.8E74 97.2 (36.2) 107.05 (145.14) 70.5 (1.79) 60.3E62 174.7 (79.3) 140.88 (191.01) 92.0 (2.34) 100.0E5K 327.9 (164.4) 144.72 (196.21) 93.5 (2.37) 100.0'All specimen lUs have a dot over the middle alphanumeric character.

4-3 Charpy Test DataTable 4-2Irradiated Charpy V-Notch impact test results for weld metal (heat BF2 ESW) specimens from the Browns Ferry 2 1200 surveillance capsuleEBJEALEAEEC3EBUEB6EAKEADOF-49.714.450.081.3119.1204.3301.6373.8(°C)(-45.4)(-9.8)(10.0)(27.4)(48.4)(95.7)(149.8)(189.9)ft-lb5.4728.0647.8922.4083.90102.15113.53110.62(J)(7.42)(38.04)(64.93)(30.37)(113.75)(138.49)(153.92)(149.98)mils6.521.540.025.063.579.088.084.5(mm)(0.17)(0.55)(1.02)(0.64)(1.61)(2.01)(2.24)(2.15)% Shear4.612.722.528.350.379.3100.0100.01All specimen lDs have a dot over the middle alphanumeric character.

Table 4-3Irradiated Charpy V-Notch impact test results for HAZ specimens from the Browns Ferry 21200 surveillance capsuleOF (°C) ft-lb (J) milsED3 -109.5 (-78.6) 2.17 (2.94) 0.5EDK -70.6 (-57.0) 22.95 (31.12) 14.0EEL -12.5 (-24.7) 45.17 (61.24) 35.5EE6 23.5 (-4.7) 62.35 (84.53) 48.0EEJ 47.3 (8.5) 92.68 (125.66) 70.0ED5 72.0 (22.2) 112.58 (152.64) 76.5EEA 181.6 (83.1) 122.64 (166.28) 84.5EET 300.9 (149.4) 141.24 (191.49) 91.5'All specimen lUs have a dot over the middle alphanumeric character.

Note: HAZ test results are not used in the BWRVIP ISP.(mm)(0.01)(0.36)(0.90)(1.22)(1.78)(1.94)(2.15)(2.32)% Shear5.115.723.635.853.080.7100.0100.04-4 5CHARPY TEST RESULTS5.1 Analysis of Impact Test ResultsFor analysis of the Charpy test data, the BWRVIP ISP has selected the hyperbolic tangent (tanh)function as the statistical curve-fit tool to model the transition temperature toughness data. Ahyperbolic tangent curve-fitting program named CVGRAPH [ 10] was used to fit the Charpy V-notch energy and lateral expansion data. Analysis methodology (e.g., definition of upper fixedshelf and lower shelf) followed the BWRVIP conventions established for analysis of all ISP data[21, 22]. The impact energy curve-fits from CVGRAPH are provided in Figures 5-1 (plateA0981-1) and 5-3 (weld BF2 ESW), and the lateral expansion curve-fits are shown in Figures5-2 (plate A098 1-1) and 5-4 (weld BF2 ESW). Because HAZ results are not used in theBWRVIP ISP, that data was not fit.For the analysis of Charpy energy test data, lower shelf energy was fixed at 2.5 ft-lbs (3.4 J).Upper shelf energy was fixed at the average of all test energies exhibiting shear greater than orequal to 95%, consistent with ASTM Standard E185-82 [3]. For analysis of the lateral expansion test data, the lower shelf was fixed at 1.0 mils; the fixed upper shelf was defined as the averageof the lateral expansion test data points at the same test temperatures used to define the fixedupper shelf energy.5.2 Irradiated Versus Unirradiated CVN Properties Table 5-1 summarizes the T30 [30 ft-lb (41 J) Transition Temperature],

T35mil1 [35 mil (0.89 mm)Lateral Expansion Temperature],

T50 [50 ft-lb (68 J) Transition Temperature],

and Upper ShelfEnergy for the unirradiated and irradiated materials and shows the change (shift) from baselinevalues. The unirradiated values of T30 and T50 were taken from the CVGRAPH fits provided inFigures 2-3 and 2-4; the unirradiated values of T35mil were previously determined in [21, 22]. Theirradiated values are from the index temperatures determined in Figures 5-1 through 5-4.Table 5-2 provides a comparison of the measured shifts to predicted shifts for plate heat A0981-1and weld heat BF2 ESW. Predicted shift is based on the formula provided in Reg. Guide 1.99Rev. 2 [6] and shown in Note 2 to the table. The fluence was input as 6.44 x 1017 n/cm2, asreported in Section 3 for the 1200 capsule.

