ML20040B068
| ML20040B068 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/31/1981 |
| From: | Ewing J, Lowe A, Pavinich W BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML15223A766 | List: |
| References | |
| BAW-1697, NUDOCS 8201250095 | |
| Download: ML20040B068 (95) | |
Text
{{#Wiki_filter:__ _ __ _ I BAW-1697
- October 1981 1
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j ANALYSIS OF CAPSULE OCIII-B FROM DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNIT 3 '= - Reactor Vessel Materials Surveillance Program - l 'I I I i I I I I
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l BAW-1697 l l October 1981
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!I ANALYSIS OF CAPSULE OCIII-B FROM I DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNIT 3
- Reactor Vessel Materials Surveillance Program -
I l 1 il i by l A. L. L we, Jr., PE 1E I I3 J. W. Ewing W. A. Pavinich W. L. Redd J. K. Schmotzer I I I !I . I B6W Contract No. 582-7109-71 l J BABCOCK & WILCOX j Nuclear Power Group I Nuclear Power Generation Division I P. O. Box 1260 Lynchburg, Virginia 24505 i l Babcock & Wilcox
i Il l l E l I SDStARY l This report describes the results of the examination of the second capsule of l the Duke Power Company's Oconee Nuclear Station Unit 3 reactor vessel sur-veillance program. The capsule was removed and examined after accumulating a fluence of 3.12 x 10 18 nvt which is equivalent to approximately 12 ef fective full power years (EFPYs). The objective of the program is to monitor the ef-I fects of neutron irradiation on the tensile and fracture toughness properties ; of the reactor pressure vessel materials by the testing and evaluation of ten- , sion, Charpy impact , and compact f racture toughness specimens. The program = l was designed in accordance with the requirements of Appendix H to 10 CFR 50 i and ASTM specification E185-73. The capsule received an average f ast fluence of 3.12 x 10 1e n/cm 2 (E > 1 MeV) ( and the predicted fast fluence for the reactor vessel T/4 location at the end j i 18 2 of 4.1 EFPY operation is 1.08 x 10 n/cm (E > 1 MeV). Based on the calcu- l lated current fast flux at the vessel inside wall and an 80% load f actor , the I p roj ec t ed fast fluence that the Oconee Unit 3 reactor pressure vessel will re-ceive in 40 calendar years' operation is 1.56 x 10 19 n/cm 2 (E > 1 MeV). The results of the tensile tests indicated that the materials exhibited nor- l mal behavior relative to neutron fluence exposure. The Charpy impact data results exhibited the characteristic behavior of shif t to higher temperature for both the 30 and 50 f t-lb transition temperatures as a result of neutron fluence damage and a decrease in upper shelf energy. These results demon-strated that the current techniques used for predicting the change in both the inc rease in the RT g and the decrease in upper shelf properties due to ir-radiation a re conservat ive. The recommended period for using the new operating limitations was extended to 15 effective full power years as a result of the second capsule evaluation. These new operating limitations are in accordance with the requirements of Appendix G of 10 CFR 50. Babcock & Wilcox
!I I il a I II 4 CONTENTS I Page
- 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . ..... 1-1
- 2. BACRCROUND . . . . . . . . . . . . . . . . . . . . . . . ..... 2-1
- 3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . ..... 3-1
- 4. PREIRRADIATION TESTS . . . . . . . . . . . . . . . . . . ..... 4-1 1
j 4.1. Tensile Tests . . . . . . . . . . . . . . . . . . ..... 4-1 4.2. Irpact Tests . . . . . . . . . . . . . . . . . . . ..... 4-1
- 5. POST-IRRADIATION TESTS . . . . . . . . . . . . . . . . . ..... 5-1 l 5.1. Thermal Monitors . . . . . . . . . . . . . . . . . ..... 5-1 i
5.2. Tensile Test Results . . . . . . . . . . . . . . . ..... 5-1 5.3. Charpy V-Notch Impact Test Results . . . . . . . . ..... 5-1 l 6. NEUTRON DOSIMETRY , . . . . . . . . . . . . . . . . . . ..... 6-1 l 6.1. Vessel Fluence . . . . . . . . . . . . . . . . . . ..... 6-2 6.2. Capsule Fluence . . . . . . . . . . . . . . . . . ..... 6-3 I 'l i
- 7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . ..... 7-1 j 7.1. Preirradiation Property Data . . . . . . . . . . . ..... 7-1 7.2. Irradiated Property Data . . . . . . . . . . . . . ..... 7-1
,g 7.2.1. Tensile Properties . . . . . . . . . . . . . .... 7-1 , a 7.2.2. Impact Properties . . . . . . . . . . . . . .... 7-2
- 8. DETERMINATION OF RCPB PRESSURE-TEMPERATURE LIMITS . . . . .... 8-1
! 9.
SUMMARY
OF RESULTS . . . . . . . . . . . . . . . . . . . . .... 9-1 l
- 10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . . .... 10-1 i
! 11. CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . .... 11-1 i APPENDIXES A. Reactor Vessel Surveillance Program - Background I B. C. Data and Information . . . . . . . . . . . . . . . .... Preirradiation Tensile Data . . . Preirradiation Charpy Impact Data A-1 B-1 C-1 I I - 111 - Babcock & Wilcox
I CONTENTS (Cont'd) Page D. Fluence Analysis Procedures . . . . . . . . . . ..... D-1 E. Capsule Dosimeter Data . . . . . . . . . . . . . ..... E-1 F. References . . . . . . . . . . . . . . . . . . . ..... F-1 I List of Tables Table 3-1. Specimens in Surveillance Capsule OCIII-B . . . . . . . ..... 3-2 3-2. Chemistry and lleat Treatment of Surveillance Materials ..... 3-3 5-1. Irradiation Tensile Properties of Capsule OCIII-B Base Metal and Weld Metal Irradiated to 3.12E18 nyt . . . . ..... 5-3 5-2. Charpy Impact Data From Capsule OCIII-B Shell Forging l W Material, Ileat ANK-191, Irradiated to 3.12E18 nyt . . . ..... 5-3 5-3. Charpy Impact Data From Capsule OCIII-B Shell Forging Material,lleat-Affected Zone, Ileat ANK-191, Irradiated g to 3.12E18 nvt . . . . . . . . . . . . . . . . . . . . ..... g 5-4 5-4. Charpy Impact Data From Capsule OCIII-B Shell Forging Material, IIcat AWS-192, Irradiated to 3.12E18 nyt . . . ..... 5-4 5-5. Charpy Impact Data From Capsule OCIII-B Shell Forging Material, IIcat-Af fected Zone, lleat AWS-192, Irradiated to 3.12E18 nyt 5-5 5-6. Charpy Impact Data From Capsule OCIII-B Weld Metal, 5 E WF-209-1B, Irradiated to 3.12E18 nyt . . . . . . . . . . .... 5-5 5-7. Charpy Impact Data From Capsule OCIII-B Correlation Monitor Material, llSST Plate 02, lleat Irradiated g to 3.12E18 nyt
. . . . . . . . . . . . . . . . . . . . . .... 5-6 5 6-1. Surveillance Capsule Detectors . . . . . . . . . . . . . .... 6-3 6-2. Neutron Fluence at Pressure Vessel Surface Through Cycle 5 . . . . . . . . . . . . . . . . . . . . . .... 6-4 =
i 6-3. Predicted Fast Neutron Fluence in Pressure Vessel . . . . .... 6-4 6-4. Neutron Fluence at Capsule Center . . . . . . . . . . . . 7-1. Comparison of Tensile Test Results
.... 6-5 g,
. . . . . . . . . . . .... 7-4 3 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Properties . . . . . . . . . . . . . . . . .... 7-5 8-1. g Data for Preparation of Pressure-Temperature Limit Curves for Oconee 3, Applicable Through 15 EFPY . . . . . .... 8-4 g
A-l. Surveillance Program Materials Selection Data for Oconee 3 . . . . . . . . . . . . . . . . . . . . . . .... A-3 A-2. Materials and Specimens in Upper Surveillance Capsules OCIII-A, OCIII-C, and OCIII-E . . . . . . . . . . . . . . .... A-4 B-1. Prcirradiation Tensile Properties of Shell Forging Material,lleat ANK-191 g
. . . . . . . . . . . . . . . . . .... B-2 g B-2. Preirradiation Tensile Properties of Weld Metal -
Longitudinal, WF-209-1 . . . . . . . . . . . . . . . . . .... B-2 I
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I I Tables (Contl Table Page C-1. Preirradiation Charpy Impact Data for Shell Forging Material - Transverse Orientation, Heat AWS-192 . . . . ..... C-2 I C-2 C-3. Preirradiation Charpy Impact Data for Shell Forging Material - HAZ, Transverse Orientation, Heat AWS-192 Preirradiation Charpy Impact Data for Shell Forging
. ..... C-3 Material - Transverse Orientation, Heat ANK-191 .... ..... C-4
'I C-4. Preirradiation Charpy Impact Data for Shell Forging Material, HAZ, Transverse Orientation, Heat ANK-191 . . ..... C-5 C-5. Preirradiation Charpy Impact Data for Weld Metal, WF-209-1 . . . . . . . . . . ... . .. . . . . ..... C-6 D-1. Dosimeter Activations . . . . . . ....... .... ..... D-5 D-2. Reactor Vessel Fast Fluence . . . . ... .. . . . . . ..... D-5 I E-1. E-2. E-3. Detector Composition and Shielding Capsule OCIll-B Neutron Dosimeters Dosimeter Activation Cross Sections E-2 E-3 E-7 I I List of Figures I Figure 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule Locations at Oconee Unit 3 . . . .. . ... . ..... 3-4 3-2. Reactor Vessel Cross Section Showing Surveillance Capsule Locations at Crystal River Unit 3 .. .... . .... 3-5 5-1. Charpy Impact Data for Irradiated Base, Metal Transverse Direction . . . . . . ................ 5-7 5 -2. Charpy Impact Data for Irradiated Heat-Af fected I Zone Material . . . . . . . . . . ..... . ... . . .... 5-8 5-3. Charpy Impact Data fer Irradiated Base Metal, I 5-4. Transverse Direction . . . . . . Charpy Impact Data for Heat-Affected Zone, Transverse Direction . . . . . . .... .. .. . . . . .... 5-9 5-10 I 5-5. 5-6. 6-1. Charpy Impact Data for Irradiated Weld Metal . ... . . .... Charpy Impact Data for Correlation Monitor Material Oconee 3 Reactor Vessel Relative Fast Fluence 5-11 5-12 6-6 6-2. Reactor Vessel Fast Flux Oconee 3, Cycles 2, 3, 4, and 5 . ... 6-7 6-3. Relative Fast Fluence Capsule OCIII-B . ..... . . . .... 6-8 8-1. Predicted Fast Neutron Fluences at Various Locations Through Reactor Vessel Wall for First 15 EFPY . . . . .... 8-5 I 8-2. 8-3. Leactor Vessel Pressure-Temperature Limit Curvea for Normal Operation - Heatup, Applicable for First 15 EFPY .... Reactor Vessel Pressure-Temperature Limit Curve for 8-6 I 8-4. Normal Operation - Cooldown, Applicable for 15 EFPY Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Hydrostatic Tests, Applicable for
. . .... 8-7 First 8 EFPY , . . . . . . . . . ................ 8-8
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I Figures (Cont'd) Figure Page A-1. Location and Identification of Materials Used in Fabrication of Oconee Unit 3 Reactor Pressure Vessel . . . ... A-5 C-1 Charpy Impact Data From Unirradiated Base Metal . . . . .... C-7 C-2. Charpy Impact Data From Unirradiated lleat-Affected Zone Material . . . . . . . . . . . . . . . . - . . . . .... C-8 C-3. Charpy Impact Data From Unirradiated Base Metal . . . . .... C-9 g C-4. Charpy Impact Data From Unirradiated IIcat-Af fected C-19 5 Zone Material . . . . . . . . . . . . . . . . . . . . . .... l C-5. Charpy Impact Data From Unirradiated Weld Metal . . . . .... C-Il C-6. Charpy Impact Data From Unirradiated Correlation Material . . . . . . . . . . . . . . . . . . . . . . . . .... C-12 9' I I I I I I I I I I I I
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- 1. INTRODUCTION This report describes the results of the examination of the secone capsule of j Duke l ower Company's Oconee Nuclear Station, Unit 3 (Oconee 3) reactor vessel survei llance program. The first capsule of the program was removed and exam-Ined after the first year of operation; the results are reported in BAW-1438 . 1 The objective of the program is to monitor the ef fects of neutron irradiation
] on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillance program for Oconee 3 was designed and furnished by Babcock & Wilcox; it is described in BAW-10006A, Revision 3.2 The program was planned to monitor the ef fects of neutron irradi- ] ation on the reactor vessel materials for the 40-year design life of the reac- ! tar pressure vessel. j The surveillance program for Oconee 3 was designed in accordance with E185-66 i and has been updated to meet the intent of the first draft of ASD1 E185-73. The program is not in compliance with Appendixes C and H to 10 CFR 50 since ! the requirements did not exist at the time it was designed. Because of this difference, additional tests and evaluations were required te ensure meeting ! the requirements of these appendixes. The recommendations for the future op-l cration of Oconee 3 included in this report do comply with these requirements. i l i i
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- 2. BACKGROUND 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 re-actors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general ef-fects of fast neutron irradiation on the mechanical properties of such low-alloy ferritic steels as SA508, Class 2, used in the fabrication of the Oconee Unit 3 reactor vessel are well characterized and documented in the literature.
