ML20246N384
| ML20246N384 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/30/1989 |
| From: | Aadland J, Lowe A, Nana A BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20246N365 | List: |
| References | |
| BAW-2075, NUDOCS 8905190420 | |
| Download: ML20246N384 (110) | |
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< i l- BAW-2075 April 1989 i
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ANALYSIS OF CAPSULE AN1-C ARKANSAS POWER & LIGHT COMPANY l
! ARKANSAS NUCLEAR ONE, UNIT 1
, -- Reactor Vessel Material Surveillance Program --
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BAW-2075 April 1989 ANALYSIS OF CAPSULE AN1-C ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT 1
-- Reactor Vessel Material Surveillance Program --
by L A. L. Lowe, Jr., PE l J. D. Aadiand A. D. Nana L. Petrusha L
l W. R. Stagg I
l B&W Document No. 77-1174843-00 (See Section 12 for document signatures)
BABC0CK & WILC0X Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935
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SUMMARY
This report describes the results of the examination of the fourth capsule (AN1-C) of the Arkansas Power & Light Company Arkansas Nuclear One, Unit I reactor vessel surveillance program. The objective of the program is to j monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materials by the testing ard evaluation of tension and Charpy impact specimens. The program was designed in accordance with the requirements of 10CFR50, Appendix H, and ASTM Specifi- <
1 cation E185-73. l The capsule received an average fast fluence of 1.46 x 10 19 n/cm2 (E > 1.0 ]
MeV) and the predicted fast fluence for the reactor vessel T/4 location at I the end of the seventh cycle is 1.50 x 1018 n/cm2 (E > 1 MeV). Based on the calculated fast flux at the vessel wall, an 80% load factor, and the planned I fuel management, the projected fast fluence that the Arkansas Nuclear One-Unit I reactor pressure vessel inside surface will receive in 40 calendar I 18 years of operation is 9.75 x 10 n/cm2 (E > 1 MeV).
The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure. The Charpy impact data results exhibited the characteristic shift to higher temperature for the 30 j ft-lb transition temperature and a decrease in upper-shelf energy. These results demonstrated that the current techniques used for predicting the change in both the increase in the RT and the decrease in upper-shelf NDT properties due to irradiation are conservative. The recommended operating period was extended to 32 effective full power years as a result of the fourth capsule evaluation. These new operating limitations are in accordance with the requirements of Appendix G of 10CFR50. A low upper-shelf fracture analysis in accordance with 10CFR50, Appendix G, demonstrated that the low upper-shelf weld metals will not restrict normal plant operations for at least 32 EFPY.
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CONTENTS Page
- 1. INTRODUCTION . . . . . ... . . . . . . . . . . . . . . . . . . . 1-1
- 2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . . . . . 3-1
- 4. PRE-IRRADIATION TESTS . . . . . . . . . . . . . . . . . ... . . . 4-1 4.1. Tension Tests . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2. Impact Tests . . . . . . . . . . . . . . . . . . . . . . . 4-1
- 5. POST-IRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . 5-1 l
l 5.1. Thermal Monitors . . . . . . . . . . . . . . . . . . . . . 5-1 5.2. Tension Test Results . . . . . . . . . . . . . . . . . . . . 5-1 5.3. Charpy V-Notch Impact Test Results . . . . . . . . . . .'._. 5-2
- 6. NEUTRON FLUENCE , . . . . . . . . . . ... ... . . . . . . . . . . 6-1 6.1. Introduction . . . . . . . . . . . . . . . . . . . . . . 6-1 l 6.2. Vessel Fluence . . . . . . . . . . . . . . . . . . . . . . 6-4
'6.3. Capsul e Fl uence . . . . . . . . . . . . . . . . . . . . . . 6-5
, 6.4. Fluence Uncertainties . . . . . . . . . . . . . . . . . . . 6-5 l
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- 7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . . . . . . 7-1 7.1. Pre-Irradiation Property Data . . . . . . . . . . . . . . . 7-1 7.2. Irradiated Property Data . . . . . . . . . . . . . . . . . 7-1 7.2.1. Tensile Properties . . . . . . . . . . . . . . . . 7-1 7.2.2. Impact Properties . . . . . . . . . . . . . . . . . 7-2 7.3. Reactor Vessel Fracture Toughness . . . . . . . . . . . . . 7-4
- 8. DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY PRESSURE -
l TEMPERATURE LIMITS . . . . . . . . . . . . . . . . . . . . . . . 8-1 L
- 9. LOW UPPER-SHELF FRACTURE TOUGHNESS ANALYSIS . . . . . . . . . . . 9-1 9.1. Background . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2. Charpy Upper-Shelf Impact Energy . . . . . . . . . . . . . 9-2 9.3. Material Fracture Toughness Properties . . . . . . . . . . 9-3 9.4. Analytical Method and Acceptance Criteria . . . . . . . . . 9-3 9.5. Fracture Analysis . . . . . . . . . . . . . . . . . . . . . 9-4 9.6. Summary . . . . . . . . . . . . . . . . . . . . . . . . . . 9-4 l
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4 Contents (Cont'd) i Page-
- 10.
SUMMARY
OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . 10-1.
11.- SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . . . . . . 11-1 -!
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' 12 . ' C ERT I F I C AT NN . . . . . . . . . . . . . . . . . . . . . . . . . . 12 - 1 APPENDIXES I A. Reactor Vessel Surveillance Program Background Data and Information . . ... . . . . . . . . . . . . . . . .. . . . A-1 ;
B. Pre-Irradiation Tensile Data . . . . . . . . . . . . . . . . . . . B-1 i C. Pre-Irradiation Charpy Impact Data . . . . . . . . . . . . . . . . C-1 D. Fluence Anklysis Methodology . . . . . . . . . . . . . . . . . . . D-1 ,
E. Capsule Dosimetry Data . . . . . . . . . . . . . . . . . . . . . . E-1 i F. References . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 1
I list of Tables l
Table l 3-1. Specimens in Surveillance' Capsule AN1-C . . . . . . . . . . . . 3-2 1 3-2. Chemical Composition and Heat Treatment of J
Surveill ance Materi al s . . . . . . . . . . . . . . . . . . . . . 3-3 5-1. Tensile Properties of Cygsule AN1-C Base Metal and Weld Metal Irradiated to 1.46 x 10 n/cm (E > 1 MeV) . . . . . . . . . . 5-3 .l 5-2. Charpy Impact Data From Capsule AN1-C, Dase 1 Metal -(C5114-1), Longitginal orientation, Irradiated to 1.46 x 10 n/cm (E > 1 MeV) . . . . . . . . . . 5-3 5-3. Charpy Impact Data From Capsule AN1-C, Base 2 Metal (C5114-1),Transv9gseOrjentation, Irradiated to 1.46 x 10 n/cm (E > 1 MeV) . . . . . . . . . . 5-4 ]
5-4. Charpy Impact Data From Capsul 9gAN1-C Heat-Affected Zone j Metal, Irradiated to 1.46 x 10 n/cm2 (E > 1 MeV) . . . . . . . 5-4 5-5. Charpy . Impact Data From3 gapsulg AN1-C, Weld Metal WF-193 Irradiated to 1.46 x 10 n/cm (E > 1 MeV) . . . . . . . . . . 5-5 5-6. Charpy Impact Data From Capsule AN1-C Correlation MonitorMategal,HgatNo.A-1195-1, to 1.46 x 10 n/cm (E > 1 MeV) . . . . . .
Irradiated
.......... 5-5 6-1. Surveillance Capsule Dosimeters . . . . . . . . . . . . . . . . 6-7 i
- 2. Arkansas Nuclear One Unit 1 Reactor Vessel Fast Flux . . . . . . 6-7 6-3. Calculated Arkansas Nuclear One Unit 1 Reactor Vessel Fluence . . 6-8 6-4. . Surveillance Capsule AN1-0 Fluence, Flux, and DPA . . . . . . . 6-9 6-5. Estimated Fluence Uncertainty . . . . . . . . . . . . . . . . . 6-9 7-1. Comparison of Capsule AN1-C Tension Test Results . . . . . . . . 7-7 7-2. Summary of ANO Reactor Vessel Surveillance Capsule Tensile Te s t Re s ul t s . . . . . . . . . . . . . . . . . . . . . . . . . . 7 -8
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Tables (Cont'd)
Table Page 7-3. Observed Vs. Predicted Changes for Cfgsule AN1-C Irradiated Charpy Impact Properties - 1.46.x 10 n/cm (E > 1 MeV) . . . . 7-9 ,
7-4. Summary of ANO-1 Reactor Vessel Surveillance Capsules Charpy '
Impact Test Results . . . . . . . . . . . . . . . . . . . . . . 7-10 7-5. Evaluation of Reactor Vessel End-of-Life Fracture Toughness and Pressurized Thermal Shock Criterion -
Arkansas Nuclear One, Unit 1. . . . . . . . . . . . . . . . . . 7-11 7-6. Evaluation of Reactor Vessel End-of-Life Upper Shelf Energy -
Arkansas Nuclear One, Unit 1 . . . . . . . . . . . . . . . . . . 7-12 8-1. Data for Preparation of Pressure-Temperature Limit Curves for Arkansas Nuclear One, Unit 1 -- Applicable Through 32 EFPY . . . 8-5 9-1. Input Data for Deformation Plasticity Failure Assessment Diagram . . . . . . . . . . . . . . . . . . . . . . . 9-5 9-2. Failure Assessment Data Points . . . . . . . . . . . . . . . . . 9-6 A-1. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials -- ANO-1 . . . . . . . . . . . . . A-3 A-2. Test Specimens for Determining Material Baseline Properties . . . A-4 A-3. Specimens in Surveillance Capsules (Designations A, C, and E) . . A-5 l A-4. Specimens in Surveillance Capsules (Designations B, D, and F) . . A-5 I
B-1. Tensile Properties of Unirradiated Shell Plate Material, Heat No. C5114-1 . . . . . . . . . . . . . . . . . . . . . . . . . B-2 B-2. Tensile Properties of Unirradiated Weld Metal, WF-193 . . . . . . B-2 l C-1. Charpy Impact Data From Unirradiated Base Material, l Longitudinal Orientation, Heat No. C5114-1. . . . . . . . . . . . C-2 C-2. Charpy Impact Data From Unirradiated Base Material, Transverse Orientation, Heat No. C3265-1 . . . . . . . . . . . . . C-3 C-3. Charpy Impact Data From Unirradiated Base Metal, HAZ, Longitudinal Orientation, Heat No. C5114-1 . . . . . . . . . C-4 C-4. Charpy Impact Data From Unirradiated Weld Metal, WF-193 . . . . . C-5 D-1. Flux Normalization Factor ....................D7 D-2. Arkansas Unit 1 Reactor Vessel Fluence by Cycle . . . . . . . . . D-8 E-1. Detector Composition and Shielding . . . . . . . . . . . . . . . . E-2 E-2. Measured Specific Activities (Unadjusted) for Dosimeters in Capsule AN1-C . . . . . . . . . . . . . . . . . . . . . . . . . . E-2 E-3. Dosimeter Activation Cross Sections, b/ atom . . . . . . . . . . . E-3 1
1 List of Fiaures Figure 3-1. Reactor Vessel Cross Section Showing Location of Capsule AN1-C in Arkansas Nuclear One, Unit 1 . . . . . . . . . . . . . . . . 3-4 3-2. Reactor Vessel Cross Section Showing Location of Arkansas One, i
Unit 1 Capsule AN1-C in Davis-Besse Unit 1. . . . . . . . . . . 3-5 3-3. Loading Diagram for Test Specimens in Capsule AN1-C . . . . . . 3-6 l
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Fioures (Cont'd)
Figure Page 5-1. Charpy Impact Data for Irradiated Plate Material, Longitudinal Orientation, Heat No. C5114-1 ... ..... . . . . . . . . . 5-6 5-2. Charpy Impact Data for Irradiated Plate Material, Transverse Orientation, Heat No. C5114-1 . .... ...... . . . . . . 5-7 5-3. Charpy Impact Data for Irradiated Plate Material, Heat-Affected Zone, Heat No. C5114-1 . . . . . . . . . . . . . . . . . . . . . 5-8 5-4. Charpy Impact Data for Irradiated Weld Metal, WF-193 . . . . . . 5-9 5-5. Charpy Impact Data for Irradiated Correlation Material, HSST PL-02, Heat No. A1195-1 . . . . . . . . . . . . . . . . . . 5-10 6-1. General Fluence Determination Methodology . . . . . . . . . . . 6-2 6-2. Fast Flux, Fluence and DPA Distribution Through Reactor Vessel Wall . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 6-3. Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 8-1. Predicted Fast Neutron Fluence at Various Locations Through Reactor Vessel Wall for 32 EFPY - Arkansas Nuclear One, Unit 1. 8-6 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Heatup, Applicable for First 32 EFPY -
Arkansas Nuclear One, Unit 1. . . . . . . . . . . . . . . . . . 8-7 8-3.
