ML16224A241
| ML16224A241 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/31/2014 |
| From: | AREVA |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16224A232 | List: |
| References | |
| L-16-227 ANP-3339 Rev 0 | |
| Download: ML16224A241 (66) | |
Text
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 ANP-3339 Davis-Besse Unit 1 Reactor Vessel Revision 0 Material Surveillance Program:
Analysis of Capsule TE1-C December 2014 AREVA Inc.
(c) 2014 AREVA Inc.
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 ................................................................................... 77 1 7.2 Charpy Impact Properties ....... ,............................................................... 7-1
AREVA Inc. ANP-3339 Revision 0
_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. REACTOR VESSEL SURVEILLANCE PROGRAM BACKGROUND DATA AND INFORMATION .............................. A-1 APPENDIX B. PRE-IRRADIATION TENSILE DATA ........................................... 8-1 APPENDIXC. PRE-IRRADIATION CHARPY IMPACT DATA. ........................... C-1 APPENDIX D. FLUENCE ANALYSIS METHODOLOGY .................................... D-1 APPENDIX E. ASTM E185-82 RVSP TECHNICAL REPORT REQUIREMENTS ........................................................................ E-1 APPENDIX F. REFERENCES ............................................................................. F-1
AREVA Inc. ANP-3339 Revision 0 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: DB-1 (TE1) RVSP Capsule Fast Fluence (E > 1 MeV) Results ................. 6-8 Table 6-8: DB-1 (TE1) RVSP Capsule Calculation Comparison ................................. 6-8 I
Table 7-1: Summary of DB-1 RVSP Capsule Tensile Test Results, Room Temperature Data ........................................................................................................ 7-2 Table 7-2: Summary of DB-1 RVSP Capsule Tensile Test Results, Elevated Temperature Data ................................................................................... 7-3 Table 7-3: Summary of DB-1 RVSP Capsule Charpy lmpactTest Results ................. 7-4 Table A-1: 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
AREVA Inc. ANP-3339 Revision 0 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
AREVA Inc. ANP-3339 Revision 0 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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page vi Nomenclature Abbreviation Definition ASME The American Society of Mechanical Engineers ASTM American Society of Testing and Materials B&W Babcock and Wilcox BPVC Boiler and Pressure Vessel Code BWR Boiling Water Reactor CFR Code of Federal Regulations CMTR Certified Material Test Report DB-1 Davis-Besse Nuclear Station Unit 1 E > 1 MeV Energy greater than 1 million electron volts E > 0.1 MeV Energy greater than 0.1 million electron volts EFPY Effective Full Power Years EOC End of Cycle EOL End of Life OF Degrees Fahrenheit FENOC f irstEnergy Nuclear Operating Company ft Foot GALL Generic Aging Lessons Learned HAZ Heat-Affected Zone J Joule Kie Stress Intensity Factor ksi Kilopound per square inch lb Pound LRA License Renewal Application MIRVP Master Integrated Reactor Vessel Surveillance Program MLE Mils of Lateral Expansion 2
n/cm Neutrons per square centimeter NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission PWR Pressurized Water Reactor RCPB Reactor Coolant Pressure Boundary RTNDT Reference Temperature, Nil Ductility Transition RV ~. Reactor Vessel RVSP Reactor Vessel Surveillance Program SER Safety Evaluation Report USE Upper Shelf Energy
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page vii
SUMMARY
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/cm 2 (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/cm 2 (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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 1-1
1.0 INTRODUCTION
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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 2-1
2.0 BACKGROUND
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
AREVA Inc: ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 2-2 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, RTNDT. 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 RTNDT 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 RTNDT 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 RTNDT 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.
