ANP-3339, Rev. 0, 'Davis-Besse Unit 1 Reactor Vessel ...Capsule TE1-C received an estimated, average...

)

Enclosure L-16-227

AREVA Report, ANP-3339, Revision 0, "Davis-Besse Unit 1 Reactor Ve,ssel Material Surveillance Program: Analysis of Capsule TE1-C"

(65 Pages Follow)

A AREVA

Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C

December 2014

AREVA Inc.

(c) 2014 AREVA Inc.

ANP-3339 Revision 0

Copyright © 2014

AREVA Inc. All Rights Reserved

AREVA Inc. ANP-3339 Revision 0

Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Page i

Contents

Page

LIST OF TABLES ........................................................................................................... 111

NOMENCLATURE ... -...................................................................................................... VI

SUMMARY ..................................................................................................................... Vll

1.0 INTRODUCTION ............................................................................................... 1-1

2.0 BACKGROUND ..................... ~ ........................................................................... 2-1

3.0 SURVEILLANCE PROGRAM DESCRIPTION .................................................. 3-1

4.0 PRE-IRRADIATION TESTS .............................................................................. 4-1

4.1 Tension Tests ......................................................................................... 4-1

4.2 Impact Tests ............................................................................................ 4-1

5.0 POST-IRRADIATION TESTS ............................................................................ 5-1

5.1 Tension Test Results, ............................................................................. 5-1

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

6.0 NEUTRON FLU ENCE ....................................................................................... 6-1

6.1 Introduction ............................................................................................. 6-1

6.2 Overview of Analytical Methodology ....................................................... 6-2

6.3 Fluence Analysis Inputs .......................................................................... 6-2

·6.3.1 Reactor Geometry ........................................................................ 6-2 6.3.2 Cycle Lengths .............................................................................. 6-3

6.4 Fluence Analysis Results ........................................................................ 6-3

6.4.1 Capsule Flue nee Rate (Time-Averaged Flux) .............................. 6-3 6.4.2 Capsule Fluence .......................................................................... 6-4 6.4.3 Lead Factor .................................................................................. 6-4

6.5 Fluence Uncertainty ............................................................... ; ................ 6-5

6.6 DB-1 Surveillance Capsule Comparison ................................................. 6-6

6.7 Fluence Analysis Conclusions ....................... : ........................................ 6-9

7.0 DISCUSSION OF CAPSULE RESULTS ........................................................... 7-1

7 .1 Tensile Properties ................................................................................... 7 71

7.2 Charpy Impact Properties ....... , ............................................................... 7-1


_Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Page ii

8.0 SUMMARY OF RESULTS ................................................................................. 8-1

APPENDIX A.

APPENDIX B.

APPENDIXC.

APPENDIX D.

APPENDIX E.

APPENDIX F.

REACTOR VESSEL SURVEILLANCE PROGRAM BACKGROUND DATA AND INFORMATION .............................. A-1

PRE-IRRADIATION TENSILE DATA ........................................... 8-1

PRE-IRRADIATION CHARPY IMPACT DATA. ........................... C-1

FLUENCE ANALYSIS METHODOLOGY .................................... D-1

ASTM E185-82 RVSP TECHNICAL REPORT REQUIREMENTS ........................................................................ E-1

REFERENCES ............................................................................. F-1

'


Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Page iii

List of Tables

Table 3-1: Specimens in Surveillance Capsule TE1-C [8] ........................................... 3-2

Table 3-2: Chemical Composition of Surveillance Materials ....................................... 3-2

Table 3-3: Heat Treatment of Surveillance Materials [8] ..................................... :·······3-2

Table 5-1: Tensile Properties of Capsule TE1-C Irradiated Base Metal and Weld Metal ................................................................................................................ 5-2

Table 5-2: Charpy·lmpact Data for Capsule TE1-C Base Metal, BCC 241, Transverse Orientation, Irradiated to 1.88 x 1019 n/cm2 (E > 1 MeV) ......................... 5-6

Table 5-3: Charpy Impact Data for Capsule TE1-C Heat-Affected Zone Metal, BCC 241, Transverse Orientation, Irradiated to 1.88 x 1019 n/cm2 (E > 1 MeV) 5-6

Table 5-4: Charpy Impact Data for Capsule TE1-C Weld Metal, WF-182-1, Irradiated to 1.88 x 1019 n/cm2 (E > 1 MeV) ................................................................ 5-7

Table 6-1: DB-1 Fuel Cycle Lengths, Cycles 1 through 7 ............................................ 6-3

Table 6-2: Capsule TE1-C Fast Fluence (E > 1 MeV) Rate Results ........................... 6-4

Table 6-3: Capsule TE1-C Fast Fluence (E > 1 MeV) Results .................................... 6-4

Table 6-4: Capsule TE1-C Lead Factors, Wetted Surface .................. ,. ....................... 6-5

Table 6-5: Three Dimensional Coordinates for DB-1 (TE1) RVSP Capsules Points of Interest .................................................................................................... 6-6

Table 6-6: DB-1 (TE1) RVSP Capsule Fast Fluence Rate (E > 1 MeV) Results ......... 6-8

Table 6-7:

Table 6-8:

Table 7-1:

Table 7-2:

Table 7-3:

Table A-1:

DB-1 (TE1) RVSP Capsule Fast Fluence (E > 1 MeV) Results ................. 6-8

DB-1 (TE1) RVSP Capsule Calculation Comparison ................................. 6-8 I

Summary of DB-1 RVSP Capsule Tensile Test Results, Room Temperature Data ........................................................................................................ 7-2

Summary of DB-1 RVSP Capsule Tensile Test Results, Elevated Temperature Data ................................................................................... 7-3

Summary of DB-1 RVSP Capsule Charpy lmpactTest Results ................. 7-4

Unirradiated Impact Properties and Residual Element Content Data of DB-1 RV Beltline Region Materials Used for Selection of Surveillance Program Materials ................................................................................................. A-2

Table A-2: Test Specimens for Determining Material Baseline Properties .................. A-3


Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Page iv

Table A-3: Specimens in Upper Surveillance Capsules (Designations A, C, and E) ... ,A-3

Table A-4: Specimens in Lower Surveillance Capsules (Designations 8, D, and F) ... A-4

Table B-1: Pre-Irradiation Tensile Properties of Shell Forging Material, 8CC 241, Transverse Orientation ........................................................................... 8-1

Table 8-2: Pre-Irradiation Tensile Properties for Weld Metal WF-182-1, Transverse Orientation .............................................................................................. 8-1

Table C-1: Pre-Irradiation Charpy Impact Data for Shell Forging Material, 8CC 241, Transverse Orientation .......................................................................... C-1

Table C-2: Pre-Irradiation Charpy Impact Data for Shell Forging Material Heat-Affected Zone, 8CC 241, Transverse Orientation ................................................ C-2

Table C-3: Pre-Irradiation Charpy Impact Data for Weld Metal WF-182-1, Transverse Orientation ............................................................................................... C-3

Table C-4: Pre-Irradiation Charpy USE and Index: Temperatures' .............................. C-3


Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Pagev

List of Figures

Figure 3-1: Reactor Vessel Cross Section Showing Location of Capsule TE1-C in Davis-Besse Unit 1 ................................................................................. 3-3

Figure 3-2: Loading Diagram for Test Specimens in Capsule TE1-C .......................... 3-4

Figure 5-1: Stress-Strain Curve for Irradiated Weld Metal Tensile Specimen SS011 in Capsule TE1-C ............................................................... : ....................... 5-3

Figure 5-2: Stress-Strain Curve for Irradiated Base Metal Tensile Specimen SS617 in Capsule TE1-C ....................................................................................... 5-4

Figure 5-3: Impact Data (Impact Energy) for Irradiated Shell Forging Material, BCC 241 ................................................................................................................ 5-8

Figure 5-4: Impact Data (Lateral Expansion) for Irradiated Shell Forging Material, BCC 241 .......................................................................................................... 5-9

Figure 5-5: Impact Data (Percent Shear) for Irradiated Shell Forging Material, BCC 241 .............................................................................................................. 5-10

Figure 5-6: Impact Data (Impact Energy) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241 ....................................................................... 5-11

Figure 5-7: Impact Data (Lateral Expansion) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241 ....................................................................... 5-12

Figure 5-8: Impact Data (Percent Shear) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241 ....................................................................... 5-13

Figure 5-9: Impact Data (Impact Energy) for Irradiated Weld Metal, WF-182-1 ........ 5-14

Figure 5-10: Impact Data (Lateral Expansion) for Irradiated Weld Metal, WF-182-1. 5-15

Figure 5-11: Impact Data (Percent Shear) for Irradiated Weld Metal, WF-182-1 ...... 5-16

Figure A-1: Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel ...................................................................................... A-5

Figure D-1: Fluence Analysis Methodology Flow Chart .............................................. D-3

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Abbreviation

ASME

ASTM

B&W

BPVC

BWR

CFR

CMTR

DB-1

E > 1 MeV

E > 0.1 MeV

EFPY

EOC

EOL OF FENOC

ft

GALL

HAZ

J

Kie

ksi

lb

LRA

MIRVP

MLE

n/cm2

NIST

NRC

PWR

RCPB

RTNDT

RV

RVSP

SER

USE

Nomenclature Definition

The American Society of Mechanical Engineers

American Society of Testing and Materials

Babcock and Wilcox

Boiler and Pressure Vessel Code

Boiling Water Reactor

Code of Federal Regulations

Certified Material Test Report

Davis-Besse Nuclear Station Unit 1

Energy greater than 1 million electron volts

Energy greater than 0.1 million electron volts

Effective Full Power Years

End of Cycle

End of Life

Degrees Fahrenheit

f irstEnergy Nuclear Operating Company

Foot

Generic Aging Lessons Learned

Heat-Affected Zone

Joule

Stress Intensity Factor

Kilopound per square inch

Pound

License Renewal Application

Master Integrated Reactor Vessel Surveillance Program

Mils of Lateral Expansion

Neutrons per square centimeter

National Institute of Standards and Technology

U.S. Nuclear Regulatory Commission

Pressurized Water Reactor

Reactor Coolant Pressure Boundary

Reference Temperature, Nil Ductility Transition

~. Reactor Vessel

Reactor Vessel Surveillance Program

Safety Evaluation Report

Upper Shelf Energy

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Page vi

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SUMMARY

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Page vii

This report describes the results of the examination of the TE1-C capsule of FirstEnergy

Nuclear Operating Company's (FENOC's) Davis-Besse Nuclear Power Station Unit f (DB-1)

reactor vessel surveillance program (RVSP). The capsule was removed at the end of the

seventh fuel cycle (EOC 7). The objective of the RVSP is to monitor the effects of neutron

irradiation on the tensile and fracture toughness properties of the reactor vessel materials via

the testing and evaluation of Charpy impact and tensile specimens. The RVSP was designed in

accordance with the requirements of 10 CFR 50 Appendix Hand ASTM E185-73.

Capsule TE1-C received an estimated, average cumulative fast fluence of 1.88 x 1019 n/cm2

(Energy greater than 1 million electron volts (E > 1 MeV)) prior to its removal at EOC 7. The

projected peak cumulative fast fluence that the DB-1 reactor pressure vessel inside wetted

surface will receive at the end-of-life (EOL) or 60 calendar years of operation (52 effective full

power years (EFPY)) is 1.70 x 1019 n/cm2 (E > 1 MeV) for the upper shell forging, upper-to

lower-shell circumferential weld, and lower· shell forging. Therefore . fluence exposure for

material specimens in capsule TE1-C prior to its withdrawal at EOC 7 is confirmed, through

analysis, to be greater than the EOL (52 EFPY) fast neutron fluence at the inside wetted surface

for the limiting reactor vessel material and less than two times the EOL fast neutron fluence at

the inside wetted surface, indicating that the TE1-C surveillance material specimens can provide

meaningful metallurgical data for the period of extended operation.