Reactor Vessel Pressure-Temp'erature Limit Curves for Normal l Operation - Cooldown, Applicable for First 32 EFPY -
Arkansas Nuclear One, Unit 1 . . . . . . . . . . . . . . . . . . 8-8 8-4. Reactor Vessel Pressure-Temperature Limit Curves for Inservice Leak and Hydrostatic Tests, Applicable for First 32 EFPY -
Arkansas Nuclear One, Unit 1. . . . . . . . . . . . . . . . . . 8-9 9-1. Failure Assessment Diagram for Arkansas Nuclear One, Unit-1 Reactor Vessel Based on Weld Metal WF-112 . . . . . . . . . . . 9-7 9-2. Safety Factors Vs. Crack Extension for Arkansas Nuclear One, Unit-1 Reactor Vessel ..................... 07 A-1. Location and Identification of Materials Used in the Fabrication of ANO-1 Reactor Pressure Vessel . . . . . . . . . . . A-6 A-2. Location of Longitudinal Welds in the Upper and Lower Shell Courses . . . . . . . . . . . . . . . . . . . . . . . . . . A-7 C-1. Charpy Impact Data From Unirradiated Base Metal, Longitudinal Orientation . . . . . . . . . . . . . . . . . . . . . C-6 C-2. Charpy Impact Data From Unirradiated Base Metal, Transverse Orientation . . . . . . . . . . . . . . . . . . . . . . C-7 C-3. Charpy Impact Data From Unirradiated Base Metal, HAZ, Longitudinal Orientation . . . . . . . . . . . . . . . . . . . . . C-8 C-4. Charpy Impact Data From Unirradiated Weld Metal . . . . . . . . . C-9 D-1. Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Capsule . .... .... . . . . . . . . . . D-9 D-2. Rationale for the Calculation of Neutron Flux in the Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . D-10 D-3. Plan View Through Reactor Core Midplane (Reference R-0 Cal cul ati on Model ) . . . . . . . . . . . . . . . . . . . . . . . . D-Il
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- 1. INTRODUCTION This report describes the results of the examination of the fourth capsule (AN1-C) of the Arkansas Pewer & Light Company's Arkansas Nuclear One, Unit 1 (ANO-1) reactor vessel material surveillance program (RVSP). The capsule was removed and evaluated after being irradiated in the Toledo Edison Company's Davis-Besse Unit I reactor as part of the Integrated Reactor Vessel Materials Surveillance Program (BAW-1543A).1 The irradiation exposure in Davis-Besse Unit 1 plus the previous irradiation in Arkansas Nuclear One, Unit 'l is the equivalent of 16.8 EFPY or approximately 22 calendar years of exposure in the Arkansas Nuclear One, Unit I reactor vessel. The capsule experienced a I9 fluence of 1.46 x 10 n/cm2 (E > 1 MeV), which is the equivalent of approxi-mately 48 effective full power years' (EFPY) operation of the Arkansas
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Nuclear One, Unit I reactor vessel. The first capsule (AN1-E) from this I
program was removed and examined after the first year of operation; the
, results are reported in BAW-1440.2 The second capsule (AN1-B) was removed L and examined after irradiation in Toledo Edison Company's Davis-Besse Unit 1 as part of the Integrated Reactor Vessel Materials Surveillance Program; the l results are reported in BAW-1698.3 The third capsule (AN1-A) of the program was removed and evaluated as part of the Integrated Reactor Vessel Materials l Surveillance Program and the results reported in BAW-1836.4 The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillance program for Arkansas Nuclear One Unit-1 was designed and furnished by Babcock & Wilcox (B&W) as described in BAW-10006A5 and conducted in accordance with BAW-1543A.I The program was planned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel.
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The surveillance program for Arkansas Nuclear One, Unit I was designed in accordance with E185-666 and thus is not in compliance with 10CFR50, Appen-dixes G7 and H8 since the requirements did not exist at the time the program l was designed, Because of the difference, additional tests and evaluations were required to ensure meeting the requirements of 10CFR50, Appendixes G and I H. The recommendations for the future operation of Arkansas Nuclear One, Unit-1 included in this report do comply with these requirements. l l
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l l 2. BACKGROUND 1
The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water-cooled reactors. The beltline region of the reactor vessel is the most critical 4 region of the vessel because it is expo;ed to neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of such low-alloy ferritic steels as SA533, Grade B used in the fabrication of the AN0-1 reactor vessel, are well characterized and documented in the literature. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition ftom brittle to ductile fracture accompanied by a reduction in the Charpy upper shelf energy value.
Appendix G to 10CFR50, " Fracture Toughness Requirements,"7 specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including antici-pated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 13, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date_
Appendix H to 10CFR50, " Reactor Vessel Materials Surveillance Program Requirements,"8 defines the material surveillance program required to monitor 2-1 Babcock &Wifcom a McDermott company
I changes in the fracture toughness properties of ferritic materials in the reactor vessel . beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the ther. mal environment. Fracture toughness test data are obtained from material specimens withdrawn periodi-cally from the reactor vessel. These data will permit determination of the i conditions under which the vessel can be operated with adequate safety.
margins against fracture throughout its service life.
A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code, Section l III, " Nuclear Power Pl ant Components."9 This method utilizes fracture I 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 E-208) or the temperature that is 60F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The I RT f a given material is used to index that material to a reference NDT.
stress intensity factor curve (K IR curve), which appears in Appendix G of ASME Section III. The K IR curve is a lower bound of dynamic, static, and crack arrest fracture 'oughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K curve, IR allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined j using these allowable stress intensity factors.
The RT NDT and, in turn, the operating limits of a nuclear power plant, can be adjusted to account for the effects of radiation on the properties of the reactor vessel 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 contain-ing prepared specimens of the reactor vessel 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 to adjust it for radiation embrittlement. This. adjusted RT is used NDT NDT to index the material to the K curve which, in turn, is used to set IR operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.
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Appendix G, 10CFR50, also requires a minimum Charpy V-notch upp tr-shelf energy of 75 ft-lbs for all beltline region materials unless it is demon-strated .that lower values of upper-shelf fracture energy will provide an adequate margin for deterioration as the result of neutron radiation. No action is required for a material that does not meet the 75 ft-lb requirement provided the irradiation deterioration does not cause the upper-shelf energy to drop below 50 ft-lbs. The regulations specify that if the upper-shelf energy drops below 50 ft-lbs it must be demonstrated in a manner approved by the Office of Nuclear Regulation that the lower values will provide adequate margins of safety.
When a reactor vessel fails to meet the 50 ft-lb requirement, a program must be submitted for review and approval at least three years prior to the time the predicted fracture toughness will no longer satisfy the regulatory requirements. The program must address the following:
A. A volumetric examination of 100 percent of the beltline materials that do not meet the requirement.
B. Supplemental fracture toughness data as evidence of the fracture -
toughness of the irradiated beltline materials.
C. Fracture toughness analysis to demonstrate the existence of equiva-lent margins of safety for continued operation.
If these procedures do not indicate the existence of an adequate margin of safety, the reactor vessel beltline may be given a thermal annealing treat-ment to recover the fracture toughness properties of the materials.
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- 3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for ANO-1 comprises six surveillance capsules designed to monitor the effects of aeutron and thermal environment on the materials of the reactor pressure vessel core 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. BAW-10006A5 includes a full description of the capsule locations and design. After the capsules were removed ' from ANO-1 in 1976 and included in the integrated RVSP, they were scheduled and irradiated in the Davis-Besse Unit I reactor as described in BAW-1543AI . During this period of irradiation, capsule AN1-C was irradiated in the bottom position in holder tube YZ as shown in Figure 3-2.
Capsule AN1-C was removed during the fifth refueling shutdown of Davis-Besse Unit 1. This capsule contained Charpy V-notch impact test specimens fabricated from one base metal (SA533, Grade B), one heat-affected-zone, a weld metal and correlation material. Tension test specimens were fabricated from the base metal and the weld metal only. The specimens contained in the capsule are described in Table 3-1, and the location of the individual l
specimens within the capsule are described in Figure 3-2. The chemical composition and heat treatment of the surveillance material in capsule AN1-C are described in Table 3-2.
All test specimens were machined from the 1/4-thickness (1/4T) location of the plate material. Charpy V-notch and tension test specimens were cut from the surveillance material such that they were oriented with their longitudi-nal axes either parallel or perpendicular to the principal working direction.
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Capsule AN1-C' contained dosimeter wires, described as follows: I Dosimeter Wire Shieldina U-Al alloy Cd-Ag alloy Np-Al alloy Cd-Ag alloy Nickel Cd-Ag alloy 0.66 wt % Co-Al alloy Cd-Ag allo.v 0.66 wt % Co-Al alloy None Fe None Thermal monitors of low-melting metals and alloys were . included in the l capsule. The metals and alloys and their melting points are as follows:
Allov Meltino Point. F 90% Pb, 5% Ag, 5% Sn 558 97.5% Pb, 2.5% Ag 580 l 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Cadmium 610
. Lead 621 1
Table 3-1. Specimens in Surveillance Caosule AN1-C Number of Test Soecimens Material Description Tension CVN(a) Impact Weld Metal, WF-193 4 8 HAZ i Heat No. C5114-1, Longit'udinal 0 8 Base Material Plate Heat No. C5114-1, Longitudinal 4 8 Transverse 0 4 Correlation Material, HSST Plate 02 0 _B Total Per Capsule 8 36 (a)CVN denotes charpy V-notch.
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Table 3-2. Chemical Composition and Heat Treatment of Surveillance Materials Chemical Composition, w/o Heat N HSST WeldMg1 Element C5114-1g) Plate 02 Id) WF-193 C 0.21 0.23 0.09 Mn 1.32 1.39 1.49 I
P 0.010 0.013 0.016 S 0.016 0.013 0.016 Si 0.20 0.21 0.51
) Ni 0.52 0.64 0.59 j Cr 0.19 --
0.06 t Mo 0.57 0.50 0.39
) Cu 0.15 0.17 0.28 l
Heat Treatment Heat No. Temo. F Time h Coolina C5114-1 1550-1600 4.5 Brine quench 1200-1225 5.0 Brine quench 1100-1150 29.0 Furnace cooled HSST PL-02 1600+75 4 Water quenched Normalized 1225+25 4 Furnace cooled at 1675F 75 1125+25 40 Furnace cooled WF-193(c) 1100-1150 29 Furnace cooled
$8)Per Certified Materials Test Report (b)Per Licensing Document BAW-1500P10 (c)Per Licensing Document BAW-1820Il (d)Per ORNL-446312 3-3 Babcock & Wilcox a McDermott cornpany
i Figure 3-1. Reactor Vessel Cross Section Showing Location of Caosule AN1-C in Arkansas Nuclear One, Unit 1
- Surveillance Capsule Holder Tube -- Capsules AN1-C, AN1-D
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- 4. 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 Appendixes G and H to 10CFR50.
I 4.1. Tension Tests Tension test specimens were fabricated from the reactor vessel 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 I 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 accor-dance with the applicable requirements of ASTM A370-77.13 For each material type and/or condition, six specimens in groups of three were tested at both room temperature and 580F. 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 preirradiation tensile specimens are given in Appendix B.
4.2. Impact Tests Charpy V-notch impact tests were conducted in accordance with the require-ments of ASTM Standard Methods A370-77 13 and E23-8214 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 to +550F. Speci-mens were removed from the baths and positioned in the test frame anvil with tongs specifically designed for the purpose. The pendulum was released I 4-1 I Babcock &Wilcox a McDermott company
manually, allowing the specimens to be broken within 5 seconds from their l
l removal from the temperature baths.
Impact test data for the unirradiated baseline reference materials are <
presented in Appendix C. Tables C-1 through C-5 contain the basis data that are plotted in Figures C-1 through C-5. <
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f Figure 3-2. Reactor Vessel Cross Sectiori Showing Location of Arkansas One Unit 1 Caosule AN1-C in Davis-Besse Unit 1 1
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Table 5-1. Tensile Properties of Capsule AN1-ggBase Metal and Weld Metal Irradiated to 146 x 10 n/cm (E > 1 MeV) k Strenath, osi Elonaation, % Red'n.
Specimen Test Temp, in Area, No. F Yield. Ultimate Uniform Total %
l' Base Metal C5114-1. Longitudinal GG703 70 79,600 99,100 10.6 26.3 63.7 GG709 300 71,300 91,000 9.8 23.4 62.7 I GG701* 400 --
93,400 -- --
61.3 GG702 580 69,200 93,100 9.6 22.9 58.5
)
( Weld Metal . WF-193 GG118 70 83,700 98,900 12.7 25.7 57.0 GG113 300 75,600 89,400 9.9 19.8 50.6 GG115 400 74,400 89,400 9.3 20.4 51.7 GG111 580 73,200 90,700 9.1 15.7 41.5
- Extensometer slipped during test.
Table 5-2. Charpy Impact Data From Capsule AN1-C, Base
' Metal (C5114-1), Longitginal orientation, Irradiated to 1.46 x 10 n/cm (E > 1 MeV)
Test Impact lateral Shear Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch %
j GG728 0 15.0 0.011 0 GG729 20 23.0 0.025 10 GG712 70 52.5 0.048 60 GG726 125 72.0 0.055 50 GG725 175 83.0 0.066 90 GG719 275 106.0 0.079 100 GG716 375 117.0 0.083 100 GG738 550 104.0 0.080 100 5-3 Babcock &WHcom a McDermott company
Table 5-3. Charpy Impact Data From Capsule AN1-C, Base Metal (C5114-1), Tgnsverge Orientation, Irradiated to 1.46 x 10 n/cm (E > 1 MeV) k Test Impact Lateral Shear Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch %
GG602 70 34.0 0.029 30 GG606 150 51.5 0.045 40 GG615 250 72.0 0.067 100 GG623 350 79.5 0.070 100 i
Table 5-4. Charpy A Impact Data From Capsule AN1 f Heatg ffected Zone Metal, Irradiated to 1.46 x 10 n/cm (E > 1 MeV) j i
Test Impact lateral Shear Specimen Temperature Energy Expansion Fracture 4 ID F ft-lbs Inch % f GG407 -60 18.5 0.015 30 <
GG424 0 27.0 0.025 30 GG442 40 39.0 0.025 60 GG409 70 68.0 0.050 40 GG416 150 66.5 0.061 100 GG413 225 65.0 0.051 100 GG406 300 60.0 0.056 100 GG415 375 71.5 0.061 100 ;
5-4 Babcock & Wilcox a McDermott company
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i 7:
- 5. POST-IRRADIATION TESTS 5.1. Thermal Monitors Capsule AN1-C contained three temperature monitor holder tubes, each
- containing five fusible alloy wires with melting points ranging from 558 to 621F. All the thermal monitors at 558, 580, 588 and 610F locations had f melted while the monitor at the 621F location showed no signs of melting in all three holder tubes. From these observations, it was concluded that the j capsule had been exposed to a peak temperature in the range of 610 to 621F
) during the reactor operating period. These peak temperatures are attributed
, to operating transients that are of short durations and are judged to have l insignificant effect on irradiation damage. Short duration operating transients cause the use of thermal monitor wires to be of limited value in determining the maximum steady state operating temperature of the surveil-lance capsules; however, it is judged that the maximum steady state operating temperature of specimens in the capsule was held within 25F of the 1/4T vessel thickness location temperature of 559 to 577F. It is concluded that the capsule design temperature may have been exceeded during operating transients but not for times and temperatures that would make the capsule data unusable.