AREVA Inc. ANP-3339 Revision 0 D,avis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 3-1 3.0 SURVEILLANCE PROGRAM DESCRIPTION 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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 3-2 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 Wt%
Material ID c Mn p s Si Ni Cr Mo Cu BCC 241 <a> 0.22 0.63 0.011 0.011 0.27 0.81 0.32 0.63 0,02 WF-182-1. Cbl 0.09 1.69 0.014 0.013 0.41 0.63 0.15 0.40 0.21 (a) Per Certified Materials Test Reports (CMTRs)
(b) Per BAW-1543, Revision 4 (Reference 8)
Table 3-3: Heat Treatment of Surveillance Materials [8]
Material ID Heat Treatment 1590°F +/- 10°F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Water Quenched BCC 241 1240°F +/- 10°F for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, Air Cooled 1125°F +/- 25°F for 15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, Furnace Cooled WF-182-1 1125°F +/- 25°F for 15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, Furnace Cooled
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program :
Analysis of Capsule TE1-C Page 3-3 Figure 3-1: Reactor Vessel Cross Section Showing Location of Capsule TE1 -C in Davis -Besse Unit 1 x
w ---tl++--hil"~~~-+--+-~ +---"~~lll--l.l==.....£1~~ y l--+-+-+-+--+--+-..........-+--+---+----+-.......---4~1--1 z
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 3-4 Figure 3-2: Loading Diagram for Test Specimens in Capsule TE1-C CORE tSIDE
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 4-1 4.0 PRE-IRRADIATION TESTS 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 10 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.
AREVA Inc. ANP-3339 Revision O Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-1 5.0 POST-IRRADIATION TESTS 5.1 Tension Test Results
\
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 TNSLTEST 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.
AREVA Inc. ANP-3339 Revision O Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-2 Table 5-1: Tensile Properties of Capsule TE1-C Irradiated Base Metal and Weld Metal Test Yield Tensile Fracture Fracture Fracture Uniform Total Reduction Specimen Material* Temp. Strength Strength Load Stress Strength Elongation Elongation in Area No. (oF) (ksi) (ksi) (lb) (ksi) (ksi) (%) (%) (%)
Weld Metal SS011 550 79.928 94.923 7765 135.115 77.572 8.39 19.45 42.6 Weld Metal SS013 ** 250 86.313 94.372 9446 198.348 94.372 -- -- 52.4 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 54.0
- (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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-3 Figure 5-1: Stress-Strain Curve for Irradiated Weld Metal Tensile Specimen SS011 in Capsule TE1-C S eoimen: SSOU lsat: Temp.: 550 F ( 287 C)
Strength Yield: 79928.
g UTS: 9Y923.
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C\l CJ ci 0.00 O.OY 0.08 0.12 O.lS 0.20 0.21! 0.28 0.32 Enginee~ing Strain
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-4 Figure 5-2: Stress-Strain Curve for Irradiated Base Metal Tensile Specimen SS617 in Capsule TE1-C Specimen: SS617 leet Temp.: 200 F ( 93 C)
Strength d Yield: 72150. (Cl UTS: 91050.
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- AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-5, 5.2 Charpy V-Notch Impact Test Results 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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-6 Table 5-2: Charpy Impact Data for Capsule TE1-C ~ase Metal, BCC 241, Tramwerse Orientation, Irradiated to 1.88 x 1019 n/cm 2 (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 19 Zone Metal, BCC 241, Transverse Orientation, Irradiated to 1.88 x 10 2
n/cm (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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-7 Table 5-4: Charpy Impact Data for Caftsule TE1-C Weld Metal, WF-182-1, Irradiated to 1.88 x 10 9 n/cm 2 (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
ANP-3339 AREVA Inc.
Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-8 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 140 i----*-r** __1 ___ .-**---
Orientation: TL Capsule: TEl-C Fluencc: TBD
. r-** ---- -
120 j_ -i- --i - ---'. --
~ 100 - -r-- ---j-- ---------- {"- ------*l
.f 80 *---"-*- ---*- --- ------ -- :-- - - -- I_-,
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- f ,_ 1--- --t-*-* --t-- f--i--t -- I
-300 -200 -100 0 100 200 300 400 500 . 600 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-9 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"2 100 - - --~----T---- --*- .--... I. *-l **- . i-*- . *---. -*-
90 ___ _j_ _______ *+--... ------* ---- -**. ----1---- -*t--~--. - - * ---------1 I . o
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- 1*~ ... ,._ .. ,.. .. _,_.