The results of the tension tests indicated that the materials exhibited normal behavior relative to

neutron fluence exposure. The Charpy impact test data exhibited the characteristic behavior of

shift to higher temperature for the 30 ft-lb transition temperature and a decrease in upper shelf

energy (USE) as a result of neutron fluence damage.

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1.0 INTRODUCTION

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Page 1-1

This report describes the results of the examination of the TE1-C capsule of FirstEnergy

Nuclear Operating Company (FENOC)'s Davis-Besse Nuclear Power Station Unit 1 (DB-1)

reactor vessel surveillance program (RVSP). Capsule TE1-C was removed at the end of the

seventh fuel cycle (about 6.55 EFPY). The first capsule of the program, capsule TE1-F, was

removed and evaluated at the end of the first fuel cycle (about 1.02 EFPY); the results are

reported in BAW-1701 (Re,ference 1). The second RVSP capsule, TE1-B, was removed and

evaluated at the end of the third fuel cycle (about 2:58 EFPY); the results are reported in BAW-_,

1834 (Reference 2). The third RVSP capsule·, TE1-A, was removed and evaluated at the end of ~

the fourth fuel cycle (about 3.33 EFPY); the results are reported in BAW-1882 (Reference 3).

The fourth RVSP capsule, TE1-D, was removed and ~valuated at the end of the sixth fuel cycle

(about 5.45 EFPY); the results are reported in BAW-2125 (Reference 4).

The objective of the RVSP is to monitor the effects of neutron irradiation on the tensile and

impact properties of reactor pressure vessel materials under actual operating conditions. The

DB-1 RVSP was developed by Babcock & Wilcox (B&W) as described in BAW-10100A

(Reference 5). The program, designed to comply with the requirements of 10 CFR 50 Appendix

H (Reference 6) and ASTM E185-73 (Reference 7), is conducted in accordance with BAW-1543

(References 8 and 9) and ASTM E185-82 (Reference 10) to the extent possible (see Appendix

E for ASTM E185-82 requirements that are not addressed in this report).

The DB-1 RVSP was originally planned to monitor the effects of neutron irradiation on th~ RV

materials for a 40-year design life of the reactor pressure vessel. The original 40-year operating

license for DB-1 will expire in 2017 (Reference 15). Testing the material in the TE1-C capsule

in accordance with ASTM E185-82, to the extent practicable, and incorporating the results in the

RVSP is consistent with Aging Management Program (AMP) Xl.M31 of the U.S. Nuclear -

Regulatory Commission's (NRC's) Generic Aging Lessons Learned (GALL) Report (Reference

11) and supports License Renewal Commitmer:it #17 (Reference 15) regarding the

management of the effects of neutron embrittlement through the period of extended operation.

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2.0 BACKGROUND

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The ability of the reactor pressure vessel to resist frac!ure is a primary factor in ensuring the

safety of the primary system in light water-cooled reactors. The RV beltline region is the most

critical region of the vessel because it is exposed to fast neutron irradiation (E > 1 MeV). The

general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels

such as SA508 Class 2, used in the fabrication of the DB-1 reactor vessel, include an increase

in ultimate and yield strength properties with a corresponding decrease in ductility after

irradiation. The most significant mechanical property changes in irradiated RV ferritic steels are

the increase in temperature for the transition from brittle to ductile fracture and the reduction in

the Charpy upper shelf impact toughness.

Appendix G to 10 CFR 50, "Fracture Toughness Requirements," (Reference 12) specifies

fracture toughness requirements for the ferritic materials of pressure-retaining components of

the reactor coolant pressure boundary (RCPB) of light water nuclear power reactors, and

provides procedures for determining the pressure-temperature limitations on operation of the

RCPB. The fracture toughness and operational requirements are specified to provide adequate

safety margins during any condition of normal operation, including anticipated operational

occurrences and system hydrostatic tests, to which the pressure boundary may be subjected

over its service lifetime. Although 10 CFR 50 Appendix G became effective in August 1973, the

requirements are applicable to all boiling water reactors (BWRs) and pressurized water reactors

(PWRs), including those under construction or in operation on the effective date.

Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements,"

defines the material surveillance program required to monitor changes in the fracture toughness

properties of ferritic materials in the RV beltline region of light water nuclear power reactors

which result from exposure to neutron irradiation and the thermal environment. Fracture

toughness test data are obtained from surveillance material specimens withdrawn periodically

from the reactor vessel. These data will permit determination of the conditions under which the

vessel can be operated with adequate safety margins against fracture throughout its service life.

A method for guarding against non-ductile fracture in reactor pressure vessels is described in

Nonmandatory Appendix G, "Fracture Toughness Criteria for Protection against Failure," of

ASME Boiler and Pressure Vessel Code (BPVC) Section Ill, "Rules for Construction of Nuclear

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Facility Components" (Reference 13) and Section XI, "Rules for lnservice Inspection of Nuclear

Power Plant Components" (Reference 14). This method utilizes fracture mechanics concepts

and the reference nil-ductility temperature, RT NDT. which is defined as the greater of the drop

weight nil-ductility transition temperature (per ASTM E208) or the temperature that is 60°F

below that at which the material exhibits 50 ft-lb and 35 mils lateral expansion. The RT NDT of a

given material is used to index that material to a reference stress intensity factor curve (K1c

curve). The K1c curve is a lower bound of static critical fracture toughness results obtained from

several heats of pressure vessel steel. When a given material is indexed to the K1c curve,

allowable stress intensity factors can be obtained for this material as a function of temperature.

Allowable operating limits can then be determined using these allowable stress intensity factors.

The RT NDT and, subsequently, the operating limits of a nuclear power plant, can be adjusted to

account for the effects of radiation on the properties of the RV materials. The radiation

embrittlement and the resultant changes in mechanical properties of a given pressure vessel

steel can be monitored by a surveillance program in which a surveillance capsule containing

prepared specimens of the RV materials is periodically removed from the operating nuclear

reactor and the specimens are tested. The increase in the Charpy V-notch 30 ft-lb temperature

is added to the original RT NDT to adjust it for radiation embrittlement. This adjusted RT NDT is

used to index the material to the K1c curve which, in turn, is used to set operating limits for the

nuclear power plant. These new limits take into account the effects of irradiation on the RV

materials.

Appendix G to 10 CFR 50 also requires a minimum initial Charpy USE of 75 ft-lbs in the

transverse direction and maintenance of Charpy USE throughout the life of the vessel of no less

than 50 ft-lb, unless it is demonstrC!ted, in a manner approved by.the Office of Nuclear Reactor

Regulation, that lower values will provide adequate margins of safety equivalent to those

required by Appendix G of Section XI of the ASME Code.

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D,avis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C

3.0 SURVEILLANCE PROGRAM DESCRIPTION

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The surveillance program is comprised of six surveillance capsules designed to monitor the

effects of neutron irradiation and the thermal environment on the materials of the reactor

pressure vessel beltline region. The capsules, which were inserted into the reactor vessel

before initial plant startup, were positioned inside the reactor vessel between the thermal shield

and the vessel wall at the locations shown in Figure 3-1. The six capsules, originally designed

to be placed two in each holder tube, are positioned near the peak axial and azimuthal neutron

flux. However, with the use of DB-1 as one of the irradiation sites of the 177-fuel assembly

(177-FA) master integrated reactor vessel surveillance program (MIRVP), the capsules are

irradiated on a schedule integrated with the capsules of the other participating reactors. This

integrated schedule is described in BAW-1543. BAW-10100A includes a full description of the

capsule design.

Capsule TE1-C was removed at the end of the seventh fuel cycle of DB-1. This capsule

contained Charpy V-notch impact test specimens fabricated from base metal (SA508, Class 2),

weld metal, and heat-affected zone (HAZ) material. Tensile specimens were fabricated from

base metal and the weld metal only. The specimens contained in the capsule are described in

Table 3-1, and the locations of the individual specimens within the capsule are shown in Figure

3-2.

All weld and HAZ specimens are made from weld metal that closely represents actual RV welds

located in the beltline region. In addition, other aspects of specimen fabrication history, such as

heat treatment, are fully representative of actual vessel beltline region material. The chemical

composition and heat treatment .of the surveillance material in capsule TE1-C are described in

Table 3-2 and Table 3-3, respectively.

The test specimens were machined from the %-thickness (% T) location of the forging material.

Charpy V-notch and tension test specimens from the RV material were oriented with their

longitudinal axes perpendicular to the principal working direction of the forging.

Capsule TE1-C contained neutron dosimeter wires and temperature monitors (see Section 3 of

BAW-1543 for material descriptions); these materials were withdrawn with the surveillance

specimens at EOG 7 in 1991. However, these materials were discarded approximately 15 years

after the capsule entered storage. Therefore, dosimetry and thermal data specific to capsule

\

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TE1-C is not available for this analysis. Alternate fluence analyses have been utilized in place

of the origin~! dosimetry.

Additional details and background of the DB-1 RVSP capsules are reported in Appendix A.

Table 3-1: Specimens in Surveillance Capsule TE1-C [8]

Number of Tension Number of CVN Material Description Material ID (Heat)

Specimens Impact Specimens

Weld Metal WF-182-1 2 12

HAZ BCC 241 (5P4086) 0 12

HAZ AKJ 233 (123X244) 0 6*

Base Metal BCC 241 (5P4086) 2 12

Base Metal AKJ 233 (123X244) 0 6*

Correlation Material HSST Plate 02 0 6*

-. _, ., Total Specimens in

4 54 -· Capsule:

* These specimens were not tested and are not included in this analysis

Table 3-2: Chemical Composition of Surveillance Materials

Material ID c Mn p s

BCC 241 <a> 0.22 0.63 0.011 0.011

WF-182-1. Cbl 0.09 1.69 0.014 0.013

(a) Per Certified Materials Test Reports (CMTRs)

(b) Per BAW-1543, Revision 4 (Reference 8)

Wt%

Si Ni Cr

0.27 0.81 0.32

0.41 0.63 0.15

Table 3-3: Heat Treatment of Surveillance Materials [8]

Material ID Heat Treatment

1590°F ± 10°F for 4 hours, Water Quenched BCC 241 1240°F ± 10°F for 5 hours, Air Cooled

1125°F ± 25°F for 15.5 hours, Furnace Cooled

WF-182-1 1125°F ± 25°F for 15.5 hours, Furnace Cooled

Mo

0.63

0.40

Cu

0,02

0.21

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Figure 3-1: Reactor Vessel Cross Section Showing Location of Capsule TE1 -C in Davis-Besse Unit 1

x

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Page 3-3

w ---tl++--hil"~~~-+--+-~ +---"~~lll--l.l==.....£1~~ y l--+-+-+-+--+--+-.......... -+--+---+----+-....... ---4~1--1

z

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Figure 3-2: Loading Diagram for Test Specimens in Capsule TE1-C

CORE t SIDE

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4.0 PRE-IRRADIATION TESTS

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Unirradiated material was evaluated for two purposes: (1) to establish a baseline of data to

which irradiated properties data could be referenced, and (2) to determine those materials ..

properties to the extent practical from available material, as required for compliance with

Appendices G and H to 1 O CFR 50.

4.1 Tension Tests

Tension test specimens were fabricated from the RV shell course forging and weld metal. The

· specimens were 4.25 inches long with a reduced section 1. 750 inches long by 0.357 inch in

diameter. They were tested on a 55,000-lb load capacity universal test machine at a crosshead

speed of 0.050 inch per minute. A 4-pole extension device with a strain gaged extensometer

was used to determine the 0.2% yield point. Test conditions were in accordance with the

applicable requirements of ASTM A370-77. For each material type and/or condition, six

specimens in groups of three were tested at both room temperature and 580°F. The tension

compression load cell used had a certified accuracy of better than ±0.5% of full scale (25,000

lb). All test data for the pre-irradiation tensile specimens are given in Appendix B.