) 5.2. Tension Test Results The results of the postirradiation tension tests are presented in Table 5-1.
Tests were performed on specimens at both room temperature and in the temper-ature range of 300 to 580F using the same test procedures and techniques used to test the unirradiated specimens (Section 4.1). In general, the ultimate strength and yield strength of the material increased with a corresponding slight decrease in ductility as compared to the unirradiated values; both effects were the result of neutron radiation damage. The type of behavior observed and the degree to which the material properties changed is within 5-1 Babcock &Wilcom a McDermott company
., o l 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 B. i 5.3. Charov V-Notch Imoact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 5-2 through 5-6 and Figures 5-1 through 5-5. The test procedures and techniques were the same as those i used to test the unirradiated specimens (Section 4.2). 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.
The results of the pre-irradiation Charpy V-notch impact tests are given in Appendix C. '
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Table 5-5. Charpy Impact Data From CapsulygAN1-C W WF-193 Irradiated to 1.46 x 10 n/cm g (eld E > Metal 1 MeV)
Test Impact lateral Shear Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch %
GG022 150 23.0 0.023 30 GG014 190 29.5 0.030 60 GG038 225 40.0 0.039 70 GG025 300 45.0 0.043 90 GG018 340 44.0 0.043 100
> GG036 375 53.0 0.051 100
, GG019 450 46.0 0.047 100 GG002 550 40.0 0.058 100 i
Table 5-6. Charpy Impact Data From Capsule AN1-C Correlation MonitorMate{ gal,HgatNo.A-1195-1, Irradiated l to 1.46 x 10 n/cm (E > 1 MeV)
Test Impact lateral Shear Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch %
GG946 70 12.5 0.013 10 GG920 100 22.5 0.018 20
) GG915 150 46.0 0.035 40 GG937 225 70.0 0.061 90 GG903 300 98.0 0.075 100 GG941 375 102.5 0.081 100 GG948 450 96.5 0.078 100 4 i GG927 550 93.5 0.082 100 5-5 Babcock & WHcox a McDermott company
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Figure 5-1. Charpy Ir. pact Data for Irradiated Plate Material, LonaituJinal Orientation. Heat No. C5114-1 100 , , ;, ;
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Fiaure 5-4. Charov Imoact Data for Irradiated Weld Metal. WF-193
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Figure 5-5. Charpy Impact Data for Irradiated Correlation Material. HSST PL-02. Heat No. A1195-1
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i Figure 5-2. Charpy Impact Data for Irradiated Plate Material, Transverse Orientation. Heat No. C5114-1 100 , , , .
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+106r Tgy (35 Met) -
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_2-___-____ . _ . _ . _ _ _ _ _ _ _ . _ _ _ - _ _
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l Figure 5-3. Charpy Impact Data for Irradiated Plate Material, Heat-Affected Zone Heat No, C5114-1 100 i . . - i
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-100 0 100 200 300 400 500 600 Test Temperature, F 5-8 Babcock &Wilcox a McDermott company
)
- 6. NEUTRON FLUENCE 6.1. Introduction l
The neutron fluence (time integral of flux) is a quantative way of expressing the cumulative exposure of a material to a pervading neutron flux over h specific period of time. Fast neutron fluence, defined as the fluence of neutrons having energies greater than 1 MeV, is the parameter that is pre-I sently used to correlate radiation induced changes in material properties.
Accordingly, the fast fluence must be determined at two locations: (1) in
[ the test specimens located in the surveillance capsule, and (2) in the wall i
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 fluence in the reactor vessel, the relative radial and azimuthal distribution of the fluence, the f fluence gradient through the reactor vessel wall, and .the corresponding material properties.
The accurate determination of neutron flux is best accomplished through the simultaneous consideration of neutron dosimeter measurements and analytically derived flux spectra. Dosimeter measurements alone cannot be used to predict the fast fluence in the vessel wall or in the test specimens because (1) they cannot measure the fluence at the points of interest, and (2) they provide only rudimentary information about the neutron energy spectrum. Conversely, reliance on calculations alone to predict fast fluence is not prudent because of the length and complexity of the analytical procedures involved. In short, measurements and calculations are necessary complements of each other and together they provide assurance of accurate results.
Therefore, the determination of the fluence is accomplished using a combined analytical-empirical methodology which is outlined in Figure 6-1 and describ-ed in the following paragraphs. The details of the procedures and methods j are presented in general terms in Appendix D and in BAW-1485P.15 l 6-1 Babcock &WHcom a McDermott company j
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I Fioure 6-1. General Fluence Determination Methodoloav MEASUREEMTS T Namm ANM.MCM. DETEMNTION T DOSIETER ACTIVITIES DOSDETER ACTIVITIES APO NEUTRON FLUX
= USTED E = v .
DEPDOENT NEUTRON 5 FL11X .
I.
I REACER TERAUE NEU1RON HISTORY N O PRE-FLUENCE DICTED FURRE '
OPERAHON I
Analytical Determination of Dosimeter Activities and Neutron Flux The analytical calculation of the space and energy dependent neutron flux in I the test specimens and in the reactor vessel is performed with the two 4
dimensional discrete ordinates transport code, DOTIV.16 The calculations employ an angular quadrature of 48 sectors (S8), a third order LeGendre polynomial scattering approximation (P3), the CASK 23E cross section set I7 with 22 neutron energy groups and a fixed distributed source corresponding to the time weighted average power distribution for the applicable irradiation g period.
In addition to the flux in the test specimens, the DOTIV calculation deter- g mines the saturated specific activity of the various neutron dosimeters located in the surveillance capsule using the ENDF/B5 dosimeter reaction cross sections.18 The saturated activity of each dosimeter is then adjusted by a factor which corrects for the fraction of saturation attained during the i
6-2 ~
Babcock &Wilcox a McDermott company t ,
dosimeter.'s actual (finite) irradiation history. Additional corrections are made to account for the following effects:
) e Photon induced fissions in V and Np dosimeters (without this correc-tion the results underestimate the measured activity).
e Fissile impurities in U dosimeters (without this correction the results underestimate the measured activity).
e Short half-life of isotopes produced in iron and nickel dosimeters (303-day Mn-54 and 71-day' Co-58, respectively). (Without this correction, the results could be biased high or low depending on the long term versus short term power histories.)
Measurement of Neutror Dosimeter Activities E The accuracy of neutron fluence predictions is improved .if the calculated neutron flux is compared with neutron dosimoter measurements adjusted for the
- effects noted above. The neutron dosimeters located in the . surveillance I- capsules are listed in Table 6-1. Both activation type and fission type dosimeters were used.
The ratio of measured dosimeter activity to calculated dosimeter activity 7
(H/C) is determined .for each dosimeter, as discussed in Appendix D. These i M/C ratios are evaluated on a case-by-case basis to assess the dependability or veracity of each individual dosimeter resMnse. After carefully evaluat-ing a'11 factors known to affect the calculations or the measurements, an average M/C ratio is calculated and defined as the " normalization factor."
! The normalization factor is applied as an adjustment factor to the D0T-calculated flux at all points of interest.
Neutron Fluence The determination of the neutron fluence from the time averaged flux requires
! only a simple multiplication by the time in EFPS (effective full-power seconds) over which the flux was averaged,1.e.
f jj (AT) = E p $3g AT 9
where 2
fjj (AT) - Fluence at (i,j) accumulated over time AT (n/cm ),
g - Energy group index, 6-3 Babcock &WIIcom a McDermott company
___m___i_______________.____._______..
2
$3 g.= Time-average flux at (i,j) in energy group g, (n/cm -sec),
AT = Irradiation time, EFPS. 4 1
Neutron fluence was calculated in this analysis for the following components over the indicated operating time:
Test Specimens: Capsule irradiation time in EFPS Reactor Vessel: Vessel irradiation time in EFPS
~
Reactor Vessel: Maximum point on inside surface extrapolated to 32 effective full power years The neutron exposure to the reactor vessel and the material surveillance specimens was also determined in terms of the iron atom displacements per j atom of iron (DPA). The iron DPA is an exposure index giving the fraction of iron atoms in an iron specimen which would be displaced during an irradia-tion. It is considered to be an' appropriate damage exposure index since
)
displacements of atoms from their normal lattice sites is a primary source of ,
neutron radiation damage. DPA was calculated based on the ASTM Standard E693-79 (reapproved 1985).I9 A DPA cross section for iron is given in the 1 ASTM Standard in 641 energy groups. DPA per second is determined by multi- l plying the cross section at a given energy by the neutron flux at that energy and integrating over energy. DPA is then the integral of DPA per second over the time of the irradiation. In the DPA calculations reported herein, the ASTM DPA cross sections were first collapsed to the 22 neutron group struc- I ture of CASK-23E; the DPA was then determined by summing the group flux times the DPA cross section over the 22 energy groups and multiplying by the time ;
of the irradiation. !
6.2. Vessel Fluence j The maximum fluence (E > 1 MeV) exposure of the AN01 reactor vessel during Cycles 6-7, was determined to be 7.3 x 10 I7 n/cm2 based on a maximum neutron flux of 1.08 x 10 10 n/cm2 -s (Tables 6-2 and 6-3). The maximum fluence occurs at the cladding / vessel interface at an azimuthal location of approximately 11 1 degrees from a major horizontal axis of the core.
Fluence data were extrapolated to 32 EFPY of operation based on two assump- <
tions: (1) the future fuel cycle operations do not differ significantly from 6-4 Bat > cock &Wilcom 1 a McDermott company
their current, designs, and (2) the latest calculated (or extrapolated) flux remains constant from that time through 32 EFPY. The extrapolation was carried out in two stages, (1) from E0C7 to EOC9, and (2) from E0C9 to 32 EFPY. In the first stage, cycle averaged fluxes are calculated using D0TIV, the current design data for cycles 8 and 9, and DOT adjoint factors for assembly-averaged power distributions. In the second stage, the 32 EFPY fluence was calculated by assuming a constant flux over the period which was equal to the average flux for cycles 8 and 9.
Relative fluence and DPA (displacement per atom) as a function of radial
, location in the reactor vessel wall is shown in Figure 6-2. Reactor vessel neutron fluence lead factors, which are the ratio of the neutron flux at the clad interface to that in the vessel wall at the T/4, T/2 and 3T/4 locations, are 1.79, 3.53, and 7.25, respectively. DPA lead factors at the same I locations are 1.59, 2.64, and 4.56, respectively. The relative fluence as a I
function of azimuthal angle is shown in Figure 6-3. A peak occurs in the fast flux (E > 1 MeV) at about 11 degrees with a corresponding value of 1.08 I x 1010 n/cm2 -s.
t 6.3. Caosule Fluence l
The capsule was irradiated for 1609.5 EFPD in the bottom holder tube position
! of Davis-Besse 1 located 11 degrees off the major horizontal axis at about 202 cm from the vertical axis of the core. The capsule was also irradiated for 345 EFPD in Arkansas Unit 1, during cycle 1, located at the 11 degree position about 211 cm from the vertical axis of the core. The cumulative I fast fluence at the center of the surveillance capsule was calculated to be 1.46 x 10 19 n/cm2 of which 5.0% was accumulated during the Arkansas I cycle 1 irradiation, and 95% was accumulated during the Davis-Besse 1 cycles 1-5 l irradiation (Table 6-4). This fluence value represent an average value for the various locations in the capsule.
6.4 Fluence Uncertainties Uncertainties were estimated for the fluence values reported herein. The results are shown in Table 6-5 and are based on comparisons to benchmark experiments, when available; estimated and measured variations in input data; and on engineering judgement. The values in Table 6-5 represent best 6-5 7 Babcock &Wilcox 3 MCDermott company 4
(
estimate values bhsed on past experience with reactor vessel fluence ana-lyses.
t 1
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Babcock &Wilcox a McDermott company
Table 6-1. Surveillance Caosule Dosimeters Lower Energy Limit for Isotope Dosimeter Reactions I3) Reaction, MeV Half-Life 54Fe(n,p)54Mn 2.5 312.5 days-58Ni(n,p)58Co 2.3 70.85 days 238U (n,f)137Cs- 1.1 30.03 years 237Np(n,f)137Cs 0.5 30.03 years (a) Reaction activities measured for capsule flux evaluation.