-300 *200 -100 0 100 200 300 400 . 500 600 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1 -C Page 5-10 Figure 5-5: Impact Data (Percent Shear) for Irradiated Shell Forging Material, BCC 241 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 80 i-- --0
...ca Cl>
J:
ti) cii) 60
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Q..
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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 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1 -C Page 5-11 Figure 5-6: Impact Data (Impact Energy) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241 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= I06.9(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"2 140 T
-~*
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0 . l*-~-,-1---..-+--
-*-~~*---.a-
- 300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-12 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"2 100 F---~- I 90 ------ ----
=~o~I~
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-300 -200 0 100 200 300 400 500 600 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-13 Figure 5-8: Impact Data (Percent Shear) for Irradiated Shell Forging Material, Heat-Affected Zone, BCC 241 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"2 120 100 - -- ~-~~-* - ---- ~---
80 --* -- I-m
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-300 -200 -100 0 100
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200
_,_, I 300 400
-~ -~- 500 600 Temperature In Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-14 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= I51.0Dcg F Tcmp@50 ft-lbs= NA Plant: DAVIS-BESSE Material; LINDE80 Heat: WF-182-1 r:*-T----1**
Orientation: NA Capsule: TE 1-C Fluence: TBD 11/cml\2 60 r***- -**-* ,
i - r- ---T*---- --
50 - --- - - ----- - - -
1 I - --i- -------
~ 40 L--* .. - . --* **----*----***---* . . -**--oJ-~------ -* ----* --J --** '"
I
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.
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- I I I *-i ***"'*-*I--** -- - ~--*-'- * .,. . .__
-300 -200 -100 0
- 100 200 300 400 500 600 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-15 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)
Tcmp.@L.E. 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 60 .......
0 50 -*---
I !.,..---*-~
/.,.,.
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.2 4o r I w
c a 30 T-*-- I *-
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0 0 _,__,_,-=i::::=,~1.-t-l- --t-t-****- ---1-r-1-t- -~~ ..... 1.,..-
-300 0 100 200 300 400 500 600 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 5-16 Figure 5-11: Impact Data (Percent Shear) for Irradiated Weld Metal, WF-182-1 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"2 120 ---*--
l r
100 --- *~--e--- -
113
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a.
40 - - - - ----* -*
()
I ~Lt 20 0 *-*;-"'\* ~---..-~~
_._._... 1 -..1 -
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-1 6.0 NEUTRON FLUENCE 6.1 Introduction 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/cm 2 (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).
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-2 6.2 Overview of Analytical Methodology 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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-3 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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-4 Table 6-2: Capsule TE1-C Fast Fluence (E > 1 MeV) Rate Results Average Fast Fluence Rate* (n/cm 2/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 2
Average Fast Fluence* (n/cm , 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/cm 2 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 10 19 n/cm 2 to 2.22 x 1019 n/cm 2 .
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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-5 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 Vessel Average Capsule Cycle R e z Fluence Rate Fluence Rate Lead (cm) (Degrees) (cm) Factor (n/cm 2/s) (n/cm 2/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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-6 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/cm 2 (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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-7 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.