4.2 Impact Tests

Charpy V-notch impact tests were conducted in accordance with the requirements of ASTM

Standard Methods A370-77 and E23-72 (1978) on an impact tester certified to meet Watertown

standards. Test specimens were of the Charpy V-notch type, which were nominally 0.394 inch

square and 2.165 inches long.

Prior to testing, specimens were temperature-controlled in liquid immersion baths, capable of

covering the temperature range from -85°F to +550°F. Specimens were removed from the

baths and positioned in the test frame anvil with tongs specifically designed for the purpose.

The pendulum was released manually, allowing the specimens to be broken within five seconds

from their removal from the temperature baths.

Impact test data for the unirradiated baseline reference materials are presented in Appendix C.

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5.0 POST-IRRADIATION TESTS

5.1 Tension Test Results \

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Page 5-1

Four tensile specimens were tested at 200°F (1), 250°F (1), and 550°F (2). The tests were

performed using an MTS servohydraulic test machine. Certified TNSL TEST software was used

to control the machine and acquire the data. All tensile tests were run using stroke control with

an initial actuator travel rate of 0.0015 inch per minute; following specimen yielding, an actuator

speed of 0.075 inch per minute was used. Load was measured with a 55 kip MTS load cell at

10,000 pounds range. Strain was measured using a MTS extensometer with 0.5 inch of

available travel. The initial and final diameter of each specimen was measured using dial

calipers. The specimen temperature was monitored throughout the duration of each test.

The loading fixture failed during testing of specimen SS609 due to aging of the fixture. The test

was suspended, and the data were deemed unusable due to plastic deformation which had

occ_urred during previous testing resulting in strain hardening. For this test, only yield data are

presented.

The extensometer slipped during testing of specimen SS013. This was due to plastic

deformation occurring at or below the contact point between the extensometer and the

specimen. It was determined that the strain data are not accurate; therefore, the data are not

reported.

l

The tensile test data were analyzed using MTADS; this certified program uses the load and

strain data in conjunction with various specimen and testing parameters to perform a standard

ASTM EB analysis. The results of the post-irradiation tension tests are presented in Table 5-1.

The corresponding stress-strain curves are shown in Figure 5-1 for weld metal specimen SS011

and in Figure 5-2 for base metal specimen SS617.

In general, the ultimate strength and yield strength of the material increased with a

corresponding slight decrease in ductility; both effects were the result of neutron radiation

damage. The type of behavior observed and the degree to which the material properties

changed are within the range of changes to be expected for the radiation environment to which

the specimens were exposed.

The results of the pre-irradiation tension tests are presented in Appendix 8.

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Table 5-1: Tensile Properties of Capsule TE1-C Irradiated Base Metal and Weld Metal

Specimen Test Yield Tensile Fracture Fracture Fracture Uniform Total

Material* Temp. Strength Strength Load Stress Strength Elongation Elongation No. (oF) (ksi) (ksi) (lb) (ksi) (ksi) (%) (%)

Weld Metal SS011 550 79.928 94.923 7765 135.115 77.572 8.39 19.45

Weld Metal SS013 ** 250 86.313 94.372 9446 198.348 94.372 -- --Base Metal (T) SS609 *** 550 67.941 -- -- -- -- -- --Base Metal (T) SS617 200 72.150 91.050 6654 144.670 66.477 9.57 37.75

* (T) =Transverse

** The extensometer slipped during testing of specimen SS013; the strain data are not accurate and are therefore not reported.

***The loading fixture failed during testing of specimen SS609 due to aging of the fixture; for this test, only yield data are presented.

Reduction in Area

(%)

42.6

52.4

-54.0

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Figure 5-1: Stress-Strain Curve for Irradiated Weld Metal Tensile Specimen SS011 in Capsule TE1-C

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S eoimen: SSOU lsat: Temp.: 550 F ( 287 C)

...

cl 0 rl

OJ I

·ifi g

d C\l

Strength Yield: 79928.

UTS: 9Y923.

CJ

0.00 O.OY 0.08 0.12 O.lS 0.20 0.21! 0.28 Enginee~ing Strain

g

8:. ::i;:

,; "' Q)

• c.. 0 ... ::I:' Cf)

0) c ... L Q)

~ ... gi

.w 0 C\l

ci 0.32

AREVA Inc.


Figure 5-2: Stress-Strain Curve for Irradiated Base Metal Tensile Specimen SS617 in Capsule TE1-C

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Page 5-4

Specimen: SS617 leet Temp.: 200 F ( 93 C)

. 0) (/) ID c.. ...,

D 0 T"t

(,l')d ~co .... c.. Q) 0) c ..... ~d

LI.I ::I'

o (\J

0

Strength Yield: 72150.

UTS: 91050.

0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.110 Engi nearing Sira in ·

d (Cl

d"o cc .... llE

~ ,; (I) Q)

• L o+' =' (I)

Cl c ..... L m c ..... Cl

.i.5 0 N

0

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5.2 Charpy V-Notch Impact Test Results

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Page 5-5,

Charpy testing was performed in compliance with ASTM E23-94a. A total of 36 Charpy

specimens were tested at various test temperatures (noted in Table 5-2 through Table 5-4).

Impact energy was measured using a NIST-certified Satec S 1-1 K impact tester with 240 ft-lb

available hammer energy and 16.97 ft/second hammer velocity; the accuracy of this Charpy

tester is ± 1 ft-lb or 5% of the dial reading, whichever is larger. Lateral expansion was

measured using a dial indicator mounted on a specialized anvil. Percent shear was estimated

by video examination and comparison with the visual standards contained in ASTM E23-94a.

Test temperature was controlled to ±2°F and monitored using circulating oil heating baths and

an ethanol cooling bath with Omega and EXACAL digital temperature controllers.

The instrumented test data for the irradiated Charpy V-notch impact specimens were analyzed

with certified CHARTEST software. The test results were plotted using certified CVGRAPH

software; the results are summarized in Table 5-2 through Table 5-4 and Figure 5-3 through

Figure 5-11.

The data show that the materials exhibited a sensitivity to irradiation within the values to be

expected from their chemical composition and the fluence to which they were exposed. Scatter

in the TE1-C HAZ data prevents a serious interpretation of the results regarding the

temperatures at which 30 ft-lb and 50 ft-lb are reached. The TE1-C base metal and weld metal

data appear to follow a smooth trend with an exception being the weld data at 250°F, which is

above the upper-shelf trend and considered to be abnormal scatter.

The results of the pre-irradiation Charpy V-notch impact tests are given in Appendix C.

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Table 5-2: Charpy Impact Data for Capsule TE1-C ~ase Metal, BCC 241, Tramwerse Orientation, Irradiated to 1.88 x 1019 n/cm2 (E>1

MeV)

Specimen Test Temperature Impact Energy Lateral Expansion Percent Shear No. (oF) (ft-lb) (J) (mils) (%)

SS649 0 9.5 12.9 7 0

SS670 20 21.5 29.2 16 5

SS641 40 33.75 45.8 29 10

SS623 69 38.5 52.2 34 35

SS629 90 54.25 73.6 48 45

SS647 100 76.25 103.4 59 60

SS665 125 77.5 105.1 61 80

SS685' 150 92.5 125.4 71 80

SS611 175 116.5 158.0 84 100

SS671 200 114.5 155.2 81 100

SS668 250 118 160.0 83 100

SS620 300 113.5 153.9 82 100

Table 5-3: Charpy Impact Data for Capsule TE1-C Heat-Affected Zone Metal, BCC 241, Transverse Orientation, Irradiated to 1.88 x 1019

n/cm2 (E > 1 MeV)

Specimen Test Temperature Impact Energy La~eral Expansion Percent Shear No. (oF) (ft-lb) (J) (mils) (%)

SS353 -50 61.5 83.4 37 0

SS328 -25 21 28.5 11 5

SS314 0 59.75 81.0 35 15

SS382 20 92.5 125.4 59 60

SS321 20 80.25 108.8 49 70

SS340 40 55 74.6 35 45

SS386 69 80 108.5 60 80

SS368 100 76.25 103.4 58 70

SS352 ' 125 120.5 163.4 81 100

SS392 150 118.5 160.7 81 100

SS344 200 78.25 106.1 72 100

SS381 300 110.5 149.8 81 100

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Table 5-4: Charpy Impact Data for Caftsule TE1-C Weld Metal, WF-182-1, Irradiated to 1.88 x 10 9 n/cm2 (E > 1 MeV)

Specimen Test Temperature Impact Energy Lateral Expansion Percent Shear No. (oF) (ft-lb) (J) (mils) (%)

SS007 40 14.75 20.0 8 0

SS020 68 12 16.3 11 10

SS091 100 18 24.4 17 35

SS018 125 21.75 29.5 20 40

SS043 150 25.5 34.6 25 50

SS050 175 27.5 37.3 26 65

SS023 175 39.25 53.2 31 80

SS082 200 44.5 60.3 40 95

SS070 225 44.5 60.3 38 95

SS006 250 55.25 74.9 51 100

SS037 300 44.5 60.3 41 100

SS041 350 46.5-- 63.0 42 100


Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Page 5-8 Analysis of Capsule TE1-C

Figure 5-3: Impact Data (Impact Energy) for Irradiated Shell Forging Material, BCC 241

A= 58.9 B == 56. 7 C = 75.05 TO:::: 88.65 D = O.OOE+OO Equntion is A+ B * (Tnnh((T-To)/(C+DT))]

Upper Shelf Energy=-! 15.6(Fb:ed) Lower Shelf Enerm=2.2(Fixed) Temp@30 fi-lbs=46.5 Deg F Temp@50 ft-lbs--76.8 Deg F

Pln11t: DA VIS-BESSE Malerietl: SA508CL2 Heat: BCC24 l Orientation: TL Capsule: TEl-C Fluencc: TBD

140 i----·-r·· __ 1

___ . -··--- . r-·· ---- -! ;

----- --- -- {"- ------·l

-- :-- - - -- I _ -,

120 j _ -i- --i - ---'. --~ 100 - -r-- ---j--.f 80 ·---"-·- ---·- --- ------

E> OJ c: w z (;

I ! 1 I --- -·-- -1-- ··----i -· ···· -·~- r-=------- ·· ·--+------·-·I

_L___ ----+---- ----!--------~ ! ! l

60 ---- --j--·----.

·I 40 .. ~.-

20 - ·1 0 . ->--·>---•---.-· •

-;--~-- -·---1--.--.. . .. ·--··· ·--·4·-··-- -··-

~ ""~- ·· .... _......_ t ..... -'1 ·- ·t f ' -t - •-, -+--!-- -·· 1• -·I · f ,_ 1--- --t-·-• --t-- f--i--t -- I

-300 -200 -100 0 100 200 300 400 500 . 600

Temperature in Deg F

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Figure 5-4: Impact Data (Lateral Expansion) for Irradiated Shell Forging Material, BCC 241

A= 42.58 B"" 41.58 C = 76:45 TO= 77.91 D = O.OOE+OO Equation is A+ B * [Tnnh((T-To)/(C+DT))]

Upper Shelf L.E.=84.2 Lower Shelf L.E.= I .O(Fixecl) Temp.@L •. E. 35 mils=63.9 Deg F

Plnnt: DA VIS-BESSE Materi:il: SA508CL2 Hent: BCC241 Oricntotion: TL Capsule: TEl-C Fluence:-YBD n/cm"2

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100 - - --~----T---- --·- . --... I. ·-l · ·- . i-·- ... ·---. -·-90 ___ _j_ _______ ·+--... ---- --· ---- -··. ----1---- -·t--~--. ----· ---------1

I . o

Ja :: ==t~_-=-L ___ ~=--=~ ----_ ----~-j-.~~------=---~ :; 60 --- - ---" -- --- ---t _] _ j .-. . ---~ -----~ i 501--- --------- --r----~ ------ +--- ------i 40 J .. --- ... --~--- ·----- ·---- --·r------r------------ ···-·--- --- -----1 ~ 30 --------·--·-· ------------·---u-·-J-----+--- ---------·--·-----.. ·;

20 ·----··-~---·-··-- -·-- ---- ,·--- -----·7--- .. ----~-----·- ---~- ·------~ I '

10 ----- - . ----- ,------ - __ :r· ---- -y ------ .... ---- -- --- -------J·. 0 · -~ -+--·•.__ -+--·+---4- · -t--·-1" -~ ····f--1··--•-· •·-- ' • • -t-·i·- •. _.__ ---~· -----·•-., .. ~•-···I· .. · 1 ·~ .. ..,._ .. ,.. .. _,_.