I The low-alloy ferritic steels used in the beltline region of reactor vessels exhihit an increase in ultimate and yield strength properties with a corre-sponding decrease in ductility after irradiation. In reactor pressure vessel steels, the nost serious mechanical property change is the increase in tempera-ture for the transition from brittle to ductile fracture accompanied by a re-duction in the upper shelf impact toughness. Appendix G to 10 C/R 50, " Fracture Toughness Requirements," 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 I pressure-temperature limitations on operation of the RCPB. The toughness and operat lonal requirements are specified to provide adequat e safety margins dur-Ing anv condition of i.ormal operation, including anticipated operational occur-rences and system hydrostatic tests, to which the pressure boundary may be sub-jected over its service lifetime. Although the requirements of Appendix C to 10 CFR 50 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 10 CFR 50, " Reactor Vessel Materials Surveillance Program Re-quirements," defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the re-actor vessel beltline region of water-cooled reactors resulting from exposure 2-1 Babcock s, Wilcox
l I to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the re-actor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life. A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code, Section III. This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RTNDT, which is defined as the- greater of the drop weight nil-ductility transition temperature (per ASTM E-208) or the tempera- = ture that is 60F below that at which the material exhibits 50 ft-lb and 35 mils lateral expansion. The RT NDT "E "* "' '" """ " * ^ material to a reference stress intensity factor curve (K curve), which ap-IR pears in Appendix G of ASME Section III. The K cune is a lown bound of IR dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K curve, al NwaMe stmss intensity f actors can be oMaind W Ws ma-IR terial as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors. The RT@T 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 re-l actor 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 containing prepared speci-mens of the reactor vessel materials is periodically removed from the operating u nuclear reactor and the specimens tested. The increase in the Charpy V-notch 50-ft-lh temperature, or the increase in the 35 mils of lateral expansion tem-perature, whichever results in the larger temperature shift due to irradiation, is added to the original RT us ra n em men . s NDT adjusted RT """ " *# ^ ' " "" NDT IR '"" ' is used to set operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials. I I 2-2 Babcock 8. Wilcox
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- 3. SURVEII. LANCE PROGRAM DESCRIPTION
, The surveillance program for oconee 3 comprises six surveillance capsules de-I signed to monitor the ef fects of neutron and thermal environment on the mate-rials of the reactor pressure vessel core region. The capsules, which were j insert ed into the reactor vessel before initial plant startup, were positioncd inside the reactor vessel between the thermal shield and the vessel wall at the j locat f ons shown in Figure 3-1. The capsules, placed two in each holder tube, a were posit ioned near the peak axial and azimuthal neutron flux. BAW-10006A,
- Revision 3, includes a full description of capsule locations and design.2 j After removal of the capsules from the Oconee 3 irradiation site and their in-1
{ clusion in the integrated reactor vessel materials surveillance program, they j were itradiated in Crystal River Unit 3. During this period capsule OCIII-B was irradiated in site WX as shown in Figure 3-2. I Capsul. OCIII-B was removed f rom Crystal River Unit 3 after cycle 2 and an accumulated fluence of approximately 3 x 10 18 nvt. This capsule contained Cha rpy V-not ch impact and tensile specimens fabricated of SA508, Class 2 steel, weld metal, and correlation steel. The specimen contained in the cap-sule are described in Table 3-1, and the chemistry and heat treatment of the surveillance materials are described in Table 3-2. All test specimens were machined f rom the 1/4-thickness location of the shell forgings. Charpy V-notch and tensile specimens from the vessel material were oriented with their longitudinal axes parallel to the principal working direc-tion of the forging; specimens were also oriented transverse to the princi-pal working direct ion. Capsule OCIII-B contained dosimeter wires, described
, as follows:
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i I Dosimeter wire Shielding U-Al alloy Cd-Ag alloy Np-Al alloy Cd-Ag alloy Nickel Cd-Ag alloy 0.66% Co-Al alloy Cd 0.66% Co-Al alloy None Fe None 3 Thermal monitors of low-melt ing eutectic alloys were included in the capsule. The eutectic alloys and their melting points are as fo llows : E Alloy Melting point, F 90% Pb, 5% Ag, 5% Sn 558 97.5% Pb, 2.5% Ag 580 97.5% Pb. 1.5% Ag, 1.0% Sn 588 Lead 621 Cadmium 610 I Table 3-1. Specimens in Surveillance Capsule OCIII-B No. of specimens Material description Tensile Charpy = Weld metal, WF-209, longitud. 2 12 Weld,liAZ l lleat A - ANK-191, longitud. 0 12 l licat B - AWS-192, longitud. 0 6 Baseline material lleat A - ANK-191, longitud. 0 0 W lleat A - ANK-191, transverse 2 12 IIcat B - AWS-192, longitud. 0 0 Ileat B - AWS-192, transverse 0 6 E 3 Correlation, hSST plate 02 0 6 Total per capsule 4 54 I 3-2 Babcock 8.Wilcox g
j II 4 l j Table 3-2. Chemistry and licat Treatment ! of Surveillance Materials !I
- Cliemical Analysis
, Correlation ! lleat Ileat material ! Element ANK-191 ") AWS-192(" WF-209-1Bt Weld meta}b) IISST-02 (A-1195-1)(C} l ' l C 0.24 0.22 0.067 0.23 ) Mn 0.72 0.58 1.58 1.39 i P 0.014 0.011 0.020 0.013 S 0.012 0.015 0.005 0.013 i S1 0.21 0.24 0.56 0.21 t Ni 0.76 0.73 0.48 0.64 Mo 0.62 0.60 0.33 0.50 5 Cu 0.02 0.01 0.30 0.17 i lie.it Treatment l
- Time, l !! eat No. Temp, F h Cooling EK-191("} 1620-1660 4.0 Water quench j 1570-1610 4.0 Water quench 4
1230-1270 10.0 Water quench 1100-1150 40.0 Furnace-cooled i AWS-192(") 1620-1660 4.0 Water quench I 1570-1610 1220-1250 1100-1150 4.0 10.0 Water quench Water quench ! 40.0 Furnace-cooled WF-209-1B 1100-1150 30.0 Furnace-cooled A-1195-1 1600 75 4.0 Water quench
- 1225!25 4.0 Furnace-cooled j 1125 25 40.0 Furnace-cooled
'i I " Per certificate of test. I Per weld certification. ORNL-4463. (d)Per plate section identification card. I ("} Normalized at 1675F 2 75F. E l I , 1 I 3-3 Babcock & Wilcox
I Figure 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule Locations at Oconec Unit 3 X Surveillance Capsule Holder Tubes - Capsules OCIII-C, OCIII-D E 3 f
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h Holder Tubes - Cap-l sules OCIII-A, OCIII-B Surveillance Capsule j lloider Tube - Capsules Z l W OCIII-E, OCIII-F I I I I 3-4 Babcock & Wilcox
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, F i;;u re 3-2. Reactor Vessel Cross Section Showing Surveillance Capsule Locations at Crystal River Unit 3 X
SURVEILLANCE f CAeSULE HOLDER TUBES II 7 g x . o x e
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F SURVEILLANCE .I SURVEILLANCE I / CieSuu R0t0tR I TUBES CAPSULE HOLDER TUBE Z l I I 3-5 Babcock & Wilcox
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- 4. PREIRRADIATION TESTS I l'nf rradiated material was evaluated for two purposes: (1) to establish a base-line of data to which irradiated properties data could be referenced, and (2) to det ermine those materials properties to the extent practical from available j material, as required for compliance with Appendixes C and H to 10 CFR 50.
4.1. Tensile Tests i l Tensile specimens were fabricated from the reactor vessel shell course plate I3 and weld metal. The subsize specimens were 4.25 inches long with a reduced i sectIon 1.750 inches 1cng by 0.357 inch in diameter. They were tested on a
- 55,000-lb-load capacity universal test machine at a crosshead speed of 0.050 inch per minute. A 4-pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accor-dance with the applicabic requirements of ASTM A370-72. 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 cer-tified accuracy of better than ! 0.5% of full scale (25,000 lb). All test
!I i data for the preirradiation tensile specimens are given in Appendix B. j 4.2. Impact Tests l Charpy V-notch impact tests were conducted in accordance with the requirements of ASn! Standard Methods A370-72 and E23-72 on an impact tester certified to > I= meet Watertown standards. Test specimens were of the Charpy V-notch type, which were nominally 0.394 inch square and 2.165 inches long. lI Prior to testing, specimens were temperature-controlled in liquid immersion baths, capable of covering the temperature range from -85 to +550F. Specimens l were removed f rom the baths and positioned in the test frame anvil with tongs specifically designed for the purpose. The pendulum (hammer) was released i manually, allowing the specimens to be broken within 5 seconds from their re-moval from the temperature baths. I 4-1 Babcock & Wilcox
I I Impact test data for the unirradiated baseline reference materials are pre-sented in Appendix C. Tables C-1 through C-5 contain the basic data which are plotted in Figures C-1 through C-5. I I I I I I I I I 1 I I! I-I I I 4-2 Babcock & Wilcox
I I I l I 5. POST-IRRADIATION TESTS 5.1. Thermal Monitors Surveillance capsule OCIII-B contained three temperature monitor holder tubes, each containing five fusible alloys with different melting points ranging from 558 to 621F. All the thermal monitors at 558, 580, and 588F had melted, while those at 610 and 621F appeared to remain in their original configuration as initially placed in the capsule. However, a close examination revealed slight I melt ing at one end of the 621F monitor and what appeared to be only slight slumpf ng of the adjacent 610F monitor. Since only a portion of the 588F moni-itors had melted it was concluded that the 610 and 621F monitors were in re-verse locations and that the 610F monitors had exhibited the incipient melting. From these data it was concluded that the irradiated specimens had been exposed to a maximum temperature in the range of 588 to less than 610F during the re-actor vessel operating period. There appeared to be no significant temperature g rad ient along the capsule length. 5.2. Tens ile Test Result s The results of the post-irradiation tensile tests are presented in Table 5-1. Tests were performed on specimens at both room temperature and 580F using the same test procedures and techniques used to test the unirradiated specimens (sect ion 4.1) . In general, the ultimate strength and yield strength of the material increased slightly with a corresponding slight decrease in ductility; bot h ef fects were the result of neutron radiation damage. The type of be-havior observed and the degree to which the material properties changed is I within the range of changes to be expected for the radiation, environment to which the specimens were exposed. The results of the preirradiation tensile tests are presented in Appendix B.