AREVA Inc. ANP-3339 Revision 0 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/cm 2/s, E > 1 MeV)
ID Cycle 1 Cycle 2 Cycle 3 Cycle4 Cycle 5 Cycle 6 Cycle 7 TE1-C 9.70E+10 1.10E+11 1.16E+11 1.19E+11 8.05E+10 6.93E+10 7.06E+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 2
Fast Fluence (n/cm , E > 1 MeV)
Capsule ID Cycle 1 Cycle 2 Cycle 3 , Cycle4 Cycle 5 Cycle 6 Cycle 7 Cumulative TE1-C 3.14E+18 2.81E+18 2.73E+18 2.79E+18 2.74E+18 2.28E+18 2.47E+18 1.89E+19 TE1-D 1.84E+18 1.73E+18 1.65E+18 1.66E+18 1.77E+18 1.44E+18 - 1.01E+19 TE1-F 1.94E+18 -- - -- -- - - 1.94E+18 Table 6-8: DB-1 (TE1) RVSP Capsule Calculation Comparison 2
Capsule Fast Fluence (n/cm , 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
AREVA Inc. ANP-3339 Revision O
, Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 6-9
- 6. 7 Fluence Analysis Conclusions 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/cm 2 (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/cm 2 (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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 7-1 7 .0 DISCUSSION OF CAPSULE RESULTS 7 .1 Tensile Properties 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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 7-2 Table 7-1: Summary of DB-1 RVSP Capsule Tensile Test Results, Room Temperature Data Test Yield Ultimate Uniform Total Reduction Matl. TE1 Fluence **
2 Temp. Strength o/o *** Strength %*** Elongation %*** Elongation %*** of Area %***
ID Capsule (n/cm ) (oF) (ksi) (ksi) (%) (%) (%)
Q)
~
Baseline 0 73* 72.3* - 90.7* -- 12.9* - 27.7* - 68.5* --
(])
(/)
c..-
F 1.96E+18 70 75.0 +3.7 95.6 +5.4 14 +8.5 26 -6.1 66 -3.6
~~
1-N
- - (.) B 5.92E+18 76 70.1 -3.0 91.1 +0.4 11 -14.7 26 -6.1 65 -5.1 ct! (.)
Qi CD 2 D 9.62E+18 70 73.8 +2.1 95.2 +5.0 10 -22.5 25 -9.7 61 -10.9
(])
(/)
ct!
CD A 1.29E+19 69 74.7 +3.3 96.4 +6.3 11 -14.7 25 -9.7 65 -5.1 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 -9.7 1U N 2 CX)
°O I
..- B 5.92E+18 76 85.5 +21.8 100.9 +17.9 10 -33.8 16 -40.1 54 -15.9
- u_
~$ D 9.62E+18 70 87.3 +24.4 103.3 +20.7 10 -33.8 25 -6.4 56 -12.8 A 1.29E+19 69 88.8 +26.5 104.1 +21.6 11 -27.2 23 -13.9 53 -17.4
- Average of several specimens
- Average cumulative fast fluence (E > 1 MeV) 2 Percent change relative to unirradiated (fluence = 0 n/cm ) material at similar test temperature
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 7-3 Table 7-2: Summary of DB-1 RVSP Capsule Tensile Test Results, Elevated Temperature Data Test Yield Ultimate Uniform Total Reduction Matl. TE1 Fluence **
2 Temp. Strength % *** Strength %*** Elongation %*** Elongation o/o *** of Area % ***
ID Capsule (n/cm ) (oF) (ksi) (ksi) (%) (%) (%)
Baseline 0 580* 64.0* - 86.3* - 14.8* - 25.7* - 65.4* -
Q)
{!!
Q) F 1.96E+18 577 66.3 +3.6 88.8 +2.9 12 -18.9 22 -14.4 59 -9.8
~
c.,....
Ill"<!" B 5.92E+18 580 66.9 +4.5 87.5 +1.4 8 -45.9 21 -18.3 57 -12.8 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 -11.3
- ?!
Q)
Ill . A 1.29E+19 580 72.2 +12.8 92.4 +7.1 10 -32.4 23 -10.5 65 -0.6 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 -4.4 (ij .,....
.._. I Q) N B 5.92E+18 580 77.8 +15.1 93.9 +12.9 8 -38.0 15 -20.2 42 -16.3 co
- ?! .,....