-300 ·200 -100 0 100 200 300 400 . 500 600



Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1 -C Page 5-10

... ca Cl> J: ti)

c ii) 0 ... Gl

Q..

Figure 5-5: Impact Data (Percent Shear) for Irradiated Shell Forging Material, BCC 241

80

60

40.

i--

A::::: 50. B = 50, C = 54.24 TO= 91.71 D = O.OOE+OO Equotion is A ·f· B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shcnr = 91.8 Plant: DA VIS-BESSE Materinl: SA508CL2 Hcnt: DCC24 l

Oricntolion: TL Copsule: TEI-C Fluence: TBD n/cm"2

--0

p ------·

20 -!----+---·------ - ---1----l----1----t---- ----1

.'

0 - -f·_.'i...._- __ ,...,..._.,~-+-..,....., .•• ~~~-t--1--__._,·--.i---1~--1---f--•--r--i--... 1~--·--- _,1-t-->-i

-300 -200 -100 0 100 200 300 400 500 600


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Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1 -C

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Page 5-11

Figure 5-6: Impact Data (Impact Energy) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241

[J) .c I 0 0 u. >-e> C) c: w z > u

140 -~·

120

100 .

80 -

60

40

A ;; 54.55 B = 52.35 C = 130.55 TO = -17.57 D = O.OOE+OO Equation is A+ B * {Tm1h((T-To)/(C+DT))]

Upper Shelf Energy= I 06.9(Fixed) Lower Shelf Encrg)=2.2(Fixed) Temp@30 ft-lbs=-83.9 Deg F Temp(({~50 ft-Jbs=-28.9 Deg F Pinnt: DAVIS-BESSE Material: SA508CL2 Heat: BCC24 I

Orientation: TL Capsule: TEl-C Fh1ence: TBD n/cm"2

T o~ ·- - _..:..-~.

1ff ----Ii)

11

~

Q..,,_ ~- --0

~

--

·--r-----

'

--·

--

I -

20 _,,//

. :::71~0 -~,,_J" .... ~--

0 ·300

. ~-. --1-1--.. -.t ....... ~--

----- _,, ____

~·--1-l·-~-,-1---..-+-- -·-~~·---.a-

-200 -100 0 100 200 300 400 500 600



Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Page 5-12

~ E c .2 II) c: m Q.

~ -£!:! Ill ..... .9

100

90

80

70

60

50

40

30

20

10

Figure 5-7: Impact Data (Lateral Expansion) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241

A= 41.85 B = 40.85 C = 125.9 TO= 13.99 D = O.OOE+OO Equation is A+ B" [Tru1h((T-To)/(C+DT))]

Upper Sbelf L.E.=82.7 Lower Shelf L.E.=1.0(Fixed) Temp.l?JL.E. 35 mils=-7.3 Deg F

Pinnt: DA VIS-BESSE Molerilll: SA508CL2 Hent: BCC24 l Orientalion: TL Capsule: TEJ-C Flucncc: TBD n/cm"2

I ------ ----F---~-

=~o~I~ 1C ____ --·- ........ --·---

--· (j)

- -- -u---d 1)

--·--~---· I .... ~ . --

~: 0

-

1= ----

-

I __ - - 7~ n

,__ .. ......--Y -1~-1--·-1--,-· 0

-300 -200

-t-.f-t-- -<-1--1--l .... --t-,__,-1--;-,-1--+--1~-•-[I-~·-!-~ --1-t--+-

0 100 200 300 400 500 600


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Figure 5-8: Impact Data (Percent Shear) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241

... m .c

UJ

120

100

80

1: 60 (I)

~ Q)

D.. 40

-

.

A= 50. B = 50. C = 63.83 TO= 27.17 D := O.OOE+OO Equntion is A+ B * (Trmh((T-To)/(C+DT))]

Temperature at 50% Shenr"" 27.2 Pinnt: DA VIS-BESSE Materinl: SA508CL2 Heal: BCC241

Orientation: TL Capsule: TEl-C Fluencc: TBD n/cm"2

--~-~~-· - ----

--· -- I-

0 ~ i'>

-i----r~ -~-~ .. ----- ----

~---

~--

---

'~

-~ 1--~~L-20

0 -300

-i.---t•_...·I'--'-

---·--

_,_, I -~ -~--1-

-200 -100 0 100 200 300 400 500 600

Temperature In Deg F

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Figure 5-9: Impact Data (Impact Energy) for Irradiated Weld Metal, WF-182-1

A = 24.98 B = 22. 77 C = 93.03 TO= 130.11 D = O.OOE+OO Equntlon is A + B * [Ttmh((T-To)/(C+DT))]

Upper Shelf Energy=47.8(Fixed) Lower Shelf Energy=2.2(Fixed) Temp@30 ft-lbs= I 51.0Dcg F Tcmp@50 ft-lbs= NA

Plant: DAVIS-BESSE Material; LINDE80 Heat: WF-182-1 Orientation: NA Capsule: TE 1-C Fluence: TBD 11/cml\2

60 r···- -··-· , !

i - r - ---T·---- -- r:·-T----1·· 50 - --- - - -1- ----- -

1-- I - --i- -------

~ 40 L--· .. - .. --· ··----·----···---· ... -··--oJ-~------ I -·· ----· --J --·· '"

i

!