- 5. 3. Charpy V-Notch Impact Test Results The test resultc f rom the irradiated Jharpy V-notch specimens of the reactor vessel beltline material and the correlation monitor material are presented in 5-1 Babcock & Wilcox
l l I l l Tables 5-2 through 5-7 and Figures 5-1 through 5-6 The test procedures and I I i t echniques were the same as those used to test the unirradiated specimens (section 4.2). The data show that the material exhibited a sensitivity to ir-radiation within the values predicted from its chemical composition and the fluence to which it was exposed. The results of the preirradiat ion Charpy V-notch impact test are given in Appendix C. 5 l I , I I I' I; Il I I I I' i I I I 5-2 Babcock 8.Wilcox l I
I i j Table 5-1. Irradiation Tensile Properties of Capsule OCIII-B j Base Metal and Weld Metal Irradiated to 3.12E18 nyt (E > 1.0 MeV) , Test Red'n rength, psi Flongation, 7. g Specimen temp, of area, j 5 No. F Yield Ultimate Unif Total % Base Metal, Transverse (ANK-191) JJ-609 69 91,300 105,600 15.1 25.1 55.9 JJ-606 585 56,000 86,900 13.9 36.9 64.0 Weld Metal (WF-209-1) PP-010 69 90,600 105,000 15.6 24.7 57.4 ) PP-009 585 37,800 85,600 15.9 25.7 55.9 I I Table 5-2. Charpy Impact Data From Capsule OCIII-B Shell Forging Material,lleat ANK-191, Irradiated to I 3.12E18 nyt (E > 1.0 MeV) Test Absorbed Lateral Shear I Specimen No. temp, F energy, ft-lb expansion, 10-3 in. fracture, Transverse Orientation I JJ-614 -60 4.5 1.5 0 JJ-620 -21 20.0 14.5 0 I JJ-684 0 29.0 22.0 0 JJ-616 0 41.0 32.0 0 JJ-623 19 16.5 15.0 2 JJ-688 30 61.0 47.5 15 JJ-661 33 81.0 60.0 35 JJ-672 40 64.0 49.0 20 JJ-642 72 63.5 58.0 20 JJ-610 140 120.5 78.5 85 JJ-618 I JJ-671 230 400 126.5 120.5 92.5 84.5 100 100 I 1 l l 5-3 Babcock 8.Wilcox l 1 l l
I Table 5-3. Charpy Impact Data From Capsule OCIII-B Shell Forging, Heat-Affected Zone, Heat ANK-191, Irradiated to 3.12E18 nyt (E > 1.0 McV) Test Absorbed Lateral Shear Specimen No. temp, F energy, ft-lb expansion, 10-3 in. fracture, 5
% 3 Transverse Orientation JJ-306 -118 9.0 4.0 0 JJ-355 -80 45.5 28.0 20 JJ-354 -60 10.0 7.0 5 JJ-379 -21 20.0 18.5 25 i JJ-351 -10 81.5 43.5 25 JJ-391 10 27.0 21.0 25 JJ-363 19 81.0 43.0 70 JJ-385 75 60.0 46.0 60 JJ-358 110 70.0 43.0 95 JJ-377 140 62.5 40.0 95 JJ-373 146 136.0 89.0 100 g
JJ-376 230 62.0 51.0 100 W JJ-365 440 54.0 47.0 103 I, Table 3-4. Charpy Impact Data From Capsule OCIII-B ShcIl Forging Material, llent AWS-192, Irradiated to 3.12E18 nyt (E > 1.0 MeV) Test Absorbed Lateral Sheat Specimen temp, energy, expansion, fracture, No. F ft-lb 10 3 in. % l e Transverse Orientation i I a KK-629 -60 8.5 4.5 0 KK-609 -21 6.0 5.0 0 KK-672 19 47.5 39.0 5 , KK-661 KK-663 75 67.5 95.0 59.0 25 g' 146 79.0 100 5 l KK-637 319 102.0 81.0 100 ' 5-4 Babcock 8.Wilcox B! 3: b
I Table 5-5. Charpy Impact Data From Capsule OCIII-B Shell Forging Material, Heat-Affected Zone,IIcat AWS-192, Irradiated to 3.12E18 nyt (E > 1.0 MeV) ,I Test Absorbed Lateral Shear I Specimen No. temp, F energy, ft-lb expansion, 10-3 in. fracture, 1 Transverse Orientation KK-319 -60 25.0 16.0 0
- KK-302 -10 80.0 49.0 30 KK-301 10 31.0 21.0 20
, FK-306 40 55.0 38,0 30 KK-304 72 119.0 77.5 100 KK-320 140 126.0 78.5 30 l Table 5-6. Charpy Impact Data From Capsule OCIII-B Weld Metal, WF-209-1B, Irradiated to 3.12E18 nyt (E > 1.0 MeV) Test Absorbed Lateral Shear I Specimen No. temp, F energy, ft-lb expansion, 10-3 in. fracture, I JJ-032 JJ-048 0 2 12.0 19.0 12.5 14.0 15 5 JJ-001 38 15.0 14.0 5 JJ-019 75 24.0 23.5 20 JJ-008 100 30.5 29.0 85 JJ-031 146 34.0 34.0 70 JJ-007 175 46.5 44 5 95 JJ-020 194 51.0 51.0 100 JJ-013 230 43.5 44.5 100 I JJ-047 JJ-003 280 440 44.0 52.0 46.5 50.5 100 100 JJ-026 580 54.5 58.5 100 I I 5-5 Babcock & Wilcox
I l 1 Table 5-7. Charpy Impact Data From Capsule OCIII-B Correlation ! Monitor Material, IISST Plate 02, IIcat Irradiated to i 3.12E18 nyt (E > 1.0 MeV) i Test Absorbed Lateral Shear , Specimen No. temp, energy, expansion, 10-3 in. fracture, E: F ft-lb % 3( JJ-918 72 21.5 16.5 5 l l JJ-916 94 19.5 16.0 10 JJ-908 110 46.0 34.5 20 JJ-902 140 52.0 , 41.5 40 JJ-914 210 93.5 73.0 100 ( JJ-911 242 97.0 75.0 100 1 Il . Il t
.I' Ii I i I! ;
l I Il i I 5-6 Babcock & Wilcox
- I
; Fip,ure 5-1. Charpy Impact Data for Irradiated Base, i Metal Transverse Direction g ; g g g ; 3 i 3 i 100{
I i a 75 - - J
- E C
I
- 50 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
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' ' ' i I I I i l I l DATA
SUMMARY
180- T,g7 _ l l m Tcy(35 mLE) 23F T CV (50 FT-La) 29F - 160 iI i
= 140 y
Tcy(30 FT-La) 12F Cy -USE (Ava) 125 FT-L8S gi NDT _31F - i _ 1 E 120- - l %
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MATERIAL SA508 CLl2 OnitmTAfton TRANSVERSE ~~ 20- . FLUENCE 3.12El8 nyt - ' HEAT No. AN K- 191 I I I I I I I I ' ' ' 0
-00 -40 0 40 80 120 160 200 240 280 320 360 400 Test TemptRATunt, F 5-7 Babcock & Wilcox
l I I' Figure 5-2. Charpy Impact Data for Irradiated IIcat-Af fected Zone Material 100s i i g ;
,- ; , ., ; i , ,
I l
" 75 - -
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- W 50 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
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200 i i i ; ; , ; g g g ; DATA
SUMMARY
-~
180-T uor ~ 9F Tcv (35 Mtt) T cy (50 n-La) 40F 160 - Tcy(30 rT-Ls) 3F Cy -USE (avs) 67 FT-L6S
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-C0 -40 0 40 80 120 160 200 240 280 320 360 400 IEST TEMPERATURE, f 5-8 Babcock & Wilcox l,
s i I Figure 5-3. Charpy Impact Data for Irradiated Base Metal, Transverse Direction I E I I I l l
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tea rn ORIENMTION TRANSVERSE-~ 20- FLUEME 3.12E18 nyt - e HEAT No. AWS-192 I I I I I I I I ' ' ' 0 I -C0 -40 0 40 30 120 160 Test TEMPERATURE, F 200 240 280 320 360 400 5-9 Babcock & Wilcox
I Figure 5-4. Charpy Impact Data for Heat-Affected Zone, Transverse Direction 100 g g g ; g , , g , , , a 75 - J s 0 50 - - - - - - - - - - - - - - - - . _ - - _ - - - - _ - - - - _ u
- e e e d 25 -
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200 , i i 1 I l- 1 I I I I E DATA Sut?ARY 180-TNOT ~ Tcy (35 mtt) - 160-T cy (50 n-ts) -- _ Tcy (30 n-ts) - Cy -USE (Avo)
.140 gy _
y NDT C
, e 5 120- e __
g
- i 8
.?100- -
a 5 - 5 80- e _ 0 ' r + l ~ 60- _ t l 40 - l MATE R I AL SA50s. CL2 i 1 - - _ - - .A. - _ _ _ _ _ _ _ _. _. _ .. t e ORIENTATION .l! LAPS. H AZ ' 20 - FLutNCE 3.12E18 NVT ' HEAT No. Aws-192 I I I I I I I I i 0 i e
-C0 -40 0 40 80 120 . 160 200 240 280 320 360 400 Test TEnrenATune, F W!
5-10 Babcock & WilCOX i
1 1 'I !I Figure 5-5. Charpy Impact Data for Irradiated Weld Metal W ~T- 1 - 1 I l 1 i T i - i l 1
~ 75 -
,I J 5 l gw ___________ __________________ i I :e a 25 - I I I I I I I I I 0
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'i I I I I I I I I I I I 0-1 lg 100 i i i ; ; ; g g y g g jE DATA SUFFARY 90 r T,3, - I lT (35 8QITCV r cy mtE) 140F (50 n a) 400F Tcy(30 n-u) 109 F
,'g Cy -USE (avo) 49/53 rT-Les 3 70F RT 360F 7
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l C I . E 3a i 20- , WELD METAL - TERIAL ORIENTATION TRANSVERSE
, 10- FLutNCE 3.12E18 nyt -
HEAT No, WF-209-l B j I I I I I I I I ' II ' ' 0
-80 -40 0 40 80 120 160 200 240 280 400 500 600 TEsf TEMPERATURE, F 5-11 Babcock & Wilcox (
1 l l
Il Figure 5-6. Charpy Impact Data for Correlation Monitor Material El l l 100 ; j g g ; , g ; ; 7 i
** 75 -
- I S
t;
=
50 -------------- - - - - - - - - - - - - - - - - - - - - - - - - - - - l u- l l 3 w i d 25 - - l l l I I I I I I I I ! 0
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[ ! E. i i
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# .02 -
l
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G i a I I I I I I I l I I I 0 l l I l l l [ [ ! DATA
SUMMARY
180- T,37 _ Tcy(35 MLE) 120F , Tcy(50 FT-ts) 126F 160 _ i Tcy(30 FT-La) 95F j ! Cy-USE (Avo) 96 FT-L85 [
.140 RT 66F -
7' NDT C
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-
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l FLutNCE
/ HEAT No. A-Il95-1 I I I I I I I I i t 0
-00 -40 0 40 80 120 160 200 240 280 320 360 400 !
Test TEMPERATURE, F l l l l 5-12 Babcock & Wilcox I
- I I
I .
- 6. NEUTRON DOSIMETRY I Fluence analysis as a part of the pressure vessel surveillance program has three primary objectives: (1) determination of maximum fluence in the pressure I vessel as a function of reactor operation, (2) prediction of pressure vessel fluenae in the future, and (3) determination of the test specimen fluence within the surveillance capsule. Vessel fluence data are used to evaluate changes in reference transition temperature and upper shelf energy levels, and to calculate pressure-temperature operation curves. Test specimen fluence data are used to establish a correlation between changes in material proper-ties and fluence exposure. To provide this information, a computer model for calculation of flux distributions in the reactor is established. The accuracy of fast flux values derived from this model is enhanced by the use of a nor-malization factor utilizing measured activity data obtained from capsule do-simeters.