Table 6-2. Arkansas Nuclear One Unit 1 Raactor Vessel Fast Flux Fast Flux (E > 1 MeV), n/cm2 -s Flux n/cm2 -s (E > 0.1 MeV)
Inside Surface Inside Surface Cycle (Max location) T/4 3T/4 (Max location)
Cycle 1A 1.39E+10 0.79E+10 0.18E+9 2.68E+10 345 EFPD Cycle IB-4 1.46E+10 0.84E+10 0.19E+10 3.11E+10 1046.5 EFPD Cycles 5 1.03E+10 0.59E+10 0.13E+10 2.15E+10 1 446.4 EFPD Cycles 6 - 7 1.08E+10 0.60E+10* 0.15E+10* 2.43E+10 844 EFPD 10 EFPY 0.91E+10 0.51E+10* 0.13E+10*
l 15 EFPY 0.91E+10 0.51E+10* 0.13E+10*
i 21 EFPY 0.91E+10 0.51E+10* 0.13E+10*
24 EFPY 0.91E+10 0.51E+10* 0.13E+10*
32 EFPY 0.91E+10 0.51E+10* 0.13E+10*
l
- Divide flux at inside surface by the appropriate lead factors on p. 6-8 to f obtain these T/4 and 3T/4 fast flux values.
l
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I 6-7
- Balscock&Wilcos a McDermott company
Table 6-3. Calculated Arkansas Nuclear One Unit 1 Reactor Vessel Fluence Fast Fluence, n/cm2 (g > 1 gey) l Cumulative Inside Surface ,
Irradiation Time (Max location) T/4 T/2 3T/4 End of Cycle 1A 0.41E+18 0.24E+18 0.12E+18 0.05E+18
(.94 EFPY)
End of Cycle 4 1.73E+18 0.98E+18 0.49E+18 0.22E+18 (3.81 EFPY)
End of Cycle 5 1.96E+18 1.10E+18* 0.56E+18* 0.27E+18*
(5.03 EFPY)
End of Cycle 7 2.69E+18 1.50E+18* 0.76E+18* 0.37E+18*
(7.34 EFPY) 10 EFPY 3.45E+18 1.93E+18* 0.98E+18* 0.48E+18*
15 EFPY 4.88E+18 2.73E+18* 1.38E+18* 0.67E+18*
21 EFPY 6.60E+18 3.69E+1B* 1.87E+18* 0.91E+18*
I 24 EFPY 7.46E+18 . 4.17E+18* 2.11E+18* 1.03E+18*
32 EFPY 9.75E+18 5.45E+18* 2.76E+18* 1.34E+18* 1
- Calculated using these 1.0 1.79 3.53 7.25 lead factors Conversion Factors Fluence (E > 1 MeV) 1.45E-21** 1.63E-21** 1.94E-21** 2.31E-21**
to DPA
- Multiply fast fluence values (E > 1 MeV) in units of n/cm2 by these factors to obtain the corresponding DPA values. q 1
6-8 Babcock & Wilcox i a McDermott company
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Total ---
1.46E+19 2.09E-2
) Table 6-5. Estimated Fluence Uncertainty n
Estimated 7
Calculated Fluence .
Uncertainty Basis of Estimate In the capsule 15% Activity measurements, cross section fission yields, satu-( ration factor, deviation from
)
average fluence value
[- In the reactor vessel 21% Activity measurements, cross at maximum location for sections, fission yields, fac-cycles 1 through 6 of tors, axial factor, capsule Arkansas Nuclear One, location, radial / azimuthal ex-Unit 'I trapolation, normalization factor I In the reactor vessel 23% Factors in vessel fluence above at the maximum location plus uncertainties for extra-for end-of-life extra- polation to 32 EFPY r polation 6-9 Babcock &Wilcom a McDermott company
Figure 6-2. Fast Flux, Fluence and DPA Distribution Throuah Reactor Vessel Wall 1.0 _
~
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l 6-10 Babcock & Wilcox I a McDermott company
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> 7. DISCUSSION OF CAPSULE RESULTS-
)
7.1. Pre-Irradiation Property Data I
A review of the unirradiated properties of the reactor vessel core beltline region materials indicated no significant deviation from expected properties except in the case of the upper shelf properties of the weld metal. Based on
!- the predicted end-of-service peak neutron fluence value at .the 1/4T vessel I
wall location and the copper content of the weld metal, it was predicted that 7
the end-of-service Charpy upper shelf energy (USE) will be below 50 ft-lb.
) This. weld was selected for inclusion in the r, surveillance program in accordance with the criteria in effect' at the time the program was designed L
for Arkansas Nuclear One, Unit 1. The applicable selection criterion was based on the unirradiated properties only.
)
i 7.2. Irradiated Property Data 7.2.1. Tensile Properties Table 7-1 compares irradiated and unirradiated tensile properties. 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 within the limits observed for similar materials.
There is some strengthening, as indicated by increases in ultimate and yield strengths and decreases in ductility properties. All changes ~ observed in the base metal are such as to be considered within acceptable limits. The changes at both room temperature and 580F in the properties of the base metal are not as large as those observed for the weld metal, indicating a lesser sensitivity of the base metal to irradiation damage. In either case, the changes in tensile properties are insignificant relative to the analysis of the reactor vessel materials at this time period in the reactor vessel service life.
7-1 Bat > cock &Wilcom a McDermott company
A comparison of the tensile data from previously evaluated capsules (Capsules AN1-E, and AN1-A) with the corresponding data from the capsule reported in this report is shown in Table 7-2. The currently reported capsule experienc-ed a fluence that is approximately twenty times greater than the first capsule.
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 in ductility as measured by both total elongation and reduction of area. The most significant observation from these data is that the weld metal exhibited greater sensitivity to neutron radiation than the base metal.
7.2.2. Imoact Properties The behavior of the Charpy V-notch impact data is more significant to the calculation of the reactor system's operating limitations. Table 7-3 1 compares the observed changes in irradiated Charpy impact properties with the predicted changes. ;
The 30 ft-lb transition temperature shift for the base metal is in relatively poor agreement with the value predicted using Regulatory Guide 1.99, Rev. ;
2.20 The predicted value is significantly greater than the measured value and with the addition of the margin makes the predicted have a large amount of conservatism. It would be expected that these values would exhibit good agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken at the 30 ft-lb temperature.
The transition temperature measurements at 30 ft-lbs for the weld metal is in good agreement with the predicted shift using Regulatory Guide 1.99, Revision <
- 2. The shift being in good agreement with the predicted value which indi-cates that the estimating technique based on the Regulatory Guide 1.99, Rev.
2, are conservative for predicting the 30 ft-lb transition temperature shifts since the method requires that a margin be added to the calculated value to provide a conservative value.
The data for the decrease in Charpy USE with irradiation showed only fair 1 agreement with predicted values for the base metal in both the longitudinal and transverse directions. The weld metal and the correlation material showed the best comparison of the measured data with the predicted value.
7-2 Babcock & Wilcox 4 MCDMmott Company
The various degrees of comparison is to be expected in view of the lack of data for medium , or high-copper-content mtterials at the fluence values that were used to develop the estimating curves. -
A comparison of the Charpy impact data from the previously evaluated capsules from Arkansas Nuclear One Unit I with the corresponding data from the capsule reported in this report is shown in Table 7-4. The currently reported data experienced a fluence that is approximately twenty times greater than the first capsule.
The base metal exhibited transition temperature shifts at the 30 ft-lb levels for the latest capsule that were similar to those of the previous two capsules. The corresponding data for the weld metal showed slightly further increase at the 30 ft-lb level as compared to the previously reported increase at the 30 ft-lb level. This may be related, in part, to a lack of further decrease in the upper-shelf energy.
Both the base metal and the weld metal exhibited decreases in the upper shelf values in a similar manner as the previous capsules. The weld metal in this capsule exhibited a decrease similar to the weld metal in the previous capsule. These data confirm that the upper shelf drop for this weld metal may have reached a stabilized condition as observed in the results of capsules evaluated by others but the difference continues to be observed.
This behavior of Charpy USE drop for this weld metal should not be considered i
indicative of a similar behavior of upper-shelf region fracture toughness properties. This behavior indicates that other reactions may be taking place within the material besides simple neutron damage. Verification of this relationship must await the testing and evaluation of the data from compact fracture toughness test specimens.
Results from other surveillance capsules also indicate that RTNDT estimating curves have greater inaccuracies than originally thought. These inaccuracies l are a function of a number of parameters related to the basic data available L
at the time the estimating curves are established. These parameters may include inaccurate fluence values, poor chemical composition values, and variations in data interpretation. The change in the regulations requiring the shift measurement to be based on the 30 ft-lb value has minimized the 1 '
7-3 Babcock &Wilcon a McDermott company
E errors that result from using the 30 ft-lb data base to predict the shift h behavior at 50 ft-lbs. ,
The design curves for predicting the shift will continue to be modified as more data become available; until that time, the design curves for predicting the RT f4DT shift as given in Regulatory Guide 1.99, Revision 2, are considered adequate for predicting the RT NDT shift of those materials for which data are not available. These curves will be used to establish the pressure-tempera- l ture operational limitations for the irradiated portions of the reactor i vessel until the time that improved prediction techniques are developed and g approved. W The lack of good agreement of the change in Charpy USE is further support of g k
the inaccuracy of the prediction curves. Although the prediction curves are conservative in that they generally predict a larger decrease in upper-shelf energy than is observed for a given fluence and copper content, the conser- -
vatism can unduly restrict the operational limitations. These data support the contention that the USE drop curves will have to be revised as more reliable data become available; until that time the design curves used to predict the decrease in USE of the controlling materials are considered conservative.
7.3. Reactor Vessel Fracture Touchness An evaluation of the reactor vessel end-of-life fracture toughness and the pressurized thermal shock criterion was made and the results are presented in g Table 7-5.
The fracture toughness evaluation shows that the controlling weld metal may have a T/4 wall location end-of-life RTf4DT f 226F based on Regulatory Guide 1.99, Revision 2, with a margin of 56F. This predicted shift is excessive since data from an Integrated Reactor Vessel Surveillance Program surveil-lance capsules exhibit measured RT f4DT significantly less for comparable g fluence values. It is estimated that the end-of-life RT f4DT shift will be 5 significantly less than the value predicted using Regulatory Guide 1.99, Revision 2. This reduced shift will permit the calculation of less restric-tive pressure temperature operating limitations than if Regulatory Guide 1.99, Revision 2, was used.
7-4 I
Babcock &Wilcox a McDermott company
Table'7-1. Comoarison of Capsule AN1-C Tension Test Results Room Temo Test Unirr. Jrrad Elevated Temo L!aitt'"/ ' Test )
Irradsb Base Metal -- C5114-1. Longitudinal-Fluence,10 I9 n/cm2 (E > 1 MeV) 0 1.46 0 1.46 Ultimate tensile. strength, ksi 94.8 99.1 91.8 93.1 0.2% yield strength, ksi 72.0' 79.6 64.5 69.2 Uniform elongation, %~ 10.0 10.6 12.0 9.6 Total elongation, % 26.7 26.3 24.3 22.9 Reduction of area, % 68.2 63.7 64.0 58.5 1
Weld Metal -- WF-193 ,
t ,
. Fluence, 10 I9 n/cm2 (E > 1 MeV) 0 1.46 0 1.46.
Ultimate tensile strength, ksi 84.6 98.9 81.4 90.7 i 0.2% yield strength, ksi- 67.6 83.7 60.4 73.2 I liniform elongation, % 12.2 12.7 10.8 9.1 Total elongation, % 28.1 25.7 21.9 15.7 Reduction of area, % 64.0 57.0 52.1 41.5 (a)The test temperature is 570F.
(b)The test temperature is 580F.
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7-7 Babcock &Wilcox a McDermott company
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The pressurized thermal shock evaluation demonstrates that the Arkansas Nuclear One, Unit I reactor pressure vessel remains below the screening l criterion limits. The increased margin below the screening criteria is the result of improved fuel management which significantly lowers the reactor vessel inside surface fl uence. Therefore, the decrease in RT PTS values
> indicate good fuel-management must be continued to assure that no additional corrective actions, as required by the PTS regulations, will be necessary prior to license expiration.
An evaluation of the reactor vessel end-of-life upper-shelf energy for each i of the materials used in the fabrication was made and the results are presented ir Table 7-6. This evaluation was .made because the weld metals used to fabricate the reactor vessel are characterized by relatively low-upper-shelf-energy and high copper contents; and, consequently, are expected
! to be sensitive to neutron radiation damage. Two methods were used to I
evaluate the radiation induced decrease in upper-shelf energy; the method of
- Regulatory Guide 1.99, Revision 2, which is the same procedure used in .
Revision 1, and the method presented in BAW-180321 which was developed specifically to address the need for an estimating method for this class of weld metals.
The method of Regulatory Guide 1.99, Revision 2, show that all of the weld metals used in the fabrication of the beltline region of the reactor vessel will have an upper shelf energy below 50 ft-lbs prior to the 32 EFPY design life based on the T/4 wall location. Regulatory Guide 1.99 method medicts a decrease below 50 ft-lbs for the controlling weld metal at the val inside wall. However, based on surveillance data and the prediction techniques presented in BAW-1803, it is calculated that none of the reactor vessel material upper shelf energies will decrease to below 50 ft-lbs during the vessel design life.
The uncertainties of the procedures used to evaluate the materials upper-shelf energies necessitates a conservative approach ~ to the problem to insure that the requirements of 10CFR50, Appendix G, are satisfied. Therefore, a l fracture analysis based on the most limiting weld metal was performed to insure that the operating limitations would not compromise the specified i
7-5 Babcock &Wilcon a McDermott company
margins of safety. The details of this fracture analysis are described in Section 9, i
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7-6 Babcock & Wilcox a McDermott company
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Table 7-3. Observed Vs. Predicted Changes for Cgsule gN1-C Irradiated i Charpy Impact Properties - 1.46 x 10 n/cm -(E > 1 MeV) t i
Predicted Material Observed RG 1.99/2(a) Predicted RG 1.99/2+M-(b)
Increase in 30 ft-lb Trans. Temo.. F Base Material (C5114-1)
Longitudinal 64 117 151 Transverse 38 117 151 i Heat-Affected Zone (C5114-1) 53 117 .. 151
> Weld Metal (WF-193) 185 206 262 Correlation Material 68 141 175 (HSST PL-02)
Decrease in Charov USE. ft-lb t
> Base Material (C5114-1)
Longitudinal 23 35 N.A.
Transverse -
16 28 N.A.
)
Heat-Affected Zone (C5114-1) 48 30 N.A.
Weld Metal (WF-193) 26 33 N.A.