For surveillance plate heat A0981-1, the measuredshift is within the value expected (e.g., the measured shift is less than predicted shift + margin);however, the measured shift for weld heat BF2 ESW is about 15% greater than the predicted shift + margin.Measured percent decrease in USE is presented in Table 5-3 and compared to the percentdecrease predicted by Figure 2 of Reg. Guide 1.99, Rev. 2. For both the surveillance plate andweld, the measured percent decrease is less than the predicted percent decrease.

5-1 Chaipy Test ResultsIrradiated Plate Heat A0981-1 CVE (BF2-120) 300250Al 2000ILl150Lu10050CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 09/23f2012 01:19 PMPage 1Coefficients of Curve IA = 72.65 B = 70.15 C = 69.29 TO = 45.34 D = O.OOE+00Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 142.8(Fixed)

Lower Shelf Energy=2.5(Fixed)

Temp@30 ft-lbs=-3.5 Deg F Temp@50 ft-lbs=22.2 Deg FPlant: BROWNS FERRY 2 Material:

SA302BM Heat: A0981-1Orientation:

LT Capsule:

120 DE Fluence:

n/cmA200-300-200 -1000 100 200 300 400 500 600Temperature in Deg FCharpy V-Notch DataTemperature

-105.-39.6.40.64.97.174.327.3050301000207090Input CVN4. 3321 .9223. 0873. 6899. 02107. 05140. 88144. 72Computed CVN4. 2913. 6636. 8467. 3691 .10117. 14139. 53142. 76Differential

.048. 2613.766. 327.9210. 091.351.96Correlation Coefficient

= .989Figure 5-1Irradiated plate A0981-1 (Browns Ferry 2 120" capsule)

Charpy energy plot5-2 Charpy Test ResultsIrradiated Plate Heat A0981-1 LE (BF2-120)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 09/23/2012 01:25 PMPage 1Coefficients of Curve IA = 46.88 B = 45.88 C = 69.83 TO = 34.81 D = O.OOE+00Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=92.8(Fixed)

Lower Shelf L.E.=l.0(Fixed)

Temp.@L.E.

35 mils=16.4 Deg FPlant: BROWNS FERRY 2 Material:

SA302BM Heat: A0981-1Orientation:

LT Capsule:

120 DE Fluence:

n/cm^2200150.3ECi 100 -500-3000 300Temperature in Deg FCharpy V-Notch Data600Temperature

-105. 30-39. 506. 3040. 1064. 0097. 20174. 70327. 90Input LE..5. 0015. 0020. 5054. 0072. 5070. 5092 .0093. 50Computed L.E.2. 6310. 76..29. 1250. 3465. 0179. 5991. 1192. 73Differential 2.374. 24-8. 623 667 49-9. 098977Correlation Coefficient

= .986Figure 5-2Irradiated plate A0981-1 (Browns Ferry 2 120" capsule) lateral expansion plot5-3 Chaipv Test ResultsIrradiated Weld Heat BF2 ESW CVE (BF2-120)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 09/23/2012 01:27 PMPage 1Coefficients of Curve 1A = 57.29 B = 54.79 C = 99.55 TO = 94.15 D = O.OOE+OOEquation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1

12. 1(Fixed)

Lower Shelf Energy=2.5(Fixed)

Temp@30 ft-lbs=39.8 Deg F Ternp@50 ft-lbs=80.9 Deg FPlant: BROWNS FERRY 2 Material:

ESW Heat: ESWOrientation:

NA Capsule:

120 DE Fluence:

n/cmA2300250120015LUz100500-3(Temperature

-49. 7014. 4050. 0081. 30119. 10204. 30301. 60373. 8030 -200 -100 0 100 200 300 400 500 600Temperature in Deg FCharpy V-Notch DataInput CVN5. 4728. 0647. 8922. 4083. 90102. 15113. 53110. 62Computed CVNDifferential 8.20.34.50.70.101.110.111.2787472674284168-2. 807.1913. 4227. 8613. 16.873. 12-1.06Correlation Coefficient

= .954Figure 5-3Irradiated weld BF2 ESW (Browns Ferry 2 1200 capsule)

Charpy energy plot5-4 Charpv Test ResultsIrradiated Weld Heat BF2 ESW LE (BF2-120)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 09/23/2012 01:29 PMPage 1Coefficients of Curve 1A = 43.63 B = 42.63 C = 108.46 TO = 85.22 D = O.OOE+00Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=86.3(Fixed)

Lower Shelf L.E.=l.0(Fixed)

Temp.@L.E.

35 mils=63.0 Deg FPlant: BROWNS FERRY 2 Material:

ESW Heat: ESWOrientation:

NA Capsule:

120 DE Fluence:

n/crn^2200150EC100500;/0n-3000300600Temperature in Deg FCharpy V-Notch DataTemperature

-49.7014.4050. 0081 .30119.10204. 30301.60373. 80Input L.E.6. 5021. 5040. 0025. 0063. 5079. 0088. 0084. 50Computed L.E.7. 5419.1730. 2542. 0956.5277.7184. 7085. 84Differential

-1. 042. 339. 7517. 096. 981 .293. 30-1. 34Correlation Coefficient

= .967Figure 5-4Irradiated weld BF2 ESW (Browns Ferry 2 1200 capsule) lateral expansion plot5-5 Charpy Test ResultsTable 5-1Effect of irradiation (E>1.0 MeV) on the notch toughness properties

-46.7 -3.5 43.2 -14.8 22.2 37.0 -25.4 16.4 41.8 137.8 142.8 5.0(LT) (-43.7) (-19.7) (24.0) (-26.0) (-5.4) (20.6) (-31.9) (-8.7) (23.2) (186.8) (193.6) (6.8)-52.5 39.8 92.3 -19.5 80.9 100.4 -29.5 63.0 92.5 116.0 112.1 -3.9(-46.9) (4.3) (51.3) (-28.6) (27.2) (55.8) (-34.2) (17.2) (51.4) (157.3) (152.0) (-5.3)Table 5-2Comparison of actual versus predicted embrittlement A0981-1 Browns Ferry 2 Plate (SA533B-1) 6.44 43.2 32.7 65.43(LT orientation)

(24.0) (18.17) (36.35)BF2 ESW Browns Ferry Unit 2 surveillance weld 6.44 92.3 40.24 80.45(51.3)kzz.3o)k 4+l.. I Z")Notes:1. See Table 5-1, AT30.2. Predicted shift = CF x FF, where CF is a Chemistry Factor taken from tables from USNRC Reg. Guide 1.99, Rev. 2 [6], based on thematerial's Cu/Ni content, and FF is Fluence Factor, fO.28-0.10 log f, where f = fluence (1019 n/cm2, E > 1.0 MeV) specified.

3. Margin Term is defined as 34°F for plate materials and 560F for weld materials, or margin equals shift (whichever is less), per Reg. Guide1.99, Rev. 2 [6].5-6 Charpy Test ResultsTable 5-3Percent decrease in upper shelf energyA0981-1 Browns Ferry 2 Plate-AO981-1 6.44 0.14 -3.63 12.0BF2 ESW Browns Ferry 2 ESW 6.44 0.20 3.36 17.8Notes:1. Calculated from Table 5-1, (Change/Unirradiated)

100. A positive number indicates a decrease in USE; a negative number indicates theUSE increased over the unirradiated value.2. Based on extrapolation of and interpolation between the curves given in Figure 2 of Reg. Guide 1.99 Rev. 2 [6].5-7 6REFERENCES

1. 10 CFR 50, Appendices G (Fracture Toughness Requirements) and H (Reactor VesselMaterial Surveillance Program Requirements),