... j

~~~ 30 I_ -__ : __ ·-- . - I ~ t ------ ~~r- · --- -i-·-i-·- ·-~ 20 t ---- - . ----.-- -· ... ---t' - ' -- .. i . .. :

0 J : ~ . I -- -. ... - .... ·- ----··--i· ... ---···-- --· -- -·- -!··---·-- -

. • I - . : f

10 i -------·---r --- ~----l

0 -J--1··-i- --..J..--f.·-.J-- --J----1-· -1---+·-~·t--_.,_-·-f--~- ·-•-··I I I ·-i ···"'·-·I--•· -- - ~--·-'- · .,..... __

-300 -200 -100 0 · 100 200 300 400 500 600


AREVA Inc.


60

50

.!n

.E 4o c .2 w c a 30

~

Figure 5-10: Impact Data (Lateral Expansion) for Irradiated Weld Metal, WF-182-1

A= 23.39 B = 22.39 C = 99.37 TO= 136.92 D = O.OOE+OO Equalio11 is A+ B * [Tnnh((T-To)/(C+DT))]

Upper Shelf L.E.=45.8 Lower Shelf L.E ... I .O(Fixed) [email protected]. 35 mils= 194.0 Deg F

Plant: DAVIS-BESSE Malerial: LINDE80 Heat: WF-182-1 Orientation: NA C;;ipsule: TE 1-C Fluencc: TB D n/cm"2

.......

0 -·---I

!.,..---·-~

/.,.,.

/ (i> 0

r I

I T-·--DQ -

·-· --q

j

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Page 5-15

--

·-

10 t __ ~ -- ,_ ........ _ _.....

~l.~ ~·----

0 -300

0

_, __ ,_,-=i::::=,~1.-t-l- --t-t-···•- ---1-r-1-t-

0 100 200 300 Temperature in Deg F

-~~ ..... 1.,..-

400 500 600

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Figure 5-11: Impact Data (Percent Shear) for Irradiated Weld Metal, WF-182-1

.... 113 (!) .c U)

1: Cl> l.) ... ¢) a.

120

100

80

60

A = 50. B == 50. C = 67 .84 TO "" 138.36 D = O.OOE+OO Equation is A+ B * [Tanh((T-To)/(C+DT))}

Tcmpcrnlure nt 50% Shenr =- 138.4 Pinnt: DA VIS-BESSE Mnleriul: LfNDE80 Heal: WF-182-1

Orientotion: NA Capsule: TEl-C Fl11ence: TBD n/cm"2

---·--

--- r ·~--e---

-1 0 -

40 --------· -·

20

·-·;-"'\• 0 -300 -200

()

I ~---..-~~ ~Lt _._._...

-100 0 100 200 300 400


l -

·-

-

1-.. 1-

500 600

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6.0 NEUTRON FLUENCE

6.1 Introduction

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The neutron fluence (time integral of flux) is a quantitative way of expressing the cumulative

exposure of a material to a neutron flux over a specific period of time. Fast neutron fluence,

defined as the fluence of neutrons having energies greater than 1 MeV, is used to correlate

radiation induced changes in material properties. Accordingly, the cumulative fast fluence must

be determined at two locations: (1) in the test specimens located in the surveillance capsule,

and (2) in the wall of the reactor vessel. The former is used in developing the correlation

between fast fluence and changes in the material properties of specimens, and the latter is used

to ascertain the point of maximum (peak) fluence in the reactor vessel, the relative radial and

azimuthal distribution of the fluence, the fluence gradient through the RV wall, and the

corresponding material properties.

A previous estimate of the expected neutron fluence for capsule TE1-C is 1.81 x 1019 n/cm2 (E >

1 MeV) (Reference 9). The projected 60-year peak neutron fluence at the inside wetted· surface

of the reactor vessel, reported in the NRC's Safety Evaluation Report (SER) for the DB-1

License Renewal Application (LRA), is 1.70 x 1019 n/cm2 (E > 1 MeV) (Reference 15).

The accurate determination of neutron flux is typically accomplished by considering both

neutron dosimeter measurements and analytically derived flux spectra. Dosimeters were

withdrawn with the surveillance material in capsule TE1-C at the end of fuel cycle 7 (EOC 7) in

1991. However after fifteen years in storage, the dosimeters were considered to no longer

provide meaningful data and were discarded; thus dosimetry data specific to capsule TE1-C is

not available for comparison to calculated flux values.

Therefore, the analytical determination of neutron fluence received by the material specimens in

the surveillance capsule prior to its withdrawal at EOC 7 is used to support the demonstration

that the effects of irradiation-induced RV embrittlement are sufficiently monitored. An NRC

approved methodology, described in the following sections, is used to calculate the neutron

fluence exposure to capsule TE1-C and the DB-1 reactor vessel. The fast neutron fluence (E >

1 MeV) is calculated in accordance with the requirements of NRC Regulatory Guide 1.190

(Reference 16). The procedures and methods are presented in detail in Appendix D of this

report and in topical report BAW-2241 NP-A (Reference 17).

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6.2 Overview of Analytical Methodology

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BAW-2241 NP-A reports a calculation-based fluence analysis methodology that is used to

accurately predict the fast neutron fluence (E > 1 MeV) in the reactor vessel using surveillance

capsule dosimetry, cavity dosimetry, or both to verify the uncertainties in the fluence predictions.

The methodology was developed through a full-scale benchmark experiment that was

performed at the DB-1 reactor. The results of the benchmark experiment demonstrated that a

fluence analysis that employs this methodology (1) has an unbiased accuracy, and (2) has an

uncertainty within the NRC Regulatory Guide 1.190 suggested one standard deviation (cr) limit

of 20% for RV beltline locations.

Neutron transport calculations in three-dimensional synthesized geometry are used to obtain

energy dependent flux distributions throughout the _core. Geometric detail is selected to

explicitly represent the surveillance capsule and the reactor vessel. An analysis ·providing the

most up-to-date fluence estimates is performed for Cycles 1 through 7. Comparisons of the

calculated fluence values for capsule TE1-C to fluence values reported for other DB-1

surveillance capsules are used to show that the calculation results are reasonable and that the

TE1-C results are consistent with the AREVA benchmark database of uncertainties.

A detailed summary of the fluence methodology and uncertainty methodology are provided in

Appendix D.

6.3 F/uence Analysis Inputs

6.3.1 Reactor Geometry

A RV cross-section showing the DB-1 surveillance capsule holder tubes and capsule TE1-C is

presented in Figure 3-1. · Capsule TE1-C was in the bottom location of surveillance specimen

holder tube YX from initial fuel loading until it was withdrawn at EOC 7. The capsule was

located 10.9° from the 1/8 core symmetry axis.

The loading of surveillance capsule TE1-C is shown in Figure 3-2. The locations of the Charpy

specimens (Figure 3-2, Groups 1-6) and tensile specimens (Figure 3-2, Group 7) were inputs for

the fluence analysis. The locations of the specimens were input as three-dimensional

coordinates (R, a, Z) relative to the origin; the radius (R) values indicate the distance out from

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the center of the core, the azimuth (9) values indicate the angle off the major axis to the

centerline of the specimen, and the axial height (Z) indicates the distance below the reactor

vessel flange.

The TE1-C capsule fluence analysis was performed with greater detail and precision than

previous RVSP capsule analyses. Mesh spacing is much finer and the surveillance capsule and

Charpy specimen details are all less than 1 % different than the actual dimensions. The tensile

specimen details have a difference slightly greater than the Charpy specimens due to their small

circular geometry at the center of the specimen. Each cycle was modeled independently rather

than being grouped together. The cross sections were also updated to match operating

conditions at DB-1 during the early fuel cycles.

6.3.2 Cycle Lengths

The fuel cycle lengths in effective full power days (EF~D) and effective full power seconds

(EFPS) for Cycles 1 through 7 are reported in Table 6-1.

Table 6-1: DB-1 Fuel Cycle Lengths, Cycles 1 through 7

Cycle Cycle Length (EFPD) Cycle Length (EFPS)

1 374.20 3.23E+07

2 296.00 2.56E+07

3 272.70 2.36E+07

4 271.74 2.35E+07

5 393.77 3.40E+07

6 380.30 3.29E+07

7 405.22 3.50E+07

6.4 F/uence Analysis Results

6.4.1 Capsule Fluence Rate (Time-Averaged Flux)

The three dimensional, synthesized, incident fast neutron fluence rate (time averaged flux, E > 1

MeV) was calculated at the center of each TE1-C specimen for Cycles 1 through 7. The

average fast neutron fluence rate (time averaged flux, E > 1 MeV) for each cycle is summarized

in Table 6-2.

<.

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Table 6-2: Capsule TE1-C Fast Fluence (E > 1 MeV) Rate Results

Average Fast Fluence Rate* (n/cm2/s, E > 1 MeV)

Cycle 1 Cycle 2 Cycle 3 Cycle4 Cycle 5 Cycle 6 Cycle 7

9.62E+10 1.09E+11 1.15E+11 1.18E+11 7.99E+10 6.88E+10 6.99E+10

*Average of all specimens for a given cycle

6.4.2 Capsule Fluence

The individual cycle fluence values are determined by multiplying the fluence rates (Table 6-2)

by the respective cycle lengths in seconds provided in Table 6-1. The cumulative fluence for

each specimen in capsule TE1-C is the sum of the individual (incremental) fluence values for

Cycles 1 through 7. The average incremental and average cumulative fluence values are

shown in Table 6-3.

Table 6-3: Capsule TE1-C Fast Fluence (E > 1 MeV) Results

Average Fast Fluence* (n/cm2, E > 1 MeV)

Cycle 1 Cycle 2 Cycle 3 Cycle4 Cycle 5 Cycle 6 Cycle 7 Cumulative

3.11E+18 2.79E+18 2.72E+18 2.78E+18 2.72E+18 2.26E+18 2.45E+18 1.88E+19

*Average of all specimens for a given cycle

The cumulative fast neutron fluence (E > 1 MeV) for specimens in capsule TE1-C ranges from

1.55 x 1019 n/cm2 to 2.26 x 1019 n/cm2• As shown in Figure 3-2, the limiting material (WF-182-1)

specimens (Charpy specimens 88006, 88007, 88018, 88020, 88023, 88037, 88041, 88043,

88050, 88070, 88082, and 88091 and tensile specimens 88011 and 88013) are located core

side and received some of the highest fluence. The cumulative fast neutron fluence (E > 1

MeV) for the WF-182-1 specimens in capsule TE1-C ranges from 2.14 x 1019 n/cm2 to 2.22 x

1019 n/cm2.

6.4.3 Lead Factor

The lead factor is defined as the ratio of the average fluence rate in the surveillance capsule

specimens to the peak fluence rate on the inside surface of the reactor vessel. A previously

estimated average lead factor for the 10.9° capsule for the seven cycles that capsule TE1-C

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was installed is 6.53, with little variation between the minimum (6.46) and maximum (6.57)

(Reference 18).

The surveillance capsule fluence rate is the average for each cycle from Table 6-2. The peak

fluence rate on the inside (wetted) surface of the vessel for each cycle was determined by

synthesizing cases to determine the height (Z) at which the inside surface maximum fluence

rate occurs and the angle (0) at which the inside surface maximum fluence rate occurs and

confirming the maximum fluence rate on the inside surface. The results are presented in Table

6-4 for the wetted surface. A radius of 217.17 cm corresponds to the wetted surface of the

vessel.

Table 6-4: Capsule TE1-C Lead Factors, Wetted Surface

R e z Vessel Average Capsule Lead Cycle (cm) (Degrees) (cm) Fluence Rate Fluence Rate

Factor (n/cm2/s) (n/cm2/s)

1 217.17 10.41 501.56 1.47E+10 9.62E+10 6.55

2 217.17 10.41 625.07 1.71E+10 1.09E+11 6.37

3 217.17 10.41 621.12 1.79E+10 1.15E+11 6.44

4 217.17 12.10 569.37 1.83E+10 1.18E+11 6.47

5 217.17 12.10 625.07 1.24E+10 7.99E+10 6.42

6 217.17 10.41 557.44 1.07E+10 6.88E+10 6.45

7 217.17 10.41 501.56 1.09E+10 6.99E+10 6.41 •'

Average: 6.44

Notes:

• The radius (R) values indicate the distance out from the center of the core

• The azimuth (8) values indicate the angle off the major axis to th~ vessel location of maximum fluence

• The axial height (Z) indicates the distance below the reactor vessel flange

6.5 Fluence Uncertainty

The lack of dosimetry data for surveillance capsule TE1-C prevents a specific uncertainty to be

determined for this analysis. The benchmark database ensures that the fluence predictions are

consistent with the 10 CFR 50.61 (Reference 19) pressurized thermal shock (PTS) screening

criteria and the Regulatory Guide 1.99 (Reference 20) embrittlement evaluations.

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The uncertainty in benchmark comparisons of calculated to measured dosimetry results has

been updated to include 35 capsule analyses, including two from the PCA "Blind Test," a

comprehensive cavity benchmark experiment, and three standard cavity analyses. The generic

calc1,1lated capsule specimen fluence uncert~inty has been determined to be unbiased and has

an estimated standard deviation of 7.0 percent (Reference 17).

See Appendix D for a more detailed discussion of the methodology.