A significant aspect of the surveillance program is to provide a correlation between the neutron fluence above 1 MeV and the radiation-induced property changes noted in the surveillance specimen. To permit such a correlation, activation detectors with reaction thresholds in the energy range of interest were placed in each surveillance capsule. The significant properties of the detectors are given in Table 6-1. Because of a long half-life (30 years) and threshold energy of 0.5/1.0 Mev, the measurements of 137 Cs production from fission reactions in 237 Np (and 23e U) are more directly applicable than the other dosimeter reactions to analytical determinations of the fast neutron (E > 1 MeV) fluence for multiple fuel cycles. I The other dosimeter reactions are useful as corroborating data for shorter time intervals and/or higher energy fluxes. Short-lived isotope activities are representative of reactor conditions only over the latter portion of the irradiation period (fuel cycle), whereas reactions with a threshold energy greater than 2 or 3 MeV do not record a significant part of the total fast flux. 6-1 Babcock & Wilcox
I The energy-dependent neutron flux is not directly available from activation I, i detectors because the dosimeters register only the integrated effect of the neutron flux on the target material as a function of both irradiation time and i ne ut ron energy. To obtain an accurate estimate of the average neutron flux Incident on the detector, several parameters must be known: the operating his- f tory of the reactor, the energy response of the given detector, and the neu-t ron spect rum at the detector location. Of these parameters, the definition of the neutron spectrum is the most difficult to obtain. Two means are essen-t f ally available to obtain the spectrum - iterative unfolding of experimental g foil data and analytical methods. Because of a lack of sufficient threshold 5 foil detectors satisf ying both the threshold energy and half-life requirements as needed for a surveillance program, calculated spectra are used in this analysis. Neutron transport calculations in two-dimensional geometry are used to calcu-late energy-dependent flux distributions throughout the reactor. Reactor con-ditions are selected to be representative of an average over the irradiation time period. Geometric detail is selected to explicitly represent the sur-veillance capsule assembly and the pressure vessel. The detailed calculational procedure is described in Appendix D. 6.1. Vessel Fluence The maximum fluence in the pressure vessel through cycle 5 was determined to 18 be 1.94 x 10 n/cm2 for neutron flux with E > 1 MeV (Table 6-2). The location of maximum fluence is a point at the cladding / reactor vessel interface at an elevation of about 79 cm above the lower active fuel boundary and at an azi-nuthal (peripheral) location of about 12 from a major axis (across flats diameter). Fluence data have been extrapolated to 15 and 32 EFPY of opera-t ion (Table 6-3) . W Relative fluence as a function of radial location in the pressure vessel is shown in Figure 6-1. Corresponding lead factors from the metal surfaces to T/4, T/2, 3T/4, and outside surface are 1.8, 3.6, 7.9, and 20. 9, respective-g, ly. Relative fluence as a function of azimuthal angle is shown in Figure 6-2. 5 l A peak occurs at about 12* which roughly corresponds to a corner of the core and to four symmetric capsule locations. The other two capsule locations cor-respond to the azimuthal minimum at about 26*. However, it should be noted 6-2 Babcock & Wilcox r I
il lI that the maximum to minimun azimuthal flux ratio is only 1.3 and that the data in Figure 6-2 do not account for flux perturbation due to the presence of a capsule. Fast neutron flux is increased by about 1.25 in the capsule due to differences in scattering and absorption cross sections between steel and water. 6.2. Capsule Fluence ,l m Fluence at the center of the surveillance capsule was calculated to be 3.12 x 10" n/cm 2, s25% was received in Oconee 3 and %75% in Crystal River 3 (Figure 6-3). These data can be t ransposed to individual test specimens by the use
- of relative fluence data in Figure 6-3. Capsule OC3-B was located in a lower holder tube position at 11* off axis and s211 cm from the core axis for 345 1
effective full power days (EFPDs) in Oconee 3. (This corresponds to the orig-j inal 177-FA holder tube design.) It was then inserted in Crystal River 3 in an upper holder tube position at 12 off axis and %202 cm from the core axis for an additional 333 EFPD. During the latter irradiation period, the capsule a was rotated 90* clockwise relative to its original design orientation (key fac-ing the reactor core). The capsule rotation complicated the determination of fast fluence for individual specimens. lI Table 6-1. Surveillance Capsule Detectors l' Isotope Detector reaction Energy, MeV half-life l 5"Fe(n p) 5Mn >2.5 312.5 days ss Ni(n,p)seCo >2.3 70.85 days 238 U(n,f)l37Cs l >1.1 30.03 years 217 Np(n,f)137Cs
>0.5 30.03 years 4 238 U(n,f)l03Ru
>1.1 39.43 days 237 i Np(n,f)l33Ru >0.5 39.43 days
!I
- 1 lI
- 6-3 Babcock 8.Wijcox e
I Table 6-2. Neutron Flu 2nce at Pressure Vessel Surface Through Cycle 5 Cycle Cycle Cumulative length, Fast flux, fluence, fluence, EFPD 10 1 n/cm2 -s 10 1e n/cm 2 10 18 n/cm 2 Fast Fluence E > 1.0 MeV Cycle 1 477.9 1.39 0.575 0.58 Avg of cycles 2, 3, 4, 1020.1 1.55 1.367 1.94 and 5 Fast Fluence E > 0.1 MeV Cycle 1 477.9 2.78 1.148 1.15 . Avg of cycles 2, 3, 4, 1020.1 3.304 2.912 4.06 and 5 Table 6-3. Predicted Fast Neutron Fluence (E > 1.0 MeV) I in Pressure Vessel Fast fluence, n/cm 2 I EFPY Inside wall T/4 3/4 T Outside wall 18 4.1- 1.94 x 10 1.08 x 10 18 2.50 x 10 17 9.8 x 10 16 15 7.3 x 10 18 4.0 x 10 18 8.8 x 10 17 4.0 x 10 17 32 1.56 x 10 " 8.7 x 10 18 1.88 x 10 18 8.6 x 10 17 E 1 I I I I 6-4 Babcock & Wilcox
i t il l 4 !I ] Table 6-4. Neut ron Fluence at Capsule Center i Cycle Cycle Cumulative length, Fast flux, fluence, fluence, j EFPD n/cm2 -s 10 18 n/cm 2 10
- n/cm 2 l
1
= Fast Fluence E > 1. 0 BdeV i
l Oconee 3, cycle 1 345 2.48 x 10' O.739 0.739 l Crystal River 3, 338 8.16 x 10 2.383 3.12 ll l l5 cycles IB, 2 Fa s *. Fluence E 0. I SteV !g Oconee 3, cycle 1 345 4.68 x 10 1 1.395 1.40 W Crystal River 3, 338 1.922 x 10 ll 5.613 7.01 l cycles IB, 2 lI il , I i l i I
- i lI l
\
I l ll i lI i !I
I Figure 6-1. Oconee 3 Reactor Vessel Relative Fast Fluence (E > 1.0 MeV) 12.0 l 1.0.
- 1. 0 _
0.9 -
- 0. 8 -
T/4
- 0. 7 -
0.6 - 0.5 - 0.4 - T/2 C 5 u_ 0.3 - H 3 3 0.2 - I g 3T/4 g x e 5 Il 0.1 - 0.08 - 0.0. I: 0.05 -
\ I 1 I i 1
'E 215 220 225 230 235 240 5 Radial Distance From Core e.xial (cm)
I 6-6 Babcock & Wilcox
W M M M M M M M M M M M M M M M M M M l ( Figure 6-2. Reactor Vessel Fast Flux (R = 217.32 cm-ID of Base Metal) Oconee 3, Cycles 2, 3, 4, and 5 l 1.46 1.40 - T E g 1.30 - T E u o a
^ i t,
x I 2 W
- 1.20 -
m m
?
o E- 1.10 - e. g l' l I I I l I f I I e I e I o 0 4 8 12 16 20 24 28 32 36 40 44
-* Deviation from Major Axis, Degrees
I' Figure 6-3. Relative Fast Fluence Capsule OCIII-B OCONEE Ill CYCLE I EXPOSURE CRYSTAL RIVER 3, CYCLES 'B & 2 EXPOSURE ' 345 EFPD 338 EFPb w I N I
.84 .83 .82 .84 .83 .82 I
l . 0 l, 1.00 .98 K 1.00 1.00 .98 I 1.18 1.18 1.15 1.18 1.18 1.15 I f CENTER FAST FLUENCE CENTER FAST FLUENCE 7.4 x 1017 n/cm2 CORE 2.38 x 10 M n/cm 2 i l l l Note: K = key. I I
- I I
I I I l I l I 6-8 Babcock & Wilcox
I I
- 7. DISC 1'SSION OF CAPSULE RESULTS I 7.1. Preirradiation Property Data I A review of the unirradiated properties of the reactor vessel core belt region Indicated no significant deviation from expected properties except in the case of the upper shelf behavior of the weld metal. Based on the predicted end-of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of this weld, it is predicted that the end-of-service Charpy upper shelf energy (USE) will be below 50 ft-Ib. This weld was selected for inclusion in the surveillance program in accordance with the criteria in ef-feet at the time the program was designed for Oconee Unit 3. The applicable selection criterion was based on the unirradiated properties only.
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 negligible. There appears to be some strengthening, as indicated by increases in ultimate and yield strength and similar decreases in dnctility properties. All changes observed in the base metal are such as I to be considered within acceptable limits. The changes at room temperature in the properties of the base met.a1 are greater than those observed for the weld metal, indicating a greater 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 period in service life. At 580F there appears to be no radiation damage to either ma-terial. This is more probably associated with the fact that the low fluence did not damage the material sufficiently so that the test temperature annealed out the damage. In fact, the weld metal exhibited a small amount of what may be interpreted as stress relieving because of the decrease in yield strength and small increase in ductility. 7-1 Babcock & Wilcox
I 7.2.2. Impact Properties The behavior of the Charpy V-notch impact data is more significant than that of tensile data to the calculation of the reactor system's operating limita-tions. Table 7-2 compares the observed changes in irradiated Charpy impact properties with the predicted changes using Regulatory Guide 1.99, Revision 1. 5 The 50-ft-lb transition temperature shift for one base metal was in good agree- g ment with the shift that would be predicted according to Regulatory Guide 1.99. W The less-than-ideal comparison for the second base metal, although conservative, may be attributed to the spread in the data of the unirradiated material com-bined with a minimum of data points to establish the irradiated curve. Under these conditions, the comparison indicates that the estimating curves in RG 1.99 for low-copper materials and at low fluence levels are conservative for predicting the 50-ft-lb transition temperature shifts. The 30-ft-lb transition temperature shift for the base metals is not in as good agreement with the value predicted according to Regulatory Guide 1.99, although it would be expected that these values would exhibit better comparison when it is considered that a major portion of the data used to develop Regula-tory Guide 1.99 was taken at the 30-ft-lb temperature. The increase in the 35-mil lateral expansion transition temperature is com-pared with the shift in RT g curve data in a manner similar to the comparison made for the 50-ft-lb transition temperature shift. These data show a very - conservative prediction from the comparison of the observed and predicted 50- W f t-Ib transition data. All the transition temperature measurements for the weld metal are in poor agreement with the predicted shift. This can be attributed to the behavior , of the upper shelf region of the weld metal as compared to the normal weld i metals for which the prediction curves were developed. The abnormal behavior of the upper shelf causes the transition temperatures to be determined for a large shift at the higher energy levels. This being the case, it would not be expected that the current prediction techniques would apply to the weld metal. The data for the decrease in Charpy USE with irradiation showed a poor agree-ment with predicted values for both the base metal and the weld metal. How-ever, the poor comparison of the measured data with the predicted value is not 1 I
- 7-2 Babcock & Wilcox -
1
I I unexpected in view of the lack of data for low- and high-copper-content ma-terials at low to medium fluence values that were used to develop the estimat-ing curves. The large amount of scatter in the heat-affected zone data make a meaningful l evaluation and comparison impractical. This is the result of too few speci-mens and the high sensitivity of the exhibited material properties to the notch location in the heat-affected zone. Results f rom other capsules indicate that the RT esdmadng mms have t1DT greater inaccuracles at the very low neutron fluence levels (51 x 10 18 n/cm ), 2 This inaccuracy is attributed to the limited data at the low fluence values and to the fact that the majority of the data used to define the curves in RG 1.99 are based on the shift at 30 ft-lb as compared to the current require-I ment of 50-ft-lb. For most materials the shif ts measured at 50-f t-lb/35 MLE are expected to be higher than those measured at 30 ft-lb. The significance of the shifts at 50 ft-lb and/or 35 MLE is not well understood at present, especially for materials having USEs that approach the 50 f t-lb level and/or the 35 MLE level. Materials with this characteristic may have to be evalu-ated at transition energy levels lower than 50 ft-lb. The design curves for predicting the shif t at 50 ft-lb/35 MLE will probably be modified as data become available; until that time, the design curves for predict ing the RT
!iDT sm as ghen in Regdatory Me 1.% am wnsMerd adequate for predicting the RTg shift of those materials for which data are not available and will continue to be used to establish the pressure-tempera-ture operational limitations for the irradiated portions of the reactor vessel.
The lack of good agreement of the change in Charpy USE is further support of the inaccuracy of the prediction curves at the lower fluence levels. Although the prediction curves are conservative in that they predict a larger drop in upper shelf than is observed for a given fluence and copper content, the con-servatism can unduly restrict the operational limitations. These data support the contention that the USE drop curves will have to be modified as more re-liable data become available; until that time the design curves used to pre-dict the decrease in USE are adequate if not at all times conservative. I y I 7-3 Babcock 8.Wilcox
I' Table 7-1. Comparison of Tensile Test Results Elevated Room temp test temp test (SP0F) Unirr Irrad Unirr Irrad Base Metal - ANK-191 Transverse Fluence, 10 18 n/cm 2 (> 1 MeV) 0 3.12 0 3.12 E Ult. tensile strength, ksi 85.4 105.6 85.0 86.9 5 0.2% yield strength, ksi 63.1 91.3 55.7 56.0 Unifonn elongation, % 14 15 12 14 i Total elongation, % 30 25 28 37 RA, % 67 56 66 64 Held Metal - WF-209-1 Fluence, 10 18 n/cm 2 (> 1 MeV) 0 3.12 0 3.12 Ult. tensile strength, ksi 90.5 105.0 87.8 85.6 0.2% yield strength, ksi 74.9 90.6 67.4 57.8 L Uniform elongation, % 12 16 11 16 Total elongation, % 28 25 21 26 RA, % 63 57 52 56 I I I I 7-4 Babcock & Wilcox 1 I
l I l li I
= Observed Vs Predicted Changes in Irradiated Tahic 7-2.
Charpy Impact Properties i j Material Observed Predicted (a) Increase in 30-ft-lb trans temp, F lg l3 Base material - transverse l i ANK-191 32 39 AWS-192 19 32 j Heat-affected zone
- ANK-191 - --
. AWS-192 -- -- j Weld metal (WF-209-1) 64 206 l Correlation material (A-Il95-1) 39 87 Increase in 50-f t-lb trans temp, F j Base material - transverse i ANK-191 41 39 ) AWS-192 7 32 ) Heat-affected zone ] ANK-191 - - AWS-192 - - I .l Weld metal (WF-209-1) 315 206 Correlation material (A-1195-1) 33 87 i { Increase in 35-MI.E trans temp, F t j l Base material - transverse g ANK-191 7 AWS-192 4 39(b) 32 lI Heat affected zone ANK-191 AWS-192 Weld metal (WF-209-1) 54 206 ID} Correlation material (A-1195-1) 40 87(b) I Decrease in Charpv USE, ft-lb Base material - transverse 1 ANK-191 23 10 I AWS-192 2 7 Heat-affected zone ANK-191 - - AWS-192 -- -- Weld metal (WF-209-1) 17 22 Correlation material (A-1195-1) 34 31 (a)These values predicted per Regulatory Guide 1.99, Revision 1. Based on the assumption that MLE as well as 50 ft-lb transi-tion temperature is used to control the shif t in RT ET
- i 7-5 Babcock & Wilcox I
4 - _ _ - . _ _ -. . . . _ . . _ . _ . . . _ _ . . . _ _ _ _ _ _ _ _ _ . _ . . _ . _ _ _ __
I I i
- 8. DETERMINATION OF RCPB PRESSURE-I TEMPERATURE LIMITS I The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Oconee 3 have been established in accordance with the requirements of 10 CFR 50, Appendix G. The methods and criteria employed to establish operating pres-sure and temperature limits are described in topical report BAW-10046.6 The objective of these limits is to prevent nonductile failure during any normal operating condition, including anticipated operational occurrences and system hydrostatic tests. The loading conditions of interest include the following:
- 1. Normal operations, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests.