Federal Register, Volume 60, No. 243, datedDecember 19, 1995.2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code(Code),Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,"

Nonmandatory Appendix G, Fracture Toughness Criteria for Protection Against Failure.3. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light- WaterCooled Nuclear Power Reactor Vessels, E706 (IF), ASTM Standards, Section 3, AmericanSociety for Testing and Materials, Philadelphia, PA, 1993.4. BWRVIP-86, Revision l-A: BWR Vessel and Internals

Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.5. 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and FuelReprocessing Plants."6. U.S. NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Revision 2, May 1988.7. "Guideline for the Management of Materials Issues,"

NEI 03-08, Nuclear Energy Institute, Washington, DC, Latest Edition.8. "Browns Ferry Steam Electric Station Unit 2 Vessel Surveillance Materials Testing andFracture Toughness Analysis",

C. Oza, GE Nuclear Energy, GENE-B 1 100639-01, Revision1, August, 1995.9. Letter from R.H. Shell (TVA) to USNRC, "Browns Ferry Nuclear Plant, Sequoyah NuclearPlant, and Watts Barr Nuclear Plant -Response to Generic Letter 92-01 (Reactor VesselStructural Integrity),"

Tennessee Valley Authority, dated July 7, 1992.10. CVGRAPH, Hyperbolic Tangent Curve Fitting Program, Developed by EPRI, Version 5.0.2,Revision 1, 3/26/02.11. "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,"

Nuclear Regulatory Commission Regulatory Guide 1.190, March 2001.12. B WR VIP-126:

B WR Vessel Internals

Project, RAMA Fluence Methodology

Software, Version 1.0. EPRI, Palo Alto, CA: 2003. 1007823.13. Letter from William H. Bateman (U.S. NRC) to Bill Eaton (BWRVIP),

Safety Evaluation of Proprietary EPRI Reports BWRVIP- 114, -115, -117, and -121 and TWE-PSE-00 l-R-001.dated May 13, 2005.6-1 References

14. Letter from Matthew A. Mitchell (U.S. NRC) to Rick Libra (BWRVIP),

"Safety Evaluation of Proprietary EPRI Report BWR Vessel and Internals

Project, Evaluation of Susquehanna Unit 2 Top Guide and Core Shroud Material Samples Using RAMA Fluence Methodology (BWRVIP-145),"

dated February 7, 2008.15. "BUGLE-96:

Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derivedfrom ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications,"

RSICCData Library Collection, DL2C- 185, March 1996.16. "VITAMIN-B6:

A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 forRadiation Transport Applications,"

RSICC Data Library Collection, DL2C-184, December1996.17. B WR VIP-I 14-A: B WR Vessel and Internals

Project, RA MA Fluence Methodolog, TheomyManual, EPRI, Palo Alto, CA: 2009. 1019049.18. B WR VIP- 121-A: B WR Vessel and Internals

Project, RA MA Fluence Methodology Procedures Manual, EPRI, Palo Alto, CA: 2009. 1019052.19. Analysis of the Vessel Wall Neutron Dosimeter from Browns Ferry Unit 2 Pressure Vessel,Southwest Research Institute, 02-4884-002, Rev. 0, September 1979.20. ASTM Standard E23, 2002, "Standard Test Methods for Notch Bar Impact Testing ofMetallic Materials,"

ASTM International, West Conshohocken, PA, 2002, DOI:10.1520/E0023-02, www.astm.org.

21. B WR VIP- 135, Revision 2: B WR Vessel and Internals

Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations.

EPRI, Palo Alto, CA: 2009.1020231.22. BWRVIP Letter 2010-238, Errata for BWRVIP-135, Rev. 2, October 15, 2010.6-2 AAPPENDIX A -DOSIMETER ANALYSISA.1 Dosimeter Material Description The 1200 Browns Ferry Unit 2 (BF2) surveillance capsule primary dosimeter materials are puremetal wires which were located within the surveillance capsule Charpy packets.

The wire typesprovided for the Browns Ferry Unit 2 surveillance program are copper, iron, and nickel. Eachwire is about three inches (7.62 cm) long with one of each type included in the two Charpyspecimen packets.A.2 Dosimeter Cleaning and Mass Measurement At the time the Charpy packets were opened, the dosimeter wires were cleaned with wipes thatwere wetted with acetone to remove loose contamination.