Correlation Material 31 36 N.A.
, (HSST PL-02) i (a)Mean value per Regulatory Guide 1.99, Revision 2, May 1988.
(b)Mean value per Regulatory Guide 1.99, Revision 2, May 1988, plus margin.
N.A. - Not applicable.
7-9 Babcock &WHcom a McDermott company
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- 8. DETERMINATION OF REACTOR COOLANT PRESSURE B0UNDARY PRESSURE - TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Arkansas Nuclear One, Unit 1 are established in accordance with the requirements of 10CFR50, Appendix G. The methods and criteria employed to establish operating pressure and temperature limits are described in topical report BAW-10046A, Rev. 2.24 The objective of these limits is to prevent nonductile failure during any normal operating condition, including antici-pated operation occurrences and system hydrostatic tests. The loading condit.ons of interest include the following:
- 1. Normal operations, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests.
- 3. Reactor core operation.
The majo" components of the RCPB have been analyzed in accordance with 10CFR50, Appendix G. The closure head region, the reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of 7
the reactor vessel'(and consequently of the RCPB) that regulate the pressure-temperature limits. Since the closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first several service periods. The reactor vessel outlet nozzle also affects the pressure-temperature limit curves of the first several service periods. l This is due to the high local stresses at the inside corner of the nozzle,
> which can be two to three times the membrane stresses of the shell. After 1
I the first several years of neutron radiation exposure, the RT f the NDT beltline region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature limits of the RCPB. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained f ;
}
8-1 Babcock &Wilcox a McDermott company
i through a point-by-point comparison of the limits imposed by the closure head region, the outlet nozzle, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressures.
The limit curves for Arkansas Nuclear One, Unit 1 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of thirty-two EFPY. The thirty-second EFPY was selected because it represents a logical sequence from the previous analysis. The surveillance capsule was withdrawn at the end of the refueling cycle when the estimated caps'ule fluence corresponds to approximately the inside surface at the end-of-extended life value. The time difference between the withdrawal i of this surveillance capsule and future operating requirements provides adequate time for re- establishing the operating pressure and temperature <
limits for subsequent periods of operation. The current surveillance capsule provides data that can be used to justify continued operation beyond the current license.
The unirradiated impact properties were determined for the surveillance beltline region materials in accordance with 10CFR50, Appendixes G and H.
For the other beltline region and RCPB materials for which the measured 4 I
properties are not available, the unirradiated impact properties and residual elements, as originally established for the beltline region materials, are ;
listed in Table A-1. The adjusted reference temperatures are calculated by I adding the predicted radiation-induced RT NDT and the unirradiated RT NDT. The predicted RT NDT is calculated using the respective neutron fluence and copper i and nickel contents. Figure 8-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel beltline region wall.
The supporting information for Figure 8-1 is described in Section 6. The neutron fluence values of Figure B-1 are the predicted fluences that have !
been demonstrated (Section 6) to be conservative. The design curves of Regulatory Guide 1.99, Rev. 2, were used to predict the radiation-induced RT values as a function of the material's copper and nickel content and NDT neutron fluence.
The neutron fluences and adjusted RT values of the beltline region NDT materials at the end of the thirty-two full-power year are listed in Table 1
8-2 Babcock &Wilcom a McDermott company
m 8-1. The neutron fluences and adjusted RT NDT values are given for the 1/4T and 3/4T vessel wall locations (T = wall thickness). The assumed RT NDT f '
the closure head region and the outlet nozzle steel forgings is 60F, in accordance with BAW-10046, Rev. 2.
The chemistry factors for two of the weld metals in the beltline region were recalculated in accordance with the procedures described in Regulatory Guide 1.99, Revision 2, Section 2.
The data used to calculate a new chemistry factor for weld metal WF-112 was obtained from the B&WOG Integrated Reactor Vessel Surveillance Program. The I -
data for weld metal WF-112 which has the weld wire Heat No. 406L44. A summary of the available data is as follows.
Capsule Weld Metal Fluence, n/cm2 RTNDT, F Reference OCl-E WF-112 1.50E+18 78 25 0Cl-A WF-112 8.95E+18 191 26
)
) OCl-C WF-112 9.86E+18 185 27 l
DB1-LG1 WF-112 8.21E+18 204 28
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The analysis of these data produced a new chemistry factor for WF-112 of 195.
I Similarly, the data used to calculate a new chemistry factor for weld metal WF-182-1 was obtained from the B&WOG-IRVSP (Integrated Reactor Vessel
! Surveillance Program). The data for weld metal WF-182-1 has the weld wire Heat No. 821T44. A summary of the available data is as follows.
! Capsule Weld Metal Fluence,n/cm2 RTNDT, F Reference DB1-F WF-182-1 1.96E+18 127 29 i
DB1-B WF-182-1 5.92E+18 125 TO DB1-A WF-182-1 1.29E+19 175 31 The analysis of these data produced a new chemistry factor for WF-182-1 of f
167.
These two chemistry factors were used as shown in Table 8-1 to determine the 8-3
> Babcock &Wilcox e a McDermott company
correct Reference Temperature values for the calculation of the pressure-temperature operating limits.
Figure 8-2 shows the ~ reactor vessel's pressure-temperature limit curve for normal heatup. This figure also shows the the core criticality limits as required by 10CFR50, Appendix G. Figures 8-3 and 8-4 show the vessel's ;
pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively. All pressure-temperature limit curves are applicable up to the thirty second EFPY. Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits of the pressure-temperature limit curves. The acceptable i pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to go 4 critical until the pressure-temperature combinations are to the right of the criticality limit curve. To establish the pressure-temperature limits for protection against nonductile failure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be adjusted by the pressure differential between
{
the point of system pressure measurement and the pressure on the reactor '
vessel controlling the limit curves. This is necessary because the reactor ;
vessel is the most limiting component of the RCPB. l i
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- 9. LOW UPPER-SHELF FRACTURE TOUGHNESS ANALYSIS 9.1. Backaround Paragraph VI. A.1, 10CFR50, Appendix G, specifies that the reactor beltline materials must have Charpy upper-shelf energy of no less than 50 ft-lb throughout the life of the vessel. Otherwise, it must be demonstrated in a manner approved by the Director of Nuclear Regulation that lower values of S upper-shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of the ASME Code.
The pressure-temperature curves presented in Section 8 are based on the
\
linear-elastic fracture mechanics methods described in Section III, Appendix l
l G of the ASME Code and the analytical procedures documented in topical report ,
BAW-10046A, Rev. 2. Accordingly, the material reference toughness data (i.e.
K curve) of Appendix G, with an upper-shelf value of 200 ksi M was used IR to calculate the allowable pressures. The radiation induced reduction in fracture toughness of the weld metals progresses as the fluence in the reactor vessel increases; concurrently, the Charpy upper-shelf impact energy j decreases. Certain of the weld metals in the reactor vessel are expected to reach values below 50 ft-lb at some time before E0L. At which time, an i
alternate low upper-shelf fracture analysis becomes necessary. The methodo-I logy for the elastic-pl astic fracturo analysis is documented in topical report BAW-10046A, Rev. 2. NRC approval of B&WOG procedures for lower upper-shelf toughness fracture analysis was received in 1986.
The purpose of this section is to provide a low upper-shelf elastic-plastic j toughness analysis based on the controlling weld in the Arkansas Nuclear One Unit-1 reactor vessel. The analysis is based on the flaw described in Section III, Appendix G of the ASME Code, and using the B&WOG elastic-plastic fracture mechanics analytical methods as approved by the NRC. The material l fracture toughness data is obtained from the B&WOG IRVSP (BAW-1920P).28 This .
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l l
9-1 Babcock &WUcox
)
a McDermott company l
i analysis is included to demonstrate the adequacy of the " low-upper-shelf-material" such an evaluation and provide validity to the pressure-temperature operating limits described in Section 8. Also, it is intended to provide an assessment of the reactor vessel integrity to both the licensee and the NRC that the reactor vessel will have adequate upper-shelf fracture toughness for the 32 EFPY design life. i 9.2. Charov Upper-Shelf Impact Eneray l
All welds are designated as shown in a cross section of the beltline region of the Arkansas Nuclear One Unit-1 reactor vessel presented in Figure A-1.
The requirement for an alternate elastic-plastic fracture analysis starts three years prior to when it is estimated that the controlling Charpy upper- 1 shelf energy will drop below 50 ft-lbs. From Table 7-6 the most limiting i weld is identified as WF-112. The Charpy upper-shelf energy at 32 EFPY is estimated as 44 ft-lbs using the procedure in RG 1.99, Revision 2, and to be j greater than 50 ft-lbs at 32 EFPY using the procedure described in BAW-1803.
It is conservative to state, based on the limited data, that the Charpy
]
upper-shelf energy is expected to decrease to a value below the 50 ft-lb level prior to 32 EFPY. Using this approach it is not necessary to accu- 1 rately predict when the weld metal drops below the 50 ft-lb regulatory value, i Weld metal WF-112 is identified in Table 7-6 as the. controlling weld metal to I exhibit a Charpy upper-shelf energy value less than 50 ft-lbs, therefore, a '
low upper-shelf toughness fracture analysis is appropriate at this time. 1 Since the regulatory requirement is based on Charpy energy level, these values are provided to determine the " trigger point" for the required low ,
Charpy upper-shelf analysis, however, for demonstrating an adequate margin on l the material toughness requires fracture toughness data, i.e. J-resistance curves, and not Charpy upper-shelf energy data. There is fracture toughness properties for weld metal WF-112. The toughness data for WF-ll2 are avail-able from the B&WOG IRVSP. Data are available for this weld metal at a 18 fluence of 6.6 x 10 n/cm2 which provides a good basis for demonstrating adequate upper-shelf toughness for 32 EFPY since the estimated T/4 fluence at 18 32 EFPY is 5.45 x 10 n/cm2 ,
9-2 Babcock &WHcom a McDermott company
9.3. Material Fracture Touahness Properties To perform an elastic-plastic fracture analysis, material J-resistance curves and corresponding tensiTe properties of the limiting welds in the reactor vessel are needed. These material data must be representative of the particular reactor vessel fluence value and corresponding operating tempera-ture range. The operating temperature range for the reactor vessel of Arkansas Nuclear One, Unit 1 is 550-570F. The estimated end-of-life fluence level at the T/4 thickness of the vessel wall is 5.45 x 10 18 n/cm2 . There are currently available several sets of irradiated fracture toughness data for WF-112 welds in the B&W data base at a fluence value of 6.6 x 10 18 n/cm2 ,
Therefore, the reactor vessel is analyzed for 32 EFPY, which corresponds to the pressure-temperature operating limit curves in Section 8, evaluation will be applicabie for 32 EFPY (the end of the design life). There are two sets of fracture toughness data available for 550F. Another two sets of data obtained at 480F are also available. All four J-R curves from these data were examined and the two larger specimens were selected as the basis for this analysis. The two sets of data have nearly identical J-R curves. A conservative power law fit was made through the selected two J-R curves and the results used to perform this evaluation.
There are several irradiated tensile data sets: the one at 580F was used for this analysis. An evaluation of the data showed closely grouped yield and ultimate strength values. The corresponding Ramberg-Osgood parameters are l calculated from these tensile data in accordance with the analytical model recommended by the ASME Code,Section XI to address the requirements of the NRC. These calculated values are presented in Table 9-1.
i 9.4. Analytical Method and Acceptance Criteria The elastic-plastic fracture mechanics analytical method, in the form of Failure Assessment Diagram (FAD), as used in this analysis is described in
! BAW-10046A, Rev. 2. The predicted instability point, and the J at the crack propagation initiation point is obtained from this analysis.
I In the temperature range of the upper-shelf toughness region, there is a I
negligible contribution of KI from thermal gradient loading, therefore, only
) the pressure load was considered. The present acceptance criteria using
)
t 9-3 Babcock & Wilcox a McDermott company
DPFAD is summarized below and a more detailed description is presented in
~
BAW-10046A, Rev. 2.
- 1. P is init > 3000 psi - where the pressure at crack initiation, Pdafined as a pressu tion initiates.
J(a,P - P init) - J R( a - 0.04 in.)
where: a - T/4 1
- 2. J(a, P = 2500 x 2) < JR( a = a , cP = 2500 x 2) where: a - T/4 ac = crack depth at the instability point. -
9.5, Fracture Analysis A fracture mechanics analysis was performed using the properties of the WF-112 weld as. the most limiting weld metal in the Arkansas Nuclear One, Unit-1 reactor vessel. The DPFAD analysis methodology was applied using the input data described in Table 9-1 for 2500 psi pressure. The results are presented !
in Figures 9-1 and 9-2 and the calculated safety margins are presented in
-Table 9-2. .
j In accordance with the acceptance criteria discussed in the preceding subsection, P (crack initiation pressure), the pressure at J=J init IC' I should be greater than 3000 psi. As shown in Table 9-2, the safety factor at i a - 0.04 in. (1 mm), where JR is equivalent to JIc, is 1.37 when the applied ;
pressure is 2500 psi. The resulting P init is 3425 psi. Thus, the first l acceptance criterion is met with an adequate margin. The safety factor at the instability point (Table 9-2) is 2.18, which is greater than the required safety factor of 2.0. Therefore, the second criterion is met.
9.6. Summary j The results of this fracture analysis demonstrate that the most limiting low i upper-shelf welds have irradiated fracture toughness characteristics which will assure adequate margins of safety in accordance with the requirements of 10CFR50, Appendix G, for fluence values equivalent to 32 EFPY operation of the reactor vessel. Consequently, the pressure-temperature limit curves calculated in Section 8 are valid for 32 EFPY without any restrictions ;
I because of upper-shelf energy requirements.