°C I
-u. D 9.62E+18 550 78.1 +15.5 94.8 +13.9 9 -30.2 18 -4.3 48 -4.4
~~
A 1.29E+19 580 79.4 +17.5 96.4 +15.9 8 -38.0 17 -9.6 49 -2.4 c 1.88E+19 550 79.9 +18.2 94.9 +14.1 8.4 -34.9 19.5. +3.7 42.6 -15.1
- Average of several specimens
- Average cumulative fast fluence (E > 1 MeV) 2
- Percent change relative to unirradiated (fluence = 0 n/cm ) material at similar test temperature
AREVA Inc. ANP-3339 Revision O Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 7-4 Table 7-3: Summary of DB-1 RVSP Caps*u1e Charpy Impact Test Results Tcvat .O.Tcv- Tcv at .0. Tcv ** Tcv at .0. Tcv **
TE1 Fluence
- Avg. CvUSE .O.CvUSE-Material 2 30 ft-lb 30 ft-lb 50 ft-lb 50 ft-lb 35MLE 35MLE Capsule (nlcm ) (oF) (oF) (oF) (ft-lb) (ft-lb)
(oF) (oF) (oF)
Baseline 0 +16 - +25 - +26 - 122 -
F 1.96E+18 -5 -21 +26 +1 +27 +1 120 -2 Base Metal B 5.92E+18 +2 -14 +41 +16 +34 +8 110 -12 (Transverse)
BCC 241 D 9.62E+18 +19 +3 +55 +30 +42 +16 117 -5 A 1.29E+19 +44 +28 +48 +23 +50 +24 118 -4 c 1.88E+19 +47 +31 +77 +52 +64 +38 116 -6 Baseline 0 -100 - -57 - -45 - 124 -
F 1.96E+18 -57 +43 -44 +13 -46 -1 115 -9 HAZ Metal B 5.92E+18 -43 +57 -21 +36 -12 +33 110 -14 BCC 241 D 9.62E+18 +1 +101 +10 +67 +6 +51 117 -7 A 1.29E+19 -66 +34 +16 +73 +19 +64 111 -13 c 1.88E+19 -84 +16 -29 +28 -7 +38 107 -17 Baseline 0 -11 - +65 - +33 - 70 -
F 1.96E+18 +116 +127 +178 +113 +143 +110 65 -5 Weld Metal B 5.92E+18 +114 :+-125 +259 +194 +191 +158 57 -13 WF-182-1 D 9.62E+18 +139 +150 +214 +149 +164 +131 54 -16 A 1.29E+19 +164 +175 +273 +208 +211. +178 62 -8 c 1.88E+19 +151 +162 N/A N/A +194 +161 48 -22 NIA:: Not Applicable, MLE:: Mils Lateral Expansion
- Average cumulative fast fluence (E > 1 MeV)
- 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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page 8-1 8.0
SUMMARY
OF RESULTS 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/cm 2 (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 10 19 n/cm 2 (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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C PageA-1 Appendix A. Reactor Vessel Surveillance Program Background Data and Information Material Selection Data 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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C PageA-2 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 Weight Charpy Impact RTNoT RV Location ID Type TNDT Energy at 10°F 50 ft-lb 35 MLE (oF)
(oF) (ft-lb) Temp. Temp.
USE Cu p s Ni (oF) (oF) (ft-lb) wt%* wt% wt% wt%
ADB 203 SA 508 Cl. 2 Nozzle Belt 50 - 61 - 134 50 0.04 0.007 0.009 -
136, 179, 130 AKJ 233 SA508 Cl. 2 Upper Shell B 20 107, 96, 81 30 - 144 20 0.04 0.004 0.006 -
60,62,47 BCC 241 SA 508 Cl. 2 Lower Shell A 50 47,62,59 27 - 118 50 0.02 0.011 0.011 -
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%)
Middle WF-182-1 Weld Circ. Seam
-20 36,33,44 62 - 81 2 0.18 0.014 0.015 -
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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C PageA-3 Table A-2: Test Specimens for Determining Material Baseline Properties Material Description Number of Test Specimens Tension Tension at CVN Compact ID (Heat No.) Type (Orientation) <al 600°F (bl Fracture (cl at 70°F Impact 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 --
8 (1/2 TCT)
WF-182-1 Weld Metal (L) 3 3 15 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 Number of Tension Number of CVN Impact (Orientation) (a) Material ID (Heat No.)
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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Analysis of Capsule TE1-C PageA-4 Table A-4: Specimens in Lower Surveillance Capsules (Designations B, D, and F)
Number of Number of Number of Yz T Material Type Material ID (Orientation) (a) Tension CVN Impact Compact Fracture (Heat No.) . Specimens (bl Specimens Specimens 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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C PageA-5 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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Analysis of Capsule TE1-C Page 8-1 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 Reduction Specimen Test Temp.