Upon receipt at the radiometric lab, thewires were visually inspected and cleaned with a lab wipe soaked in pure ethanol.

The wiresegments were then examined under a low magnification optical microscope.

There wasevidence of oxidation indicating the need for chemical etching and further cleaning.

This wasaccomplished by soaking the Fe wire segments in a 4N solution of hydrochloric acid until theoxidation was etched from the surface.

Similarly, the Ni and Cu wires were immersed in a 2Nsolution of nitric acid solution.

The wires were then rinsed with distilled water, wiped once morewith ethanol, and then allowed to dry in air at room temperature.

The wires then exhibited aclean, shiny appearance.

Figures A- 1 through A-6 show low-power magnifications of thedosimetry wires as they were found prior to cleaning, and after cleaning and coiling.

In general,the iron and copper wires had experienced the most oxidation, while the nickel wires wererelatively clean at the time of recovery from the Charpy packets.The total mass of each wire was measured using a Mettler Toledo XS 105DU analytical digitalbalance.

Table A-I lists the results of these measurements, as well as the identification assignedto each dosimeter.

The dosimeters identifications were assigned as the Charpy packet numbersfollowed by the type of dosimeter material.

As previously mentioned, the wires were tightly coiled for subsequent counting and weighing.

Each wire was wrapped around a thin metal rod to form a coil of approximately 0.5 inch (12.7mm) diameter or less, which yields a reasonable approximation to a point source geometry at thedistance the dosimeter wires are placed from the gamma detector.

The coiled wire segments werepressed firmly against a hard surface to flatten the coil to yield the best counting geometryA-1 Appendix A -Dosimeter AnalysisG9 CuG9CuFigure A-IPacket G9 Cu dosimeter wire: prior to cleaning (left); and after cleaning/coiling (right)G9FeG9 FeFigure A-2Packet G9 Fe dosimeter wire: prior to cleaning (left); and after cleaning/coiling (right)G9NiG9 Ni.Figure A-3Packet G9 Ni dosimeter wire: prior to cleaning (left); and after cleaning/coiling (right)A-2 Appendix A -Dosimeter AnalysisGID CuCUFigure A-4Packet GIO Cu dosimeter wire: prior to cleaning (left); and after clean ing/coiling (right)GlO FeFeFigure A-5Packet G10 Fe dosimeter wire: prior to cleaning (left); and after cleaning/coiling (right)G10 NiNiOFigure A-6Packet GIO Ni dosimeter Wire: prior to cleaning (left); and after cleaning/coiling (right)A-3 Appendix A -Dosimeter AnalysisTable A-1Wire dosimeter massesG9 Cu361.65G9 Fe 159.02G9 Ni 320.05G10 Cu 360.86G10 Fe 160.34G10 Ni 318.77A.3 Radiometric AnalysisRadiometric analysis was performed using high resolution gamma emission spectroscopy.

In thismethod, gamma emissions from the dosimeter materials are detected and quantified using solid-state gamma ray detectors and computer-based signal processing and spectrum analysis.

Thespecifications of the gamma ray spectrometer system (GRSS) are listed in Table A-2. The GRSSfeatures a hyper pure germanium (HPGe) detector that is housed in a lead-copper shield toreduce background count rates. Standard background subtraction procedures were used.GRSS calibration was performed using a National Institute for Standards and Technology (NIST) traceable mixed gamma quasi-point source. The Canberra analysis software provides thecapability for energy resolution and efficiency calibration using specified standard sourceinformation.

Calibration information is stored on magnetic disk for use by the spectrographic analysis software package.Since detector efficiency depends on the source-detector

geometry, a fixed, reproducible geometry must be selected for the gamma spectrographic analysis of the dosimeter materials.

Forthe dosimeter wires, the counting geometry was that of a quasi-point source (coiled wire) placedfive inches (12.7 cm) vertically from the top surface of the detector shell. In this way, extendedsources up to 0.5 inch (1.27 cm) can be analyzed with a good approximation to a point source.The coiled wires were well within the area needed to approximate a point source geometry.