9-4 Babcock &Wilcos a McDermott company
I l Table 9-1. Input Data for Deformation Plasticity Failure Assessment Diaaram j I n (Ramberg-Osgood Constant).......... 9.3950 Alpha (Ramberg-Osgood Const.)........ 1.6630 Poisson Ration....................... 0.3000 E (Youngs Modulus, MSI).............. 27.4500 YS (Yield Strength, KSI)............. 74.0000 US (Ultimate Strength, KSI).......... 93.2000 A (Crack Depth, Inches).............. 2.1250 L (Crack Length, Inches)............. 12.7500 T (Thickness, Inches)................ 8.5000 RI (Inside Radius)................... 85.5000 HI (Calibration Function)............ 6.6792 P (Pressure, psi).................... 2500.0000 F l a w Ty p e . . . . . . . . . . . . . . . . . . . . . . . . . . . . No. 2*
I
- Semi-Elliptical Surface Crack in cylinder I
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I I 9-5 Babcock &Wilcox I a McDermott company
Table 9-2. Failure Assessment Data Points DELTA-A Sr' Kr' S.F.
0.0400 0.3297 0.7143 1.3706 0.0700 0.3309 0.6193 1.5677 0.1000 0.3322 0.5671 1.7017 0.1300 0.3334 0.5327 1.7959 0.1600 0.3347 0.5078 1.8714 0.1900 0.3361 0.4887 1.9288 0.2200 0.3374 0.4735 1.9764 0.2500 0.3388 0.4611 2.0144 0.3000 0.3411 0.4447 2.0634
(
0.3500 0.3435 0.4322 2.1005 i 0.4000 0.3460 0.4222 2.1274 I
0.4500 0.3485 0.4142 2.1468 0.5000 0.3511 0.4076 2.1610 1 0.5500 0.3537 0.4022 2.1711 t
0.6000 0.3564 0.3977 2.1775 j 0.6500 0.3592 0.3939 2.1812 ;
0.7000 0.3621 0.3907 2.1823*
0.7500 0.3650 0.3881 2.1816 0.8000 0.3680 0.3859 2.1791 0.8500 0.3711 0.3840 2.1752
- Instability point.
9-6 Sabcock & WHcox a McDermott Company
Figure 9-1. Failure Assessment Diagram for Arkansas Nuclear One, Unit-1 Reactor Vessel Based on Weld Metal WF-112 1.28 Initiation Point e.% <
,' Instability Point
/
/
e.72 Mr /
l .
e.48 ,'
/
/
/
e.24 -
/
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e.88 e.ee e.31 8.62 e.92 1.23 1.54 Sr Figure 9-2. Safety Factors Vs. Crack Extension for Arkansas Nuclear One. Unit-1 Reactor Vessel 25 24-23 -
2.2 -
2.t - n bility 2- f Minimum e Safety Factor of 2 g i .s -
{ is -
l .7 -
'}
g i .s -
is -
~
Initiation Point is -
l.
1 12 - Minimum Safety Factor for
'8-na = 0.04 in. (initiation ofcrackgrowth) i , i i i . i i 0 0.2 0.4 06 08 DELAT c.6tt 9-7 Babcock &t' Uten a McDermott company
_ _ _ _ _ . . - _ . ___--____-_________-_-.m_-___-_-_m.m.-. _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _m.______m__._ -a_
L L
l
- 11. SURVEILLANCE CAPSULE REMOVAL SCHEDULE L
Based on the postirradiation' test results of Capsule AN1-C the following schedule is recommended for the examination of the remaining capsules in the .
Arkansas Nuclear One, Unit 1 RVSP:
-EvaluationSchedule(a)
_ Estimated Vesse}gFluenge, 10 n/cm Estimated gpsule Estimated Date Capsule ID Fluence, 10 n/cm2 Surface- T/4 Date Available(b)
- l. AN1-D(c) 2.1 0.34 0.19 1996 L
AN1-F(d) 1.4 0.27 0.15 1992
-(8)The schedule is in accordance with BAW-10006A and E185-82 as modified by BAW-1543a, Rev. 2, Addendum 1.
(b)The estimated date is based on a 0.8 plant operation factor.
(c)This capsule is designated as a standby and may not be evaluated.
(d)This capsule is designated as a standby and may not be evaluated. It may be re-inserted at a later date to accumulated additional fluence relative l
to license renewal.
11-1 Babcock &Wilcon a McDermott company
___-_--_-__ _ ___-_-___ _ _ a
1
- 10.
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in the fourth surveil-lance capsule, AN1-C, removed for evaluation as part of the Arkansas Nuclear One, Unit 1 Reactor Vessel Surveillance Program, led to the following conclu-sions: -
- 1. The capsule received an average fast fluence of 1.46 x 10 19 n/cm2 (E > 1.0 MeV). The predicted fast fluence for the reactor vessQ T/4 location at the end of the tenth fuel cycle is 2.15 x 10 n/cm2 (E > 1 MeV).
- 2. The fast fluence of 1.46 x 10 19 n/cm2 (E > 1 MeV) increased the RT of the capsule reactor vessel core region shell materials a mabumof185F.
- 3. Based on the calculated fast flux at the vessel wall, an 80% load factor 'and the planned fuel management, the projected fast fluence that the Arkansas Nuclear One, Unit I reactor pressure vessel insifg surfgce will receive in 40 calendar year's operation is 9.75 x 10 n/cm (E > 1 MeV).
- 4. The increase in the RT for the shell plate material was not in good agreement with tkN predicted by the currently used design curves of RT versus fluence (i.e., Regulatory Guide 1.99, Revision 2), b$Tthe prediction techniques are conservative.
i
l
- 6. The low upper-shelf energy fracture analysis demonstrated that the most limiting weld metal has adequate irradiated toughness proper-ties to assure safe operation to 32 EFPY.
- 7. The 2T values decreased for 32 EFPY because of an decrease in the esEIkated E0L fluence values and are below the PTS screening criteria.
l 8. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in weld metal Charpy upper-shelf proper- .
ties due to irradiation are conservative. l l
l 10-1 Babcock &WHeox a McDermott company
- 9. The analysis of the neutron dosimeters demonstrated that the
~
analytical techniques used to predict the neutron flux and fluence were accurate.
I 10. The capsule design operating temperature may have been exceeded
} during operating transients but not for times and temperatures that l would make the capsule data unusable.
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ll 10-2 Babcock & WHcox a McDermott company
- 12. CERTIFICATION The specimens were tested, and the data obtained from Arkansas Power & Light Company Arkansas Nuclear One Unit 1, reactor vessel surveillance Capsule AN1-C were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H.
Y . AE 7Ahvil/fB A. L. Lowe,"Jr., P.E. (/ v Date Project Technical Manager This report has been reviewed for technical content and accuracy.
L.'B. Gross, P.E. (Material Analysis) ~ Date M&SAdnit k
D. A. Nitti (Fluence ~ Analysis)
Vf7/99
/ Date Performance Analysis Unit
( 62m~
K. K on, P.E. (Fracture Analysis) 6M Dat'e M&SA it Verification of independent review.
[ 0 -C ' 27![f A. D.'McKim, Manager Date M&SA Unit This report has been approved for release.
d J. F. Walters 'l Ddte
' Program Manager 12-1 Babcock &WHcom a McDermott company
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Table A-2. Test Soecimens for Determining Material Baseline Procerties Number of Test Specimens Tensile Material Description 70F 570F(a) Charpy Heat GG (C5114-1)
Base metal Transverse direction 3 3 26 Longitudinal direction 3 3 27 HAZ Transverse direction 3 2 27 Longitudinal direction J 3 _21 Total 12 11 106 Heat HH (C5144-2)
Base metal <
Transverse direction 3 3 27 Longitudinal direction 3 3 27 HAZ .
Transverse direction 3 3 27 Longitudinal direction 3 3 26 Total 12 12 107 Weld metal Longitudinal direction 3 3 27 l
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A-4 Babcock & Wilcox a McDermott company
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I APPENDIX A Reactor Vessel Surveillance Program Background Data and Information I
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I A-1 Babcock &WHcom I a McDermott company
'1. Material Selection Data l'
The data used to select the materials for the specimens in the surveillance
]
program,- in accordance with E185-66, are shown in Table A-1. The locations :
of these' materials within the reactor vessel are shown in Figures A-1 and-A-2. ]
-2. Definition of Beltline Reaion t .- .The beltline' region of ANO-1 was defined in accordance with the data given in BAW-10006A 5 1
- 3. Caosule Identification The. capsules used. in the ANO-1 surveillance program are -identified below by j identification number, type, and location.
Caosule Cross Reference Data .l Number Tyjg AN1-A I AN1-B 11 l
AN1-C I '
AN1-D II AN1-E I AN1-F II 4 Specimens per Surveillance Caosule-See Tables A-2 A-3 and A-4.
A-2 Babcocic&Wilcoa a McDermott company
Table A-3. Specimens in Surveillance Caosules (Designations A. C and El No. of Test Soecimens Material Description Tensile Charpy Weld metal, WF-193 4 8 HAZ A, Heat No. C5114-1, 0 8 longitudinal Base material -- Plate A Heat No. C5114-1: Longitudinal 4 8 Transverse 0 4 Correlation, HSST plate 02 0 8 Total per capsule 8 36 Table A-4. Soecimens in Surveillance Caosules (Designations B. D. and F) i N_p. of Test Specimens <
1 Material Description Tensile Charny HAZ B, heat no. C5114-2, 4 10 l longitudinal Base material -- Plate B Heat No. C5114-2: Longitudinal 4 10 Transverse 0 8 Correlation, HSST Plate 02 0 8 l
Total per capsule 8 36 l
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A-5 Babcock &Wilcox i a McDermott company
- - - - --__-_ _ A
f l
l Figure A-1. Location and Identification of Materials Used in the Fabrication of ANO-1 Reactor Pressure Vessel M
M t
l l
f y.
l AYN 131 Nozzle Belt '
W+ WF 182-1 C 5120-2'
' upper Shell
- - WF 18 #C 14-2 s h
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$* WF 112
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WF-18 + C 5120-1
- C 5114-1. Lower Shell ,
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A-6 Babcock &WHcox a McDermott company
l t
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APPENDIX B i Pre-Irradiation Tensile Data I
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . . _ _ _ . . . - J
i Table B-1. Tensile Properties of Unirradiated Shell Plate Material . Heat No. C5114-1 Test Specimen Temp, Strenath, osi Elonaation. % Reduction of No. F Yield Ultimate Uniform Total Area %
Longitudinal GG708 RT 71,430 94,500 10.3 27.9 68.8 GG711 RT 72,870 94,930 8.3 26.4 68.5 GG717 RT 71,590 94,120 11.3 26.7 67.3 Mean RT 71,960 94,850 10.0 26.7 68.2 Std dev'n 790 320 1.5 1.1 0.8 GG704 570 64,110 91,160 11.2 22.9 63.6 GG705 570 64,220 92,070 12.8 25.7 63.6 GG715 570 65,290 92,270 12.0 24.3 64.9 Mean 570 64,540 91,830 12.0 24.3 64.0 Std dev'n 650 590 0.8 1.4 0.8 <
Table B-2. Tensile Properties of Unirradiated Weld Metal WF-193 Tcst ,
Specimen Temp, Strenath, osi Elonaation, % Reduction of i No. F Yield Ultimate Uniform Total Area %
GG101 RT 67,800 84,850 13.3 28.6 64.7 <
GG105 RT 67,790 84,980 10.8 27.1 62.9 GG114 RT 67,080 84,040 12.7 28.6 64.3 Mean RT 67,560 84,620 12.2 28.1 64.0 Std dev'n 410 510 1.23 0.9 1.0 GG102 570 62,390 82,400 9.9 21.4 51.1 GG109 570 59,110 80,850 11.4 22.1 50.4 GG112 570 59,640 80,870 11.2 22.1 54.7 Mean 570 60,380 81,370 10.8 21.9 52.1 Std dev'n 1,760 890 0.8 0.4 2.3 B-2 Babcock & Wilcox a McDermott company
Fiqurg p_,- location of Longitudinal ygyg, in voDer and lower She11 cOurseg W
27.6*
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Shel.t 27.6e E
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Lower 7 ,
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A-7
- bcock awggcon
- Mcoer,n,it Company
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APPENDIX C l
Pre-Irradiation Charpy Impact Data
(
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Table C-1. Charpy Impact Data From Unirradiated Base Material, 1 Longitudinal Orientation, Heat No. C5114-1
, Absorbed Lateral Shear Specimen Test Temp, Energy, Expagsion, Fracture, No. F ft-lb 10 in. %
G6707 319 132 68 100 708 319 131 68 100 741 319 124 67 100 GG703 199 136 66 100 735 199 130 60 100 717 198 124 64 100 GG748 142 136 69 100 747 140 129 65 100 756 140 136 68 100 GG744 86 108 64 80 ;
732 85 108 61 82 (
706 85 104 57 75 GG745 50 70 48 25 751 50 82 57 30 f 752 50 102 63 45 GG749 36 68 48 18 750 35 73 49 15 755 35 71 51 12 GG754 20 21 16 2 753 20 30 21 3 746 19 53 38 5
(
GG733 0.9 48 32 4 724 0.3 43 30 2 ,
704 0.3 60 43 8 l GG736 -39 46 31 3 ,
739 -39 37 24 2 722 -39 15 8.5 0 i
l i
s 7
C-2 I Rabcock &Wilcox a McDermott company
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Table C-2. Charpy Impact Data From Unirradiated Base Material, L-Transverse Orientation. Heat No. C3265-1 t' Absorbed Lateral Shear Specimen Test Temp, Energy, Expagsion, Fracture, N o '. F ft-lb 10 in. %.