6.6 DB-1 Surveillance Capsule Comparison

To support the assessment of the fluence uncertainty for capsule TE1-C, previo·us fluence

results and estimates for DB-1 RVSP capsules TE1-C, TE1-D, and TE1-F were revievved. The

previous analyses report average cumulative fluence values for the center of the capsules. An

earlier TE1-C cumulative fluence value of 1.81 x 1019 n/cm2 (E > 1 MeV) was estimated based

on calculated fluence rates for Cycles 1 through 4 and the fluence from a capsule that was

located above capsule TE1-C for Cycles 5 through 7 using the fluence tracking system; the

fluence tracking system only calculated fluence values through Cycle 6 and estimated later

cycles (Reference 9).

Capsule TE1-D was irradiated in DB-1.for Cycles 1 through 6. The capsule was located in the

. top holder tube position at 26.9° off the major horizontal axis at approximately 202 cm from the

vertical axis of the core (Reference 4). Capsule TE1-F was irradiated in DB-1 for Cycle 1. The

capsule was located in the lower holder tube position at 26.9° off the major horizontal axis at

approximately 202 cm from the vertical axis of the core (Reference 1 ). The points of interest,

and their respective three-dimensional coordinates, relative to the flange mating surface, are

listed in Table 6-5.

Table 6-5: Three Dimensional Coordinates for DB-1 (TE1) RVSP Capsules Points of Interest

,

Capsule ID R 9 z

(cm) (Degrees) (cm)

TE1-C -202 10.9 520.07

TE1-D -202 26.9 443.87

TE1-F -202 26.9 520.07

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The three dimensional synthesized fluence rates (time averaged flux) at the center of capsule

TE1-C for Cycles 1 through 7, capsule TE1-D for Cycles 1-6, and capsule TE1-F for Cycle 1,

are shown in Table 6-6.

The individual cycle fluence values are determined by multiplying the fluence rates provided in

Table 6-6 by the respective cycle lengths in seconds provided in Table 6-1. The cumulative

fluence for each surveillance capsule is the sum of the cycles during which the capsule was in

the core. The resulting incremental and cumulative fluence values are shown in Table 6-7.

The previously calculated cumulative fluence values, a range for the .updated fluence values

based on the previously calculated fluence values and one standard deviation (see Appendix

D), and the updated cumulative fluence values of capsules TE1-C, TE1-D, and TE1-F are

presented in Table 6-8. The average fluence for the three surveillance capsules lies within the

expected range providing further verification that the calculated capsule specimen fluence

uncertainty has an estimated standard deviation of 7.0 percent.


Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Page 6-8

Tabl,e 6-6: DB-1 (TE1) 'RVSP Capsule Fast Fluence Rate (E > 1 MeV) Results

Capsule Fast Fluence Rate ·(n/cm2/s, E > 1 MeV)

ID Cycle 1 Cycle 2 Cycle 3 Cycle4 Cycle 5 Cycle 6

TE1-C 9.70E+10 1.10E+11 1.16E+11 1.19E+11 8.05E+10 6.93E+10

TE1-D 5.69E+10 6.77E+10 6.99E+10 7.05E+10 5.21E+10 4.38E+10

TE1-F 6.01E+10 -- -- -- -- --

Table 6-7: DB-1 (TE1) RVSP Capsule Fast Fluence (E > 1 MeV) Results

Capsule Fast Fluence (n/cm2, E > 1 MeV)

ID Cycle 1 Cycle 2 Cycle 3 , Cycle4 Cycle 5 Cycle 6 Cycle 7

TE1-C 3.14E+18 2.81E+18 2.73E+18 2.79E+18 2.74E+18 2.28E+18 2.47E+18

TE1-D 1.84E+18 1.73E+18 1.65E+18 1.66E+18 1.77E+18 1.44E+18 -TE1-F 1.94E+18 -- - -- -- - -

Table 6-8: DB-1 (TE1) RVSP Capsule Calculation Comparison

Capsule Fast Fluence (n/cm2, E > 1 MeV)

ID Previously Calculated Fluence Expected Range Updated Fluence ·I

TE1-C 1.81E+19 1.65E+19to 1.99E+19 1.89E+19

TE1-D 9.62E+18 8.75E+18 to 1.06E+19 1.00E+19

TE1-F 1.96E+18 1.78E+18 to 2.15E+18 1.94E+18

Cycle 7

7.06E+10

---

Cumulative

1.89E+19

1.01E+19

1.94E+18

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, Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C

6. 7 Fluence Analysis Conclusions

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Page 6-9

The average cumulative fluence exposure to surveillance capsule TE1-C specimens prior to its

removal at EOC 7 has been calculated to be 1.88 x 1019 n/cm2 (E > 1, MeV) using current \

methods, refined models, tighter meshing, and core follow data for each cycle.

The peak cumulative neutron fluence for the reactor vessel at end-of-life (EOL), 52 effective full

power years (EFPY) is 1.70 x 1019 n/cm2 (E > 1 MeV) at the inside wetted surface for upper

shell forging (AKJ 233), upper-to-lower shell circumferential weld (WF-182-1), and lower shell

forging (BCC 241). This peak wetted surface value corresponds to the EOL value for the lower

shell forging reported in the NRC's SER for the DB-1 LRA (Reference 15).

Fluence exposure for material specimens in capsule TEl-C prior to its withdrawal at EOC 7, is

confirmed through analysis to be greater than the EOL (52 EFPY) RV fast neutron fluence (E >

1 MeV) at the inside wetted surface for the limiting material. The fluence experienced by

material specimens in capsule TE1-C before its withdrawal is less than two times the peak 52

EFPY projected fluence; therefore the materials in the capsule provide meaningful metallurgical

data for the period of extended operation.

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7 .0 DISCUSSION OF CAPSULE RESULTS

7 .1 Tensile Properties

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The post-irradiation tensile data from surveillance capsules TE1-F, TE1-B, TE1-A, TE1-D, and

,TE1-C and the unirradiated (baseline) specimens are compared in Table 7-1 for tests at

approximately room temperature (69°F to 76°F) and in Table 7-2 for tests at elevated

temperatures (550°F to 580°F).

The general behavior of the tensile properties as a function of neutron irradiation is an increase

in both ultimate and yield strength and a decrease !n ductility as measured by both total

elongation and reduction of area. At both room temperature· and elevated temperature, the

ultimate and yield strength changes in the base metal as a result of irradiation and the

corresponding changes in ductility are considered to be within the limits observed for similar

materials. The changes at both room temperature and elevated temperature in the properties of

the weld metal are generally larger than those observed for the base metal, indicating a greater

sensitivity of the weld metal to irradiation damage.

7 .2 Charpy Impact Properties

The post-irradiation Charpy impact results for surveillance capsules TE1-F, TE1-B, TE1-A, TE1-

D, and TE1-C and the unirradiated (baseline) specimens are compared in Table 7-3.

The TE1-C Charpy impact test data exhibited the characteristic behavior of shift to higher

temperature for the 30 ft-lb transition temperature relative to the results of the unirradiated

specimens. The 30 ft-lb temperature shift for the TE1-C base metal was greater than those of

the previously analyzed capsules. The HAZ material for the TE1-C capsule had a temperature

shift at the 30 ft-lb level that was less than those of previously analyzed capsules. The

temperature shift at the 30 ft-lb level for the TE1-C weld metal was approximately 7% lower as

compared to the shift for capsule TE1-A, which exhibits the highest 30 ft-lb temperature shift for

the weld specimens. The base metal, HAZ material, and the weld metal specimens all exhibited

reductions in the upper shelf values comparable to that observed in the previous capsules.

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Table 7-1: Summary of DB-1 RVSP Capsule Tensile Test Results, Room Temperature Data

Matl. TE1 Fluence ** Test Yield Ultimate Uniform Total Reduction

ID Capsule (n/cm2)

Temp. Strength o/o *** Strength %*** Elongation %*** Elongation %*** of Area (oF) (ksi) (ksi) (%) (%) (%)

Q) Baseline 0 73* 72.3* - 90.7* -- 12.9* - 27.7* - 68.5* ~ (])

> F 1.96E+18 70 75.0 +3.7 95.6 +5.4 14 +8.5 26 -6.1 66 (/)

c..-~~ 1-N

B 5.92E+18 76 70.1 -3.0 91.1 +0.4 11 -14.7 26 -6.1 65 :::::- (.) ct! (.) Qi CD 2 D 9.62E+18 70 73.8 +2.1 95.2 +5.0 10 -22.5 25 -9.7 61 (]) (/) ct! A 1.29E+19 69 74.7 +3.3 96.4 +6.3 11 -14.7 25 -9.7 65 CD

Baseline 0 73* 70.2* - 85.6* - 15.1* -- 26.7* - 64.2*

(ii ..- F 1.96E+18 70 82.5 +17.5 98.1 +14.6 15 -0.7 25 -6.4 58 1U N 2 CX)

B 5.92E+18 76 85.5 +21.8 100.9 +17.9 10 -33.8 16 -40.1 54 ..-°O I - u_

~$ D 9.62E+18 70 87.3 +24.4 103.3 +20.7 10 -33.8 25 -6.4 56

A 1.29E+19 69 88.8 +26.5 104.1 +21.6 11 -27.2 23 -13.9 53

* Average of several specimens Average cumulative fast fluence (E > 1 MeV) Percent change relative to unirradiated (fluence = 0 n/cm2

) material at similar test temperature **

%***

--

-3.6

-5.1

-10.9

-5.1

-

-9.7

-15.9

-12.8

-17.4

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Table 7-2: Summary of DB-1 RVSP Capsule Tensile Test Results, Elevated Temperature Data

Matl. TE1 Fluence ** Test Yield Ultimate Uniform Total Reduction

ID Capsule (n/cm2)

Temp. Strength % *** Strength %*** Elongation %*** Elongation o/o *** of Area (oF) (ksi) (ksi) (%) (%) (%)

Baseline 0 580* 64.0* - 86.3* - 14.8* - 25.7* - 65.4* Q) {!!

F 1.96E+18 577 66.3 +3.6 88.8 +2.9 12 -18.9 22 -14.4 59 Q)

~ c.,.... 5.92E+18 580 66.9 +4.5 87.5 +1.4 8 -45.9 21 -18.3 57 Ill"<!" B

i=N ::::: (.) -Ill(.) Qi Ill D 9.62E+18 550 69.5 +8.6 91.9 +6.5 9 -39.2 22 -14.4 58 ::?! Q) A 1.29E+19 580 72.2 +12.8 92.4 +7.1 10 -32.4 23 -10.5 65 Ill . Ill co

c 1.88E+19 550 67.9 +6.1 - - - - - - -

Baseline 0 580* 67.6* - 83.2* - 12.9* - 18.8* - 50:2*

F 1.96E+18 577 73.1 +8.1 90.0 +8.2 11 -14.7 21 +11.7 48 (ij .,.... .._. I Q) N

::?! co B 5.92E+18 580 77.8 +15.1 93.9 +12.9 8 -38.0 15 -20.2 42

.,.... °C I -u. D 9.62E+18 550 78.1 +15.5 94.8 +13.9 9 -30.2 18 -4.3 48 ~~

A 1.29E+19 580 79.4 +17.5 96.4 +15.9 8 -38.0 17 -9.6 49

c 1.88E+19 550 79.9 +18.2 94.9 +14.1 8.4 -34.9 19.5. +3.7 42.6

* Average of several specimens

*** Average cumulative fast fluence (E > 1 MeV) Percent change relative to unirradiated (fluence = 0 n/cm2

) material at similar test temperature **

% ***

-

-9.8

-12.8

-11.3

-0.6

-

-

-4.4

-16.3

-4.4

-2.4

-15.1

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Table 7-3: Summary of DB-1 RVSP Caps·u1e Charpy Impact Test Results

TE1 Fluence * Tcvat Material 30 ft-lb

Capsule (nlcm2) (oF)

Baseline 0 +16

F 1.96E+18 -5

Base Metal B 5.92E+18 +2 (Transverse)

D 9.62E+18 +19 BCC 241

A 1.29E+19 +44

c 1.88E+19 +47

Baseline 0 -100

F 1.96E+18 -57

HAZ Metal B 5.92E+18 -43

BCC 241 D 9.62E+18 +1

A 1.29E+19 -66

c 1.88E+19 -84

Baseline 0 -11

F 1.96E+18 +116

Weld Metal B 5.92E+18 +114

WF-182-1 D 9.62E+18 +139

A 1.29E+19 +164

c 1.88E+19 +151

• NIA:: Not Applicable, MLE:: Mils Lateral Expansion Average cumulative fast fluence (E > 1 MeV)

.O.Tcv-30 ft-lb

(oF)

--21

-14

+3

+28

+31

-+43

+57

+101

+34

+16

-+127

:+-125

+150

+175

+162

Tcv at .0. Tcv ** Tcv at .0. Tcv ** 50 ft-lb 50 ft-lb 35MLE 35MLE

(oF) (oF) (oF) (oF)

+25 - +26 -

+26 +1 +27 +1

+41 +16 +34 +8

+55 +30 +42 +16

+48 +23 +50 +24

+77 +52 +64 +38

-57 - -45 --44 +13 -46 -1 -21 +36 -12 +33

+10 +67 +6 +51

+16 +73 +19 +64 "

-29 +28 -7 +38

+65 - +33 -+178 +113 +143 +110

+259 +194 +191 +158

+214 +149 +164 +131

+273 +208 +211. +178

N/A N/A +194 +161

•• •••

a Tevis defined as the change in temperature for a given Charpy property relative to the unirradiated (baseline) material a CvUSE is defined as the change in Upper Shelf Energy relative to the CvUSE of the unirradiated (baseline) material

Avg. CvUSE .O.CvUSE-(ft-lb) (ft-lb)

122 -120 -2

110 -12

117 -5

118 -4

116 -6

124 -115 -9

110 -14

117 -7

111 -13

107 -17

70 -65 -5

57 -13

54 -16

62 -8

48 -22

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8.0 SUMMARY OF RESULTS

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The analysis of the reactor vessel material contained in surveillance capsule TE1-C, removed

for evaluation as part of the DB-1 RVSP, led to the following conclusions:

1. The capsule received an average cumulative fast fluence of 1.88 x 1019 n/cm2 (E > 1. 0

MeV).

2. Based on the calculated fast flux at the RV wall, the projected peak fast fluence that the

DB-1 RV upper shell forging, upper-to-lower-shell circumferential weld, and lower shell

forging inside surface will receive in 52 EFPY of operation is 1.70 x 1019 n/cm2 (E > 1

MeV).

3. The results of the tension tests indicated that the materials exhibited normal behavior

relative to neutron fluence exposure. The ultimate and yield strength changes in the

TE1-C base metal as a result of irradiation and the corresponding changes in ductility

are considered to be within the limits observed for similar materials. The changes in the