- 3. Reactor core operation.
The major components of the RCPB have been analyzed in accordance with 10 CFR 50, Appendix G. The closure head region, the reactor vessel outlet nozzle, anit the beltline region have been identified as the only regions of the reac-tor vessel, and consequently of the RCPB, that regulate the pressure-tempera-I ture limits. Since the closure head region is significantly stressed at rel-atively low temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first sev-eral service periods. The reactor vessel outlet nozzle also affects the pres-vure-temperature limit curves of the first several service periods. 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 the first several years of neutron radiation exposure, the RT * * "" #*E " "# * ~ NDT als will be high enough that the beltline region of the reactor vessel will I start to control the pressure-temperature limits of the RCPB. period for which the limit For the service curves are established, the maximum allowable pres-sure as a function of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, the outlet noz-zie, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressures. 8-1 Babcock & Wilcox
I The limit curves for Oconce Unit 3 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the 15th full-power year. The 15th full-power year was selected because it is estimated that the third surveillance capsule will be withdrawn at the end of the refueling c ycle when the estimated fluence corresponds to approxi-mately the 18th full-power year. The time dif ference between the withdrawal of the second and third surveillance capsule provides adequate time for re-establishing the operating pressure and temperature limits for the period of operation between the third and fourth surveillance capsule withdrawals. The unirradiated impact properties were determined for the surveillance belt-line region materials in accordance with 10 CFR 50, Appendixes G and H. For the other beltline region and RCPB materials for which the measured properties are not available, the unirradiated impact properties and residual elements, as originally established for the beltline region materials, are listed in 5 Table A-1. The adjusted reference temperatures are calculated by adding the predicted radiation-induced t2RT # ""
- NDT.
e pr cted NDT t.RT is calculated using the respective neutron fluence and copper and phos-phorus contents. Figure 8-1 illustrates the calculated peak neutron fluence at several locatione through the reactor vessel beltline region wall as a f unct Ion o f exposure time. The supporting information for Figure 8-1 is de-scribed in BAW-10100.7 The neutron fluence values of Figure 8-1 are the pre-dicted fluences, which have been demonstrated (section 6) to be conservative. The design curves of Regulatory Guide 1.99* were used to predict the radiation- E induced tRT values as a function of the material's copper and phosphorus NDT content and neutron fluence. The neutron fluences and adjusted RT valu s f the beltline region materials NDT at the end of the 15th full-power year are listed in Table 8-1. The neutron fluences and adjusted RT ET 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-10046P.6 Figure 8-2 shows the reactor vessel's pressure-temperature limit curve for normal heatup. This figure also shows the core criticality limits as required
- Revision 1, January 1976.
8-2 Babcock & Wilcox
I by 10 CFR 50, 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 16th ef fective full-power year. Protection I against nonductile failure is ensured by maintaining the coolant pressure be-low the upper limits of the pressure-temperature limit curves. The acceptable 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 crit-ical until the pressure-temperature combinations are to the right of the crit-icality limit curve. To establish the pressure-temperature limits for protec-i tion against nonductile failure of the RCPB, the limits presented in Figures 8-2 t hrough 8-4 must be adjusted by the pressure differential between the i I point of system pressure measurement and the pressure on the reactor vessel u controlling the limit curves. This is necessary because the reactor vessel is the most limiting component o f the RCPB. I I I i I I E 8-3 Babcock 8.Wilcox
- I i
I ,l , 1 Table 8-1 Data for Preparation of Pressure-Tenperature Linit Curves for Oconee 3, Applicable Throuch 15 EFPY
. . l ave r 1 . t n
"I"~A"d"J l heut rue. f l ., cot e at otrr r o.1 et 15 IIPY b
?:; T et e lusted M p7 at
' ' "' ir i sv, e g_ 15 EFPY, 'F end en 1% FrPY, F' Mat eri al ident t f t 4e 1 m g . ,
kl 7 F
- At 3 / 6T At ! !sT At 14T At i t4T At 1/ 4T Heat ha i Tspe rerl n Im elan JL cm degrece ! ,. t i ,,9 % At I /=T RDM4680 S A 509, C1 2 lxwc r nor t l e 5e l . - +M' O 11 C#!O 1.0 9F. l d 7. 04 E 17 55 27 115 A7
.sW S- 19 2 SA5G8. C1 2 Lpper shell -- -- - +20 0.01 U.Ull s.05119 9.2 7F l ? 15 !7 55 37
+20 s.02 0.014 6. O'>E ! H 9.2 7 E l 7 45 21 65 41 I ANK-191 SA508 C1 1 1.ower she ll -- - --
WF-200 Weld 1.'pper c ircus seam 121 -- Yes +20 ' (b) (b) 3. 03E 18 7. Ge E 17 !!7 56 117 76 WF -07 Weld Fid-c i rc ten se an ( 7 57,) -43 -- Yes +20'* ($) (b) 4. 05 E 18 9.2 7E 17 169 61 189 101
+2d * (b) (h) 9. 2 7 f.1, 110 - ! ?g kP-70 Weld Mid-c i e cian seas (2 5%) -63 - ha -- --
WF- l h 9- 1 WelJ leve r c i r c um ne am -249 -- Yes +20 (M (H 2. 2 < * :6 5.19F I S 9 4 29 .,
Per SAW-10046P. Marc h 1976.
Per BAW-1511P. Oct ober 19M0. Per Reputatory Guide 1. 99. Revi sion 1. I I C0 t O W to CT O O O X" 9'
=
O O X M M M M - - M
- . - . . . - - - - - - _ . . . _ . .. -_. . - -- . _ . . - ~ - -
M M M M M M M M M- M M M M Figure C-1. Predicted Fast !!eutron Fluences at Various Locations Through Reactor Vessel 1lall for First 15 EFPY (Oconee 3) 8 7.3 x 1018 nyt 7 - t 1 6 - A w 5 C
- 2 4.0 x 1018 nvt
- S INSIDE SURFACE 4 _
a M E i C c 3 - S 1/4T LOCATION i E E 2 m 8.8 x 10I7 nyt 1 n 3/4T LOCAT10N 4.0 x 1017 nyt g f OUlstBE SURF C,E b 0 "- i i i i i i i i i i I o 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 x Time. EFPY i I l
Figure 8-2. Reactor Vessel Pressure-Temperature Linit Curves f or Normal operation - heatup , Applicable fc r First 15 FFPY (Oconee 3) 2400 THE ACCEPTABLE PRE SSURE-TE MPE R ATUk E COMBINAll0NS ASSUMED RT MDi* I ARE BELUw AND 10 T t!! RIGHI Of THE LlMil CURWf(S) { g 2200 - eELillME REciOM I/4T 189 DO N0' INCLUDE T:1E PRESSURE DIFFERENTIAL BE iwE E N BELTLINE REGION 3/4T 130 THE POINT Of SYSTEM PRtSSURE MEASUREMENT AND THE CLOSURE HEAD REGION 60 PRESSURE ON THE REACTOR vfSSEL RE GION CONIROLLING 2000 - OUTLET N0llLE 60 THE LIMIT CURVI, OR ANY ADDITIONAL MARGlM OF SAFETY na FOR POSSIBLE INSTRUMENT ERROR. f 1800 - Pressure, Temp, Point psig _ F n> j 1600 - A 395 70 a B 625 266 C 625 273 1400 - D 660 274 E" E 2250 390 F 1190 374 APPLICABLE FOR HEATUP ce g 1200 - G 2250 430 RATES uP 10 100F/> , a F T f 1000 a 800 - 0 0
- BC 600 -
400 - -- A
, 200 -
m er i f I ' I I ' 8 0 8 8 320 360 400 440 O 40 80 120 160 200 240 280
- r P Reactor Vessel CO0lant Temperature F I
=:
O O mum num man man muu muu uma em mum uns uma mas uma men mum
Finure E-3. Reactor Vessel Pressure-Temperature Limit Curve :o r ?:o rmal Operat ion - Cooldown, Applicable 'or 15 EFPY (Oconee 3) 2400 ASSUMED PT F MDT. 2200 - BELTLINE REGION 1/4T 189 8ELTLIME REGION 3/4T 130 CLOSURE HEAD REGION 60 2000 - OuTLti =0ZztE 60 Y
PRESSURE, TEMP, 1800 -
y P01hT PSIG f
$ A 230 70 g 1600 -
B 7i0 220 7 C 1440 280 i - 0 2250 315 C e 1400 - c E i 1200 -
- S
$ 1000 -
(Aret: caste FOR C00t00=> g RATES UP TO 100F/h j U g 800 -
- =
i 8 600 - THE ACCEPTABLE PRESSURE-TEMPERATURE COMaiMATiONS ARE BELOW AND TO THE PlGHT OF cn 400 - m THE LIMIT CURVE (s). THE LIMIT CURVES DO NOT i
$ INCLUDE THE PRESSUFE DfffEdENTIAL BETwEEN THE E P INT OF SYSTEM PRESSURE MEASUREMENT AND THE 200 -
A F PRESSURE ON THE REACIOR VESSEL REGION CONTROL- j P LING THE LIMil CURit, OR ANY ADDITIONAL MARGIN
, , , , OF SAFETY joR POSSIBLE,1NSTRUMEMI FRROR.
8 40 80 120 160 200 240 280 320 360 M React 0r Vessel CO0lant Temperature, F
Figure 8-4. Reactor Vessel Pressure-Temperat ure Limit Curve for Inservice . eak and llydrostatic Tests, Applicable for first 8 EFPY (Oconee 3) 2600 ASSUMED RT f NDT' p 2400 - BELTLINE REGION 1/4T 189 BELTLINE REGION 3/6 T 130 2200 CLOSURE HEAD RfGION 60 OUILET N0llti (0
~
08 TE MP,
. PRFSSURE.
g POINI PSib d i 200 70 J 1800 - A g g 515 150
'd C 625 216 APPLICABLE FOR HEATUP AND COOLDOWN R A TE S UP S 1600 -
o 625 245
"- 10 100F/h (<50F IN ANY E 740 248 30-MIN. PERIOD) 5* F 2500 374 1400 - = =
a g 1200 - E
; 1000 -
U m E 800 - E 600 - f C D E CF 400 - THE ACCEPTABLE PRESSURE-TEMPERATURE COMBINATIONS arf BELOW AND TO THf o RIGHT OF THC LIMIT CdRVE(s). 1HE L l;4 d i CURVES DO NOT iMCLUDF lHf PRESSURE R 200 - olFFEstNTIAL BETWEEN THE P0INI OF SYSTEM PRESSURE MiASURIMINT AND THf 9* PRtSSURE ON THE REACTOR VESSEL REGION CONTROLLING lHF LIMil CURv(, OR ANY
, , ApolTIONALMARGlN OF SAFETY,FOR P O S $ 1 B L F, INSTRUMENT , ERROR. ,
g 40 80 120 160 200 240 280 320 360 400
- Reactor VeS$el CO0lant Temperature, F M M W M M M M M M M M M M M M M M
M M
I
- I I
- 9.
SUMMARY
OF REST'LTS 1 he analysis of the reactor vessel material contained in the second survell-lance capsule OCIII-B of the Oconee Nuclear Station Unit 3 pressure vessel surveillance program led to t he following conclusions:
- 1. The capsule received an average fast fluence of 3.12 x 10 " n/cm 2 1
(E
- 1 MeV). The predicted fast fluence for the reactor vessel T/4 loca-tion at the end of 4.1 EFPY is 1.08 x 10 18 n/cm 2 (E > 1 MeV).
- 2. The fast fluence of 3.12 x 10 18 n/cm2 (E > 1 MeV) increased the RT ET f the capsule reactor vessel core region shell materials a maximum of 41F.
- 3. Based on a comparison of the fast flux in the surveillance capsule to that at the vessel wall and an 80% load f actar, the calculated projected f ast fluence that the Oconee Unit 3 reactor pressure vessel will receive in 40 calendar years' operation is 1.56 x 10 n/cm2(E > 1 MeV).
- 4. The increase in the RT for the base plate material was in good agree-ment for one of the two base metals with that predicted by the currently used design curves of ARTg . versus fluence, liowever, for the other base metal and correlation material the design curves over-predicted the shif t s.
I 5. The increase in t he RT g for the weld metal was not in good agreement with that predicted by the currently used design curves of ARTg versus I fluence because of the unusual behavior of the material in the upper-shelf region.
- 6. The current techniques used for predicting the charge in Charpy impact upper shelf properties due to irradiation are conservative.
,l " 7. The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.