GG603 320 88 62 100 605 319 84 60 100 618 322 94 64 100 GG632 259 91 61 100 638 259 94 63 100
!; 640' 258 92 61 100' GG610 200 86 58 100 616 200 96 58 100 l 624 200 101 60 100 GG634 141 89 60 94 i 639 140 90 61 100 1
633 138 100 61 97 GG631 111 66 46 75 636 111 53 43 45 630 110 80 55 70 r
l GG626 80 54 40 35 612 80 56 42 45 607 79 48 32 30 L GG637 61 60 42 30 635 61 38 29 20 629 60 64 44 45 30 GG625 37 32 2-l GG614 30 32 24 1 l
GG608 -38 28 16 1 620 -40 28 16 1 613 -40 14 5 0 C-3 Babcock &WHcom a McDermott company
i 1
Table C-3. Charpy Impact Data From Unirradiated Base Metal, j
, HAZ. Longitudinal Orientation. Heat No. C5114-1 Absorbed Lateral Shear i Specimen Test Temp, Energy, Expagsion, Fracture, No. F ft-lb 10 in. %
GG423 321 83 62 100 434 320 96 58 100 438 320 83 57 100 GG447 205 104 56 100 445 202 116 59 100 3 449 198 123 63 100 l
GG453 142 107 63 100 450 140 95 54 100 i 448 140 107 58 100 i GG426 81 70 44 100 420 80 74 44 94 l'
-403 80 84 46 100 GG452 56 79 43 100 446 55 77 42 90 f
455 54 73 45 92 GG451 46 86 44 94 2 456 45 69 43 98 GG412 31 52 33 35 j 417 30 66 35 85 422 30 46 33 65 GG411 '
-10 43 21 65 1 444 -10 40 21 30 418 -10 53 31 18 GG433 -38 32 18 18 419 -39 32 17 15 425 -39 44 25 12 I
1 I
C-4 Babcock &WHcom a McDermott company
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L F
L Table C-4. Charov Imoact Data From Unirradiated Weld Metal. WF-193 i i
Absorbed Lateral Shear i
. Specimen Test Temp, Energy, Expagsion, Fracture, No. F. -ft-lb- 10 in. %
GG011 320 77 59 100 044 320 73 61 100 035 319 70 60 100 GG023 201 76 49 100 006 200 72 52 100 031 200 66 55 100 GG055 141 67 55 96 051 140 67 46 97 049 139 71 53 100 GG050 109 72 51 97 056 109 67 59 82 045 108 64 46 85
-GG001 87 63 52 75 L 021 82 56 48 65 i 034 82 50 38 55
- GG053 60 35 32 20 1
047 60 28 25 25 046 60 49 38 55 GG005 30 42 35 28 012 29 38 29 40 024 29 42 42 38 3
GG052 1 16 14 8 054 1 33 25 10 048 0 37 31 6 i
GG028 -38 22 15 4 013 -39 12 9 2 027 -39 20 12 3 i
C-5 Bat > cock &Wilcon a McDermott company
Figure C-1. Charpy Impact Data From Unirradiated Base ;
. Metal. Longitudinal Orientation E I I I I ~ l 1 1 1 1 1
- 75 - -.
k w
sw -_-_____e _____________________
5 E 25 - -
O I I I I I I I I I I -4
=
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E
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= (
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O I I I I I I I l I ' I 200 . . , ; , y y , , g j q l
DATA SUPERY i
180. T,,, +10F _,,
Tey (35 nLt) 13F yg _Tey (50 rv.La) 8F _
j Te , (30 rt La) 31F Cy -USE (Avs) 132 ft.1bs -;
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20 22 Fwsw: None .
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1 Figure C-2. Charpy Impact Data From Unirradiated Base Metal. Transverse Orientation 4 d
a 100 g
- l. l l 1 1 - 1 1 i
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e v0s _ _ _ _ _ _ _ _ _ _ _ _
5 5 25_
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- ' ' ' I I I I I I I i i DATA SiffERY 130.T, +10F _
Tey (35 m.r) 64F 1
) 160 icv (50 n-ts) . 74F _
Te , (30 n-La) 17F Cy -USE (avo) 96 ft.1bs i .140 RT +30F -
7 8
NDT C
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20 FLuemet Hone l
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-DQ -40 0 40 80 120 , 160 200 2'O 280 320 360 400 Test Tve enArunt, F t
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4 Figure C-3. Charpy Impact Data From Unirradiated Base Metal. HAZ. Longitudinal Orientation 100 i 3 gg - g - 3 ; 3 g i i s
- 75 j-
= .
3 50 E .
3 25 - -
g i I I I i 1 1 I I I I
. 08 ; g. g y g ; g i g g j E
s - .
t _:= -
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) APPENDIX D
) Fluence Analysis Methodology
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- 1. Analytical Method 1
A semiempirical method is used to calculate the capsule and vessel flux. The method employs explicit modeling of the reactor vessel and internals and uses an average core power distribution in the discrete ordinates transport code DOTIV, version 4.3. DOTIV calculates the energy and space dependent neutron flux for the specific reactor under consideration. This semiempirical method is conveniently outlined in Figures D-1 (capsule flux) and D-2 (vessel flux).
The two-dimensional transport code DOTIV was used to calculate the energy-and space-dependent neutron flux at all points of interest in the reactor j system. DOTIV uses the discrete ordinates method of solution of the Boltz- '
mann transport equation and has multi-group and asymmetric scattering {
capability. The reference calculational model is an R-0 geometric represen- i tation of a plan view through the reactor core midplane which includes the core, core liner, coolant, core barrel, thermal shield, pressure vessel, and j concrete. The material and geometry model, represented in Figure D-3, uses one-eight core symmetry. In order to include reasonable geometric detail l within the computer memory limitations, the code parameters are specified as P3 order of scattering, S8 quadrature, and 22 energy groups. The P3 order of )
scattering adequately describes the predominately forward scattering of neutrons observed in the deep penetration of steel and water media, as 1 demonstrated by the close agreement between measured and calculated dosimeter I activities. The S8 symmetric quadrature has generally produced accurate results in discrete ordinates solutions for similar problems, and is used j routinely in the B&W R-0 DOT analyses.
1 Flux generation in the core was represented by a fixed distributed source I which the code derives based on a 235 0 fission spectrum, the input relative power distribution, and a normalization factor to adjust flux level to the desired power density.
Geometrical Configuration For modeling purposes, the actual geometrical configuration is divided into three parts, as shown in Figure D-3. The first part, Model "A," is used to ;
generate the energy-dependent angular flux at the inner boundary of Model "B," which begins at the outer surface of the core barrel. Model A includes D-2 Batacock &WHcom a McDermott company
I a detailed representation of the core baffle (or liner) in R-0 geometry that has been checked for both metal thickness and total metal volume to ensure
~
that the DOT approximation to the actual geometry is as accurate as possible I for these two very important parameters. The second, Model B, contains an explicit representation of the surveillance capsule and associated compo-nents. The B&W Owners Group's Flux Perturbation Experiment 32 verified that the surveillance capsule must be explicitly included in the D0T model: used for capsule and vessel flux calculations in order to obtain the desired accuracy. The magnitude of the perturbations in the fast flux due to the presence of the capsule was determined in the Perturbation Experiment to be g
'as high as 47% at the capsule center and as high as 10% at the inner surface of the reactor vessel. Detailed explicit modeling cf the capsule, capsule I holder tube, and internal components is therefore incorporated into the D0T calculational models. The third, Model "C," is similar to Model B except i that no capsule is included. Model C is used in determining the vessel flux in quadrants that do not contain a surveillance capsule; typically these l
l quadrants contain the azimuthal flux peak on the inside surface of the reactor vessel.
An overltp region of approximately 32.5 cm or 17 radial intervals is specifi-ed between Model A and Models B or C. The width of this overlap region, which is fixed by the placement of the Model A vacuum boundary and the Model B boundary source, was determined by an iterative process that resulted in close agrennent between the overlap region flux as predicted by Models A and I B or C. She outer boundary was placed sufficiently far into the concrete shield (cat ity wall) that the use of a " vacuum" boundary condition does not cause a perturbation in the flux at the points of interest.
Macroscopic Cross Sections I Macroscopic cross sections, required for transport analyses, are obtained with the mixing code GIP. Nominal compositions are used for the structural I metals. Coolant compositions were determined using the average boron concen-tration over a fuel cycle and the bulk temperature of the region. The core region is a homogeneous mixture of fuel, fuel cladding, structure, and cool ant.
I I D-3 Babcock &Wilcox a McDermott company
I The cross-section library presently used is the (22-neutron group and 18-gamma group) CASK 23E coupled set. The dosimeter reaction cross sections are based on the ENDF/B5 library, and are listed in Table E-3. The measured and calculated dosimeters activities are compared in Table D-1.
Distributed Source The neutron population in the core during full power operation is a function of neutron energy, space, and time. The time dependence is accounted for in l the analysis by calculating the time-weighted tverage neutron source, i.e.
the neutron source corresponding to the time-weighted average power distribu- g tion. The effects of the other two independent variables, energy and space, are accounted for by using a finite but appropriately large number of discrete intervals in energy and space. In each of these intervals the neutron source is assumed to be invariant and independent of all other variables. The space and energy dependent source function can be considered as the product of a discretely expressed " spatial function" and a magnitude coefficient, i.e.
Sv43g = [v/K PD 3, x [RPD 43X g] (0-1) magnitude spatial where:
Sv$3g
- Energy-and space-dependent neutron source, n/cc-sec, v /K = Fission neutron production rate, n/w-sec, l
= Average power density in core, w/cc, RPD - Relative power density at interval (i,j), unitiess,
$3 l
Xg - Fission spectrum, fraction of fission neutrons having energy in group "g,"
l 1 - Radial coordinate index, j - Azimuthal coordinate index, g - Energy group index.
The spatial dependence of the flux is directly related to the RPD distribu-tion. Even though the entire (eighth-core symmetric) RPD distribution is modeled in the analysis, only the peripheral fuel assemblies contribute D-4 I
Babcock & Wilcox a McDermott company
significantly to the ex-core flux. The axial average pin-by-pin RPD distri-bution is calculated on a quarter-core symmetric basis for 8 to 12 times during each core cycle for the entire capsule irradiation period. The time-weighted average RPD distribution is 'used to generate the normalized space and energy dependency of the neutron source. Calculations for the energy and space dependent, time-averaged flux were performed for the midpoint of each DOT interval throughout the model. Since the reference model calculation produced fluxes in.the R-0 plane that are averaged over the core height, an axial correction factor was required to adjust these fluxes to the capsule elevation. The factor used (1.14) was prescribed in BAW-1485P.152 1.1. Caosule Flux and Fluence Calculation As discussed above, the D0TIV code was used to explicitly model the capsule assembly and to calculate the neutron flux as a function of energy within the !
i capsule. The calculated fluxes were used in the following equation to obtain calculated activities for comparison with the measured data. The calculated
[ activity for reaction product Dj, in (FCi/gm') is:
i Dj =
N fj b on (E) # (E) E Fj (1-e- Ajtj) e 'A i(T 7j )
(D-2)
)
(3.7 x 104)An E j where:
t N = Avogadro's number, An - Atomic weight of target material n, i
fi - Either weight fraction of target isotope in n-th material or the -
fission yield of the desired isotope, on(E) - Group-averaged cross sections for material n (listed in Table l E-3) p(E) - Group averaged fluxes calculated by D0TIV analysis, Fj = Fraction of full power during j-th time interval, tj l A
> i = Decay constant of the ith isotope, T - Sum of total irradiation time, i.e., residual time in reactor, and the wait time between reactor shutdown and counting times, 7j - Cumulative time from reactor startup to end of j-th time period.
D-5 Bat > cock &WHcom a McDermott company
tj = Length of.the j-th time period ,
Adjustments were made to the calculated dosimeter activities to correct for the effects listed below:
! Short half-life adjustments to Ni and Fe dosimeter activities 238 Photofission adjustments to 0 and 237 Np dosimeter activities 238 Fissile impurity adjustments to U dosimeter activities After making these adjustments the calculated dosimeter activities were used with the corresponding measured activities to obtain the flux normalization factors:
Dj (measured) )
C1- ,
Dj (calculated)
These normalization factors were evaluated, averaged, and then used to adjust a the '.alculated test specimen flux and fluence to be consistent with the dosimeter measurements. Additionally, the normalization factor was used to 1 update the average normalization factor which had been derived from previous i analyses.' The updated normalization factor was then used to adjust the 4
calculated vessel flux and fluence. The flux normalization factors are given i in Table D-1. ;
- 2. Vessel Fluence Extrapolation i For past core cycles, fluence values in the pressure vessel are calculated <
as . described above. Extrapolation to future cycles is aquired to predict 1 the useful vessel life. Three time periods are considered in the extrapola-tion: 1) operation to date for which vessel fluence has been calculated, 2) 3 future fuel cycles for which PDQ calculations have been performed, and 3) future cycles for which no analyses exist.