TheHPGe detector was calibrated for efficiency using the NIST traceable source. The accuracy ofthe efficiency calibration was checked using a gamma spectrographic analysis of the NISTtraceable mixed gamma source. The isotopes contained in the source emit gamma rays whichspan the energy response of the detector for the dosimeter materials.

These measurements showthat the efficiency calibration is providing a valid measurement of source activity.

Theacceptance criteria for these measurements are that the software must yield a valid isotopicidentification, and that the quantified activity of each correctly identified isotope must be withinthe uncertainty specified in the source certification.

Validation of system performance wasperformed prior to starting the counting tasks, and upon completion of all counting work forBF2. The counting system performance was acceptable in each case, indicating that the countingsystem properties did not change during the course of the counting procedure.

A-4 Appendix A -Dosimeter AnalysisTable A-3 shows the counting schedule established for this work. There was no requirement fororder of counting, since the dosimeter materials still contained sufficient quantities of activation products to allow accurate radio assay. Counting times were more than sufficient to achieve thedesired statistical accuracy for gamma emissions of interest in all cases.Neutrons interact with the constituent nuclei of the dosimeter materials, producing radionuclides in varying amounts depending on total neutron fluence, its energy spectrum, and the nuclearproperties of the dosimeter materials.

Table A-4 lists the reactions of interest and their resultant radionuclide products for each element contained in the dosimeters.

These are threshold reactions involving an n-p or n-a interaction.

Finally, Table A-5 presents the primary results of interest for flux and fluence determination.

Thespecific activity units are in dps/mg, which normalizes the activity to dosimeter mass. Theactivities are specified for a useful reference date/time, which in this case is the BF2 plantshutdown date and time. This reference date/time was specified as February 26, 2011, at12:00:00 AM EST.Table A-2GRSS specifications DetectorCanberra Model GC1518Energy Resolution 1.8keV @ 1.33 MeVDetector Efficiency (relative to a 3 inch x 3 inch 15% at 1.3 MeV(7.62 cm x 7.62 cm) Nal crystal)Amplifier/Multichannel Analyzer Canberra DAS-1000Computer System Intel i5-2500 CPU at 3.30 GHz,, 2.91 GB MainMemory, 931 GB Hard Disk, 17-inch Monitor, HPLaserJet PrinterSoftware Canberra Apex v 1.2A-5 Appendix A -Dosimeter AnalysisTable A-3Counting schedule for the dosimeter materials G9 Cu08/18/122:04 PM86,400G9 Fe 08/17/12 1:34 PM 86,400G9 Ni 08/20/12 6:51 AM 86,400G10 Cu 08/22/12 8:05 AM 86,400G10 Fe 08/21/12 7:48 AM 86,400G10 Ni 08/23/12 8:47 AM 86,400Table A-4Neutron-induced reactions of interestTable A-5Results of the radiometric analysisG9 Cu60Co1.34E-0113.711.90G9 Fe 54Mn 3.44E-01 80.04 2.56G9 Ni 58Co 9.31 E+00 1076.30 2.74G10 Cu 60Co 1.34E-01 13.74 1.90G10 Fe 54Mn 3.69E-01 85.15 2.56G10 Ni 58Co 9.65E+00 1120.09 2.76February 26, 2011 at 12:00:00 AM EST is the reference date and time.A-6 The Electric Power Research Institute, Inc. (EPRI, www.epri.com) conducts research and development relating to the generation, deliveryand use of electricity for the benefit of the public. An independent, nonprofit organization, EPRI brings together its scientists and engineers as well as experts from academia and industry to help address challenges in electricity, including reliability, efficiency, affordability, health, safetyand the environment.

EPRI also provides technology, policy and economicanalyses to drive long-range research and development

planning, andsupports research in emerging technologies.

EPRI's members represent approximately 90 percent of the electricity generated and delivered inthe United States, and international participation extends to more than30 countries.

EPRI's principal offices and laboratories are located inPalo Alto, Calif.; Charlotte, N.C.; Knoxville, Tenn.; and Lenox, Mass.Together..

Shaping the Future of Electricity Programs:

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