I 8. The thermal monitors indicated that the capsule design was satisf actory I for maintaining the specimens within the desired temperature range. 9_i Babcock & Wilcox
- g ,
I l I j 10. SURVEILLANCE CAPSULE REMOVAL SCilEDULE t I l 1 Based on the postirradiation test results of capsule OCIII-B, the following schedule is recommended for examination of the remaining capsules in the Oconee 3 reactor vessel surveillance program: Evaluation schedule 1 Est capsule Est d a te' (b) i L,apsule tluence, Estimated EFPY data ! ID(") n/cm2 Surface 1/4T available OC I I I-C 1.2 > 10 13 18 33 1984 { OCITI-D 2.2 x 10 19 32 58 1986 0CIII-E Standby -- -- -- i OCIII-F Standby -- -- -- i i ; j (a) Capsules contain weld metal specimens. l j (b)These dates do not represent the earliest dates that I data will be available for the materials that control the oper.iting limitations. Similar materials are in-E cluded as part of the B6W Integrated Reactor Surveil-lg lance Program, which will provide necessary data on a j timely basis. II i l t 1 'I I ,I i iI 10-1 Babcock & Wilcox iI 1
I I \ I \ lI. CERTIFICATION i The specimens were tested, and the data obtained from Oconee Nuclear Station Unit 3 surveillance capsule OCIII-B were evaluated using accepted techniques and established standard methods and procedures in accordance with the re-qui rements of 10 CFR 50, Appendixes G and 11. I
. .. f,
{' ,, l> l A. L. Lowe, Jr., ME.' Date Project Technical Manager I This report has been reviewed for technical content and accuracy. I y lll 'l
- d. . L Sm t i I / D6te Com onent Engineering l
I lI I I lI : 11-1 Babcock a.Wilcox
i 4 lI i a l !I l lI 4 r I I lI
- I i
1 !I APPENDIX A Reactor Vessel Surveillance Program - Background Data and Information I I I I !I
!I l a
iI i lI A-1 Cabcock & Wilcox
I
- 1. Katerial Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E-185-66, 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 Region The beltline region of Oconee Nuclear Station, Unit 3 was defined in accord-ance with the data given in BAW-10100A.7
- 3. Capsule Identification 1he capsules used in the Oconee 3 surveillance program are identified below by identification number and type.
Capsule Cross Reference Data ID No. Type OCIII-A V OCIII-B VI OCIII-C V OCIII-D VI W OCIII-E V OCIII-F VI
- 4. Spec imenti _per Survei1 lance Capsule i See T.iblen A-2 and A-3.
I I I I I I A-2 Babcock 8. Wilcox
M M M M M M M m M M M M M M _ _ _ _ _M
-M _ _ _M. _ _ _ _ .M Table A-1. Surveillance Program !bterials Selection Data for Oconee 3 Charpy data, C ,
Core P Transverse b'IIII"* "IdPI*n, sterial-- - ----- -ident. RT '"III' I region to weld L ngitudinal 50 NDT' T t. Ilar No TYPE location CL, cm NDT' _9 10F, fr-Ib ft-lb 35 MLE l'S E F Cu P S Ni RD.4-4680 SA503 C1 2 Lower nozzle -- -- 111, 117, 103 -- - - -- 0.13 0.010 0.009 0.80 belt 112, 101, 49 l N45-192 SA508 C1 2 Upper shell -- 40 76, 82, 46 -- - -- -- 0.01 0.011 0.015 -- i 85, 77, 78 l i Ah-191 SA508 C1 2 lower shell -- 40 39, 50, 66 - - -- -- 0.02 0.014 0.012 -- 49, 83, 43 WF-200 Weld Circum seam 123 -- Jn, 35, 26 -- -- -- -- 0.26 0.010 0.015 - l VF-67 Weld Circum seam -63 -- 29, 35, 30 -- - -- -- 0.27 0.014 0,017 -- l (75%) ! WF-70 Weld Circum asam -63 -- 39, 35, 44 -- - -- -- 0.27 0.014 0.011 -- (25%) l WF-169 Weld Circum seam -249 -- 36, 43, 42 - -- - - 0.106 0.014 0.013 -- I I L.2 l 1 1 1
\
I 1 I W r m i T j O
, O i O i u-l 90 l q ,
W O I x ' i 1 } ,
1 l l I Table A-2. Materials and Specimens in Upper Surveillance Capsules OCIII-A, OCIII-C g nd OCIII-E ___ No. of specimens Mate rial description Tensile Charpv We ld me t a l , W F- 209- 1 2 12 lleat-affected zone (il AZ) llea t A -- AFK- l91, longitud. 0 12 Baseline material l Ilea t A - ANK-191, longitud. 0 9 Ile a t A - ANK-191, transverse 2 12 Ileat B - AWS-192, transverse 0 9 l 3 Tot al per capsule 4 54 ' I Table A-3. Materials and Specimens in lower I Surveillance Capsules OCIII-B, g' _ OCIII-D, and OCIll-F g 1 No. of specimens Material descript ion Tensile Cha rpy. Weld me t al , WF-209-1, longitud. 2 12 Weld, ilAZ lleat A - ANK- 191, longitud. 0 12 ilea t B - AWS-192, longitud. 0 6 Baseline material llea t A - ANK-191, transverse 2 12 t lleat B - AWS-192, transverse 0 6 , l g Correlation llSST plate 02 0 6 W Total per capsule 4 54 l I I, I' A-4 Babcock & Wilcox
I i l Figure A-1. Location and Identification of Materials Used ( in Fabrication of Oconee Unit 3 Reactor Pressure Vessel I -
,I t L
I (
\
RDM-4680 (Lower Nozzle Belt) I [ WF- 200 g AWS-192 (Upper Shell) I k - WF-67 (75% ID)
~ WF-70 (25% OD)
ANK- 191 (Lower Shell) I I l n WF- 169 417525-1 (Dutchman) I t I I A-5 Babcock & Wilcox
l I I I il I i il I il } i i i !I I .Ii.
- APPENDIX B Preirradiation Tensile Data I
I I lI I i il i i lI i l 15 i, )g B-l Babcock 8 WilCOX ig i L - _ _ . _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ . _ __.
I Tabic B-1. Preirradiation Ten.aile Properties of Shell Forging Material, lleat ANK-191 Test Red'n S t ren gt hnpsi Elon ga t lon , % Specimen temp, o f area, i No. F Yield Ult. Unif. Total % Transverse JJ-603 RT 70,070 85,060 13.4 30.7 66.9 l 604 RT 60,940 86,520 14.8 31.1 68.4 W 608 RT 58,310 84,730 13.3 29.3 65.1 Mean RT 63,110 85,440 L3.83 30.37 66.8 Std dev'n 5,040 780 0.68 0.77 0.55 JJ-607 580 56,200 85,300 12.26 28.6 66.1 l 610 580 56,220 85,580 13.82 28.6 61.I B 611 580 54,600 84,150 9.6 28.6 67.5 l Mean 580 55,670 85,010 11.89 28.6 66.57 l Std dev'n tS ,0 620 1.74 0 0.66 W I i I. I l Table B-2. Preirradiation Tensile Properties of Weld Il Me ta l - Mgit ud inal, WF-209-1 _ Test Red'n Specimen temp, "b "E^ ' "' of area, No. l[ F Yield Ult. Unif. Total % 5[ ! JJ-002 RT 74,630 90,460 12.5 29.3 63.2 g; l 004 RT 73,540 89,110 13.1 27.1 63.6 5; l 018 RT 76,720 91,910 11.9 27.9 62.0 l Mean RT 74,960 90,490 12.5 28.1 62.9 i Std dev'n 1,320 1,140 0.49 0.91 0.68 t JJ-001 580 67,540 87,700 10.83 20.7 52.9 g 011 580 66,480 88,300 11.54 22.1 013 580 68,210 87,620 10.33 21.4 53.0 50.3 5 Mean 580 67,410 87,870 10.9 21.4 52.07 g std dev'n +5 712 303 0.50 0.57 1.25 gL l Il llf B-2 Babccck 8. Wilcox a g,
- ,n _ _ _ . . - - , - . _ - - . _ _ - - - - - - . - _ . - - - . . _ . - -- ,
l I i <I I I i i l ,I I il i fg APPENDIX C E Preirradiation Charpy Impact Data I .i I 1 t ,!I I 1 1 I !I I i I !I l c-1 Babcock & Wilcox !I
1 I Table C-1. Preirradiation Charpy impact Oa t . for Shell Forging Kit erial - Transverse Orientat ien, llea t AWS-192 Test Absorbed I.a t e r a l Shear 5 Specimen
~
t e rnp . energy, expanston, tracture, No. F tt-lb 10~I in. 7 KK 665 360 98 77 100 1 641 360 10.1 70 100 l 614 159 109 12 100 l KK h/6 282 111 Il 100 l 677 280 106 /1 100 674 279 104 10 100 KK 619 200 115 71 100 627 199 108 68 100 669 199 115 /2 100 KK h81 140 119 71 100 659 140 92 71 90 679 1 19 113 67 100 684 1 19 114 71 100 KK h?2 81 64 51 8 E g 651 80 83 62 25 648 80 99 /1 35 l KK h/5 61 80 65 40 g' h/1 60 80 64 ') 5 5 678 60 64 54 25 h10 60 50 44 15 KK hl4 '40 51 4) 6 W h 10 40 40 36 'l 6 19 40 59 50 12 FK h5/ 6 17 '10 si ' l 6Li O 51 42 4 666 0 21 18 (1 l KK 681 -59 2 1 0 l 682 -60 10 1 26 680 -60 'l 2 () I i ! I I (,-y Babcock & Wilcox
I I lEg Table C-2. Preirradiation Charpy Impact Data for Shell l Forging Material - HAZ, Transverse Orienta-tion, Heat AWS-192 l Test Absorbed Lateral Shear Specimen fracture, ]lE jg No. temp, F energy, ft-lb expansion, 10-3 in. % l l KK 325 361 135 65 100 336 360 153 77 100 l 318 358 128 63 100 j ig KK 343 278 118 76 100 i 339 278 125 76 100 lg 344 277 132 76 100 )g KK 333 199 65 37 100 lg 310 199 122 66 100 j 313 197 78 53 100 I KK 346 139 131 77 100 lgg - 337 139 132 67 100 j 348 139 121 71 100 KK 312 80 91 46 94 ll
- E 307 79 76 41 85
) 335 49 63 36 98 ll != KK 308 316 40 40 141 72 67 42 100 35 l 327 40 66 40 45 l KK 341 20 93 70 40 i 340 20 104 77 65 345 19 82 64 30 l l KK 317 0.5 106 64 45 1 324 0.1 43 31 18 326 0.1 43 32 40 KK 309 -40 30 22 20 j 330 - 39 80 51 28 lg KK 342 -39 27 21 2 !g 347 -40 52 35 5 I 338 -40 38 28 3 l lI I c_3 Babcock 8.Wilcox
I Table C-3. Preirradiation Charpy Impact Data for Shell l ay Forging Material - Transverse Orientation, _ lieat Aa;g_19 j ____ Te s t Absorbed Lateral Shear l Specimen temp, energy, expansion, fracture, No. F ft-lb 10-3 in. % JJ 651 362 172 75 100 l 640 361 144 66 100 621 357 128 72 100 JJ 701 284 1 16 73 100 705 283 140 73 100 700 277 135 71 100 JJ h15 200 150 71 100 625 200 153 61 100 637 200 141 73 100 JJ 704 140 130 74 88 703 140 126 74 85 l 702 140 118 71 70 W
- JJ 076 80 123 71 55 611 79 130 78 65 l 657 79 129 74 60 =
JJ 670 41 77 62 5 l 6hS 4L 71 58 4 612 41 86 66 10 JJ 708 25 68 55 5 g 706 24 74 62 10 g 707 24 57 46 8 I 675 21 58 48 6 641 20 47 41 4 JJ 622 0.9 35 26 1 692 0.9 56 44 2 680 0.5 t8 15 0 JJ 698 -39 3 2 0 697 -40 8 8 0 g 699 -41 0 l 14 12 5 I I I I r-4 Babcock 8. Wilcox
1 l it i l i Table C-4. Preirradiation Charpy Impact Data for Shell Forging Material, HAZ, Transverse Orienta-tion, Heat ANK-191 l Test Absorbed Lateral Shear Specimen temp, energy, expansion, fracture, ll j5 _ No. F ft-lb 10-3 in. % JJ 313 360 86 53 100 j 315 359 106 6; 100
- 335 358 73 54 100 1
)g JJ 328 201 102 54 100 g 323 201 92 59 100 308 200 83 49 100 g JJ 322 80 76 42 85 g 319 80 98 51 100 344 80 74 45 80 ll JJ 357 40 66 44 85 j5 332 40 57 33 45 1 339 40 55 30 55 i l JJ 301 20 46 29 70 1 304 20 36 27 30 4 i JJ 331 0 35 29 35 1 324 0 33 26 23 f 337 0 47 33 35 i j JJ 424 -59 27 20 2 i 4^' -60 56 44 4 !l= 42' -61 67 35 6 1 il !I I !I 1 !I 4 b !I lI Babcock & Wilcox jg C-5 1
I Table C-5. Preirradiation Charpy Impact Data for Weld Metal, WF-209-1 Test Absorbed Lateral Shear
- Specimen temp, energy, expansion, fracture, No. F ft-lb 10-3 in. %
JJ 063 359 57 49 100 083 357 58 48 100 g 089 357 65 53 100 5 JJ 091 201 57 46 100 053 201 63 50 100 g 084 199 78 57 100 g JJ 093 140 60 46 100 060 140 69 52 100 065 140 61 49 100 JJ 092 110 64 46 100 088 110 58 46 100 l 077 109 47 37 96 W JJ 064 80 61 40 98 095 80 38 32 95 081 79 48 35 85 JJ 090 40 28 28 10 069 40 33 29 15 075 40 21 20 12 JJ 061 -40 17 16 5 050 -40 16 14 2 057 -39 19 18 6 i I< I-I l
\
I I c-6 Babcock & Wilcox l
l
!I Figure C-1. Charpy Impact Data From Unirradiated Base Metrl 100 I I I I I I i T 6 'I 5 .