(oF) Strength Strength Elongation Elongation of Area No.
(ksi) (ksi) (%) (%) (%)
SS601 73 75.6 91.9 12.7 27.0 67.3 SS603 73 69.4 90.0 13.1 27.2 67.0 SS604 73 71.9 90.3 13.0 28.8 71.1 Mean 73 72.3 90.7 12.9 27.7 68.5 Std. Dev. 0 3.12 1.02 0.21 0.99 2.29 SS606 580 64.4 86.3 14.4 25.7 65.4 SS611 580 64.4 86.3 13.6 26.0 63.7 SS615 578 63.1 86.3 16.3 25.5 67.0 Mean 580 64.0 86.3 14.8 25.7 65.4 Std. Dev. 1.15 0.75 0 1.39 0.25 1.65 Table B-2: *Pre-Irradiation Tensile Properties for Weld Metal WF-182-1, Transverse Orientation Yield Ultimate Uniform Total Reduction Specimen Test Temp.
(oF) Strength Strength Elongation Elongation of Area No.
(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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Analysis of Capsule TE1-C Page C-1 Appendix C. Pre-Irradiation Charpy Impact Data Table C-1: Pre-Irradiation Charpy Impact Data for Shell Forging Material, BCC 241, Transverse Orientation Absorbed Energy Lateral Expansion Shear Fracture Specimen No. Test Temp. (°F)
(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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page C-2 Table C-2: Pre-Irradiation Charpy Impact Data for Shell Forging Material Heat-Affected Zone, BCC 241, Transverse Orientation Absorbed Energy Lateral Expansion Shear Fracture Specimen No. Test Temp. (°F)
(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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis,of Capsule TE1-C Page C-3 Table C-3: Pre-Irradiation Charpy Impact Data for Weld Metal WF-182-1, Transverse Orientation Absorbed Energy Lateral Expansion Shear Fracture Specimen No. Test Temp. (°F)
(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* (ft-lb)*
Base Metal BCC 241 (Transverse) +16 +25 +26 122 HAZ Metal BCC 241 -100 -57 -45 124 Weld Metal WF-182-1 -11 +65 +33 70 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.
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page D-1 Appendix D. Fluence Analysis Methodology Analytical Methodology 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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page D-2 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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Analysis of Capsule TE1-C Page D-3 Figure D-1: Fluence Analysis Methodology Flow Chart Assembly x Assembly RPD pinxpin I Reactor Geometry Materinls of Fission Spectrum Construction by Fissile Isotope Distribution History I
I BUGLE-93
-..i
+
SORREL code DORT models Cross Section Library
~,
Time-averaged Time-averaged L.j GIPCode r
Radial Source Axial Source S0 (R.6,E)
H Cross sections Dosimetry Counting ...... DORT Analysis R6andRZ
~
and Analysis (NESI) Data to Calculate
,, Absolute Magnitude Results Power History Synthesized (saturation) 3DResults l\*Ieasured Calculated Dosimeter 4 Dosimetry ~
Acti\ities B&\VOG Activities y
C/M :~
I NO Benchmark Analysis +
Bias and Uncertainty
...... Statistical Analysis ~
Validate Bias
...... Validate Uncertainty I t Apply Bias Final Plant .... Removal Function .... YES Validation Specific Fluences "" and Specify "" Acceptable Uncertainty
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Analysis of Capsule TE1-C Page D-4 Synthesized Three Dimensional Results 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
AREVA Inc. ANP-3339 Revision 0 Davis-Besse Unit 1 Reactor Vessel Material Surveillance Program:
Analysis of Capsule TE1-C Page D-5 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 < <Pupdated < 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:
.
prtIVIUUS <rpda1rui ~ <D pnMOus X 1.098995 1.098995
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Analysis of Capsule TE1-C Page E-1 Appendix E. ASTM E185-82 RVSP Technical Report Requirements 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.
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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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Analysis of Capsule TE1-C Page F-1 Appendix F. References
- 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 (BAW-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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Analysis of Capsule TE1-C Page F-2
- 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.