properties of the TE1-C weld metal are generally larger than those observed for the base

metal, indicating a greater sensitivity of the weld metal to irradiation damage.

4. The Charpy impact test data exhibited the characteristic behavior of shift to higher

temperature for the 30 ft-lb transition tem'perature and a decrease in upper shelf energy

as a result of neutron fluence damage.

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Appendix A. Reactor Vessel Surveillance Program Background Data and Information

Material Selection Data

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The data used to select the materials for the specimens in the surveillance program, in

accordance with E-185-73, are shown in Table A-1. The locations of these materials within the

reactor vessel are shown in Figure A-1.

Definition of Beltline Region

The beltline region of DB-1 was defined in accordance with the data given in BAW-10100A.

Capsule Identification

The ID, type, and location of the capsules used in the DB-1 RVSP are identified below:

Capsule Cross Reference Data

Cap!;mle ID Type Location

TE1-A Ill Upper

TE1-B IV Lower

TE1-C Ill Upper

TE1-D IV Lower

TE1-E Ill Upper

TE1-F IV Lower

Specimens for Determining Material Baseline

See Table A-2.

Specimens Per Surveillance Capsule

See Table A-3 and Table A-4.

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Table A-1: Un irradiated Impact Properties and Residual Element Content Data of DB-1 RV Beltline Region Materials Used for Selection of Surveillance Program Materials

Transverse Charpy Chemical Composition

Drop Longitudinal Impact Data Material Material

RV Location Weight Charpy Impact RTNoT

ID Type TNDT Energy at 10°F 50 ft-lb 35 MLE USE

(oF) Cu p s Ni (oF) (ft-lb) Temp. Temp.

(ft-lb) wt%· wt% wt% wt% (oF) (oF)

ADB 203 SA 508 Cl. 2 Nozzle Belt 50 - 61 - 134 50 0.04 0.007 0.009 -

AKJ 233 SA508 Cl. 2 Upper Shell B 20 136, 179, 130

30 - 144 20 0.04 0.004 0.006 -107, 96, 81

BCC 241 SA 508 Cl. 2 Lower Shell A 50 60,62,47

27 - 118 50 0.02 0.011 0.011 -47,62,59

Upper WF-232 Weld Circ. Seam - 25,31,35 - - - - 0.14 0.011 0.007 -

(ID 9%)

Upper WF-233 Weld Circ. Seam - 43,30,26 - - - - 0.22 0.015 0.016 -

(OD 91%)

WF-182-1 Weld Middle

-20 36,33,44 62 81 2 0.18 0.014 0.015 Circ. Seam - -

Lower WF-232 Weld Circ. Seam - 25,31,35 - - - -- 0.14 0.011 0.007 -

(ID 12%)

Lower WF-233 Weld Circ. Seam - 43,30,26 - - - - 0.22 0.015 0.016 -

(OD 88%)

Note: Values listed in the most recent TE1 capsule report for capsule TE1-D are reported.

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Table A-2: Test Specimens for Determining Material Baseline Properties

' Material Description Number of Test Specimens

ID (Heat No.) Type (Orientation) <al Tension Tension at CVN Compact at 70°F 600°F (bl Impact Fracture (cl

Base Metal (T) 3 3 15 --Base Metal (L) 3 3 15 --

BCC 241 (5P4086) HAZ (T) 3 3 15 --HAZ (L) 3 3 15 --Total: 12 12 60 --

Base Metal (T) 3 3 15 --Base Metal (L) 3 3 15 --

AKJ 233 (123X244) HAZ (T) 3 3 15 --HAZ (L) 3 3 15 --Total: 12 12 60 --

WF-182-1 Weld Metal (L) 3 3 15 8 (1/2 TCT) 4 (1 TCT)

Notes: a. {T) = Transverse, (L) = Longitudinal b. Test temperature to be the same as irradiation temperature c. Test temperature to be determined from shift in impact transition curves after irradiation

exposure

Table A-3: Specimens in Upper Surveillance Capsules (Designations A, C, and E)

Material Type Material ID (Heat No.) Number of Tension Number of CVN Impact (Orientation) (a) Specimens Specimens

Weld Metal WF-182-1 2 12

Weld, HAZ (T) BCC 241 (5P4086) -- 12

Weld, HAZ (T) AKJ 233 (123X244) -- 6

Base Metal (T) BCC 241 (5P4086) 2 12

Base Metal (T) AKJ 233 (123X244) -- 6

Correlation Material HSST Plate 02 -- 6

Total per Capsule: 4 54

Note: a. {T) = Transverse

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Table A-4: Specimens in Lower Surveillance Capsules (Designations B, D, and F)

Material Type Material ID Number of Number of Number of Yz T

(Orientation) (a) (Heat No.) Tension CVN Impact Compact Fracture Specimens Specimens . Specimens (bl

Weld Metal WF-182-1 2 12 8

\fl!eld, HAZ (T) BCC 241 (5P4086) -- 12 --Base Metal (T) BCC 241 (5P4086) 2 12 --

Total per Capsule: 4 36 8

Notes: a. (T) =Transverse b. Compact fracture specimens pre-cracked per ASTM E399-72

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Figure A-1 : Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel

ADB-203 (Lower Nozzle Belt)

BCC241 (Lower Shell)

Dutchman

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Appendix B. Pre-Irradiation Tensile Data

Table B-1: Pre-Irradiation Tensile Properties of Shell Forging Material, BCC 241, Transverse Orientation

Yield Ultimate Uniform Total r Specimen Test Temp.

Strength Strength Elongation Elongation No. (oF)

(ksi) (ksi) (%) (%)

SS601 73 75.6 91.9 12.7 27.0

SS603 73 69.4 90.0 13.1 27.2

SS604 73 71.9 90.3 13.0 28.8

Mean 73 72.3 90.7 12.9 27.7

Std. Dev. 0 3.12 1.02 0.21 0.99

SS606 580 64.4 86.3 14.4 25.7

SS611 580 64.4 86.3 13.6 26.0

SS615 578 63.1 86.3 16.3 25.5

Mean 580 64.0 86.3 14.8 25.7

Std. Dev. 1.15 0.75 0 1.39 0.25

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Reduction of Area

(%)

67.3

67.0

71.1

68.5

2.29

65.4

63.7

67.0

65.4

1.65

Table B-2: ·Pre-Irradiation Tensile Properties for Weld Metal WF-182-1, Transverse Orientation

Specimen Test Temp. Yield Ultimate Uniform Total Reduction Strength Strength Elongation Elongation of Area No. (oF)

(ksi) (ksi) (%) (%) (%)

SS003 73 70.6 85.6 14.8 26.0 63.7

SS007 73 69.7 85.6 15.4 27.3 64.7

Mean 73 70.2 85.6\ 15.1 26.7 64.2

Std. Dev. 0 0.64 0 0.42 0.92 0.71

SS009 582 64.4 80.6 14.8 20.0 50.1

SS015 582 67.8 83.1 I 11.4 I 17.4 I 49.7

SS016 579 70.6 85.9 12.5 18.9 50.9

Mean 580 67.6 83.2 12.9 18.8 50.2

Std. Dev. 1.73 3.10 2.65 1.73 1.31 0.61

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Appendix C. Pre-Irradiation Charpy Impact Data

Table C-1: Pre-Irradiation Charpy Impact Data for Shell Forging Material, BCC 241, Transverse Orientation

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Specimen No. Test Temp. (°F) Absorbed Energy Lateral Expansion Shear Fracture

(ft-lb) (mils) (%)

SS642 -100 5.0 9 0

SS616 -79 5.5 10 0 . SS636 -40 17.5 14 0

SS609 -2 19.5 18 0

SS617 0 16.5 16 0

SS621 +21 39.0 33 2

SS666 +40 53.0 45 15

SS667 +40 73.0 57 20

SS672 +40 88.0 69 60

SS643 +70 76.0 60 25

SS646 +70 87.0 70 25

SS652 +74 109.0 79 85

SS627 +106 99.0 74 80

SS663 +130 111.5 85 90

SS686 +171 120.0 88 100

SS656 +213 128.5 92 100

SS658 +278 116.0 89 100

SS681 +338 113.5 88 100

SS630 +585 113.0 83 ( 100

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Table C-2: Pre-Irradiation Charpy Impact Data for Shell Forging Material Heat-Affected Zone, BCC 241, Transverse Orientation

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Specimen No. Test Temp. (°F) Absorbed Energy Lateral Expansion Shear Fracture (ft-lb) (mils) (%)

SS331 -120 27.0 19 0

SS330 -100 21.0 15 0

SS327 -100 19.0 13 0

SS307 -80 30.5 16 0

SS309 -80 60.0 36 0

SS310 -80 28.0 17 2

SS325 -59 67.0 37 20

SS346 -40 56.0 31 10

SS320 -20 62.0 37 25

SS337 -20 94.0 54 30

SS341 -2 97.5 57 60

SS329 +40 114.5 69 40

SS305 +74 133.0 76 90

SS333 +106 135.5 88 100

SS304 +130 110.5 77 100

SS315 +176 138.5 82 100

SS335 +223 110.0 79 100

SS343 +338 . 112.0 83 100

SS322 +406 135.5 84 100

SS348 +578 101.0 78 100

(

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Table C-3: Pre-Irradiation Charpy Impact Data for Weld Metal WF-182-1, Transverse Orientation

Specimen No. Test Temp. (°F) Absorbed Energy Lateral Expansion Shear Fracture (ft-lb) (mils) (%)

SS046 -80 15.5 16 '

0

SS060 -40 16.0 15 2

SS077 -2 37.5 35 10 ~

SS084 -2 28.0 27 25

SS053 0 33.0 29 20

SS055 0 33.5 29 15

SS027 +40 40.0 40 50

SS028 +40 40.0 38 35

SS029 +40 37.5 34 15

SS071 +70 45.5 44 50

SS081 +70 58.0 55 70

SS092 +74 55.0 56 75

SS056 +130' 70.5 - 64 100

SS067 +145 36.5. 35 40

SS036 +169 69.5 64 100

SS063 +223 72.5 71 100

SS085 +228 66.5 65 100

SS016 +338 72.0 70 100

SS040 +583 68.5 72 100

Table C-4: Pre-Irradiation Charpy USE and Index Temperatures

\ Tcv (°F) Tcv (°F) T CV (°F) Avg.CvUSE Material at 30 ft-lb* at 50 ft-lb* at 35 MLE*

Base Metal BCC 241 (Transverse) +16 +25 +26

HAZ Metal BCC 241 -100 -57 -45

Weld Metal WF-182-1 -11 +65 +33

Tcv = Charpy index temperature, MLE =Mils Lateral Expansion, CvUSE = Cha,.Py Upper Shelf Energy

* Values listed in the most recent TE1 capsule report for capsule TE1-D are reported.

(ft-lb)*

122

124

70

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Appendix D. Fluence Analysis Methodology

Analytical Methodology

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The primary analytical tool used in 'the determination of the flux and fluence exposure to the

capsule specimens is the two-dimensional discrete ordinates transport code DORT. The

primary technique used to verify the accuracy and uncertainty in the flux and fluence i§ a

benchmark to measured data. Fluence results from other DB-1 (TE1) capsules are used in

benchmark comparisons.

The DB-1 RVSP capsule TE1-C was located in the reactor vessel at 10.9° (off of the major axis).

for Cycles 1 through 7. The power distributions in the Cycle 1 through 7 irradiations were

symmetric both in 0 and Z. That is, the axial power shape is roughly the same for any angle,

and the azimuthal power shape is the same for any height. This means that the flux at some

point (R, 0, Z) can be considered to be a separable function of (R, 0) and (R, Z). Therefore,

irradiation for Cycles 1 through 7 can be modeled u~ing the standard synthesis procedures in

BAW-2241 NP-A (Reference 17).

Figure D-1 depicts the analytical procedure that is used to determine the fluence accumulated

over Cycles 1 through 7. As shown in the figure, the analysis is divided into several tasks:

generation of the neutron source, development of the DORT geometry models, calculation of

the macroscopic material cross sections, synthesis of the results, and estimation of the

calculational bias, the calculational uncertainty, and the final fluence. Each of these tasks is

discussed in greater detail in the following sections.

Generation of the Neutron Source

The time-averaged space and energy-dependent neutron source for Cycles 1 through 7 was .•

calculated using the SORREL code. The effects of burnup on the spatial distribution of the

neutron source are accounted for by calculating the cycle average fission spectrum for each

fissile isotope on an assembly-by-assembly basis and by determining the cycle-average specific

neutron emission rate. This data is then used with the normalized time weighted average pin

by-pin relative power density (RPO) distribution to determine the space and energy-dependent

neutron source. The azimuthally-averaged, time-averaged axial power shape in the peripheral

assemblies is used with the fission spectrum of the peripheral assemblies to determine the

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neutron source for the axial DORT run. These two neutron source distributions are input to

DORT as indicated in Figure D-1.

Development of the Geometrical Models

The system geometry models for DORT are developed using standard interval size and

configuration guidelines. The R9 model extends radially from the center of the core to a point

inside the reinforced concrete of the reactor cavity and azimuthally from the major axis to 45°.

Th~ surveillance capsule was modeled explicitly in the RS model. The axial model extends from

below the active core region to the reactor vessel flange mating surface above the active core

region. Both geometry models were developed using AREVA procedures for modeling and

were consistent with previous analyses. The geometrical models either meet or exceeded all

guidance criteria concerning interval size that are provided in Regulatory Guide 1.190. In all

cases, cold dimensions were used. The geometry models are input to the DORT code as

indicated in Figure D-1.

Calculation of Macroscopic Material Cross Sections

In accordance with BAW-2241 NP-A, the BUGLE-96 cross section library is used. The GIP code

was used to calculate the macroscopic energy-dependent cross sections for all materials used

in the analysis - radially from the core out through the cavity and into the concrete and axially

from below the active core region to the RV flange mating surface above the active core region.

The ENDF/B-VI dosimeter reaction cross sections are used to generate the response functions

that are used to calculate the DORT-calculated "saturated" specific activities.

DORT Analyses