I a 75 y .
- 2 l V 50 --------- --------------------
i '
=
5 5 25 - -
' I I I I I I I I 0
.08 j j ; , j g ;
l 9 l .i ,g ( t - I ; .; e .06 - -
=
2 . j.04-l , l
,5 2
i .02 - 5 I I I I I I I I I I 0 i 200 i i i , ; ; ; ; , l , ; l DATA SUM %RY l 180 - T,g, +20F _ lg Tcv (35 mtt) +5F
+12F
!g TCV (50 ri-ts) 160 - Tcy(30 FT-ts) "" , j Cy -USE (avc.) 148 FT-L83 .
; m 140 gy +20F . _ ; f NDT a C 9
- i, ,
e
','120 " .
8 -
. a 100 -
i
- I
' 80-V
- i S I 60-
;g ,
40 ] I MArta AL SA508 CL2 - l 20
- MONE -
J , FLutNCE { HEAT No. ANK-191 l 0 I I I I I I I I ' ' ' l -00 -40 0 40 80 120 160 200 240 280 320 360 400 ] Test TEmetaarunt, F C-7 babcock & WilCOX i
I: 1 rigure C-2. Charpy Impact Data From Unirradiated lleat-Affected Zone >!aterial ICOI I I i f ~1 i T i i i 7
" 75 -
l 0
- 5 a
j 50 _ _ _ _ _ _ _ p___._______._________.______ E
, m Y
- m 25 - ,
l 1 1 1 I I I I I I I 0
.0a j ;
I
; , j g g i ; ; . i =
i .0s- . -
. i i
.04 -
g - . - _ _ _ _ , _ _ , _ - - - _ _ _ _ _ _ _ _ _ _ e , S d .02 -* - 5 5 I I I I I I I I I I I 0 2% i i ,
; g g ; ; , , g DATA
SUMMARY
180-T Not + OF _ Tcv (35 mtt) +37F 160 I,Y (50 ri-La) +32F Icy (30 ri-ta) -30F Cy-USE (Ave,) 92 FT-Las '
= 140 RT +20F -
l C f120-E E
$100- ,
g a . g, 5 * . i 80- _ l 3 . 9 t
~ 60- .
40- .
- MATrntAL SA508.CL2 ORIENTATION TRANSVERSE 20- FLUENCE NOME kui No. ANK-'93 I I I I I _._1 I I i i i 0
-80 -40 0 40 80 120 160 200 240 280 320 360 400 i.
Test TEMPERATURE, F i c-8 Babcock & Wilcox ,
I .i i
'I
- Figure C-3. Charpy Impact Data From Unirradiated Base Metal i
100 g i j g - l g i i i w 75 - - J O j
= 50 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
i . 1 3 w Js 25 -
- I ! I I I I '
] 0 ) .08 ;
; g g g ; g g g l 4
- I ? ! .
5.06- - 2 *
= .
4
- i " .04- . -
! 5 . - _ _ _ __________ ____ ___.-___
-5 f.02 .
I :- i 3 ! I I I I I I I I I I I O 1 4 200 i i i ; ; ; ; g g g i i DATA
SUMMARY
! 180- T,g, +20F _
j Tcy (33 mtt) 18F 160 TCV (50 n-La) 37F - j Tcy (30 n-ts) -9F l Cy -USE (ass) 112 FT/LBS 4
' 140 +20F --
RT'DT P j h" j -
- -5120- .
f k . ' 100- . '. - r, . i d 80- . - C 4 g r . l ~~ 60+- , - i 40- * - l'
- NTragat SA508,CL2
[ ~ ~ - ~ -- ~ - - - - - -~'~~ ~ ORIENTATION TRANSVERSE -- j 20 FLut=cg NONE - HEAT No. AWS-192 0 I I I I I ! I I ' ' '
-00 -40 0 40 80 120 160 200 240 280 320 360 400 j Test TEMPERATURE, F i
C-9 Babcock & Wilcox
I Figure C-4. Charpy Impact Data From Unirradiated lleat-Affected Zone Material 1* I l t - i i i i i i
" 15 - -
E 2 y 50 - - - - - - - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - e *
- a J 25 -
e
- I I I I I I I I I I "
0
.08 g g g g j ; j y j ;
. , i ,
i , .
*
- i i f,g _
" 04 - * . -
3.
~
-___.2--________-
t R $ d .02- * - S
.3 0 I L I I I I I I I I I 2 ' ' 3 I I i l l l l l DATA
SUMMARY
180- T,3, +2 ' Tcy(35 MLE) -40F E W 160 ICV (50 st-ts) -32F -- T.v s (30 FT-La) -56F . Cy -USE (ava) 125 FT-Las
=
J 140 RT +20F - NOT
? * .
\ s .
$120- e
=
1 y * . 3 100- - 5 f 80- , . 0 , I * *
.' 60- .
. - 2- - - - - - - - - _ - _ _ - - - - . - - - - - - -
~ ~
e 08,M.2 MATERIAL l e OnithTAt m TRAN5 VERSE -- l 20~ FLornce NONE - Hr.Ar No. AWS-192 0 I I I I I I I I ' ' '
-80 -40 0 40 80 120 160 200 240 280 320 360 400 Test TtartnAtuar, F C-10 Babcock 8.Wilcox
!I !I i Figure C-5. Charpy Ir.1 pact Data Frora Unirradiated k' eld Metal i IN I I -~' T ~ ~~~;i
- i 1 i 1 6 1 7s -
'I e C
; 50 - _ _ _ _ _ _ _ __ _ _ _ __ ___ .___ _ _ _ ______-- i 1
l
- J 25 -
l 0 I ' i 1 I i ! I I I I I
. 08 I I I I i 1 1 I l l I i
I ) E .06-
- p .l $ .
3
- I 1
4 i '" .04 i 5 e {.02-I I I I I I I I I I I 0 l 100 i i i ; ; ; j g l l J l jlg is --- DATA
SUMMARY
-20F l 90 - T,37 _
+sIF Icv (35 mot) l gop Tcv (50 F r-ts) +s 5 F _
'cv (30 n-ti) W .
l Cy-USE ( Avo) 66 FT-Lp_ i 10 RT +25F *
- y wnT -
4 C .
- 1 ,
i $ FQ~ ! C . 1 5 j j50___________ ______.________________ 65 40 - - C !
- r .
I - 30 __----_ __ ________________--- 1 1 j
,I 20-MATERIAL Onigniaison WELD METAL TRANSVERSE 10 F Lutact NOME - ! HEAT No. WF-209-18_
0 1 1 1 L i 1 ! 1 i _ _2 i j -C0 -40 0 40 80 120 160 200 240 280 320 360 400 , itsi TEMPCRATURE, I ' l l C-11 SabC0Ck 8. WilCOX
I } k l Figure C-6 Charpy impact Dat a From t'nirrad iat eil l Correlation Material i 100; J l -{- - } - - -} g g ; g - p- ) ", 15 - i i C _ . _ . _ . _ _ _ _ _ _ _ - _ _ _ _ _ _ . _ _ . - _ _ _ .
- sc - - - - - _
" l l 0 J :s -
0 I I I I I I I I I I E
'08 I I I I I ' l ~I I
I I i
, , M
' .06-- -
}.04-4 j .02-l
.n . E 3
o_ l 1 1 l 1 L i I i I I
' ' ' I I I i 1 I T I 2*J , PATA 511MMARY I -
180~ I gy icy(% ntr) ( FT4'} ' - -- 160" CV Icy(50 st-te) 56F Cy -U5E (avs) 130 FT-t63 140 pg { I?0p
~
h l%- W 5 u", 80v-C. I
~ 60-g ,g,,,g SA533,Gr81 t
Onitulation 20
~
FLusscr Nout ___ HEAT No, A-1195 1 0 I I l-.-- I I I I-. -- I 1 i L
- l'0 40 0 40 80 170 160 700 240 280 320 360 400 Test itMPERafudt, f C-12 babcock & WilCOX
i lE lI i 'I I i l 1 I I APPENDIX D Fluence Analysis Procedures I I I I I . I I , I ! D-1 Babcock & Wilcox
Il l
- 1. Analyt ical Met hod Energy-dependent neutron fluxes at the detector locations were determined by a discret_ ordinates solut ion of t he Bolt zmann t ranspo-t equation with the I
W two-dimensional code DOT 3.5.3 The Crystal River 3 and Oconee 3 reactors were nh >d e : ml f rom the core out to the primary concrete shield in R-theta geometry (based on a plan view along t he core midplane and one-eighth core symmetry in the azimuthal (theta) d i me ns io n ] . Also included was an explicit nodel o f a <:urveillance capsule assembly in the downcomer region. The reactor model conta ? ned t he following regions: core, liner, bypass coolant, core barrel, in-let coolant, thernal shield, inlet coolant (downcemer), pressure vessel, cav-ity, and concrete shield. Input data to the code included a pin-by-pin time l averaged power distribution, CASK 23E 22-group microscopic neutron cross sec- e tions", Sa order angular quadrature, and P 3 expansion of t he scattering cross sectton matrix. Reactor conditions - power distribution, temperature, and pressure - were analyzed over the irradiation period. Because of computu storage limitations, it was necessary to use two geometric models to cover the distance f rom core to primary shield. A boundary source output from model A (core through dowucomer regiou) was used as 2nput to model B (thermal shield te primary shield), which included the capsule assembly. In those cases where the capsule " shadowed" the maximum flux location in the pres-i sure vessel, a model C (model B without a capsule assembly) was used to obtain vesse' flux unperturbed by the presence of a capsule. For a reactor without surveillance capsules, additional models A and C were calculated. Thus, the = effeet of the specific power distribution in that reactor on vessel fluence was accounted for. Flux output f rom the IX)T3.5 calculations required only an axial distribution correction to protide absolute values. An axial shape factor (local-to-aver- W age axial flux ratio) was obtained from fuel burnup distributions in those peripheral fuel assemblies nearest the capsule location. This procedure as-sumes that axial fast flux shape in the capsule and the pressure vessel is the same as axial power distribution in the closest fuel assemblies. In the 17 7-FA reactor geomet ry this is considered to be a conservative assumption because axial shape should tend to flatten as distance from the core increases. This factor was 1.1 averaged aver an elevation interval corresponding to the I D-2 Babcock & Wilcox
I .__ : l i 1 i i i i capsule length for the capsule exposures. For t he Oconee 3 reactor vessel calculations, the axial f actor has a maximum value of 1.17. tI The calculation described above provides the neut ron flux as a function of I energy at t he capsule position. The cr.lculateo Ilux is used in the following equations to obtain calculated dosir.ater activities D (pCi/g): I D t g Tn3.7 x 10 g i E n( M j=1 j
-)1 t -A g(T-T )
1 where i N = Avogadro 's number, I lg A + atomic weight of t arget material, n, n
- 5 l
r = either weight f raction of target luotope in nth material 1 or fission yield of desired isotope, I
- g o"(E) = group-averaged cross sections for material n (listed in Ig Table D-3),
i 1 4(E) = group-averaged fluxes calculated by DOT analysis, F = f raction of full power during j th time interval, t . A = decay constant of ich material,
=
t interval of power history, T = sum of total irradiation time, i.e., residual time in re-actor, and wait time between reactor shutdown and count ing, T = cumulative time from reactor startup to end of jth time z period, i.e., T = t . k=1 i 'l The normalizing constant C* Is cotained as the ratio of measured activity to D. With C specified, the neutron fluence greater than 1 MeV can he calcu-g lated from l 1s MeV M l ?(E > 1.0 MeV) = C [ 4(2) [ Ft ! E=1 j=1 , where M is the number of irradiation time intervala; the other values are de-1 j fined. Although this normalization is strictly correct only at the capsule location, it was considered applicable to all locations in the host and donor I reactors because of the similarity between reactors and calculational models. I ... _ 4 E E E E $ . Il D-3 Babcock & Wilcox
1 I B&W 177-FA reactors have essentially the same configurations and materials. In the calculational model, pressure vessel ar.d capsule are separated by only , 15-cm of water and it is very unlikely that any significant change in accuracy would occur over that distance.
- 2. Vessel Fluence Extrapolation For up-te-date operation, fluence values at the pressure vessel surface are calculated as described above. Fluence extrapolation for future operation is required for prediction of vessel life based on metallurgical correlation and reactor operational conditions. W For tnis analysis, because of the lack of the anticipated balfla flux informa-tion, the projected pressure vessel fluence values at 15 and 32 EFPY have each been calculated as the sum of three fluence values: (1) the fluence value pre-viously reported f rom capsule OC3-A, (2) the fluence value calculated as the result of cycles 2, 3, 4, and 5, and (3) cycles 2, 3, 4, and 5 values extrapo-lated for the time required to reach 15 and 32 EFPY, Current and extrapolated reactor vessel fluence calculations are shown in Table D-2.