For the Arkansas Unit I analysis, time period I is through cycle 5, time period 2 covers cycles 6 and 7, and time period 3 covers from the end of cycle 7 through 32 EFPY. The flux and fluence for time period 2 was esti-mated by calculating the vessel flux using an adjoint-DOT calculational 1 procedure with the appropriate assembly-average power distributions and '
integrating these values over time period 2. The extrapolation of the :
fluence through time period 3 was accomplished by assuming that the average D-6 Sabcock&Wilcos a McDermott company
Figure D-1. Rationale for the Calculation of Dosimeter i
Activities and Neutron Flux in the Caosule DOF/B4 0055 SECTIONS GEDERY & OlMDRARRE POER DISTRI-90F/B5 DOS M TER REAC- FIR PODEL A DDT BUTIONS SINCE ltR CRDSS SECTIONS CAPSlLE INSER-
, TION (PDQ)
GIP ,,
- S(RREL 4 CROSS SECTIONS CDDE y
7 TDE AVE DISTRI-GEDETRY & 00T 4 < BUTED 50lRCE Sv EMDRATlRE --> KDEL A (E, R, e )
, KDEL B sr ir ir
/ DDT 4 ANGLLAR FUJX
> KDEL B c AT BARREL ,
> PGER HIST (RY i DOSDETER OF CAPSLLE
> ACTIVITIES (PRHIST CIDE) u
> FINAL 1
CALCLLATED ACTIVITIES a
PEASLRED DOSDETER ACTIVITIES AXIAL CORRECTION FACTOR i
<r 1 r U CAPSlLE FLUX 70RP%LIZATI(N FACT (R
> c M/C RATIO I
D-9 Babcock &Wilcom a McDermott company
Figure D-2. Rationale for the Calculation of Neutron Flux in the Reactor Vessel GEDETRY & G.lADRATWE -
Fm KDEL "A" DDT PWER DISTRI-MICROSCOPIC CROSS BUTIONS SINCE SECTIONS E)OF/B4 STARTUP (PDQ DOF/B5 m EQUIVALENT) ir ir ir MACROSCOPIC GOSS SWREL CIDE SECTIONS BY E-GROLP
" GIP" CG)E TIE AVE DISTRIBUTED SOLRCE Sv 1 (E, R, e ) '
ir ir +
DOT 4 0(DEL A) j FOR DDT KDEL B m C <
(
ir v DOT 4 K) DEL B ANG1AR FLUX AT !
4 NO/m KDEL C +- BARREL SLRFACE I l
i
(
l PORMALIZATIM FACitR FR04 AXIAL (IRRECTION CAPSULE FlMMI NMLYSIS FACTDR (FROM TE DIAERAM ON THE PREVIOUS PAGE) w sr v TIE -AVERAGE VESSEL F1.UX AT 1%XIMH VESSEL LOCATION (E, R,0)
D-10 Babcock & Wilcox a McDermott company
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flux during period 3 was equal to the average flux for period 2 (cycles 6 and 7).
Table D-1. Flux Normalization Factor Calculate Flux Measured Activity, (a) Activity,{a) Normalization 4Ci/a #Ci/a Factor 54Fe(n,p)54Mn 1116.45 1370.21 0.947(b) 58Ni(n,p)58Co 1972.42 2455.90 1.016(c) 238U (n,f)137Cs 17.97 15.26 1.178 7 237Np(n,f)137Cs 98.57 95.36 1.034 Averaged: 1.044(d)
(") Average of four dosimeter wires.
(b) Average of four ratios (one for each dosimeter wire) corrected by
> short half-life factor of 1.134.
) (c) Average of four ratios (one for each dosimeter wire) corrected by short half-life factor of 1.236.
(d) Average of all four dosimeters was selected as the normalization constant.
l D-7 Babcock &Wilcox a McDermott company I
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i Table D-2. Arkansas Unit 1 Reactor Vessel Fluence by Cvele i
Vessel Fluence, n/cm2 (c)
Incremental Cumulative Vessel- lux, Cycles' Time, EFPY Time, EFPY n/cm s Incremental Cumulative 1A 0.94 .94 1.39E+10 0.41E+18 0.41E+18 IB 2.86 3.81 1.46E+10 1.32E+18 1.73E+18 5 1.22 5.03 '1.03E+10 0.23E+18 1.96E+18 6-7 2.31 7.34 1.08E+10 0.73E+18 2.69E+18 2.66 10.00 0.91E+10(a) 0.76E+18(b) 3.45E+18(b) 5.00 15.00 0.91E+10(a) 1.43E+18(b) 4.88E+18 (b) 6.00 21.00 0.91E+10(a) 1.72E+18(b) 6.60E+18(b) 3.00 24.00 0.91E+10(a) 0.86E+18(b) 7.46E+18(b) .
l 8.00 32.00 0.91E+10(a) 2.29E+18(b) 9.75E+18(b)
-(a) Maximum neutron flux.at inside surface of reactor vessels, based on fuel cycle designs for future cycles 8 and 9, used for extrapolation of 4 fluence to future times. !
(b) Extrapolated values.
(c) Peak fluence at inside surface of reactor vessel. l 1
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i APPENDIX E i Capsule Dosimetry Data
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E-1 Babcock &Wilcox a McDermott (cmpany
I Table E-1 lists the characteristics of the neutron dosimeters. Table E-2
~
shows the measured activity per gram of target material (i.e., per gram of uranium, nickel, etc.) for the capsule dosimeters. Activation cross sections for the various materials were flux-weighted with the 235 U fission spectrum shown in Table E-3.
Table E-1.
Detector Material Detector Composition and Shieldina
% Taraet Shieldina Reaction l
238 238U (n,f)137Cs U-Al
~
10.38% 0 Cd-Ag 237 237 Np(n,f)137Cs Np-Al Ni 1.44%
67.77% 58Ni Np Cd-Ag Cd-Ag 58Ni(n,p)58Co I t 59 59 Co-Al 0.66% 00 Cd Co(n,y)60Co Co-Al 0.66%5900 None 59Co(n,y )60C0 54 54 Fe 5.82% Fe None Fe(n,p)54Mn Table E-2. Measured Specific Activities (Unadjusted) >
for Dosimeters in Caosule AN1-C Dosimeter Activity. (4Ci/am of Tarcet)
Detector Material Dosimeter Reaction CD1 CD2 CD3 CD4 58 Ni Ni(n,p)58Co 1656.80 1693.60 2204.46 2334.82 54 Fe fe(n,p)S4Mn 941.45 966.16 1264.46 1293.73 U-Al 238U (n, f)l37Cs 17.97(a) ,, ,, ,,
Np-Al 237Np(n,f)l37Cs 98.57(a) ,, ,, ,,
I (a) Corrected average of the measured activities for all four dosimeters.
I I
I E-2 I
Babcock & Wilcox a McDermott company
Table ' E-3. Dosimeter Activation Cross Sections, b/ atom (a) 237 238U (n,f) 58 54 G Energy Range, MeV Np(n,f) Ni(n,p) Fe(n,p) {
1 12.2 - 15 2.323 1.051E+0 4.830E-1 4.133E-1 2 10.0 - 12.2 2.341 9.851E-1 5.735E-1 4.728E-1 3 8.18 - 10.0 2.309 9.935E-1 5.981E-1 4.772E-1 4 6.36 -
8.18 2.093 9.110E-1 5.921E-1 4.714E-1 5 4.96 -
6.36 1.542 5.777E-1 5.223E-1 4.321E-1 6 4.06 -
4.96 1.532 5.454E-1 4.146E-1 3.275E-1 7 3.01 -
4.06 1.614 5.340E-1 2.701E-1 '2.193E-1 8 2.46 -
3.01 1.689 5.325E-1 1.445E-1 1.080E-1 9 2.35 -
2.46 1.695 5.399E-1 9.154E-2 5.613E-2 10 1.83 -
2.35 1.676 5.323E-1 4.856E-2 2.940E-2 i
11 1.11 -
1.83 1.596 2.608E-1 1.180E-2 2.948E-3 12 0.55 -
1.11 1.241 9.845E-3 1.336E-3 6.999E-5 13 0.111 -
0.55. 2.352E-1 2.436E-4 5.013E-4 6.419E-8 14 0.0033 - G.111 1.200E-2 6.818E-5 1.512E-5 0 t
(a)ENDF/B5 valgg that have been flux weighted (over CASK energy groups) i based on a U fission spectrum in the fast energy range plus a 1/E shape in the intermediate energy range.
l E-3 Babcock &Wilcon a McDermott company
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I i APPENDIX F References
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l F-1 Babcock & Wilcox a McDermott company
1
- 1. ' A. L. Lowe, Jr., et al., Integrated Reactor Vessel Material Surveillance
~
Program, BAW-1543A. Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, May 1985, and Addendum 1, July 1987. l
- 2. A. L. Lowe, Jr., et al., Analysis of Capsule ANI-E from Arkansas Nuclear One, Cait 1, Reactor Vessel Materials Surveillance Program, BAW-1440, j Babcock & Wilcox, Lynchburg, Virginia, April 1977.
- 3. A. L. Lowe, Jr., et al., Analysis of Capsule ANI-B from Arkansas Nuclear l One, Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1698, Babcock & Wilcox, Lynchburg, Virginia,-November 1981. !
l
- 4. A. L. Lowe, Jr., et al., Analysis of Capsule ANI-A from Arkansas Nuclear One, Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1836, 1 Babcock & Wilcox, Lynchburg, Virginia, March 1989.
- 5. G. J. Snyder and G. S. Carter, Reactor Vessel Material Surveillance !
Program, Revision 3, BAW-10006A. Revision 3, Babcock & Wilcox, Lynchburg, Virginia, January 1975. !
1
- 6. American . Society for Testing and Materials, Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors, A185-66, j November 1966.
- 7. Code of Federal Regulation, Title 10, Part 50, Fracture Toughness Requirements for~ Light-Water Nuclear Power Reactors, Appendix G, Fracture Toughness Requirements. >
- 8. Code of Federal Regulation, Title 10, Part 50, Fracture Toughness Requirements for Light-Water Nuclear Power Reactors, Appendix H, Reactor !
Vessel Material Surveillance Program Requirements.
- 9. American Society of Mechanical Enginee~s r (ASME) Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure (G-2000).
I
- 10. K. E. Moore and A. S. Heller, Chemistry of 177-FA B&W Owners' Group Reactor Vessel Beltline Welds, BAW-1500P, Babcock & Wilcox, Lynchburg, l
Virginia, September 1978.
! 11. J. D. Aadland, Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor F-2 I Babcock &WHcom a McDermott company
Vessel and Surveillance Program Materials Information, BAW-1820, Babcock
& Wilcox, Lynchburg, Virginia, December 1984.
- 12. Heavy Section Steel Technology Program, Semiannual Progress Report for Period Ending February 28,1969, ORNL-4463, Oak Ridge National Labora-tory, Oak Ridge, Tennessee, January 1970.
- 13. American Society for Testing and Materials, Methods and Definitions for Mechanical Testing of Steel Products, A370-77, June 24,1977.
I
- 14. American Society for Testing and Materials, Methods for Notched Bar Impact Testing of Metallic Materials, E23-82, March 5,1982.
\ .
- 15. S. Q. King, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAW- l' 1485P. Revision 1, Babcock & Wilcox, Lynchburg, Va., April 1988.
- 16. B&W's Version of D0TIV Version 4.3, One- and Two-Dimensional Transport l l Code System," Oak Ridge National Laboratory, Distributed by the Radia- l tion Shielding Information Center as CC-429, November 1,1983.
- 17. " CASK-40-Group Coupled Neutron and Gamma-Ray Cross Section Data,"
Radiation Shielding Information Center, DLC-23E.
I
- 18. Dosimeter File ENDF/B5 Tape 531, distributed March 1984, National Neutron Data Center, Brookhaven National Laboratory, Upton, Long Island, NY.
- 19. American Society of Testing Materials, Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA), E693-79 (Reapproved 1985).
- 20. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Reaulatory Guide 1,99. Revision 2, May 1988.
- 21. A. S. Heller and A. L. Lowe, Jr., Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds, BAW-1803, Babcock & Wilcox, Lynchburg, Virginia, January 1984.
- 22. Code of Federal Regulations, Title 10, Part 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.
- 23. K. E. Moore and A. S. Heller, B&W 177-FA Reactor Vessel Beltline Weld F-3 Babcock & Wilcox a McDermott company
i Chemistry Study, BAW-1799, Babcock & Wilcox, Lynchburg, Virginia, July )
~
1983.
24.- H. W. Behnke, et al., Methods of Compliance With Fracture Toughness and i Operational Requirements of Appendix G to 10CFR50, BAW-10046A. Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, June 1986. j
- 25. A. L. Lowe, Jr., et al., Analyses of Capsule OCI-E Duke Power Company Oconee Nuclear Station -- Unit-1, Reactor Vessel Materials Surveillance Program, BAW-1436, Babcock & Wilcox, Lynchburg, Virginia, September 1977.-
- 26. J. D. Aadland, et al., Analyses of Capsule OCI-A Duke Power Company i
Oconee Nuclear Station -- Unit-1,~ Reactor Vessel Materials Surveillance l Program, BAW-1837, Babcock & Wilcox, Lynchburg, Virginia, August 1984.
- 27. A. L. Lowe, Jr., et al . , Analyses of Capsule OCI-C from Duke Power Company Oconee Unit-1, Reactor Vessel Materials Surveillance Program, BAW-2050, Babcock & Wilcox, Lynchburg, Virginia, October 1989. l 1
- 28. A. L. Lowe, Jr., et al., Analysis of Capsule DB1-LG1, Babcock & Wilcox I
Owners Group, In'tegrated Reactor Vessel Materials Surveillance Program, BAW-1920P, Babcock & Wilcox, Lynchburg, Virginia, October 1986.
- 29. A. L. Lowe, Jr., et al., Analysis of Capsule TEl-F from Toledo Edison Company, Davis-Besse Nuclear Power Station, Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1701, Babcock & Wilcox, Lynchburg, Virginia, January 1982.
- 30. A. L. Lowe, Jr., et al ., Analysis of Capsule TEl-B from Toledo Edison ,
Company, Davis-Besse Nuclear Power Station, Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1834, Babcock & Wilcox, Lynchburg, Virginia, May 1984.
- 31. A. L. Lowe, Jr., et al., Analysis of Capsule TEl-A from Toledo Edison Company, Davis-Besse Nuclear Power Station, Unit 1, Reactor Vessel Materials Surveillance Program, BAW-1882, Babcock & Wilcox, Lynchburg,
_ Virginia, September 1985.
- 32. N. L. Snider and L. A. Hassler, B&WOG Flux Perturbation Experiment at l .
F-4 Babcock &WHcom a MCDermott Comparty
ORNL, Measured and Calculated Dosimeter Results, BAW-1886, Babcock &
Wilcox, Lynchburg, Virginia, September 1985.
l t
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l
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F-5 Babcock &Wilcox a McDermott company
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