The cross sections, geometry, and appropriate source are combined to crea~e a set of DORT

models (RS and RZ) for the Cycle 1 through 7 analyses. Each RS DORT run utilizes a P3

Legendre expansion of scattering cross sections, seventy directions (S10), and the appropriate

boundary conditions. The RZ models also use a scattering cross section P3 Legendre

expansion, seventy directions (S10), with the appropriate boundary conditions. A theta

weighted flux extrapolation model is used, and all other requirements of Regulatory Guide 1.190

that relate to the various DORT parameters are either met or exceeded for all DORT runs.

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Figure D-1: Fluence Analysis Methodology Flow Chart

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Assembly x Assembly RPD I Reactor Geometry Materinls of

Fission Spectrum pinxpin

I Construction

Distribution by Fissile Isotope

History

I BUGLE-93

+ DORT models Cross Section

-..i SORREL code Library

~,

Time-averaged Time-averaged L.j GIPCode

Radial Source Axial Source

S0(R.6,E) r

H Cross sections Dosimetry ... DORT Analysis ~

Counting ... R6andRZ -

and Analysis (NESI) Data to Calculate

Absolute , , Magnitude Results

Power History Synthesized (saturation) 3DResults

.. l\·Ieasured Calculated Dosimeter

4 Dosimetry Acti\ities ~ Activities I ..

y C/M :~ NO B&\VOG

Benchmark + Analysis Bias and ...... Statistical

~ Validate ... Validate

Uncertainty Analysis Bias ... Uncertainty

I t Apply Bias

Final Plant .... Removal Function .... YES Validation Specific Fluences "" and Specify "" Acceptable

Uncertainty

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Synthesized Three Dimensional Results

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The DORT analyses produce two sets of two-dimensional flux distributions, one for channels of

vertical cylinders and one for radial planes. The vertical cylinders, which are referred to as RZ

planes, are defined as planes bounded axially by water below the active core region to the RV

flange mating surface above the active core region and radially by the center of the core out into

the concrete cavity shield. The horizontal planes, referred to as the RS planes, are defined as

the planes bounded radially by the center of the core and a point located in the concrete cavity

shield, and azimuthally by the major axis and the adjacent 45° radius. The vessel flux varies

significantly in all three cylindrical-coordinate directions (R, e, Z). Under the assumption that the

three-dimensional flux is a separable function, the two-dimensional data sets are mathematically

combined to estimate the flux at all three-dimensional points (R, e, Z) of interest. The synthesis

procedure outlined in Regulatory Guide 1.190 forms the basis for the AREVA flux-synthesis

process.

Uncertainty

The fluence rates, time-averaged flux values, and thereby the fluence values throughout the

DB-1 reactor and vessel, are calculated with the DORT discrete ordinates computer code using

three-dimensional synthesis methods. The basic theory for synthesis is described in Section

3.0 of BAW-2241NP-A. The DORT three-dimensional synthesis results are the bases for the

fluence predictions using the AREVA "Semi-Analytical" (calculational) methodology.

The embrittlement evaluations in Regulatory Guide 1.99 and 10 CFR 50.61 for the PTS

screening criteria apply a margin term to the reference· temperatures. The margin term includes

the product of a .confidence factor of 2.0 and the mean embrittlement standard deviation. The

factor of 2.0 implies a very high level of confidence in the fluence uncertainty as well as the

uncertainty in the other variables contributing to the embrittlement. The lack of meaningful data

from the dosimetry in capsule TE1-C would not directly support this high level of confidence,

since. the dosimetry was discarded after 15 years of storage with no measurements made.

However, as the same methodology is used, the calculational uncertainties in the updated

·fluence predictions for capsule TE1-C are supported by 728 dosimeter measurements and

thirty-nine benchmark comparisons of calculations to measurements as shown in Appendix A of

BAW-2241 NP-A The calculational uncertainties are also supported by the fluence sensitivity

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evaluation of uncertainties in the physical and operational parameters, which are included in the

vessel fluence uncertainty. The dosimetry measurements and benchmarks, as well as the

fluence sensitivity analyses, in BAW-2241 NP-A are sufficient to support a 95 percent confidence

level, with a confidence factor of± 2.0, in the fluence results for specimens in capsule TE1-C

from the "Semi-Analytical" methodology.

The AREVA generic uncertainty in the capsule dosimetry measurements has been determined

to be unbiased and has an estimated standard deviation of 7.0 percent for the qualified set of

dosimeters. The AREVA generic uncertainty for benchmark comparisons of capsule dosimetry

calculations relative to the measurements indicates that any benchmark bias in the greater than

1.0 MeV results is too small to be uniquely identified. The estimated standard deviation

between the calculations and measurements is 9.9 percent. This implies that the root mean

square deviation for the AREVA calculations of the TE1-C capsule fluence should be

approximately 9.9 percent in general and bounded by ± 20.0 percent for a 95 percent

confidence interval with thirty-nine independent benchmarks.

The AREVA generic calculated capsule specimen fluence uncertainty has been determined to

be unbiased and has an estimated standard deviation of 7.0 percent. In order to compare the

updated calculations to past calculations, this standard deviation must be applied to both

calculations. Therefore the uncertainty due to both would be "(0.072+0.072) = 0.098995.

Looking at a ratio of the updated fluence to the previously calculated fluence, the uncertainty

would be applied as follows:

1 < <P updated < 1+0.098995 1+0.098995 - <P previous - 1

A range for the updated fluence is determined by multiplying through by the previous calculated

fluence resulting in the following:

<I> . prtIVIUUS <<I> rpda1rui ~ <D pnMOus X 1.098995

1.098995

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Appendix E. ASTM E185-82 RVSP Technical Report Requirements

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As discussed in Section 1.0 of this report, the DB-1 RVSP is conducted in accordance with

ASTM E 185-82 (Reference 10) to the extent possible. Section 11 of ASTM E 185-82 lists the

information that shall be provided in a RVSP capsule report; several requirements listed in

ASTM E185-82 Section 11 that are not included in this report are described below, as required

by Section 11.6 of ASTM E185-82.

• Description of the TE 1-C neutron dosimeters and temperature monitors and the

corresponding neutron dosimeter measurements and temperature monitor results are

required per ASTM E185-82 Sections 11.3.3.1, 11.4.5, and 11.5.2. The neutron

dosimeters and temperature monitors were discarded after the TE1-C capsule was

removed from the reactor vessel (see Sections 3.0 and 6.0 of this report); therefore the

corresponding data are not available, and the description ·of these components is no

longer relevant to the results in this report.

• Reporting the neutron fluence (> 0.1 MeV and 1 MeV) for the surveillance specimens is

required per ASTM E185-82 Section 11.4.5.2. Neutron fluence > 1 MeV is reported in

this document. Neutron fluence > 0.1 MeV typically is not used to assess the radiation

embrittlement of RV materials via adjusted reference temperature calculations and

fracture mechanics analyses. Therefore, this information is not included in this report.

• Extrapolation of the neutron flux and fluence results to the surface and % T location of

the reactor vessel at the peak fluence location is required per ASTM E185-82 Section

11.5.1, and the determination of the lead factors between the specimen fluence and the

peak vessel fluence at the surface and% T location is required per ASTM E185-82

Section 11.2.4. Extrapolation of the fracture toughness properties to the surface and %

T locations of the reactor vessel at the peak fluence locations is required per ASTM

E185-82 Section 11.5.3. This work supports radiation embrittlement calculations which

are not included in this report.

• Reporting the adjusted reference temperature for each surveillance material is required

per ASTM E185-82 Section 11.4.2.3. The adjusted reference temperatures for these

materials are not calculated in this report.


Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program: Analysis of Capsule TE1-C Page E-2

• Several details regarding the Charpy and tension test instrumentation and results are not

included in this report. These details are as follows:

o Certification and calibration of all equipment and instruments used in conducting

the tests (required per ASTM E185-82 Section 11.3.3.2).

o Trade name and model of the gripping devices used for the tension tests

(required per ASTM E185-82 Section 11.4.1.1)

o Method of yield strength measurement (required per ASTM E185-82 Section

11.4.1.4)

o Description of the procedure used in the inspection and calibration of the Charpy

impact tester (required per ASTM E185-82 Section 11.4.2.1)

o Fracture surface appearance (required per ASTM E185-82 Section 11.4.2.2)

/

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Appendix F. References

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1. AREVA Docum~nt 77-I 132285-00 (BA W-I 70I), "Analyses of Capsule TEI-F, The Toledo Edison Company, Davis-Besse Nuclear Power Station Unit I, Reactor Vessel Materials Surveillance Program," January I 982.

2. AREVA Document 77-I I745I6-00 (BA W-I834), "Analyses of Capsule TEI-B, The Toledo Edison Company, Davis-Besse Nuclear Power Station Unit I, Reactor Vessel Material Surveillance Program," May I 984.

3. AREVA Document 77-1159086-0I (BA W-I882, Revision I), "Analyses of Capsule TEI-A, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit 1, Reactor Vessel Material Surveillance Program," June 1989.

4. AREVA Document 77-2125-00 (BA W-2I25), "Analysis of Capsule TEI-D, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit I, Reactor Vessel Material Surveillance Program," December 1990.

5. AREVA Document 43-IOIOOA-OO (BA W-IOIOOA), "Reactor Vessel Material Surveillance Program, Compliance with IO CFR 50, Appendix H, for Oconee Class Reactors," February I975.

6. Code of Federal Regulations, Title IO, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

7. ASTM EI85-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," ASTM International, I973.

8. AREVA Document 43-I543-04 (BAW-I543, Revision 4), "Master Integrated Reactor Vessel Surveillance Program," February I993.

9. AREVA Document 43-I543S-I I (BA W-1543(NP), Revision 4, Supplement 6-A), "Supplement to the Master Integrated Reactor Vessel Surveillance Program," June 2007.

10. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E 706 (IF)" ASTM International, July 1982.

I I. U.S. Nuclear Regulatory Commission, NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report," December 2010, NRC Accession Number MLI03490041.

12. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture.Toughness Requirements."

13. ASME Boiler and Pressure Vessel Code, Section III, Division I -Appendices, "Rules for Construction of Nuclear Facility Components," The American Society of Mechanical Engineers (latest version approved by 10 CFR 50.55a).

14. ASME Boiler and Pressure Vessel Code, Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," The American Society of Mechanical Engineers (latest version approved by IO CFR 50.55a).

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15. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the License Renewal of Davis-Besse Nuclear Power Station,'' September 2013, NRC Accession Number ML13248A267.

16. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,'' March 2001, NRC Accession Number ML010890301.

17. AREVA Document 43-2241NPA-002 (BA W-2241NP-A, Revision,2), "Fluence and Uncertainty Methodologies,'' April 2006.

18. AREVA Document 77-2108-01 (BA W-2108, Revision 1), "Fluence Tracking System," May 1992.

19. Code of Federal Regulations, Title 10, Part 50.61, "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events."

20. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,'' May 1988, NRC Accession Number ML003740284.

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