I I I . I l I I I . I D-4 Babcock & Wilcox
I il
- i Table D-1. Dosimeter Activations A B Measured Calculated C = A/B
' activity, g activity, Normalization Reaction LCi/g pCi/g constant 5"Fe(n,p)'"Mn 921.2 1212.63 0.76 U *Ni(n,p) s eCo 1952.0 2596.8 0.75 23e U(n,f) "Cs 2.986 3.295 0.91(b)
- 2 "Np(n,f)I37Cs 10.64 20.645 0.90 238 3 U(n,f)1 Ru 121.9 143.02 C.85 237 l Np(n.f)l"3Ru 688.8 755.2 0.91 i
i l (a)The average of the neasured dosimeter activity values ig has been corrected to account for the activity remain-jg ing from the first irradiation to facilitate the com-j parison with the calculated value for the second irradi-ation, lE j 5 (b)A valve of 0.91 was selected as a conservative normali-i zation constant for subsequent calculations. 'I !I .i Table D-2. Reactor Vessel Fast Fluence (E > 1.0 MeV) Calculated Cycle Cumulative fitat , 10 10 fluence, fluence, i Cycle E EFPY EFPY EFPD n/cm2 -s 10 18 n/cm 2 10 18 n/cm 2 g W I 1.3 1.3 478 1.39 0.57 0.57 2, 3, 4, and 5 4.1 2.8 1020 1.55 1.37 1.94 Assumed future 15 10.9 3981 1.55 5.34 7.28 ) cycles are !g sin 11ar to 32 17 6209 1.55 8.32 15.6 g 2, 3, 4, and 5 I lI !I I D-5 Babcock & Wilcox i _ _ _ _ m m _ _ _ _ _ - - - - -- - - - - - - -. _ _ -- - , - - -
l lI f 1, I I t !I I II lI i 1 i lI 1 I APPENDIX E 3 Capsule Dosimeter Data }I I l t 1 i l I 1 !I i lI I lI E-1 f babcock & WilCOX _ _ _ . _ _ _ _ _ _ _,__,-e. _ , _ -- - - - - - - -.-------Y'7* CN-
i I Table E- 1 lists the composition of the threshold detectors and the cadmium thickness used to reduce competing thermal reactions. Table E-2 shows the ' cycle 1 measured activity per gram of target material (i.e., per g, ram of ura-nium, nickel, etc.) corrected for the wait time between irradiation and count-
- ing. Activation cross sections ior the various material were flux-weighted I 235 with a U fission spectrum (T ble E-3).
1 l I Table E-1. Detector Composition and Shielding j !!onit o rs shielding Reaction % isotopic 10.38 U-Al 23e U(n , f) 1 Cd-Ag 0.02676 in. Cd 99.27 237 i 1.44 Np-Al Cd-Ag 0.02676 in. Cd Np(n,f) 100 58 Ni Cd-Ag 0.02676 in. Cd Ni(n,p)58Co 67.77 59 0.56% Co-Al Cd-Ag 0.040 in. Cd Co(n,Y) Co 100 0.56% Co-Al None 5'Co(n,y)6 Co 100 Fe None 5"Fe(n ,p) 5"Mn 5.82 i I I I l I I i I I I E-2 Babcock & Wilcox
1 Table E-2. Capsule OCIII-B Neutron Dosimeters ( ) Postirrad. Nuclide pCi/g of LCi/g of Monitor wt, g Reaction Nuclide act, LCi material target (c.d) BD-5 23t 23e "' 5 U-Al 0.0564 U(n,f)FP Zr 0.5254 9.32 90.4 103 Ru 0.6666 11.8 115 137 Cs 0.02302 0.408 3.96 1""Ce 0.2941 5.22 50.6 106 Fu 0.1405 2.49 24.2 i 237 Np-Al 237 '5 0.0536 Np(n,f)FP Zr 0.4319 8.06 560 1 3 Ru 0.4545 8.48 589 137 Cs 0.01744 0.325 22.6 0 1""Ce 0.1840 3.43 238 1 'Ru 0.0S673 1.62 112 se 58 Ni 0.1289 Ni(n,p)seCo Co 164.3 1280 1,880 ' 6U Ni(n.p)6 Co 8 Co 0.3842 2.98 11.4 C.' *1 5'Co(n,y)6 Co 60 ) (0.625 in.) 0.0197 Co 25.44 1290 231,000 1 in Cd .i l l
- Co-Al (0.5 in.) 0.0157 5 Sco (n ,y ) 6 Co 'U Co 20.82 1330 237,000
- Fe 0.1605 s"Fe(n.p)5"Mn 5"rin 8.297 51.7 888 58 Fe(n,y)5'Fe j n3 5'Fe 25.07 156 47,300 j @[ *d'e e g BD-6 o -
23s 23e U (n , f) FP '5 { l[ U-AL 0.0458 Zr 0.3278 7.16 69.5 103 l 35 Ru 0.3994 8.72 84.7 i g o 1 S i i x Ru 0.08561 1.87 18.2 l r l 1
l 1 l l Table E'. (Cont'd) L Postirrad. Nuclide ;.Ci/g of LCi/g of Monitor wt, g Reaction Nuclide act, :.Ci material target (c.d) l , r j 137 Cs 0.01460 0.319 3.10 : 1"'Ce 0.1F60 4.06 39.4 237 237 Np-Al 0.0644 Np(n,f)FP 2' 5 Zr 0.4539 6.84 475 l 1C3 Ru 0.4896 7.37 512 l 106 Ru 0.08533 1.28 89.2 137 Cs 0.1686 0.254 17.6
""Ce 0.1900 2.86 199 58
, Ni 0.1345 Ni(n,p)5'Co se Co 127.8 950 1,400 6C '0 l Ni (n.p) 6 0Co Co 0.3005 2.23 8.54 1 - [ Co-Al (0.625 in.) 0.0208 5'Co(n,y)*0Co 'C Co 23.26 1120 200,000
! in Cd ,
59 Co(n.y)6UCo j Co-Al (0.5 in.) 0.0156 6C Co 18.73 1200 214,000 Fe 0.0208 5"Fe(n.p)5"Mn 5"Mn 6.284 42.9 738 se Fe(n,y)5'Fe 5'Fe 21.82 149 45,200 J BD-7 l 238 U-Al 0.0626 23eU(n,f)FP '5 Zr 0.8450 13.5 131 3 3 Ru 1.052 16.8 163 03 106 m ku 0.2162 3.45 33.5 E[ 137 o Cs 0.03174 0.507 4.92
$$ l""Ce 0.4552 7.27 70.6
*N P -Al 0.0771 37 Np(n,f)FP '5 Zr 0.9396 12.2 846
-.E FI 1 3 1.001 o Ru 13.0 902 X
l M M M M M M M M M M M M M M l Table E-2. (Cont'd) Postirrad. Nuclide LCi/g of LCi/g of Ponitor wt, g Reaction Nuclide act, LCi material target (c,d) 166 Ru 0.1944 2.52 175 l37 Cs 6.3405 0.442 30.7 l'"Ce 0.3951 5.12 356 58 se N1 0.1320 Ni(n.p)seCo Co 223.5 1690 2,509 6U 'U Ni(n.p)6UCo Co 0.4507 3.41 13.0 Co-Al (0.625 in.) 0.0160 5'Co(n,y)6"Co ' Co 28.96 1810 323,000 in Cd ' Co-Al (0.5 in.) 0.0195 5'Co(n,y)6 Co 6 Co 34.77 1780 318,000 Fe 0.1543 5"Fe(n,p)S"Mn 5"Mu 10.44 67.7 1,160 se m Fe(n,y)5'Fe 5'Fe 35.63 231 70,000 d BD-8 23e '5 U-Al 0.0565 2 s eU(n,f)FP Zr 0.5769 10.2 99.1 1 3 Ru 0.7295 12.9 125 ! 106 Ru 0.1517 2.68 26.1 137 Cs 0.02202 0.390 3.78 1"lce 0.3249 5.75 55.8 237 237 Np(n, '5 Np-Al 0.0548 f) FP Zr 0.5209 9.50 660 as 1 3 i
- Ru 0.5934 10.8 752 Ef l06 o Ru 0.1032 1.88 131 137
$E Cs 0.01843 0.336 23.4 2"
E Ce 0.2156 3.93 273 Ni 0.1 se se [f Ni (t!, p) s eCo co 172.0 1370 2,030 l 1
l I J j l i l ? i I Table E-2. (Cont'd) t Postirrad. Nuclide aci/g of ..Ci/g o' Monitor wt, p Reaction Nuclide act, material (b) target tc,d)
- t. C i_
63 Ni(n,p)*2Co 6C Co 0.3556 2.84 10.8 i Co-Al (0.625 in.) 0.0199 5 Cc(n,y)'0Co 60 Co 19.35 13.30 232,000 i in Cd ' l Co-Al (0.5 in.) 0.0145 5'Co(n,y)*UCo 8C Co 25.90 1300 238,000 l Fe 0.1577 5"Fe(n p)5"Mn 5"Mn S.798 1370 2,030 ' 5'Fe(n, )S*Fe 5'Fe 26.86 2.84 10.8 1 1 i (a) Analyses performed at Lynchburg Research Center. (b)This column is the disintegration rate per gram of wire using the postirradiation weight . This 6C column is the disintegration rate per gram of target nuclide, viz., 238U, 237 Np, 58 N1, i Ni, 5'Co, 5.Fe, 5*Fe. i 1 l (d)The following abundances and weight percents were used to calculate the disintegration rate per gram of target nuclide:
'3'U 10.38 wt %; 99.2% isotopic t
i 237 ) Np 1.44 wt %; 100% isotopic t 58 , Ni 100 wt *; ; Ni 67.77% isotopic, 6C Ni 26.16% isotopic I Co 0.56 wt i; 5'Co 100% isotopic i I Fe 100 wt %; 5'*Fe 5.82% isotopic, 58 Fe 0.33% isotopic, to cu C1" O O O llEl* 9' I
=
0 C x mas an en am == == ==
== == == == == == == == == m. ==
i ) i l t
- Table E-3. Dosimeter Activation Cross Sections ss sections, barns / atom Energy range,
= 237 2n se gg 54 G MeV Np U Fe 1 12.2-15.0 2.323 1.050 0.4830 0.4133 1 2 10.0-12.2 2.341 0.9851 0.5735 0.4728 1
2.309 0.9935 0.5981 0.4772 I 3 8.18-10.0 1 4 6.36-8.18 2.093 0.9110 0.5921 0.4714 5 4.96-6.36 1.541 0.5777 0.5223 0.4321 6 4.06-4.96 1.532 0.5454 0.4146 0.3275 7 3.01-4.06 1.614 0.5340 0.2701 0.2193 8 2.46-3.01 1.689 0.5272 0.1445 0.1080 9 2.35-2.46 1.695 0.5298 0.09154 0.05613 10 1.83-2.35 1.677 0.5313 0.04856 0.02940 l 11 1.11-1.83 1.596 0.2608 0.01180 0.002948 '; 12 0.55-1.11 1.241 0.009845 0.000677 -- I 13 0.111-0.55 0.2341 0.0002432 -- -- 14 0.0033-0.111 0.006928 -- -- I i l 1 l i il I !I i l , E-7 Babcock & WilCOX i__.__
1 iI 1 l I 1 I I !I f i I
- I I
iI i e I l
- APPENDIX F 1
References 4 1 ! l i I I I I F-1 babcock & Wilcox
I A. L. Lowe, Jr., et al., Analysis of Capsule OC111-A From Duke Power Company, W Oconee L, clear Statlon, Unit 3 Reactor Vessel Materials Surveillance Program, < BAW-l'.38, Babcock 6 Wilcox, July 1977. l ' ( G.J. Snyder and G. S. Carter, Reactor Vessel Material Surveillance Program, , BAW-10006A, Rev. 3, Babcock & Wilcox, January 1975. DOT 3.5 - Two-Dimensional Discrete Ordinates Radiation Transport Code, (CCC-2 76) , WANL-TME-1982, Oak Ridge Nat ional Laboratory , December 1969. CASK 23E-40-Group Coupled Neutron and Gamma-Ray Cross Section Data, DIC-23E, Radiat ion Shiciding Information Center. W C. L. Whitmarsh, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAU-1485, Babcock & Wilcox, June 1978. = H. S. Palme and II. W. Behnke, Methods of Compliance With Fracture Toughness and Operational Requirements of Appendix G to 10 CFR 50, BAW-10046P, Babcock
& Wilcox, March 1976.
7 H. S. Palme, G. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material I Surveillance Program - Compliance With 10 CFR 50, Appendix H, for Oconee-Class Reactors, BAW-10100A, Babcock & Wilcox, February 1975. I I I I. I1 I I F-2 Babcock & Wilcox _- -_ _. . - . _ _}}